ML18016A889

Forwards Rev 99-1 to Plant EALs for NRC Review & Approval, Per 10CFR50,App E.Encl Provides Comparison of Currently Approved EALs & Proposed Rev 99-01.Approval of EALs Prior to June 1999,requested.With Four Oversize Drawings

ML18016A889
Person / Time
Site:	Harris
Issue date:	04/01/1999
From:	Scarola J CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-99-024, HNP-99-24, NUDOCS 9904090082
Download: ML18016A889 (253)

Text

CATEGORY 1 REGULAT KY INFORMATION DISTRIBUTIO 'YSTEM (RIDS)

ACCESSION NBR:9904090082 DOC.DATE: 99/04/01 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1,'Carolina 05000400 AUTH.NAME AUTHOR AFFILIATION SCAROLA,J. Carolina Power 6 Light Co.

,RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Desk)

SUBJECT:

Forwards rev 99-1 to plant EALs for NRC review E approval, per 10CFR50,App E.Encl provides comparison of currently approved EALs &. propsoed rev 99-01.Licensee requests approval of EALs prior to June 1999.

DISTRIBUTION CODE: A045D COP1ES RECEIVED: LTR l ENCL j SIZE: g 7 TITLE: OR Submittal: Emergency Preparedness Plans, Implement'g Procedures, C T

NOTES:Application for permit renewal filed. 05000400 RECIPIENT COPIES RECIPIENT COPIES 0 ID CODE/NAME LTTR ENCL ID CODE/NAME" LTTR ENCL PD2-1 PD 1 1 FLANDERS,S 1 1 INTERN E CENTER NRR/DRPM/PERB 0.~2 1 2

1 IRO/HAGAN,D.

NUDOCS -ABSTRACT

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1 1 EXTERNAL: NOAC NRC PDR 1 1 (P GAL G ) '1~5 D

N NOTE TO ALL NRIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD) ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED: LTTR 9 ENCL 9

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~ ae CÃGLt Carolina Power & Light Company James Scarola PO Box 165 Vice President New Hill NC 27562 Harris Nuclear Plant u,pR -1 1999 SERIAL: HNP-99-024 10 CFR 50, Appendix E United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 EMERGENCY ACTION LEVELREVISION 99-1

,

Dear Sir or Madam:

In accordance with 10 CFR 50, Appendix E, Carolina Power & Light Company (CP&L) is submitting the Harris Nuclear Plant Emergency Action Level (EAL) Revision 99-1 for your review and approval. These emergency action levels have been discussed and agreed upon by CP&L, the State of North Carolina and local governmental authorities as required by 10 CFR 50, Appendix E, IV.B. Two (2) full size copies of the EAL Flow Path (side 1 and side 2) are included with this letter.

Enclosure 1 to this letter provides a comparison of the currently approved EALs and the proposed revision 99-1. Enclosure 1 also provides a summary of the 10 CFR 50.54(q) evaluation performed by CP&L, including the basis for the determination that the changes do not decrease the effectiveness of the Emergency Plan.

The Harris Nuclear Plant Emergency Action Levels Comparison with NUREG-0654 Attachment 2 document has been revised (Rev. 99-1) and is provided as Enclosure 2 to this letter.

It is CP&L's intent to implement the EAL Revision 99-1 following receipt of NRC approval. CP&L currently plans to be prepared to implement the EALs prior to June 1999. Therefore, CP&L requests the NRC's review of these EALs in support of this schedule.

Questions regarding this matter may be referred to Mr. J. H. Eads at (919) 362-2646.

Sincerely, 9904090082 99040i PDR ADQCK 05000400 -.

F PDR MGW Enclosures Mr. J. B. Brady (NRC Seniorlresidenr Inspeeror, HNP)

)~g,n Can+~

Mr. Rich Laufer (NRR Project Manager, HNP)

Mr. L. A. Reyes (NRC Regional Administrator, Region II) gayitjo ~

~> 5413 Shearon Harris Road New Hill, NC Tel 919 362-2502 Fax 919 362-2095 sg

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01

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~Back 'round and Scope: '.'his EAL Revision is being approved for submittal to the NRC to obtain their approval, as required by 10CFR50, Appendix E, Section IV.B, prior to implementation. Implementation will be accomplished through incorporation into a revision to PLP-201 and PEP-110 following receipt of NRC review and approval.

This EAL Revision 99-01 provides for improvements within the following four categories:

A. Clarification of terminology

1. "Site Emergency" classification title was updated to "Site Area Emergency."

2. Upgraded text of "Security Alert" and "Security Emergency" related assessments with more descriptive and equivalent criteria. Also, added classification criteria for Civil Disturbances.

3. Side 1, Column 2, added reference to use of either MCB Trip switch satisfies the criteria (ATWS assessment).

4. T.S. corrective actions for Fuel Damage and RCS Leakage Unusual Events have been clarified through inclusion of those actions which are allowed (in place of reference to the T.S.)

5. "Inability to comply with Tech Spec shutdown requirements" Unusual Event Category name changed to .

"Other plant or equipment problems." The new name provides a better description of events in this category, due to addition of EAL 8-2-1 (Inadvertent Criticality) and relocation of Turbine Rotating Component Failure from category 10 to this category. Similarly renamed UE Category 10 from "Other Plant Hazards or Events" to "Other Hazards."

6. UE Category 3, "note" format of break clarification was changed to a parenthetical phrase to be consistent with other and similar Unusual Event clarifying information.

7. UE Category 6, changed lettered elements of these ICs to bullets for consistency within other EALs.

8. UE Category 9, provided definition of conditions equivalent to a Hurricane and re-worded Tornado aspect B. Addition of EAL reference numbers to enhance communication and aid in offsite understanding of events.

1. EAL Reference numbers have been added to EALs (in place of the grid coordinates contained in the previous revision). These numbers are now shown for each action step (rounded rectangle) associated with "EAL Exceeded." Also, added a reference to the EAL numbering scheme. These EAL reference numbers aid in communication of events between facilities and are linked to an EAL Reference Manual provided to offsite authorities which aids them in understanding the event, in layman's terms, with graphical representations as applicable. An aEAL Ref. No." designator has been added to the EAL Status Board (upper right hand come'r of both sides of the flow path).

2. An additional action step (rounded rectangle) has been added to all General Emergency terminus items to indicate which EAL has been exceeded.

3. Added Unusual Event category numbering (1 -11). These correspond to the EAL numbering scheme.

C. Added Turbine Bldg. Drains radiation monitor to EAL Table 1 and EAL Table 5 as an additional potential liquid effluent release point.

D. Provide reactivity control related EAL assessment which was more complete and consistent with regard to FSAR accident analysis. Also, to be consistent with NUREG-0654 event classification criteria and event definitions.

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Comparison (A. Clarification of Terminology):

Ref. Rev. 96-02 Rev. 99-01:

SAEs Event Classification title "Site Site Area Emergency.

Emergency" GE SECURITY EMERGENCY AS A HOSTILE FORCE HAS TAKEN CONTROL OF PLANT 7-1-4 DEFINED BY THE SECURITY PLAN EQUIPMENT AND/OR VITALAREA(S) SUCH THAT PLANT with LOSS OF PHYSICAL CONTROL PERSONNEL ARE UNABLE TO OPERATE EQUIPMENT OF THE PLANT REQUIRED TO MAINTAINSAFETY FUNCTIONS NUMARC/NESP-007 Rev. 2 Item HG1 (used as basis for the above revised EAL):

Security Event Resulting in Loss Of Abilityto Reach and Maintain Cold Shutdown EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Loss of physical control of the control room due to security event.

2. Loss of physical control of the remote shutdown capability due to security event.

BASIS This IC encompasses conditions under which hostile force has taken physical control of vital area required to reach and maintain safe shutdown.

NEI 97-03 Final Draft Rev. 3 October 1998 ITEM HG1 TEXT (used in the development of the HNP text:

Security Event Resulting in Loss Of Physical Control of the Facility.

EXAMPLE EMERGENCY ACTION LEVEL'.

A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.

BASIS

... Loss of physical control of the control room or remote shutdown capability alone may not prevent the ability to maintain safety functions per se....

SAE SECURITY EMERGENCY AS EITHER OF THE FOLLOWING SECURITY EVENTS WITHIN A 7-1-3 DEFINED BY THE SECURITY PLAN VITALAREA:

with SUCCESSFUL PENETRATION BOMB DISCOVERED WITHIN A VITALAREA OF VITALAREAS, and ACTUALOR POTENTIALLYAFFECTING SAFETY RELATED IMMINENTPOTENTIAL FOR EQUIPMENT OFFSITE RAD RELEASE CONFIRMED INTRUSION INTO A VITALAREA BY A HOSTILE FORCE NUMARC/NESP-007 Rev. 2 Item HS1 (used as basis for the above revised EAL):

Security Event in a Plant Vital Area.

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Intrusion into plant vital area by a hostile force.

2. Other security events as determined from (sire specific) Safeguards Contingency Plan.

NEI 97-03, Final Draft, Rev. 3 (October 1998), ITEM HS1 added "BOMB discovered within the VITAL AREA potentially affecting (site specific) Safety Related Equipment" as an additional Example EAL.

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Comparison (A. Clarification of Terminology):

Ref. Rev. 96-02 Rev. 99-01:

Alert SECURITY EMERGENCY AS EITHER OF THE FOLLOWING SECURITY EVENTS WITHIN 7-1-2 DEFINED BY THE SECURITY PLAN THE PA:

with BOMB DISCOVERED WITHIN THE PA POTENTIALLY AFFECTING SAFETY RELATED EQUIPMENT a) NO SUCCESSFUL PENETRATION IMMINENTTHREAT OF, OR ACTUALINTRUSION INTO OF VITALAREAS, or THE PA BY A HOSTILE FORCE b) PENETRATION OF VITALAREAS but NO ACTUALOR IMMINENT POTENTIAL FOR OFFSITE RAD RELEASE NUMARC/NESP-007 Rev. 2 Item HA4 (used as basis for the above revised EAL):

Security Event in a Plant Protected Area.

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2)

1. Intrusion into plant protected area by a hostile force.

2. Other security events as determined from (site specific) Safeguards Contingency Plan.

NEI 97-03, Final Draft, Rev. 3 (October 1998), ITEM HA4 added "BOMB discovered within the PROTECTED AREA potentially affecting (site specific) Safety Related Equipment" as an additional Example EAL, UE A SECURITY ALERT HAS BEEN CONFIRMED SECURITY EVENT WHICH INDICATES A 7-1-1 DECLARED AS DEFINED IN THE POTENTIAL DEGRADATION IN THE LEVEL OF SAFETY OF SECURITY PLAN THE PLANT AS INDICATED BY:

UNAUTHORIZED ALTERATIONOR TAMPERING HAS OR IS OCCURRING AFFECTING SAFETY RELATED EQUIPMENT HOSTAGE / EXTORTION SITUATION THATTHREATENS TO INTERRUPT NORMAL PLANT OPERATIONS CIVIL DISTURBANCE ONGOING BETWEEN THE SITE BOUNDARY AND THE PROTECTED AREA NUMARC/NESP-007 Rev. 2 Item HU4 (used as basis for the above revised EAL):

Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant.

EXAMPLE EMERGENCY ACTION LEVELS: (1 or 2) 1.Bomb device discovered within the plant Protected Area and outside the plant Vital Area.

2. Other security events as determined from (site specific) Safeguards Contingency Plan.

NEI 97-03 Final Draft Rev. 3 October 1998 ITEM HU4 TEXT (used in the development of the HNP text:

EXAMPLE EMERGENCY ACTION LEVEL:

1. Security events as determined from (site specific) Safeguards Contingency Plan and reported by the (site specific) security shift supervisor.

BASIS:

....Consideration should be given to the following events:

~ SABOTAGE has or is occurring affecting Safety Related Equipment

~ HOSTAGE / EXTORTION situation that threatens to interrupt NORMALPLANT OPERATIONS

~ CIVIL DISTURBANCE ongoing between the site perimeter (or other site specific nomenclature) and PROTECTED AREA

~ Hostile STRIKE ACTION at the facility which threatens to interrupt NORMAL PLANT OPERATIONS (judgment based on behavior of Strikers and/or intelligence received)

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Comparison (A. Clarification. of Terminology):

Ref. Rev. 96-02 Rev. 99-01:

UE VALID"HIGH ALARM"ON ANY OF EAL 1-1-1 1-1-1 THE MONITORS IN EAL TABLE 5 VALID"HIGH ALARM"ON ANY OF THE MONITORS IN EAL AND THE RELEASE HAS NOT BEEN TABLE 5 TERMINATED.

(UNUSUAL EVENT EXISTS UNTIL AND EFFLUENT DISCHARGE IS THE RELEASE HAS NOT BEEN TERMINATED.

TERMINATED AND ALL REQUIRED NOTIFICATIONS ARE MADE, (UNUSUAL EVENT EXISTS UNTIL EFFLUENT DISCHARGE IS WHICHEVER IS LATER) TERMINATED AND ALL REQUIRED NOTIFICATIONS ARE MADE)

UE RCS SPECIFIC ACTIVITYEXCEEDS RCS SPECIFIC ACTIVITYEXCEEDS TECHNICAL 2-2-1 TECHNICAL SPECIFICATION 3.4.8 SPECIFICATION 3.4.8 LIMITS FOR DE I-131 OR GROSS WITH LCO ALLOWED CORRECTIVE RADIOACTIVITY.

ACTION TIME ELAPSED (FOR DE I-131 THE EAL IS NOT EXCEEDED UNLESS THE 48 HOUR TIME INTERVAL, OR FIG. 3.4-1 LIMITS ARE EXCEEDED.)

UEs LOSS OF REACTOR COOLANT OR 4-2-1 ANY RCS PRESSURE BOUNDARY LEAKAGE 4-2-1 PRIMARY TO SECONDARY LEAKAGE

& IN EXCESS OF TECHNICAL SPECIFICATION 3.4.6.2 WITH LCO 4-3-1 ANY OTHER RCS LEAKAGE IN ACCESS OF 4-3-1 ALLOWED CORRECTIVE ACTION TECHNICAL SPECIFICATION 3.4.6.2 WITH THE 4 HOUR TIME ELAPSED CORRECTIVE ACTIONS NOT SATISFIED.

EALs MCB MANUALREACTOR TRIP MCB MANUALREACTOR TRIP SUCCESSFUL (EITHER 8-1-2 SUCCESSFUL SWITCH) 8-1-3 UE Category title "INABILITYTO COMPLY Category title "OTHER PLANT OR EQUIPMENT PROBLEMS" MATRIX WITH TECH $ PEC $ HUTDOWN REQUIREMENTS" UE Category title "OTHER PLANT Category title "OTHER HAZARDS" MATRIX HAZARD$ OR EVENT$

UE (NOTE: A BREAK IS A LEAK WHICH MAIN STEAM LINE OR FEEDWATER LINE BREAK 3-3-1 EXCEEDS THE OPERATORS ABILITY (A BREAK IS A LEAK WHICH EXCEEDS THE OPERATORS TO SHUTDOWN THE PLANT IN A ABILITYTO SHUTDOWN THE PLANT IN A CONTROLLED CONTROLLED MANNER OR TO NOT MANNER OR TO NOT EXCEED TECH. SPEC COOLDOWN EXCEED TECH. SPEC COOLDOWN LIMITS)

LIMITS)

MAIN STEAM LINE OR FEEDWATER LINE BREAK UE 6-1-1 A)

B) ..

UE 6-2-1 A)

B) - E) ..

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Comparison (A. Clarification of Terminology):

Ref. Rev. 96-02 Rev. 99-01:

UE 2) HURRICANE OR TORNADO EAL 9-2-1 9-2-1 CROSSING THE EAB SUSTAINED WIND SPEED AT 10 METERS OF 74 MPH OR 8

GREATER 9-3-1 EAL 9-3-1 TORNADO REPORTED WITHIN THE EAB.

Change Co'mparison (B. Addition of EAL Reference Numbers):

Ref. Rev. 96-02 Rev. 99-01:

All "Added to Document" 1. EAL Reference numbers FOR EACH "EAL Exceeded" rounded rectangle.

2. "EAL Ref. No." designator has been added to the EAL Status Board (upper right hand corner of both sides of the flow path).

3. Unusual Event Matrix Column numbers have been added.

4. Explanation of EAL Reference numbers added to side 2 which reads as follows:

EAL REFERENCE NUMBERS (X-Y-Z):

X = CATEGORY (1 - 11)

Y = IDENTIFIER WITHIN CATEGORY Z = CLASSIFICATION (1-4) 1 = UNUSUAL EVENT 2 = ALERT 3 = SITE AREA EMERGENCY 4 = GENERAL EMERGENCY GES "Added to Document" GENERAL EMERGENCY EXCEEDED DECLARE GENERAL DECLARE GENERAL EMERGENCY EMERGENCY MAKE NOTIFICATIONS MAKE NOTIFICATIONS RE-EVALUATE EALS AS RE-EVALUATE EALS AS CONDITIONS WARRANT CONDITIONS WARRANT UE TURBINE ROTATING COMPONENT Item wording is unchanged, however, the item was moved from 8-3-1 FAILURE RESULTING IN.... UE Category 10 and renamed Category 8.

AII Alpha-Numeric Grid system listed at "Removed from Document" outside border of both sides of EAL Flow Path Page 5 of 11

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HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Comparison=(C. Liquid Effluent Release Assessment): .

Ref. Rev. 96-02 Rev. 99-01:

Tbl 1 "Added to Document" TURBINE BLDG DRAIN REM-1MD-3528 2LL276 RMD3528A Tbl 5 "Added to Document" 8. TURBINE BLDG DRAINS MONITOR Change Comparison (D. Reactivity Control Related EAL Assessments):

Ref. Rev. 96-02 Rev. 99-01:

Alert UNCONTROLLABLEBORON "Removed from Document" Refer to UE 8-1-1 below DILUTIONwith PLANT NOT IN MODE 6 with DILUTION EVENT LASTING )

15 MIN SAE UNCONTROLLABLEBORON "Removed from Document" Refer to UE 8-1-1 below DILUTION with PLANT IN MODE 6 with DILUTION EVENT LASTING > 35 MIN UE "Added to Document" INADVERTENTCRITICALITY- EXTENDED AND 8-1-1 UNPLANNED SUSTAINED POSITIVE STARTUP RATE (THIS DOES NOT INCLUDE CRITICALITYEARLIER THAN ESTIMATED DURING PLANNED REACTOR STARTUPS)

NEI 97-03 FlnaI Draft Rev. 3 October 1998 ITEM SU8 TEXT (used In the development of the text for the above r'evised EAL):

EXAMPLE EMERGENCY ACTION LEVEL

1. An extended and" UNPLANNED positive period or sustained positive startup rate observed on nuclear

'-instrumentation.: .

BASIS, This IC addresses inadvertent criticality events. While the primary concern of this IC is criticality eve'nts that occur in Cold Shutdown or Refueling modes, the IC is applicable in other modes in which inadvertent criticalities are possible. This IC serves as a precursor to the Fission Product Barrier Matrix or the Cold Shutdown / Refu'cling Safety. Function Matrix. Inadvertent criticalities indicate a potential degradation of the level of safety of the plant, warranting an Unusual Event classification. This IC excludes that occur during planned reactivity changes associated with reactor startups (e.g., criticality inadvertent-,'riticalities earlier than estimated).....

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HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 rogiam Requirements:

~ 10CFR50.47(b)(4) A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

~ NUREG-0654,Section II, D.1, An emergency classification and emergency action level scheme as set forth in Appendix 1 must be established by the licensee. The specific instruments, parameters or equipment status shall be shown for establishing each emergency class, in the in-plant emergency procedures. The plan shall identify the parameter values and equipment status for each emergency class.

~ NUREG-0654,Section II, D.2, The initiating conditions shall include the example conditions found in Appendix 1 and all postulated accidents in the Final Safety Analysis Report (FSAR) for the nuclear facility.

~ NRC Emergency Preparedness Position (EPPOS) on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1, as implemented in HNP EAL Revision 96-01.

Change Assessment (A. Clarification of terminology):

1. The restoration of the title of the Site Area Emergency event classification (from Site Emergency) assists offsite agencies involved in the Harris EP program through achievement of consistency in terminology between utilities with emergency planning zones in the sate of North Carolina. The classification title change is a minor change and does not introduce any potential confusion or complication.

2. The previously approved EAL criteria for "Security Alert" and "Security Emergency" relied heavily on accurate communication between the HNP Security force and personnel responsible for event classification. The terminology was based on defined terms within the security program and procedures. The use of the term "Security Alert" as being an "Unusual Event" and a "Security Emergency" as being an "Alert" or higher emergency classification allowed significant potential for confusion. Also, the criteria were quite cryptic and non-descriptive.

The proposed criteria remain consistent with the example initiating conditions of NUREG-0654. A review of industry best practices associated with these conditions resulted in the choice of terminology closely resembling that of the second issuance of NEI 97-03, Final Draft Rev. 3 (October 1998) for use in describing the appropriate conditions for each event classification associated with security events.

~ The revised criteria for Unusual Event conditions continue to correspond to the NUREG-0654 condition of "Security threat or attempted entry or attempted sabotage." Events within this category would be classified the same under both the old and the new criteria, with the exception of the "Civil Disturbance" which would not have been an Unusual Event under the old criteria. Even though there has been no activity of this nature at the Harris Plant, the "Civil Disturbance" criteria provides for consistency with the industry standard (NEI 97-

03) and could potentially be associated with an "attempted entry." The NEI 97-03 reference to "Hostile Strike Action" was not included due to the fact that HNP is a non-union plant in a right to work state.

~ The revised criteria for Alert conditions continue to correspond to the NUREG-0654 condition of "Ongoing security compromise." HNP chose to utilize terminology of "Imminent threat of or actual intrusion into the PA by a hostile force" as a clarification and more conservative version of the NEI terminology.

~ The revised criteria for Site Area Emergency conditions continue to correspond to the NUREG-0654 condition of "Imminent loss of physical control of the plant" through more descriptive and anticipatory references to events of significance within plant "Vital Areas."

~ The revised criteria for General Emergency conditions continue to correspond to the NUREG-0654 condition of "Loss of physical control of the plant" through descriptive terminology to include maintenance of "Safety Functions" as the definition of "control of the facility."

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Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Assessment (A. Clarification of teiminology):

3. Side 1, Column 2, added reference to use of either MCB Trip switch as clarification of the original intent of the criteria of "MCB Manual Reactor Trip Successful" (ATWS assessment). This assessment is evaluating "continued power generation" aspect of the NUREG-0654 criteria, not electronics or logic of the circuitry. The successful use of either MCB trip switch provides for rapid plant shutdown and suspension of power generation.

Event classification criteria is not altered by this change.

4. T.S. corrective actions for Fuel Damage and RCS Leakage Unusual Events have been clarified through inclusion of those actions which are allowed (in place of reference to the T.S.). This insures that the intent is clearly understood without need for an external reference. This also eliminates the potential inappropriate allowance of consequential actions of Tech Specs as an element of the intended "corrective actions." Event classification criteria are not altered by this change.

5. "Inability to comply with Tech Spec shutdown requirements" Unusual Event Category name changed to "Other plant or equipment problems." The new name provides a better description of events in this category, due to addition of EAL 8-2-1 (Inadvertent Criticality) and relocation of Turbine Rotating Component Failure from category 10 to this category. Similarly renamed UE Category 10 from "Other Plant Hazards or Events" to "Other Hazards."

The new category names do not alter the conditions or criteria which result in the event classifications. The new names have been reviewed for clarity, provide consistency with the new EAL numbering scheme and do not introduce confusion.

6. UE Category 3, "note" format of break clarification was changed to a parenthetical phrase to be consistent with other and similar Unusual Event clarifying information. Event classification criteria is not altered by this change.

7. UE Category 6, changed lettered elements of these ICs to bullets for consistency within other EALs. Event classification criteria is not altered by this change.

8. UE Category 9, provided definition of conditions equivalent to a Hurricane and re-worded Tornado aspect. This provides clarification in that NOAA and Weather Services do not define "Hurricane" at inland locations such as HNP. Also, reworded "Tornado crossing the EAB" to "Tornado reported within the EAB" as more symptomatic.

Event classification criteria is not altered by this change.

Change Assessment (B. Addition of EAL'eference numbers):

The introduction of an EAL numbering scheme does not alter the parameters, symptoms, events or conditions which result in any of the four (4) event classifications. The numbering scheme has been evaluated for human factors considerations and has been determined to provide a benefit in communication of event conditions without introduction of extra effort or activities. The addition of the numbering scheme, and related flow path changes improves the effectiveness of the flowpaths for event classification without interference in any previously committed emergency plan related capabilities.

Change Assessment (C. Liquid Effluent Release Assessment):

The listing of effluent monitors has been expanded to be more comprehensive. The Turbine Bldg. Drains radiation monitor was added to EAL Table 1 and EAL Table 5 as an additional potential liquid effluent release point. During primary to secondary leak or tube rupture scenarios, secondary system leakage could accumulate in the turbine building drain system. The radiation monitor for this path has therefore been added to the potential effluent points listing.

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HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange Assessment (D: Reactivity Control Related EAL Assessments):

This revision of the HNP EALs includes the results of a close look at reactivity control events and the appropriate event classification of these events. The revised EAL scheme for these events is more in line with the NUREG-0654 criteria, event classification definitions, and FSAR descriptions and analysis of these events.

HNP EALs are based on NUREG-0654, example initiating conditions, and have been amended to improve certain EAL conditions utilizing NUMARC/NESP-007, Rev. 2, methodologies in accordance with Reg Guide 1.101, Rev. 3 and EPPOS ¹1.

The current HNP EALs contain two (2) event classifications for Boron Dilution related events. These are:

~ Site Area Emergency for an uncontrollable dilution event lasting greater than 35 minutes in mode 6

~ Alert for an uncontrollable dilution event lasting greater than 15 minutes at other times The Boron Dilution event classifications were not specified within either NUREG-0654 or NESP-007. These EAL conditions were added to the HNP EAL scheme in Revision 5 to the Emergency Plan as a more specific criteria of "Other events." This was done in response to:

~ Current (1994 and early 1995) industry concern and investigation into potential consequences of Boron dilution events

~ Simulation runs and training on HNP procedures and EALs being performed at the Seabrook station simulator (the HNP simulator was not yet functional)

~ Incorporation of FSAR chapter 15 reactivity control accident analysis.

Since this time, Generic Letter 85-15, Inadvertent Boron Dilution Events, was issued to all PWR Licenses regarding the staff position resulting from the evaluation of Generic Issue 22. This letter stated in part that "The staff determined that while power excursions during boron dilution events are possible if the operator does not take any ction and sufficient volume of dilution water is available, the excursion should be self limiting. The staff analyses ndicate that these type of boron dilution transients should not exceed the staff's acceptance criteria." Also, the letter states that "the consequences are not severe enough to jeopardize the health and safety of the public .. "

HNP FSAR Sections 4.3.1, 9.3.4, 15.4.6 and NUREG-800, Standard Review Plan, for section 15.4.6, describes the Chemical and Volume Control (CVCS) malfunctions and specifies that the operator is provided sufficient time to correct the situation in a safe and orderly manner. These events are classified as ANS Condition II Incidents. ANS Condition II Incidents are accommodated with, at most, a shutdown of the reactor with the plant capable of returning to operation after corrective action. Fuel damage is not expected during ANS Condition II Events.

Per the current HNP FSAR Section 15.4.6.2, Analysis of Effects and Consequences, item 2, Dilution During Refueling, An uncontrolled boron dilution accident cannot occur during refueling as a result of a reactor coolant makeup system malfunction. This accident is prevented by administrative controls which isolate the RCS from the potential source of unborated water... " Plant Program PLP-629, Reactivity Management Program, insures this requirement is met.

The boron dilution criteria specified in the previous HNP EALs, based on the above factors, do not meet the Alert and Site Area Emergency class descriptions of NUREG-0654s.

~ ALERT - "Events are in process or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fraction of the EPA Protective Action Guideline exposure levels."

~ SITE AREA EMERGENCY - "Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases not expected to exceed EPA Protective Action Guideline exposure levels except near site boundary."

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HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 hange'Assessment; (D: Reactivity-Control':Related EAL Assessments): -'-

The HNP evaluation of reactivity control (born dilution events) concluded that the existing Alert and Site Area Emergency EAL criteria were not consistent with the plant design basis and event classification objectives.

The NUREG-0654 class description for an Unusual Event is "Unusual events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occur" was consistent with an uncontrolled reactivity mismanagement event of consequence.

A boron dilution event is one type of reactivity mismanagement event. Per FSAR section 15.4 a slow uncontrolled RCCA bank withdrawal event exhibits similar characteristics, and from an accident analysis perspective, is more limiting. These type of reactivity control events may be precursors to other events, and as such may warrant an Unusual Event declaration. As such, the revised HNP EALs selected terminology contained in the second issuance of NEI 97-03, Final Draft, Rev. 3 (October 1998), Item SU8 for a replacement EAL condition.

This new EAL provides a site specific example of other events which meet the class description of an Unusual Event.

NUREG-0654 and NUMARC/NESP-007, Rev. 2, do not contain any specific criteria associated with this condition. As such, the NEI 97-03 terminology was selected, due to it providing a clear and measurable symptomatic assessment of the condition (resulting from either chemical poison, control rod, or any reactivity change).

Escalation to higher event classification would occur via either loss of function associated with in-ability to compensate for a dilution or other undesired/uncontrolled positive reactivity addition, fission product barrier analysis, or judgment (EAL 11-1-2).

SOER 94-2, Boron Dilution Events in Pressurized Water Reactors, was reviewed to confirm the validity of these EAL hanges. Both of the conditions presented (rapid boron dilution during shutdown operations due to stagnant or low-low areas, and slow dilution events) correspond to potential pre-cursor conditions. The revised HNP EAL scheme properly addresses the event and potential escalation. The revised EAL scheme also continues to comply with the anticipatory intent of classification activities.

The Robinson plant is the other PWR operated by CP8 L. The Robinson plant does not have a boron dilution related event classification, and as such the deletion of the existing Alert and Site Area Emergency EALs does not introduce any inconsistency within the utility.

Change Assessme'nt Conclusion The proposed EAL changes represent a positive impact on assessment and classification of events through clarification of terminology and achievement of consistency.

Effectiveness of the event classification process is enhanced through this revision. The ability of the plant staff to consistently and properly classify occurrences and events at or near the plant are enhanced. The changes do not introduce any adverse affects on other activities Justification:

Not Applicable as this change does not represent a degradation to the program.

Page 10 of 11

~ ~ ~

Document

Title:

HNP Emergency Action Level Flow Path No: EAL Rev.: 99-01 eferences:

~ 10CFR550.47(b)(4)

~ NUREG-0654,Section II, Items D.1 8 D.2

~ NUMARC/NESP-007, Rev. 2, Methodology for Development of Emergency Action Levels, January 1992

~ Reg Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors, Rev. 3.

~ NRC EPPOS No. 1 and No. 4

~ Generic Letter 85-05, Inadvertent Boron Dilution Events

~ NUREG-0800, Section 15.4.6, Rev. 1

~ FSAR sections 4.3.1, 9.3.4.1 and 15.4.6

~ EMF-98-054(P), Rev. 1, Harris Cycle 9 Safety Analysis Report, Section 7.6

~ Design Basis Document (DBD-314) HNP Parameters Document, Table 2.15

~ PLP-629, Reactivity Management Program

~ Memorandums SHNPPA-84-1217 8 SHNPPA-85-142

~ SOER94-2, Boron Dilution Events in Pressurized Water Reactors Note: Per 10CFR50, Appendix E, Section IV.B, this revision of the EALs was discussed, reviewed, and agreed on by authorities from the State of N.C. and the four EPZ County Emergency Management offices.

Page11 of11

'NP-99-024 Enclosure 2 NUREG-0654, Attachment 2, Example Initiating Conditions Cross Reference to HNP Emergency Action Levels, Rev. 99-1 This revision includes EAL changes associated with:

1. Replacement of "Boron Dilution" related event classification criteria with "Inadvertent Criticality" criteria.

2. Updated the terminology associated with assessment and classification of security related events,

3. Added Turbine Building Drain as a potential radiological liquid effluent pathway.

4. Specified meteorological conditions in place of "Hurricane."

5. Clarified Fuel Damage Indication and Loss of Reactor Coolant Unusual Event conditions references to "LCO corrective action time."

NOTES:

1. EAL Reference numbers have been added for:

~ ease of reference (replaces previous coordinate system) & improved communication of EAL conditions

~ improved interface with off-site agencies via an eHNP EAL Reference Manual" developed for this purpose.

2. "Site Emergency" classification title was updated to "Site Area Emergency."

3. Appearance/format of this document has been revised as a result of updates to computer word processing software. These changes are not marked by rev.

Bars.

4. Affected sections 8 text are indicated with a rev. bar in the right margin Prepared by:

Approved by:

Supervisor - merg ncy Preparedness Date NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 1 of 75 I

PREFACE The philosophy applied, directed by NUREG-0654, while building the HNP Emergency Action Level Flowpath is that the classification of emergencies should be anticipatory to allow early notification for events that could escalate into major specific events and major releases.

In many cases the NUREG-0654 recommendation was based upon the actual or predicted loss of one or more of the Fission Product Barriers. Instead of attempting to develop a group of procedures that would address the specifics recommended in NUREG-0654, HNP took the approach that the loss of a Fission Product Barrier should be analyzed. Action should be taken based upon the number of Barriers that have been breached or are in jeopardy of being breached.

In order to accomplish this task, the plant had to determine what indications would show that any single Fission Product Barrier had been breached or potentially could be breached (is in jeopardy). This task was accomplished and resulted in the development of the first column of Side 1 of the EAL Flowpath.

The Site Emergency Coordinator can quickly assess the status of the three Fission Product Barriers by answering the questions listed on the flowchart. In this manner, if the Fission Product Barriers are breached or in jeopardy (potential for breach is present), the event can be quickly escalated to the correct classification.

Once all of the Fission Product Barriers are examined, the Flowpath is completed to determine if any other reason exists that would require the classification of an Emergency Action Level (EAL). In using this method, we handle the major emergencies, followed by other types of emergencies which could become major emergencies.

The average SRO licensed individual is expected to take less than 15 minutes to go through the entire Flowpath and determine that an Emergency Classification is, or is not, warranted. In the case of an Unusual Event, Alert, or Site Area Emergency condition, the time delay is acceptable. If conditions are quickly deteriorating, a faster method of classification is needed.

Because of this, the Site Emergency Coordinator is directed on the Flowpath to declare a General Emergency as soon as it is determined to exist. The declaration of a Site Area Emergency, Alert, or an Unusual Event may be slightly delayed, as necessary, to finish the Flowpath and to find out if a higher level classification exists.

The Flowpath is designed to look at the worst case events first, then the other events in descending order of importance. Some consideration was given to the layout of the path, so some of the potential events were moved together to group related conditions in order to make the path flow in a more logical fashion.

The EAL Flowpath follows the same rules and conventions that the EOP (Emergency Operating Procedure) Flowpaths follow. This is an aid to the users in that they do not have to learn two conventions.

NUREG-0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99.1 Pago 3 of 75

PREFACE (continued)

Entry Point X provides a generic starting point for evaluating the Emergency Action Levels. Entry into Abnormal Operating Procedures or Emergency Operating Procedures provides direction to initiate monitoring of the EALs from this entry point to determine if the off normal situation warrants declaration of an emergency.

The NRC Emergency Preparedness Position (EPPOS) on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1 (EPPOS No.1) has been applied to the Emergency Action Levels to provide clarification, or improvements, to the EALs and Initiating Conditions.

The HNP EALs have had reference numbers assigned. These are in the format (X-Y-Z) where X = Category. (1 - 11), Y = Identifier Within Category, and Z =

Classification (1-4). The Classification numbers are 1 = Unusual Event, 2 =

Alert, 3 = Site Area Emergency, and 4 = General Emergency. These EAL reference numbers aid in communication of events between facilities and are linked to an EAL Reference Manual provided to offsite authorities which aids them in understanding the event, in layman's terms with graphical representations as applicable.

FISSION PRODUCT BARRIER ANALYSIS A. General Each of the Fission Product Barriers is analyzed to determine if it is breached or in jeopardy. The first indication of either event results in declaring the specific barrier to be either breached or jeopardized. From an EAL declaration standpoint, it does not matter whether the barrier is breached or in jeopardy.

However, it effects actions executed by other plant documents; therefore, the breach indicators are examined before the jeopardy indicators.

The assessment of plant conditions, performed as a part of plant operators progressing through the Emergency Operating Procedure (EOP) Network, identifies indications that would prove one or more of the Fission Product Barriers to be breached. The plant staff has analyzed the EOP Network to determine those points at which any one (or more) of the barriers indicates a breach. These points are identified in the EAL Flowpath as entry points T, U, and V. If a Fuel breach is indicated in the EOP Network, the EOP Network orders the operators to enter the EAL Flowpath at ENTRY POINT T. The same process is used for entry points U (RCS breached) and V (Containment breached).

NUREG.0654, Att. 2, Exampfa IC Cross Reference to HNP EALs, Rev. 99-1 Pago 4 of 75 I

II.A. FISSION PRODUCT BARRIER ANALYSIS, General (continued)

EOP network references to EAL Flowpath Entry Points T, U, and V include the following conditions:

~ Containment Hydrogen (H2) concentration > 4% results in Entry Points T and V for Fuel and Containment breach. Four percent H2 is indicative of zirconium water reaction associated with inadequate core cooling and fuel damage. The Containment is classified as breached with H2 in Containment being at the flammability threshold. This classification occurs well below the explosive limit of 18.2%.

~ Procedure steps that have identified a Steam Generator Tube Rupture reference EAL Entry Point U, for a breach of the Reactor Coolant System fission product barrier.

~ If the procedures identify an increase in Reactor Auxiliary Building radiation levels, or a loss of primary coolant outside Containment, then EAL Entry Point V is also referenced for a loss of the Containment fission product barrier.

~ Functional Restoration Procedures that direct the operations staff to vent the Reactor Coolant System (RCS) to Containment, or to otherwise initiate a bleed path from the RCS to Containment refer to EAL Entry Point U for a breach of the RCS fission product barrier.

These entry points serve two purposes:

1. They force a reentry into the EAL Flowpath in case the Emergency Action Level may need to be upgraded.

2. Since the entry point, as determined by the EOP's, has already determined that one of the Fission Product Barriers is breached, it reduces the time necessary to arrive at the correct Emergency Classification.

The assessment of the three fission product barriers is performed using the first column of Side 1 of the EAL flow path. The decision blocks used to evaluate the fission product barriers are grouped together such that the Fuel, then Reactor Coolant System, and lastly the Containment are evaluated.

Some parameters or conditions are indicative of a breach of more than one barrier.

In these cases, two or all three barriers affected are indicated in the action block following the applicable response from the flow path decision block.

The parameters used to evaluate the barriers, and the bases for utilizing these parameters are described in the following three sections (B through D).

NUREG.0654, An. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 5 of 75

II.B. FISSION PRODUCT BARRIER ANALYSIS FUEL ANY RAD MONITOR EAL TABLE 1 IN HIGH ALARM? - If none of the Rad monitors are in alarm, the next 3 decision blocks can be bypassed which minimizes the time needed to go through the Flowpath.

PLANT VENT STACK 41 WRGM EFFL CHNL >3.6 E5 IJCi/sec? - This is an indication that all 3 FPBs are breached. The stack effluent monitor would exceed this level if the Containment airborne concentration of radioactivity was due to an RCS activity of 300 IJCi/cc (I-131) with a 50 gpm RCS leak. Containment is assumed breached via the purge system running at 1500 CFM with no credit taken for cleanup systems. This includes the dilution effects that are predicted to occur during the release through the Plant Vent Stack release path.

3. EITHER CNMT HI RANGE ACCIDENT MON >17.5 R/HR? - The CNMT monitors would not indicate this level of radiation unless a fuel breach and an RCS breach had occurred. This radiation level is based upon 300 pCI/cc RCS activity (I-131) (Alert level) and 40 gpm leakage from the RCS to Containment in addition to the 10 gpm allowable by Technical Specifications.

ANY EAL TABLE 2 MONITOR >1000 TIMES NORMAL? - The 1000 x )

normal EAL Table 2 value was taken directly from the NUREG-0654 recommendations.

5. WAS ENTRY POINT AT T? - If the entry point was at T, the fuel fission product barrier is indicating breached, based upon the EOP Network determination. The remainder of the fuel fission product barrier evaluation is bypassed.

6. GFFD INCREASED >1.0E5 CPM IN 30 MINUTES? - An increase of this magnitude indicates that the fuel fission product barrier is breached. The set point is below the NUREG-0654 "Alert" classification (based on 1%

failed fuel in 30 minutes or 5% failed fuel). Only two set points were provided by Westinghouse on the Gross Failed Fuel Detector (the lower set point was used for the Unusual Event Declaration).

RCS ACTIVITY(I-131 DOSE EQUIVALENT)>300 IjCI/cc? - The value of 300 IjCI/cc (I-131) was taken directly from NUREG-0654.

CORE COOLING CSF RED? - A Red on this Critical Safety Function (CSF-2) would be due to either core temperatures above 1200EF or core temperatures above 730 'F and RVLIS less than 39%. In either case, the fuel is in jeopardy.

NUREG-0654, Att. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 6 of 75

II.C. FISSION PRODUCT BARRIER ANALYSIS RCS EOP PATH-2 ENTERED? - This is an indication of a SG tube rupture with safety injection (RCS initial leak rate of >120 gpm). This indicates that the RCS is breached. If EOP PATH 2 has been entered, the SG radiation monitor levels are an indication that both the fuel and the RCS barriers are breached. The limit of 20 mR/HR is based upon having 300 pCi/cc RCS activity (1-131) leaking at 40 gpm into a Steam Generator. A 40 gpm leak rate was utilized as a conservative threshold to allow for an assumed 10 gpm of RCS leakage to the Containment atmosphere (total RCS leakage therefore equal 50 gpm).

CNMT LEAK DET RAD MON NOBLE GAS CHNL >8.0 E-3 pCi/cc? - With normal activity in the RCS, if the Containment Leak Detection Radiation Monitor noble gas channel increases to greater than 8.0 E-3 ljCI/cc, the RCS is leakin'g at a rate greater than 40 gpm in addition to the Tech.

Spec. limit of 10 gpm. The RCS is then classified as being breached.

3. WAS ENTRY AT POINT U? - If the entry point, into the EAL Network, was at entry point U, then the EOP Flow Path has already determined that the RCS boundary is breached and the FPB status board was so indicated upon EAL entry. Time is saved by bypassing other steps which evaluate the RCS fission product barrier. Steps which evaluated more than one barrier were not bypassed.

4. RCS LEAKAGE >50 GPM? - This is an indication of an RCS breach, regardless of the activity level in the system. This was taken directly from N U REG-0654.

INTEGRITY CSF MAGENTA OR RED? - If the RCS Integrity Critical Safety Function (CSF-4) does not indicate green or yellow, the RCS FPB is in jeopardy. This would occur when RCS temperature is <240'F and low temperature overpressure or cooldown limitations were exceeded.

This ensures that a pressurized thermal shock event will be classified at least as an Alert.

NUREG.0654, Att. 2, Exarnpto IC Cross Roforonco to HNP EALs, Rov. 99 1 Pago 7 of 75

II.D. FISSION PRODUCT BARRIER ANALYSIS CONTAINMENT WAS ENTRY AT POINT V? - The EOP Flow Path has ready determined that the CNMT FPB has been breached and was indicated so upon EAL flow path entry. Steps to evaluate the Containment FPB are bypassed.

IS CNMT PHASE A OR VENT ISOLATION REQUIRED? - If plant conditions are such that isolation of the containment is designed to occur, then an evaluation of the effectiveness of the isolation is performed. If a pathway from the containment exists, the FPB is declared breached.

Releases from Containment through the secondary plant are evaluated separately and are not considered in this assessment. I BOTH FUEL AND RCS INTACT ON FPB STATUS BOARD? - If either the Fuel or RCS fission product barriers are in jeopardy or breached the status of the containment penetrations is evaluated to determine if a pathway exists. Releases from Containment through the secondary plant are evaluated separately and are not considered in this assessment.

PRIMARY TO SECONDARY LEAKAGE IN ANY SG >10 GPM? - In accordance with NUREG-0654 the Containment is not considered breached by stuck open SG safeties or PORVs or non-isolable secondary system breaks unless there is a release pathway caused by primary to secondary leakage in the affected SG. The example criteria of >10 gpm primary to secondary leakage for steam breaks from NUREG-0654 was applied to the related secondary plant valve problems. If leakage does not exceed this threshold, the related evaluations are bypassed.

5. AFFECTED SG SAFETY VALVES SHUT? - An open Steam Generator )

Safety Valve is one indication of a Main Steam break outside of Containment and with >10 gpm primary to secondary leakage the Containment is declared breached.

6. AFFECTED SG PORV SHUT? - This is normally the case following a Reactor Trip. The PORVs may open momentarily, but quickly close as the energy is dissipated. If the PORV is open, the ability to close the valve or it's block valve is evaluated. If the valve can not be shut and >10 gpm primary to secondary leakage exists, the Containment is declared breached.

UNISOLABLE STEAM AND/OR FEED BREAK OUTSIDE CNMT IN AFFECTED SG? - An unisolable steam and/or feed break outside of the Containment is a breach of the Containment FPB if a release pathway via primary to secondary leakage exists. A note is provided to aid in distinguishing between a minor leak, such as valve packing, and a break which could produce a significant radiological release.

NUREG-0654, Att. 2, Examplo IC Cross Roforonco to HNP EALs, Rov. 99.t Pago 8 of 75

II.D. FISSION PRODUCT BARRIER ANALYSIS CONTAINMENT (continued)

8. SG PRESS >1230 PSIG? - If SG pressure is below 1230 PSIG, then the SGs are acting as a normal heat sink. Following a Reactor trip, the SG pressure rapidly increases, but remains below 1100 PSIG. If it increases above this, the PORVs and Safeties lift to restore the pressure. If the pressure cannot be maintained below 1230 PSIG, a SG tube rupture has occurred that is severe enough to challenge SG integrity. The RCS will already be considered breached because EOP PATH-2 will be entered.

The SG level is then evaluated to determine if an overfill condition is occurring. A value of 82.4% was chosen as a threshold to declaring the Containment as being in jeopardy.

CNMT >3 PSIG? - If Containment pressure is greater than 3 psig, a LOCA ~

greater than normal charging capacity may have occurred since this is the action point for Safety Injection. At this pressure the Containment, per design, is not breached. A jeopardy status is assigned. This question will also result in an Alert declaration for main steam or feedwater ruptures inside containment.

III. EXAMPLE INITIATINGCONDITIONS The Fission Product Barrier analysis provides a symptomatic assessment of plant conditions to address and classify events which warrant declaration of an emergency. There are, however, other conditions, which would not be identified through evaluation of the fission product barriers, that warrant declaration of an emergency. The remainder of Side 1 and all of Side 2 provide questions to determine if the initiating conditions are such that an emergency condition exists.

The flow path was developed to insure that all applicable initiating conditions were evaluated and duplication of assessments were minimized, where applicable. This insures timely evaluation of the conditions. Similar events are grouped together on the flow path for ease of understanding; however, all steps of the flow path are analyzed to insure that all initiating conditions are fully evaluated.

Unusual Events are declared when conditions warrant, and a higher level declaration is not needed. Once the Flowpath is completed, if a declaration of an Alert, Site Area Emergency, or General Emergency is not needed, the Site Emergency Coordinator is directed to evaluate against the Unusual Event Matrix.

This Matrix is located at the bottom of the Flowpath on Side 2.

If a higher level classification is in effect, the Unusual Event Matrix is not examined in order to expedite initiation of the actions required by the higher level declaration.

A Notification of Unusual Event as defined in NUREG-0654 is referred to as "NOUE" in Attachment 1 and is equivalent to the HNP "Unusual Event" classification.

NUREG.0654, Att. 2, Exampte ICCross Referenceto HNP EALs, Rev. 99-t Pago 9 of 75

IV. ATTACHMENTS Attachments to this document provide a cross reference between NUREG-0654 example initiating conditions and the HNP method for classifying the condition.

Attachment 1 lists Notification of Unusual Event (MOUE) conditions.

Attachment 2 lists Alert conditions.

Attachment 3 lists Site Area Emergency conditions.

Attachment 4 lists General Emergency conditions.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 10 of 75

Attachment 1 Sheet 1 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE . Example IC Item No.: 1 Emergency Core Cooling System (ECCS) initiated and discharge to vessel.

HNP EAL Initiatin Condition:

N/A EXPLANATION:

This Initiating Condition was deleted by Rev. 96-1, corresponding to PLP-201, Revision 26, in response to implementation of EPPOS No. 1

1) Planned discharges to the vessel are required to comply with technical specification and ASME Boiler and Pressure Vessel Code,Section XI, surveillance requirements for pump and valve testing. These events are planned, do not challenge the plant, and do not represent an emergency condition.

2) Inadvertent discharge of ECCS to the vessel, in and of itself, does not represent an emergency condition (the event would be reportable in accordance with 10CFR50.72).

3) Required ECCS actuation may be an indicator of an RCS barrier challenge. Challenges to the RCS barrier are adequately addressed in Appendix 1 of NUREG-0654 under the example ICs for Unusual Event ¹5, Alert ¹5, and Site Area Emergency ¹1. The HNP initiating conditions which correspond to each of these are:

a. NUREG-0654, Unusual Event ¹5 "Exceeding either primary/secondary leak rate technical specification or primary system leak rate technical specification" is assessed in the existing HNP EALs through the Unusual Events 4-1-1, 4-2-1 and 4-3-1. I

b. NUREG-0654, Alert ¹5 "Primary coolant leak rate greater than 50 gpm" is assessed in the existing HNP EAL 2-1-2 via "RCS leakage > 50 GPM" indicating RCS breached on the Fission Product Barrier status board. An RCS breach results in declaration of an Alert emergency classification, or higher.

c. NUREG-0654, Site Area Emergency ¹1 "Known loss of coolant accident greater than makeup pump capacity" is assessed in the existing HNP EALs on side 1 of the Flow Path via the Fission Product Barrier (FPB) analysis.

NUREG.0654, Att. 2, Exemple IC Cross Reference to HNP EALs, Rev. 99-1 Page 11 of 75

Attachment 1 Sheet 2 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 2 Radiological effluent technical specification limits exceeded.

HNP EAL Initiatin Condition: EAL Reference No. 1-1-1

1) GASEOUS OR LIQUID EFFLUENT(S) EXCEEDING TECHNICAL SPECIFICATIONS 1-1-1 VALIDHIGH ALARMOCCURS ON ANY OF THE MONITORS IN EAL TABLE 5 AND THE RELEASE HAS NOT BEEN TERMINATED. (UNUSUAL EVENT EXISTS UNTIL EFFLUENT DISCHARGE IS TERMINATEDAND ALL REQUIRED NOTIFICATIONS ARE MADE)

EXPLANATION:

Table 5 lists the plant effluent monitors. The alarm setpoints for these monitors are set below the T. S.

effluent limit. If the alarm setpoint is exceeded, a T.S. Limit is being approached and an Unusual Event is declared if automatic isolation does not occur. The criteria for terminating NOUE extends the condition until notifications are completed.

NUREG.0654, Att. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99.1 Pago 12 of 75

Attachment 1 Sheet 3 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 3 Fuel damage indication. Examples:

a. High off gas at BWR air ejector monitor (greater than 500,000 uCi/sec; corresponding to 16 isotopes decayed to 30 minutes; or an increase of 10,000 uCi/sec within a 30 minute time period).

b. High coolant activity sample (e.g., exceeding coolant technical specifications for iodine spike).

c. Failed fuel monitor (PWR) indicates increase greater than 0.1% equivalent fuel failures within 30 minutes.

HNP EAL Initiatin Conditions: EAL Reference Nos. 2-1-1 and 2-2-1

2) FUEL DAMAGE INDICATION 2-1-1 GROSS FAILED FUEL DETECTOR INDICATES AN INCREASE GREATER THAN 2E4 CPM WITHIN 30 MINUTES.

2-2-1 RCS SPECIFIC ACTIVITYEXCEEDS TECHNICAL SPECIFICATION 3.4.8 LIMITS FOR DOSE EQUIVALENT 1-131 OR GROSS RADIOACTIVITY. (FOR DOSE EQUIVALENT I -131 THE EAL IS NOT EXCEEDED UNLESS THE 48 HOUR TIME INTERVAL, OR FIG. 3.4-1 LIMITS ARE EXCEEDED)

EXPLANATION:

a. HNP is a PWR and this item is applicable only to BWR plants.

b. EAL 2-2-1 addresses NUREG Item 3b. Technical Specifications permit corrective action when the Limiting Condition of Operation of the Technical Specification is exceeded. This HNP wording clarifies the intent to include corrective action steps as being an integral part of the Technical Specification.

The terminology associated with these allowed actions was clarified in EAL rev. 99-1 to insure the consequential actions as a result of failure to satisfy Tech. Specs. (such as be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) would not be confused with "corrective actions." Exceeding Technical Specification limits for a period designated in the action statement is an analyzed condition of the plant and does not represent an emergency. This is consistent with the basis included in NRC EPPOS ¹1 for deletion of Unusual Events related to technical specifications.

c. EAL 2-1-1 addresses NUREG Item 3c an increase of 2 x 10'PM within thirty minutes in the reading of the Gross Failed Fuel Detector is indication that fuel is starting to fail. Westinghouse provided this value along with a higher value which is used for the Alert Classification. I NUREG-0654, Att. 2, Exarnpfo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 13 of 75

Attachment 1 Sheet 4 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 4 Abnormal coolant temperature and/or pressure or abnormal fuel temperatures outside of technical specification limits.

HNP EAL Initiatin Condition: EAL Reference No. 8-1-1

8) OTHER PLANT OR EQUIPMENT PROBLEMS 8-1-1 INABILITYTO REACH REQUIRED SHUTDOWN (MODE 3) CONDITION WITHIN TECH. SPEC. TIME LIMITS EXPLANATION:

Exceeding Technical Specification limits for the period designated in the action statement is an analyzed condition of the plant and does not, by itself, represent an emergency. If plant conditions are outside of technical specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

When the plant cannot be SHUTDOWN (Mode 3) within the allowable action statement time, then declaration of an Unusual Event is warranted. This is consistent with the guidance provided in NRC EPPOS No. 1.

NUREG-0654, Att. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 14 of 75

Attachment 1 Sheet 5 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 5 Exceeding either primary/secondary leak rate technical specification or primary system leak rate technical specification HNP EAL Initiatin Conditions: EAL Reference No. 4-2-1 & 4-3-1

4) LOSS OF REACTOR COOLANT 4-2-1 ANY RCS PRESSURE BOUNDARY LEAKAGE 4-3-1 ANY OTHER RCS LEAKAGE IN EXCESS OF TECHNICAL SPECIFICATION 3.4.6.2 WITH THE 4 HOUR CORRECTIVE ACTIONS NOT SATISFIED.

EXPLANATION:

Technical Specification 3.4.6.2 addresses RCS allowed leakage. Technical Specifications, for other than Pressure Boundary leakage permit corrective action when the Limiting Condition of Operation of the Technical Specification is exceeded. The HNP EAL terminology clarifies the intent to include corrective action steps as being an integral part of the Technical Specification. Exceeding Technical Specification limits for a period designated in the action statement is an analyzed condition of the plant and does not represent an emergency. The terminology associated with these allowed actions was clarified in EAL Rev. 99-1 to ensure the consequential actions as a result of failure to satisfy Tech. Specs. (such as be in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) would not be confused with "corrective actions." This is consistent with the basis included in EPPOS ¹1 for deletion of Unusual Events related to technical specifications.

NUREG-0654, Att. 2, Exarnpfo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 15 of 75

Attachment 1 Sheet 6 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 6 Failure of a safety or relief valve in a safety related system to close following reduction of applicable pressure.

HNP EAL Initiatin Condition: EAL Reference Nos. 3-2-1 & 4-1-1

3) LOSS OF SECONDARY COOLANT OR COOLING 3-2-1 FAILURE OF A SG SAFETY OR PORV TO FULLYRESET AFTER OPERATION.

4) LOSS OF REACTOR COOLANT 4-1-1 FAILURE OF A PRESSURIZER SAFETY OR PORV TO CLOSE FOLLOWING REDUCTION OF APPLICABLE PRESSURE.

EXPLANATION:

At HNP, the valves of concern are the Safety and Relief valves on the Steam Generator or Pressurizer.

Any other relief valve that could malfunction is easily isolable and discharges to a closed system.

However, the above listed valves could result in a challenge to the plant safety systems and require notification of an Unusual Event.

NUREG-0654, Att. 2, ExampIo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 16 ot 75 I

Attachment 1 Sheet 7 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 7 Loss of offsite power or loss of onsite AC power capability.

HNP EAL Initiatin Condition: EAL Reference Nos. 5-1-1 and 5-2-1

5) LOSS OF POWER 5-1-1 LOSS OF ALL OFFSITE POWER, OR 5-2-1 LOSS OF BOTH EMERGENCY DIESEL GENERATORS.

EXPLANATION:

The loss of offsite power capability is as stated. The loss of both Emergency Diesel Generators affects the redundancy of the safety related on site power, the power of concern. Loss of only one Diesel Generator is covered by the Tech. Spec. Action statement. Therefore, if both Diesel Generators are lost, an Unusual Event is declared regardless of the status of the Main Generator or offsite power, the normal power supplies to the safety related onsite power.

NUREG.0654, Att. 2, ExarnpIo IC Cross Roferance to HNP EALs, Rav. 99.1 Pago 17 of 75 I

Attachment 1 Sheet 8 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 8 Loss of containment integrity requiring shutdown by Technical Specifications.

HNP EAL Initiatin Condition: EAL Reference No. 8-1-1

8) OTHER PLANT OR EQUIPMENT PROBLEMS 8-1-1 INABILITYTO REACH REQUIRED SHUTDOWN (MODE 3) CONDITION WITHIN TECH. SPEC. TIME LIMITS EXPLANATION:

Exceeding Technical Specification limits for the period designated in the action statement is an analyzed condition of the plant and does not, by itself, represent an emergency. If plant conditions are outside of Technical Specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

When the plant cannot be Shutdown (mode3) within the allowable action statement time, then declaration of an Unusual Event is warranted. This is consistent with the guidance provided in NRC EPPOS No. 1.

NUREG-0654, Att. 2, Example IC Cross Reference to HNP EALs, Rov. 99-t Pago 18 ot 75

Attachment 1 Sheet 9 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 9 Loss of engineered safety feature or fire protection system function requiring shutdown by Technical Specifications (e.g., because of malfunction, personnel error, or procedural inadequacy).

HNP EAL Initiatin Condition: EAL Reference No. 8-1-1

8) OTHER PLANT OR EQUIPMENT PROBLEMS 8-1-1 INABILITYTO REACH REQUIRED SHUTDOWN (MODE 3) CONDITION WITHIN TECH. SPEC. TIME LIMITS EXPLANATION:

Exceeding Technical Specification limits for the period designated in the action statement is an analyzed condition of the plant and does not, by itself, represent an emergency. If plant conditions are outside of Technical Specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

When the plant cannot be Shutdown (mode3) within the allowable action statement time, then declaration of an Unusual Event is warranted. This is consistent with the guidance provided in NRC EPPOS No. 1.

NUREG.0654, Att. 2, Examplo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 19 of 75

Attachment 1 Sheet 10 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 10 Fire within the plant lasting more than 10 minutes HNP EAL Initiatin Condition: EAL Reference No. 10-1-1

10) OTHER HAZARDS 10-1-1 FIRE WITHIN THE PROTECTED AREA NOT EXTINGUISHED WITHIN 15 MINUTES OF CONTROL ROOM NOTIFICATION OR VERIFICATION OF CONTROL ROOM ALARM (THIS DOES NOT INCLUDE FIRES WITHIN OFFICE AREAS, TRASH BIN FIRES, H2 TANK VENT STACK FIRES EXTINGUISHED PER OP-152.02, OR OTHER SMALL FIRES OF NO PLANT SAFETY CONSEQUENCE).

EXPLANATION:

The guidance provided in NUMARC/NESP-007, Unusual Event HU2,has been applied to this EAL as outlined in NRC EPPOS ff1. This includes the change in length of time criteria.

The Protected Area includes the plant and all areas within the security fence.

H2 Vent Stack Fires have been exempted from event declaration due to their being a preplanned evolution and in accordance with plant design.

The HNP additional guidance better addresses the purpose of this IC - to define the magnitude and extent of fires that may be potentially significant precursors to damage to safety systems.

Escalation to higher emergency classifications would occur through the EAL flow path, Side 1, assessments of fires. If the fire may affect safety related equipment, an Alert would be declared. If that fire resulted in a complete loss of any safety related function, a Site Area Emergency would be declared.

NUREG 0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 20 of 75

Attachment 1 Sheet 11 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 11 Indications or alarms on process or effluent parameters not functional in Control Room to an extent requiring plant shutdown or other significant loss of assessment or communication capability (e.g.,

plant computer, Safety Parameter Display, System, all meteorological instrumentation).

HNP EAL Initiatin Condition: EAL Reference Nos. 6-1-1, 6-2-1 and 6-3-1

6) LOSS OF MCB ANNUNCIATORS, ERFIS, OR COMMUNICATIONS CAPABILITY 6-1-1 UNPLANNED LOSS OF )75% OF MCB ANNUNCIATORS (ALBs) FOR >15 MINUTES AS DEFINED BY:

e ¹ MODE 1-4, TOTAL ALBs = 30 MODES 5-6, TOTAL ¹ALBs = 20 (ALB 1, 2, 4-13, 15, 22, 23, EITHER 24 OR 25 BASED ON EDG OPERABILITY, 26-28, 5 30) 6-2-1 INABILITYOF ERFIS TO PERFORM ITS INTENDED FUNCTION FOR A CONTINUOUS I PERIOD QF 4 HOURS, OTHER THAN PREPLANNED REMOVAL FORM SERVICE FOR MAINTENANCEOR MODIFICATIONPURPOSES, WHILE IN MODES 1, 2, 3, OR 4 AS DEFINED BY:

~ FAILURE OF BOTH CPUS.

~ FAILURE OF BOTH DATA CONCENTRATORS.

~ FAILURE OF BOTH DATA DISCS.

~ INABILITYTO DISPLAY SPDS IN THE CONTROL ROOM.

o INABILITYTQ UPDATE CURRENT DATADISPLAYS IN THE CONTROL ROOM.

(THIS IS NOT TO BE CONSTRUED AS A FAILURE OF A SINGLE VARIABLEOR SMALL DATA SUBSET).

6-3-1 FAILURE OF BOTH SITE TELEPHONE AND EMERGENCY (HE8 EC) TELEPHONE SWITCHES.

EXPLANATION:

The loss of Process or Effluent parameters that are important to safety are listed in the Tech. Spec. 3.3.

Technical Specifications designate, in the action statement, corrective actions and time limits to accomplish these actions. These are analyzed conditions of the plant and do not, by themselves, represent an emergency (Ref. NRC EPPOS ¹1).

a) If plant conditions are outside of technical specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

b) When the plant cannot be shutdown (Mode 3) within the allowable action statement time, then declaration of an Unusual Event is warranted The loss of alarms resulting in a loss of significant assessment capability has been defined in a manner consistent with NUMARC/NESP-007 (Refer to NUREG-0654 ALERT ¹14 for further details).

Due to the shift in emphasis from classification based upon dose assessment to classification based upon plant conditions, loss of meteorological instrumentation is no longer considered to meet the threshold of an Unusual Event. This is consistent with the guidance provided in NRC EPPOS No. 1.

Preplanned removal of ERFIS plant computer is allowed by Technical Specifications and does not result in a loss of assessment capability and is exempted from event declaration.

All of the other items addressed by the NUREG item, are listed above.

NUREG.0654, An. 2, ExampIe IC Cross Reference to HNP EALs, Rev. 99-1 Page 21 ot 75 I

Attachment 1 Sheet 12 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 12 Security threat or attempted entry or attempted sabotage HNP EAL Initiatin Condition: EAL Reference No. 7-1-1

7) SECURITY THREAT 7-1-1 CONFIRMED SECURITY EVENT WHICH INDICATES A POTENTIAL DEGRADATION IN THE LEVEL OF SAFETY OF THE PLANT AS INDICATED BY:

~ UNAUTHORIZED ALTERATION OR TAMPERING HAS OR IS OCCURRING AFFECTING SAFETY RELATED EQUIPMENT.

~ HOSTAGE/EXTORTION SITUATION THAT THREATENS TO INTERRUPT NORMAL PLANT OPERATIONS.

~ CIVIL DISTURBANCE ONGOING BETWEEN THE SITE BOUNDARY AND THE PROTECTED AREA.

EXPLANATION:

HNP EAL Revision 99-1 updated the terminology used for assessment of this EAL condition. The referenced terminology continues to correspond to the NUREG-0654 condition and more clearly describes the conditions which correspond to this event classification. The revised terminology closely resembles that of the second issuance of NEI 97-03, Final Draft Rev. 3 (October 1998), Unusual Event HU4, for use in describing the appropriate conditions for each event classification associated with security events.

Notes:

~ The NEI 97-03 reference to "Hostile Strike Action" was not included due to the fact that HNP is a non-union plant in a right to work state.

~ Even though there has been no activity of this nature at the Harris Plant, the NEI 97-03 reference to "Civil Disturbance" criteria was included to provide consistency with the industry standard (NEI 97-03) and could potentially be associated with an "attempted entry."

NUREG 0654, Att. 2, Exampfe IC Cross Referertce to HNP EALs. Rev. 99.1 Pago 22 of 75

Attachment 1 Sheet 13 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 13 Natural phenomenon being experienced or projected beyond usual levels:

a. Any earthquake felt in-plant or detected on station seismic instrumentation.

b. 50 year flood or low water, tsunami, hurricane surge, seiche.

c. Any tornado on site.

d. An hurricane HNP EAL Initiatin Condition: EAL Reference Nos. 9-1-1, 9-2-1 and 9-3-1

9) NATURAL PHENOMENA 9-1-1 INDICATIONOF ANY TWO VALIDSEISMIC SYMPTOMS LISTED ON EAL TABLE 6.

9-2-1 SUSTAINED WIND SPEED AT 10 METERS OF 74 MPH OR GREATER 9-3-1 TORNADO REPORTED WITHIN THE EAB.

EXPLANATION:

a. Symptoms of an earthquake are listed on EAL Table 6.

b. These NUREG-0654 items are not applicable to HNP EALs and were deleted by revision 96-01 I implemented by Emergency Plan Revision 26.

Basis:

1) Hi level (50 year flood) is not applicable to HNP EALs as it not a threat to the plant (ref. HNP FSAR sections 2.4 and 3.4).

a) The Harris plant is bounded by the Main and Auxiliary Reservoirs b) The maximum water level in either reservoir taking into account for 500 year flood levels (probable maximum flood (PMF)) coincident with wave run up from design wind velocity (123 mph), probable maximum precipitation (PMP), storm water drainage and runoff, is below the plant grade.

c) As described in the FSAR, all safety related structures will not be jeopardized as a result of maximum still water level or wave run-up resulting from PMF, or storm water accumulated at the plant site due to a PMP, and therefor, it will not be necessary to bring the reactor to a cold shutdown for flood conditions.

d) HNP Tech. Specs. do not provide Limiting Conditions For Operations associated with any maximum levels in the reservoirs. The maximum levels in the reservoirs have been analyzed and determined not to constitute adverse conditions of the plant and therefor do not represent an emergency.

2) Low level (drought) conditions in the main and auxiliary reservoirs are analyzed TS 3.7.5 conditions.

a) Exceeding technical specification limits for the period designated in the action statement is an analyzed condition of the plant and does not, by itself, represent an emergency.

b) The Unusual Event IC "Inability to reach required shutdown condition within Technical Specification Limits" would provide appropriate event classification in the event that the applicable Tech. Spec. time limits could not be complied with.

c) If plant conditions are outside of technical specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

3) Tsunami does not apply to HNP due to its geographical location, approximately 140 miles from the Atlantic Ocean. (The only areas of the U. S. that are susceptible to tsunamis are those bordering on the Pacific Ocean or the Gulf of Mexico, Ref. HNP FSAR 2.4.6)

4) At HNP the only dynamic mechanisms considered to be credible for the production of high water levels is the probable maximum wind discussed in the assessment of no flooding potential and lack of EAL impact above. Therefor, hurricane surge and seiche water levels do not apply to HNP EALs.

c. NOAA and Weather Services do not define "Hurricane" at inland locations such as HNP. The meteorological conditions equivalent to hurricane conditions are provided in EAL 9-2-1.

d. Tornado is addressed in EAL 9-3-1. At HNP the "affected" area was extended to the EAB.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99.1 Page 23 of 75

Attachment 1 Sheet 14 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 14 Other hazards being experienced or projected:

a. Aircraft crash on-site or unusual aircraft activity over facility.

b. Train derailment on-site.

c. Near or onsite explosion.

d. Near or onsite toxic or flammable gas release.

e. Turbine rotating component failure causing rapid plant shutdown HNP EAL Initiatin Condition: EAL Reference Nos. 10-2-1, 10-3-1, 10-4-1 and 8-3-1

10) OTHER HAZARDS 10-2-1 AIRCRAFT, TRAIN OR OTHER VEHICLE CRASH THAT MAY DAMAGE PLANT STRUCTURES CONTAINING FUNCTIONS OR SYSTEMS REQUIRED FOR SAFE SHUTDOWN OF THE PLANT 10-3-1 UNPLANNED EXPLOSION WITHIN THE PROTECTED AREA RESULTING IN VISIBLE DAMAGETO PERMANENT STRUCTURES OR EQUIPMENT 10-4-1 UNPLANNED TOXIC OR FLAMMABLEGAS RELEASE WITHIN EAB (REFERENCE EAL TABLE 7)

8) OTHER PLANT OR EQUIPMENT PROBLEMS 8-3-1 TURBINE ROTATING COMPONENT FAILURE RESULTING IN A REACTOR TRIP, CASING PENETRATION, OR SIGNIFICANT DAMAGE TO MAIN GENERATOR SEALS.

EXPLANATION:

The HNP Emergency Action Levels utilize the basis from the NUMARC/NESP-007, Recognition Category H, Hazards and Other Conditions Affecting Plant Safety, were utilized in development of more definitive descriptions for these Initiating Conditions. This was done utilizing the NRC EPPOS ¹1 "Other Changes" for the NUMARC/NESP-007.

a8b. HNP EAL10-2-1 above, utilizes the NESP-007 Unusual Event HU1, Natural and Destructive I Phenomena Affecting the Protected Area, item 4 guidance for the plant specific terminology. The EAL is intended to address such items as plane or helicopter crash, or train crash, that may potentially damage plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant vital area, the event may be escalated to Alert.

c. HNP EAL 10-3-1 above, utilizes the NESP-007 Unusual Event HU1, Natural and Destructive Phenomena Affecting the Protected Area, item 5 guidance for the plant specific terminology. For this EAL only those explosions of sufficient force to damage permanent structures or equipment within the protected area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to near-by structures and materials. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for declaration. The HNP specific step has been modified to include the word "Unplanned." This is done because periodically planned explosions occur which would result in an Unusual Event declaration when one was not warranted (Ex.

Plugging SG tubes with explosive plugs) and is consistent with NUMARC/NESP-007 terminology of "an unanticipated explosion".

(continued on next page)

NUREG.0654, Att. 2, Exarnpfo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 24 of 75

Attachment 1 Sheet 15 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 14 (continued)

d. HNP EAL 10-4-1 above, utilize the NESP-007 Unusual Event HU3, Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant guidance for the plant specific terminology. At times, planned releases of toxic or flammable gases occur. These are controlled releases and must be done to continue safe and efficient plant operations. HNP does not consider that this type of release should require an Unusual Event declaration because it is done on purpose and in a controlled manner.

- EAL Table 7 lists Toxic, Flammable, and Asphyxiant Gases stored in bulk at HNP. It also provides guidelines for evaluating the applicability of this initiating condition.

- Events occurring outside of the Exclusion Area Boundary do not require Notification of an Unusual Event if they are not part of the site, as defined in the FSAR, and do not directly affect plant operations.

This is the reason that the above statements include the qualification that the event must have occurred inside of the EAB (Exclusion Area Boundary).

e. HNP EAL 8-3-1 utilizes the NESP-007 Unusual Event HU1, Natural and Destructive Phenomena Affecting the Protected Area, item 6 guidance for the plant specific terminology to address main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. This EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure:

a) An Alert would be declared if the turbine rotating component failure resulted in a missile which impacted another plant structures or components within the power block with the reactor in modes 1, 2, 3 or 4 (EAL 10-2-2).

b) A Site Area Emergency would be declared if the above conditions were present, and a safety related equipment or structure was affected (EAL 10-2-3).

or in conjunction with a Steam Generator Tube Rupture related assessments.

Specific references to "unusual aircraft activity over the facility" and train "derailment" have been eliminated as a result of added guidance included with the EALs described above.

Basis:

1) The effect of train derailments and unusual aircraft activity over the facility are adequately addressed by the revised EAL descriptions based on NUMARC/NESP-007 documentation.

2) The Harris Plant receives spent fuel shipments by rail car. Unusual Event EAL 10-2-1 in the column "Other Hazards" provides for classifications associated with damage to the plant from rail traffic.

a) The spent fuel shipping plan contains emergency measures, notification and reportability requirements, and identifies personnel/staff to respond to "threats" associated with the transportation of spent fuel.

b) Site emergency plan implementation, emergency staff activation, and event classifications would occur for extreme events through:

1. Unusual Event IC for SEC Judgments, or

2. Alert EAL 11-1-2 for "Airborne rad levels indicate severe degradation in radioactive material control," or SEC Judgment that an Alert declaration is warranted.

3) The modified criteria avoid unwarranted event declarations which do not meet the intent of the Unusual Event condition as prescribed by NUREG-0654 through utilization of assessments and bases associated with NUMARC/NESP-007 EAL methodology and NRC EPPOS ¹1.

NUREG.0654, Att. 2, ExampIo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 25 of 75

Attachment 1 Sheet 16 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: MOUE Example IC Item No.: 15 Other plant conditions exist that warrant increased awareness on the part of a plant operating staff or state and/or local offsite authorities or require plant shutdown under technical specification requirements or involve other than normal controlled shutdown (e.g., cooldown rate exceeding Technical Specification limits, pipe cracking found during operation).

HNP EAL Initlatin Condition: EAL Reference Nos. 11-1-1, 8-1-1 and 8-2-1

11) SITE EMERGENCY COORDINATOR JUDGMENT 11-1-1 OTHER PLANT CONDITIONS EXIST THAT WARRANT INCREASED AWARENESS ON THE PART OF THE PLANT OPERATING STAFF, CHATHAM COUNTY, HARNETT COUNTY, LEE COUNTY, WAKE COUNTY OR THE STATE OF NORTH CAROLINA.

8) OTHER PLANT OR EQUIPMENT PROBLEMS 8-1-1 INABILITYTO REACH REQUIRED SHUTDOWN (MODE 3) CONDITION WITHIN TECH.

SPEC. TIME LIMITS 8-2-1 INADVERTENTCRITICALITY- EXTENDED AND UNPLANNED SUSTAINED POSITIVE STARTUP RATE (THIS DOES NOT INCLUDE CRITICALITYEARLIER THAN ESTIMATED DURING PLANNED REACTOR STARTUPS)

EXPLANATION:

Exceeding Technical Specification limits for the period designated in the action statement is an analyzed condition of the plant and does not, by itself, represent an emergency. If plant conditions are outside of Technical Specification limits and those conditions do result in a degradation in the level of plant safety, other initiating conditions would trigger an appropriate classification within an acceptable time frame.

When the plant can not be Shutdown (mode3) within the allowable action statement time, then declaration of an Unusual Event is warranted. This is consistent with the guidance provided in NRC EPPOS No. 1.

The item referring to Offsite authorities is as described in NUREG-0654.

An inadvertent criticality initiating condition (EAL 8-2-1) was added to the HNP EALs as an additional plant or equipment problem which is a site specific example of other events which meet the class description of an Unusual Event.. This EAL was added in Revision 99-1 using terminology consistent with selected terminology contained in the second issuance of NEI 97-03, Final Draft, Rev. 3 (October 1998). Item SU8 This EAL replaces previous EAL conditions associated with mode dependent boron dilution events.

Inadvertent Boron dilution events are ANS Condition II events and are "self-limiting." As such, event classifications above an Unusual Event would not be appropriate. The revised wording incorporates boron dilution events and inadvertent rod withdrawal events, the latter being the more limiting from an accident analysis perspective..

Escalation to higher event classification would occur via either toss of function associated with in-ability to compensate for a dilution (EAL 8-2-2 or 8-2-3), fission product barrier analysis, or judgment (EAL 11-1-2).

NUREG.0654, Att. 2, Exarnpte IC Cross Reference to HNP EALs, Rev. 99-1 Page 26 of 75

Attachment 1 Sheet 17 of 18 NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 16 Transportation of contaminated injured individual from site to offsite hospital HNP EAL Initiatin Condition:

N/A EXPLANATION:

This Initiating Condition was deleted by Rev. 96-1, corresponding to PLP-201, Revision 26, in response to implementation of NRC EPPOS ¹1.

BASIS: This event does not meet the threshold of the emergency class and is not a precursor to a more serious event as described in NRC EPPOS No. 1 NUREG 0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 27 of 75

Attachment 1 Sheet18of18 )

NOTIFICATION OF UNUSUAL EVENT CROSS REFERENCE NUREG-0654 EAL: NOUE Example IC Item No.: 17 Rapid depressurization of PWR secondary side HNP EAL Initiatin Condition: EAL Reference Nos. 3-1-1', 3-3-1 and 3-4-1

3) LOSS OF SECONDARY COOLANT OR COOLING 3-1-1 RAPID DEPRESSURIZATION OF SG SECONDARY SIDE.

3-3-1 MAIN STEAM OR FEEDWATER BREAK.

(A BREAK IS A LEAK WHICH EXCEEDS THE OPERATORS ABILITYTO SHUTDOWN THE PLANT IN A CONTROLLED MANNER OR TO NOT EXCEED TECH SPEC COOLDOWN LIMITS) 3-4-1 SG BLOWDOWN LINE BREAK (MODES 1, 2, & 3).

EXPLANATION:

This item complies with the NUREG.

HNP added the rupture of a blowdown line as a specific method of secondary plant depressurization that would not result in an Alert, but would require notification of an Unusual Event.

HNP added the parenthetical information to assist personnel in determining if loss of secondary events should be considered a "break" for EAL purposes, as opposed to a small leak. This provides better consistency in classifying events and avoid minor leaks which do not meet the intent, nor NUREG-0654 definition of, an Unusual Event. This information matches a note for the same purpose from side 1, column 1 NUREG.0654, Att. 2, Example IC Cross Reference to HNP EAt.s, Rov. 99-1 Pago 28 of 75

Attachment 2 Sheet 1 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 1 Severe loss of fuel cladding.

a. High off gas at BWR air ejector monitor (greater than 5 Ci/sec; corresponding to 16 isotopes decayed 30 minutes).

b. Very high coolant activity sample (e.g., 300 IfCI/cc equivalent of 1-131).

c. Failed fuel monitor (PWR) indicates greater than 1% fuel failure within 30 minutes of 5% total fuel failures HNP EAL Initiatin Condition: EAL Reference No. 2-1-2 2-1-2 1 FPB BREACHED/JEOPARDIZED EXPLANATION:

Refer to Fission Product Barrier Analysis, section II.B, for assessment of the Fuel Fission Product Barrier analysis:

a. HNP is a PWR and this item is applicable only to BWR plants.

b. The Fission product Barrier Analysis contains a step asking if RCB Dose Equivalent 1-131 activity is >

300 IfCI/cc. If the answer is YES, then the Fuel Fission Product Barrier is declared to be breached.

With one fission product barrier breached, an Alert would be declared.

c. The Fission Product Barrier Analysis contains a step to evaluate the Gross Failed Fuel Detector (Side 1, column 1). The value used for the determination that the fuel FPB is breached was supplied by Westinghouse.

NUREG 0654, Att. 2, Exampfe IC Cross Reference to HNP EALs, Rev. 99-1 Page 29 of 75

Attachment 2 Sheet 2 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 2 Rapid gross failure of one steam generator tube with loss of offsite power.

HNP EAL Initiatin Condition: EAL Reference No. 2-1-2 2-1-2 1 FPB BREACHED/JEOPARDIZED EXPLANATION:

Refer to Fission Product Barrier Analysis, section II.C, for assessment of the RCS Fission Product Barrier analysis:

Several of the RCS Fission Product Barrier questions would result in declaration of the RCS breached if a Steam Generator tube leak occurred.

o This would happen with a leak rate much less than the design leakage associated with the failure of one tube.

~ An RCS leak rate in excess of 50 gpm or entering EOP PATH-2 (Steam Generator Tube Rupture response) would result in the declaration of an Alert condition. This is done whether or not offsite power is available.

NUREG.0654, Att. 2, Examplo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 30 of 75

Attachment 2 Sheet 3 of 20 (

ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 3 Rapid failure of steam generator tubes (e.g., several hundred gpm primary to secondary leak rate)

HNP EAL Initiatin Condition: EAL Reference No. 2-1-2 2-1-2 1 FPB BREACHED/JEOPARDIZED EXPLANATION:

Refer to Fission Product Barrier Analysis, section II.C, for assessment of the RCS Fission Product Barrier analysis:

Several of the RCS Fission Product Barrier questions would result in declaration of the RCS breached if a Steam Generator tube leak occurred.

~ This would happen with a leak rate much less than the design leakage associated with the failure of one tube.

~ An RCS leak rate in excess of 50 gpm or entering EOP PATH-2 (Steam Generator Tube Rupture response) would result in the declaration of an Alert condition.

NUREG.0654, Att. 2, ExampIa IC Cross Ra(atenco to HNP EALs, Rev. 99-1 Pago 31 of 75

Attachment 2 Sheet4of20 )

ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 4 Steam line break with significant (e.g., greater than 10 gpm primary to secondary leak rate (PWR) or MSIV malfunction causing leakage (BWR).

HNP EAL Initiatin Condition: EAL Reference No. 2-1-2 2-1-2 1 FPB BREACHED/JEOPARDIZED EXPLANATION:

Refer to Fission Product Barrier Analysis, section II.D, for assessment of the Containment Fission Product Barrier analysis:

Primary to secondary leakage in any SG >10 gpm with a nonisolable steam and/or feed break outside containment would result in the Containment being classified as breached. An Alert would be declared for one fission product barrier breached (EAL 2-1-2).

The HNP EALs evaluate feed line breaks in addition to steam line breaks.

Additionally if the primary to secondary leak rate exceeds 50 gpm the RCS would also be classified as breached and the emergency classification would be upgraded to Site Area Emergency due to a breach of two fission product barriers (EAL 2-1-3).

The MSIV malfunction aspect of this NUREG item is not applicable to HNP, a PWR plant.

NUREG.0554, Att. 2, Exampfo IC Cross Roforonco to HNP EALs, Rov. 99.1 Pago 32 of 75 (

Attachment 2 Sheet 5 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT IC Item No.: 5 Primary coolant leak rate greater than 50 gpm.

EXPLANATION'xample HNP EAL Initlatin Condition:

2-1-2 1 FPB BREACHED/JEOPARDIZED EAL Reference No. 2-1-2 Refer to Fission Product Barrier Analysis, section II.C, for assessment of the RCS Fission Product Barrier analysis:

The RCS barrier assessment specifically asks if the RCS leakage is greater than 50. It also refers to other plant indications that would indicate a breach of the RCS barrier. If any indicator shows that the RCS is breached or in jeopardy, an Alert is declared unless a higher level declaration is warranted.

The 50 gpm leak rate is based on the requirements of NUREG-0654.

NUREG 0654, Att. 2, Exarnplo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 33 of 75

Attachment 2 Sheet 6 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 6 Radiation levels or airborne contamination which indicate a severe degradation in the control of radioactive materials (e.g., increase of factor of 1000 in direct radiation readings within facility).

HNP EAL Inltiatin Condition: EAL Reference Nos. 2-1-2 and 11-1-2 2-1-2 1 FPB BREACHED/JEOPARDIZED 11-1-2 AIRBORNE RAD LEVELS INDICATE SEVERE DEGRADATION IN RADIOACTIVE MATERIALCONTROL EXPLANATION:

Refer to Fission Product Barrier Analysis:

~ Section II.B, for assessment of the Fuel Fission Product Barrier analysis - Area radiation monitors in the vicinity of RCS fluids (letdown line related) are contained in EAL Table 2. If any of these monitors are 1000 times normal then the fuel fission product barrier is classified as breached.

o Section II.C, for assessment of the RCS Fission Product Barrier analysis - an increase of a factor of 500 (1/2 of the NUREG-0654 criteria) on the CNMT LEAK DET RAD MONITOR, Noble Gas Channel (in Containment), would result in a reading of >8E-3 ijCI/cc and the RCS would be classified as breached.

In either case, an Alert is declared due to one fission product barrier breached (EAL 2-1-2).

All other cases would be addressed through the evaluation of airborne radiation levels indicating severe degradation in radioactive material control and a declaration of an Alert (EAL 11-1-2). I NUREG 0654, Att. 2, Exampto IC Cross Rororonco to HNP EALs, Rov. 99-1 Pago 34 ot 75

Attachment 2 Sheet 7 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 7 Loss of Offsite power and loss of all onsite AC power (see Site Area Emergency for extended loss).

HNP EAL Initiatln Condition: EAL Reference No. 5-1-2 5-1-2 1A-SA OR 1B-SB NOT ENERGIZED EXPLANATION:

Power availability to the safety related buses is used as the screening criteria for this condition. The safety related AC buses can receive power from the turbine through the Unit Auxiliary Transformers, from offsite power through the Start Up Transformers, (or backfeed through the Unit Auxiliary Transformers) or from the Emergency Diesel Generators. If both 1A-SA and 18-SB are deenergized then an Alert, at a minimum, is declared.

An extended loss (>15 minutes) or loss of secondary system feed flow with reduction in core inventory would result in a higher level emergency declaration (refer to EALs 5-1-3 or 5-1-4). I NUREG 0654, Att. 2, Exampte IC Cross Reference to HNP EALs, Rev. 99-1 Pago 35 of 75

Attachment 2 Sheet 8 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 8 Loss of all onsite DC power (see Site Area Emergency for extended loss).

HNP EAL Initiatin Condition: EAL Reference No. 5-2-2 5-2-2 LOSS OF ALL ON-SITE ESF DC BUSSES (125VDC 1ASA AND 1BSB)

EXPLANATION:

ESF (Engineered Safety Features) DC is the plant-specific name for vital DC power. If it is lost, then an Alert is declared.

If the loss is extended (greater than 15 minutes), the Alert is upgraded to a Site Area Emergency (EAL 5-2-3).

NUREG.0654, An. 2, Exarnpto IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 36 of 75 I

Attachment 2 Sheet 9 of 20 )

ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 9 Coolant pump seizure leading to fuel failure.

HNP EAL Initiatin Condition:

N/A EXPLANATION:

In accordance with NRC EPPOS ¹1 this Initiating Condition is unnecessary because the concern is the fuel failure and not the seizure of the pump.

The condition is adequately addressed under (NUREG-0654) Alert ¹1.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EAt.s. Rev. 99.1 Pago 37 of 75 I

Attachment 2 Sheet 10 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 10 Complete loss of any function needed for plant cold shutdown HNP EAL Inltiatin Condition: EAL Reference No. 8-2-2 8-2-2 COMPLETE LOSS OF ANY FUNCTION LISTED ON EAL TABLE 3 EXPLANATION:

EAL Table 3 provides guidance on what functions are needed to achieve shutdown. Ten (10) functions are listed and are identified as being required for cold shutdown (Modes 4 - 5) or Hot Shutdown (Mode 3).

The loss of function condition is evaluated on Side 1. A loss of power condition could result in a loss of a function. The loss of power (AC or DC) is addressed separately in the flow path. A loss of a function, if due to the loss of power is not classified per this section of the flow path, it is assessed under the cause (the loss of power).

If the function is required for Hot Shutdown a higher classification is warranted (EAL 8-2-3) otherwise an Alert is declared.

NUREG 0554, An. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 38 of 75

Attachment 2 Sheet 11 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 11 Failure of the reactor protection system to initiate and complete a scram which brings the reactor subcritical.

HNP EAL Initiatin Condition: EAL Reference No. 8-1-2 8-1-2 ATWS WHILE IN MODE 1 OR 2 EXPLANATION:

At HNP this condition is termed an Anticipated Transient Without Shutdown (ATWS).

If an ATWS event has occurred and the manual reactor trip from the Main Control Board was successful, using either switch, an Alert is declared.

If the manual reactor trip was not successful, the event is upgraded to a Site or General Emergency depending on the status of the Fuel Fission Product.

The NUREG item refers to bringing the reactor subcritical. An ATWS from other than Modes 1 or 2 would not be associated with bringing the reactor subcritical, thus the addition of the criteria "while in Mode 1 or 2."

The HNP Main Control Board has 2 independent Manual Reactor Trip switches. Operation of either switch accomplishes the desired result.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 39 ot 75

Attachment 2 Sheet 12 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 12 Fuel damage accident with release of radioactivity to containment or fuel handling building HNP EAL Initiatin Condition: EAL Reference Nos. 2-4-2 and 2-5-2 2-4-2 DAMAGE TO SPENT FUEL with ANY SPENT FUEL POOL RAD MONITOR > 100 mR/HR 2-5-2 PLANT IS IN MODE 6 with VALIDCNMT VENT ISOL ACTUATION EXPLANATION:

DAMAGE to spent fuel in the Fuel Handling Building is assessed to determine if a single or multiple assemblies are affected. Damage to /dropping a single assembly results in declaration of an Alert.

~ Trigger points for the FHB area radiation monitors were calculated to provide a symptomatic indication of damage the affected assembly(s). These readings are projected to exist at the Radiation Monitor closest to the dropped assembly. This would result in an Alert being declared (EAL 2-4-2).

~ The 100 mR/HR set point is based on dropping one spent fuel assembly and is used to actuate the FHB emergency ventilation system. Damage to multiple assemblies would be associated with higher radiation levels (700 mR/HR) and would result in escalation to a Site Area Emergency (EAL 2-4-3).

Fuel damage accident in Containment could only occur during Mode 6 (Refueling), so the sequence is bypassed if the plant is not in Mode 6.

~ If fuel was damaged during refueling, a minor release would result in a Containment Ventilation Isolation and an Alert would be declared (EAL 2-5-2). The Containment Ventilation Actuation signaI is established based on the activity release that would occur if one spent fuel assembly was dropped after removal from the core. Damage to multiple assemblies would be associated with higher radiation levels (6.5 R/HR) and would result in escalation to a Site Area Emergency (EAL 2-5-3)..

NUREG-0654, An. 2, Exemple IC Cross Reference to HNP EALs, Rev. 99-1 Paffo 40 of 75

Attachment 2 Sheet 13 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 3 Fire potentially affecting safety systems.

HNP EAL Initiatln Condition: EAL Reference No. 10-1-2 10-1-2 FIRE MAYAFFECT SAFETY RELATED (ESF) EQUIPMENT EXPLANATION:

If a fire has occurred and it may affect Safety Related equipment an Alert (EAL 10-1-2) is declared.

If the results of the fire include a loss of any safety related function, the event classification would be escalated to a Site Area Emergency (EAL 10-1-3)

NUREG-0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Page 41 of 75

Attachment 2 Sheet 14 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 14 Most or all alarms (annunciators) lost.

HNP EAL Initlatin Condition: EAL Reference No. 6-1-2 6-1-2 LOSS OF >75% OF MCB ANNUNCIATOR'S (ALB's) for >15 minutes (with additional conditions as described below).

EXPLANATION:

The guidance provided in NUMARC/NESP-007 has been applied to this EAL as outlined in NRC EPPOS

¹1.

The Main Control Board annunciators are referred to as ALB's (Annunciator Light Boxes).

Loss of greater than seventy-five (75) percent of the Main Control Board (MCB) annunciators is used to quantify "Most." A 15 minute threshold to exclude transient or momentary losses of annunciation is included.

An "unplanned" criteria has been incorporated to exclude scheduled maintenance and testing activities (except for the Site Area Emergency criteria).

Credit is provided for the availability of computer based indication equipment (SPDS/ ERFIS plant computer) to mitigate the impact from the loss of alarm capabilities such that .

1) An un-planned, extended, loss of most annunciators, with no transient in progress, would result in an Alert declaration if ERFIS data is not available (EAL 6-1-2)

2) An un-planned, extended, loss of most annunciators, with a transient in progress, would result in an Alert declaration if ERFIS data is available (EAL 6-1-2).

EOP Path 1 entry (Reactor Trip or Safety Injection), Turbine runbacks greater than 25% and power oscillations greater than 10% are utilized to define the criteria for "significant transients" A note indicates the total number of ALBs to aid in determining 75%. Also when shutdown 10 ALBs are not associated with operating/operable systems and the total number of ALBs should be considered as 20.

An extended loss of most annunciators, with a transient in progress, without ERFIS data available, would result in a Site Area Emergency declaration (EAL 6-1-3). This would apply whether or not the annunciators were out of service as a pre-planned activity or not.

this would only result in an Unusual Event (EAL 6-1-1) if ERFIS data is available.

NOREG.0654, An. 2, Exompfo IC Cross Roforonoo to HNP EALs, Rov. 99-1 Pago 42 of 75

Attachment 2 Sheet 15 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 15 Radiological effluents greater than 10 times technical specification instantaneous limits (an instantaneous rate which, if continued over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, would result in about 1 MR at the site boundary under average meteorological conditions).

HNP EAL Initiatin Condition: EAL Reference No. 1-1-2 1-1-2 MONITOR IN EAL TABLE 5 READING )10 TIMES THE HIGH ALARMSETPOINT EXPLANATION:

EAL Table 5 lists the plant liquid and gaseous effluent radiation monitors. The setpoints for these monitors are less than or equal to the Tech Spec instantaneous limits.

An Alert is declared if any of these monitors exceeds its alarm setpoints by a factor of ten as listed within the NUREG crieria.

NUREG-0654, Att. 2, ExampIe IC Cross Reference to HNP EALs, Rev. 99-t Pago 43 of 75

Attachment 2 Sheet 16 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 16 Ongoing security compromise.

HNP EAL Initiatin Condition: EAL Reference No. 7-1-2 7-1-2 EITHER OF THE FOLLOWING SECURITY EVENTS WITHIN THE PA:

~ BOMB DISCOVERED WITHIN THE PA POTENTIALLYAFFECTING SAFETY RELATED EQUIPMENT IMMINENTTHREAT OF, OR ACTUALINTRUSION INTO THE PA BY A HOSTILE FORCE EXPLANATION:

HNP EAL Revision 99-1 updated the terminology used for assessment of this EAL condition.

The referenced terminology continues to correspond to the NUREG-0654 condition and more clearly describes the conditions which correspond to this event classification. The revised terminology is based on the guidance provided in NUMARC/NESP-007, Alert HA4, as outlined in NRC EPPOS if1.

The revised terminology closely resembles that of the second issuance of NEI 97-03, Final Draft Rev. 3 (October 1998), ITEM HA4 which added "BOMB discovered within the PROTECTED AREA potentially affecting (site specific) Safety Related Equipment" as an additional Example EAL.

"IMMINENTTHREAT OF, OR ACTUAL"was used in the HNP terminology in place of "ACTUAL" as being more in-line with the HNP Safeguards Contingency Plan and related security related procedures and training. This is more conservative than the term of "actual."

Discovery of a bomb or intrusion into a Vital Area would result in escalation to a Site Area Emergency (EAL 7-1-3).

NUREG 0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 44 of 75

Attachment 2 Sheet 17 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 17 Severe natural phenomena being experienced or projected.

a. Earthquake greater than OBE levels.

b. Flood, low water, tsunami, hurricane surge, seiche near design levels.

c. Any tornado striking facility.

d. Hurricane winds near design basis level.

HNP EAL Initiatin Condition: EAL Reference Nos. 9-1-2 and 9-2-2 9-1-2 ANY TWO INDICATORS OF A SEISMIC EVENT LISTED ON EAL TABLE 6 with ANY YELLOW LIGHT ON TRIAXIALRESPONSE SPECTRUM ANNUNCIATOR LIT 9-2-2 ADVERSE WEATHER with TORNADO HAS HIT THE POWER BLOCK or PROJECTED OR MEASURED SUSTAINED WIND SPEED AT 10 METERS > 90 MPH EXPLANATION:

a. EAL Table 6 lists all available plant indications of a seismic event including indication of tremors or vibration. Any two of these indications are adequate for the operators to determine if an OBE has occurred. A yellow light on the Triaxial Response Spectrum Annunciator indicates that the event has exceeded 70% of the OBE level, a Red annunciator indicates that the event has exceeded the OBE level. If a Yellow annunciator is lit, but not a Red one, then an Alert is declared. If a Red annunciator is lit, then the OBE level has been reached or exceeded and an assumption that an SSE has occurred and the event is upgraded to a Site Area Emergency (EAL 9-1-3). The event classificaiton at 70% of OBE is conservative.

b. Flood, low water, tsunami, hurricane surge, seiche are not applicable to HNP. Refer to explanation for NUREG-0654, MOUE Item 13, and FSAR sections 2.4 and 3.4.

c8d. If Adverse weather occurs, and projected or measured wind speeds exceed 90 MPH, then an Alert is declared (EAL 9-2-2). The 90 MPH wind speed is based on the 100 year reoccurrence described in the FSAR. If wind speeds increase to 100 MPH the event would be escalated to a Site Area Emergency (EAL 9-2-3).

A tornado could conceivably strike the power block without registering wind speeds of greater than 90 MPH, at the Met. Tower, so a specific question is asked concerning the chance of tornado.

NUREG.0654, An. 2, Exampto IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 45 of 75 j

Attachment 2 Sheet 18 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 18 Other hazards being experienced or projected

a. Aircraft crash on facility

b. Missile impacts from whatever source on facility

c. Known explosion damage to facility affecting plant operation.

d. Entry into facility environs of uncontrolled toxic or flammable gases

e. Turbine failure causing casing penetration HNP EAL Initiatin Condition: EAL Reference Nos. 10-2-2 and 10-3-2 10-2-2 AIRCRAFT CRASH, MISSILE IMPACT OR UNPLANNED EXPLOSION INSIDE POWER BLOCK.

10-3-2 UNCONTROLLED OR UNPLANNED RELEASE OF TOXIC OR FLAMMABLEGAS INTO POWER BLOCK REF EAL TABLE 7 EXPLANATION:

a, b, & c. These conditions result in declaration of an Alert (EAL 10-2-2) unless Safety Related equipment or structures are affected with the plant not in Cold Shutdown, which would result in escalating the emergency classification to a Site Area Emergency (EAL 10-2-3).

d. This condition results in declaration of an Alert (EAL 10-3-2) unless the and the gas is flammable or lack of access is a safety problem with the plant not in Cold Shutdown, which would result in escalating the emergency classification to a Site Area Emergency (EAL 10-3-3).

EAL Table 7 lists Toxic, Flammable, and Asphyxiant Gases stored in bulk at HNP. It also provides guidelines for evaluating the applicability of this initiating condition.

e. This Initiating Condition was deleted by Rev. 96-1, corresponding to PLP-201, Revision 26, in response to implementation of EPPOS No. 1. A turbine failure resulting in casing penetration would result in declaration of an Unusual Event (EAL 8-3-1).

Basis:

1) HNP is a PWR and as such a penetration of the turbine casing will not result in a radiological release unless it occurs in conjunction with a steam generator tube rupture.

The radiological release and/or steam generator tube rupture would be identified, and classified, by the radiological ICs or the Fission Product Barrier ICs.

2) Escalation of emergency classification, above the Unusual Event level, for events I associated with missile damage from turbine rotating component failures which penetrate the casing, would occur as follows: I a) An Alert would be declared if the turbine rotating component failure resulted in a missile which impacted other plant structures or components within the power block with the reactor in modes 1, 2, 3 or 4 (EAL 10-2-2).

b) A Site Area Emergency would be declared if the above conditions were present, and a safety related equipment or structure was affected (EAL 10-2-3).

NUREG.0654, An. 2, Exampfo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 46 of 75

Attachment 2 Sheet 19 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 19 Other plant conditions exist that warrant precautionary activation of Technical Support Center and placing near-site Emergency Operations Facility and other key emergency personnel on standby.

HNP EAL Initiatin Condition: EAL Reference No. 11-1-2 11-1-2 AIRBORNE RAD LEVELS INDICATE SEVERE DEGRADATION IN RADIOACTIVE MATERIALCONTROL, or ANY PLANT CONDITION EXISTS THAT IN THE JUDGEMENT OF THE SUPERINTENDENT - SHIFT OPERATIONS OR SITE EMERGENCY COORDINATOR, WARRANTS AN ALERT DECLARATION EXPLANATION:

A separate step is provided for the discretion to declare an Alert (EAL 11-1-2) for airborne radiation levels, even if specific levels addressed in other steps have not been exceeded.

Other plant conditions are Judgment calls as provided in (EAL 11-1-2) as described in the NUREG.

An uncontrolled Boron dilution assessment had previously been included in the HNP EALs because it was an early indication of the potential loss of plant shutdown margin which could result in an unplanned criticality.

The EAL was removed in Revision 99-1, and was replaced with an inadvertent criticality assessment (EAL 8-2-1) using terminology contained in the second issuance of NEI 97-03, Final Draft, Rev. 3 (October 1998), Item SU8.

Inadvertent Boron dilution events are ANS Condition II events and are "selt-limiting." As such, event classifications above an Unusual Event would not be appropriate.

Escalation to higher event classification would occur via either loss of function associated with in-ability to compensate for a dilution (EAL 8-2-2 or 8-2-3), fission product barrier analysis, or judgment (EAL 11-1-2).

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99.1 Pago 47 of 75

Attachment 2 Sheet 20 of 20 ALERT CROSS REFERENCE NUREG-0654 EAL: ALERT Example IC Item No.: 20 Evacuation of Control Room anticipated or required with control of shutdown systems established from local stations.

HNP EAL Initiatin Condition: EAL Reference No. 10-4-2 10-4-2 CONTROL ROOM EVAC REQUIRED OR ANTICIPATED EXPLANATION:

This is as described in the NUREG.

If the Control Room must be evacuated, control is shifted to the Auxiliary Control Panel (ACP) which contains all of the controls needed to maintain the plant in Hot Shutdown or to conduct a controlled cooldown to Cold Shutdown.

If the evacuation is required or anticipated and the Auxiliary Control (Shutdown) Panel is not operational with the Control Room evacuated for >15 minutes, the Alert is upgraded to a Site Area Emergency (EAL 1 0-4-3).

NUREG-0654, Att, 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 48 of 75

Attachment 3 Sheet 1 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 1 Known loss of coolant accident greater than makeup pump capacity.

HNP EAL Initiatin Condition: EAL Reference No. 2-1-3 2-1-3 2 FPB's BREACHED/JEOPARDIZED EXPLANATION:

The High Head Centrifugal Charging Pumps at HNP can supply several hundred gpm flow at normal operating pressure with significantly higher flow rates if system pressure decreases. A Reactor Coolant System leak of this magnitude would result in a Site Area EmergeIIcy (EAL 2-1-3) declaration as a result of several potential paths. With the plant at normal operating temperature, an RCS teak of this magnitude will result in a rise of Containment pressure to greater than 3.0 psig within a few minutes. Reactor Coolant System leakage >50 gpm and >3.0 psig in containment results in classifying two fission product barriers as breached or jeopardized and a resultant Site Area Emergency event declaration. Any of six other methods for identification of an RCS Breach and at least two other methods for identification of a second Fission Product Barrier being breached or jeopardized would also be possible dependent on the specifics associated with the RCS leak.

A separate situation is possible if no CSIPs are available. In this situation the leak rate at which makeup pump capacity would be exceeded would be much lower. This is also adequately addressed in the HNP EAL scheme via the loss of function Initiating Condition which would require declaration of a Site Area Emergency for the loss of Charging Capability function, a function required for Mode 3. (EAL 8-2-3).

This NUREG-0654 example Initiating Condition is one in which several relatively unique characteristics of the Harris plant, and the integration of the operations Emergency Operating Procedures with the EAL scheme, result in a unique flow path logic. The NRC, through a Regional Initiative inspection (Report 50-400/91-20), evaluated the HNP EAL scheme in this area and determined that the Shearon Harris EAL flow paths for emergency detection and classification results in correct and sufficiently prompt emergency classifications, and which are in agreement with the guidance in NUREG-0654.

NUREG-0654, Att. 2, ExampIo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 49 of 75

Attachment 3 Sheet 2 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 2 Degraded core with possible loss of eoolable geometry (indicators should include instrumentation to detect inadequate core cooling, coolant activity and/or containment radioactivity levels).

HNP EAL Initiatin Condition: EAL Reference No. 2-1-3 2-1-3 2 FPB's BREACHED/JEOPARDIZED EXPLANATION:

Refer to Fission Product Barrier Analysis. Several indicators would be available to identify a breach or jeopardy condition for two or possibly all three fission product barriers. This condition would be associated with a loss of coolant accident. For example: the RCS could indicate breached by RCS leakage > 50 gpm (or CNMT leak detection rad monitor > 8E-3 ljCI/cc). Fuel could indicate jeopardized by elevated thermocouple temperatures and/or reduced RVLIS level or breached by Gross Failed Fuel Detector 7E+5 CPM increase within 30 minutes. Containment may indicate jeopardized due to pressure > 3 psig.

If all three of the fission product barriers were breached or jeopardized this condition would be classified as a General Emergency (EAL 2-1-4).

NUREG-0654, Att. 2, Exampto IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 50 of 75

Attachment 3 Sheet 3 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 3 Rapid failure of steam generator tubes (several hundred gpm leakage) with loss of offsite power.

HNP EAL Initiatin Condition: EAL Reference No. 2-1-3 2-1-3 2 FPB's BREACHED/JEOPARDIZED EXPLANATION:

RCS leakage in excess of 50 gpm would require that the RCS be declared to be breached. In addition, if the SG tube leak was "several hundred gpm", the Containment would indicate Jeopardized due to the high SG press (1230 PSIG) coupled with the high SG level (82.4%). In addition, lifting of a SG safety is likely with loss of offsite power and the Containment would be classified as breached. This would show two FPB's breached/jeopardized resulting in a Site Area Emergency (EAL 2-1-3) declaration.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 51 of 75 I

Attachment 3 Sheet 4 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 4 BWR steam line break outside containment without isolation HNP EAL Initiatin Condition:

N/A EXPLANATION:

HNP is a PWR and this item is applicable only to BWR plants NUREG.0654, Att. 2, Examplo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 52 of 75

Attachment 3 Sheet 5 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 5 PWR steam line break with greater than 50 gpm primary to secondary leakage and indication of fuel damage.

HNP EAL Initiatin Condition: EAL Reference No. 2-1-3 2-1-3 2 FPB's BREACHED/JEOPARDIZED EXPLANATION:

RCS leakage in excess of 50 gpm would require that the RCS be declared breached.

A steam line break coincident with primary to secondary leakage > 10 gpm would require the Containment to be declared breached.

These conditions would show at least two FPB's breachedfjeopardized resulting in at least a Site Area Emergency (EAL 2-1-3) declaration, regardless of the status of the fuel.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Page 53 of 75

Attachment 3 Sheet 6 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 6 Loss of offsite power and loss of onsite AC power for more than 15 minutes HNP EAL Initiatln Condition: EAL Reference No. 5-1-3 5-1-3 1A-SA AND 1B-SB LOST FOR >15 MIN EXPLANATION'he 6.9 KV Emergency Busses, 1A-SA and 1B-SB are normally powered by the Main Generator or by off-site power. If normal power is lost, these busses are powered directly by the Emergency Diesel Generators. Therefore, if 1A-SA and 1B-SB are lost, all on-site and offsite AC power has been lost.

When this condition lasts for >15 minutes, a Site Area Emergency (EAL 5-1-3) is declared.

NUREG.0654, Att. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99.1 Pago 54 of 75

Attachment 3 Sheet 7 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 7 Loss of all vital onsite DC power for more than 15 minutes.

HNP EAL Initiatin Condition: EAL Reference No. 5-2-3 5-2-3 LOSS OF ALL ON-SITE ESF DC BUSSES (125VDC 1ASA AND 1BSB) for > 15 MINs EXPLANATION:

ESF (Engineered Safety Features) DC is the plant-specific name for vital on-site DC power. If this DC power supply is lost for greater than fifteen minutes, a Site Area Emergency(EAL 5-2-3) is declared.

NUREG 0654, Att. 2, Examplo IC Cross Roforonco to HNP EALs, Rov. 99 1 Pago 55 of 75

Attachment 3 Sheet 8 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 8 Complete loss of any function needed for plant hot shutdown.

HNP EAL Initiatin Condition: EAL Reference No. 8-2-3 8-2-3 COMPLETE LOSS OF ANY FUNCTION LISTED ON EAL TABLE 3 for MODE 3 (excluding function lost due to loss of all AC or all DC power)

EXPLANATION:

EAL Table 3 is a listing of the plant functions required for hot or cold shutdown. Mode 3 is "Hot Standby" which is the plant condition where RCS temperature is greater than 350 'F and the Reactor is subcritical.

This equates to the NUREG term "HOT SHUTDOWN."

A loss of power condition could result in a loss of a function. The loss of power (AC or DC) is addressed separately in the flow path. A loss of a function, if due to the loss of power is not classified per this section of the flow path, it is assessed under the cause (the loss of power).

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs. Rev. 99-1 Page 56 of 75

Attachment 3 Sheet 9 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 9 Transient requiring operation of shutdown systems with failure to scram (continued power generation but no core damage immediately evident).

HNP EAL Initiatin Condition: EAL Reference No. 8-1-3 8-1-3 ATWS WHILE IN MODE 1 OR 2 with MANUALREACTOR TRIP NOT SUCCESSFUL (EITHER SWITCH)

EXPLANATION:

At HNP this condition is termed an Anticipated Transient Without Shutdown (ATWS).

If an ATWS event has occurred and the manual reactor trip from the Main Control, using either switch, was not successful the event is classified as a Site Area Emergency (EAL 8-1-3)

If the Fuel FPB is breached (core damage) the event would be escalated to a General Emergency per EAL 8-1-4.

The note following the declaration explains that the Site Area Emergency exists only as long as the rods remain out of the core (i.e., until the Reactor Trip is successfully executed, or the rods are fully inserted by other means).

The HNP Main Control Board has 2 independent Manual Reactor Trip switches. Operation of either switch accomplishes a "scram.".

NUREG 0654, Att. 2, ExampIe IC Cross Reference to HNP EALs, Rev. 99-1 Pago 57 of 75

Attachment 3 Sheet 10 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 10 Major damage to spent fuel in containment or fuel handling building (e.g., large objects damages fuel or water loss below fuel level).

HNP EAL Initiatin Condition: EAL Reference Nos. 2-3-3, 2-4-3 and 2-5-3 2-3-3 SPENT FUEL POOL LEVEL (1 FT ABOVE TOP OF FUEL 2-4-3 DAMAGETO SPENT FUEL with ANY SPENT FUEL POOL AREA RAD MON > 100 mR/HR 2-5-3 PLANT IS IN MODE 6 with VALIDCNMT VENT ISOL ACTUATION and BOTH CNMT Hl RANGE ACCIDENT MONITORS > 6.5 R/HR EXPLANATION:

If the level is less than one foot above the spent fuel assemblies, the spent fuel is about to become uncovered and a Site Area Emergency is declared (EAL 2-3-3).

Damage to spent fuel in the Fuel Handling Building is assessed to determine if a single or multiple assemblies are affected. Trigger points for the FHB area radiation monitors were calculated to provide a symptomatic indication of damage the affected assembly(s). The 700 mR/HR criteria is based on the expected dose rate from dropping more than 1 spent fuel assembly with an escalation to a Site Area Emergency (EAL 2-4-3).

Fuel damage accident in Containment could only occur during Mode 6 (Refueling), so the sequence is bypassed if the plant is not in Mode 6.

~ The Containment Ventilation Actuation signal is established based on the activity release that would occur if one spent fuel assembly was dropped after removal from the core.

~ DAMAGE to more than one assembly would warrant upgrading the classification (EAL 2-5-3). The 6.5 R/HR reading is based on the expected reading from the radiation monitors due to dropping two spent fuel assemblies that have just been removed from the core.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 58 of 75

Attachment 3 Sheet 11 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 11 Fire compromising the functions of safety systems HNP EAL Initiatin Condition: EAL Reference No. 10-1-3 10-1-3 COMPLETE LOSS OF ANY SAFETY RELATED (ESF) FUNCTION DUE TO FIRE EXPLANATION:

This is as described in the NUREG.

NUREG.0654, Att. 2, Exarnpfo IC Cross Roforonco to HNP EALs, Rov. 99.1 Pago 59 of 75

Attachment 3 Sheet 12 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 12 Most or all alarms (annunciators) lost and plant transient initiated or in progress HNP EAL Initiatin Condition: EAL Reference No. 6-1-3 6-1-3 LOSS OF >75% OF MCB ANNUNCIATORS (ALBs), with AFFECTED SYSTEMS ERFIS DATA NOT AVAILABLEand SIGNIFICANT TRANSIENT (EOP PATH 1 ENTERED, >25%

RUNBACK, OR >10% POWER OSCILLATIONS), and ALBs LOST FOR > 15 MINUTES EXPLANATION:

The guidance provided in NUMARC/NESP-007 has been applied to this EAL as outlined in NRC EPPOS

¹1.

The Main Control Board annunciators are referred to as ALB's (Annunciator Light Boxes).

Loss of greater than seventy-five percent of the Main Control Board (MCB) annunciators is used to quantify "Most."

EOP Path 1 entry (Reactor Trip or Safety Injection), Turbine runbacks greater than 25% and power oscillations greater than 10% are utilized to define the criteria for "significant transients."

A 15 minute threshold to exclude transient or momentary losses of annunciation is included A note indicates the total number of ALBs to aid in determining 75%. Also when shutdown 10 ALBs are not associated with operating/operable systems and the total number of ALBs should be considered as 20.

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 60 of 75 I

Attachment 3 Sheet 13 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 13

a. Effluent monitors detect levels corresponding to greater than 50 mR/HR for 1/2 hour or greater than 500 mR/HR W.B. for two minutes (or five times these levels to the thyroid) at the site boundary for adverse meteorolo

b. These dose rates (listed in 13a) are projected on other plant parameters (e.g., radiation level in containment with leak rate appropriate for existing containment pressure) or are measured in the environs.

c. EPA Protective Action Guidelines are projected to be exceeded outside the site boundary.

HNP EAL Initlatin Condition: EAL Reference Nos. 1-1-3 8 1-2-3 1-1-3 PROJECTED DOSE >50 MREM TEDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA with ESTIMATED DURATION OF RELEASE >30 MINs, or PROJECTED DOSE >250 MREM THYROID CDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA with ESTIMATED DURATION OF RELEASE >30 MINs, or PROJECTED DOSE >500 MREM TEDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA with ESTIMATED DURATION OF RELEASE >2 MINs, or PROJECTED DOSE >2500 MREM THYROID CDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA with ESTIMATED DURATION OF RELEASE >2 MINs 1-2-3 MEASURED WHOLE BODY DOSE RATE >50 MREM/HR AT OR BEYOND SITE BOUNDARYwith ESTIMATED DURATION OF RELEASE >30 MINs, or MEASURED I-131 EQUIVALENTCONC. >1.9E-7 uCi/cc AT OR BEYOND SITE BOUNDARY

'ith ESTIMATED DURATION OF RELEASE >30 MINs, or MEASURED WHOLE BODY DOSE RATE >500 MREM/HR AT OR BEYOND SITE BOUNDARY with ESTIMATED DURATION OF RELEASE >2 MINs, or MEASURED I-131 EQUIVALENTCONC. >1.9E-6 uCi/cc AT OR BEYOND SITE BOUNDARY with ESTIMATED DURATION OF RELEASE >2 MINs EXPLANATION'.

The Projected Dose values (EAL 1-1-3)comply with NUREG-0654 as modified for use of TEDE and CDE values per EPA-400 and NRC EPPOS ¹1.

b. The measured values (EAL 1-2-3) are consistent with the recommendations stated in the NUREG.

The Thyroid dose values are listed in Equivalent 1-131 concentration which corresponds to the 250 MREM/HR and 2500 MREM/HR, utilizing the dose conversion factors contained in the HNP dose projection methodology in order to speed up the reporting process.

c. With implementation of the revised Protective Action Guidelines contained in EPA-400 the projected dose of 1 Rem whole body and 5 Rem thyroid have been deleted since these criteria now require a higher Emergency Classification (refer to NUREG-0654 EAL General Emer enc, item 1. I This is consistent with the guidance issued by the NRC as EPPOS ¹1.

NUREG-0664, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99.1 Pago 61 of 75

Attachment 3 Sheet 14 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 14 Imminent loss of physical control of the plant HNP EAL Initiatin Condition: EAL Reference No. 7-1-3 7-1-3 EITHER OF THE FOLLOWING SECURITY EVENTS WITHIN A VITALAREA:

~ BOMB DISCOVERED WITHIN A VITALAREA POTENTIALLYAFFECTING SAFETY RELATED EQUIPMENT

~ CONFIRMED INTRUSION INTO A VITALAREA BY A HOSTILE FORCE EXPLANATION:

HNP EAL Revision 99-1 updated the terminology used for assessment of this EAL condition.

The revised criteria for Site Area Emergency conditions continue to correspond to the NUREG-0654 condition of "Imminent loss of physical control of the plant" through more descriptive and anticipatory references to events of significance within plant "Vital Areas." The revised terminology is based on the guidance provided in NUMARC/NESP-007, Site Area Emergency HS1, as outlined in NRC EPPOS ff1, with the addition of a BOMB discovered within the VITALAREA potentially affecting Safety Related Equipment. This was included in the second issuance of NEI 97-03, Final Draft Rev. 3 (October 1998),

which added "BOMB discovered within the VITALAREA potentially affecting (site specific) Safety Related Equipment" as an additional Example EAL.

Loss of plant control would result in escalation to a General Emergency (EAL 7-1-4).

NUREG.0654, Att. 2, ExarnpIe IC Cross Reference to HNP EALs, Rev. 99.1 Pago 62 of 75

Attachment 3 Sheet15of18 )

SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 15 Severe natural phenomena being experienced or projected with plant not in cold shutdown.

a. Earthquake greater than SSE levels.

b. Flood, low water, tsunami, hurricane surge, seiche greater than design levels, or failure of protection of vital equipment at lower levels.

c. Sustained winds or tornadoes in excess of design levels.

HNP EAL Initiatin Condition: EAL Reference Nos. 9-1-3 and 9-2-3 9-1-3 ANY RED LIGHT ON TRIAXIALRESPONSE SPECTRUM ANNUNCIATOR LIT 9-2-3 SUSTAINED WIND SPEEDS AT 10 METERS ) 100 MPH EXPLANATION:

a. EAL Table 6 lists all available plant indications of a seismic event including indication of tremors or vibration. If any two of these indications are present the operators determine if an OBE or SSE has occurred. HNP has no specific annunciation associated with assessment of seismic activity exceeding the SSE level. If a Red annunciator is lit, then the OBE level has been reached or exceeded,and it is assumed that an SSE may have been exceeded and a Site Area Emergency (EAL 9-1-3) is declared.

This is conservative.

b. Flood, low water, tsunami, hurricane surge, and seiche are not applicable to HNP. Refer to explanation for NUREG-0654, NOUE Item 13, and FSAR sections 2.4 and 3.4.

c. During adverse weather, sustained wind speeds of greater than 100 MPH warrant declaration of a Site Area Emergency (EAL 9-2-3). The plant has been designed to withstand 100 MPH winds. A Site Area Emergency declaration at wind speeds in excess 100 MPH is conservative and is based on the maximum reliable reading from the Anemometer located at the plant meteorological station.

NUREG.0654, An. 2, Example IC Cross Roforonco to HNP EALs, Rov. 99.1 Pago 63 of 75 I

Attachment 3 Sheet 16 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 16 Other hazards being experienced or projected with plant not in cold shutdown.

a. Aircraft crash affecting vital structures by impact or fire.

b. Severe damage to safe shutdown equipment from missiles or explosion.

c. Entry of uncontrolled flammable gases into vital areas. Entry of uncontrolled toxic gases into vital areas where lack of access to the area constitutes a safety problem.

HNP EAL Initiatln Condition: EAL Reference Nos. 10-2-3 and 10-3-3 10-2-3 AIRCRAFT CRASH, MISSILE IMPACT OR UNPLANNED EXPLOSION INSIDE POWER BLOCK with PLANT IN MODES 1, 2, 3, OR 4 with SAFETY RELATED EQUIPMENT OR STRUCTURE AFFECTED 10-3-3 UNCONTROLLED OR UNPLANNED RELEASE OF TOXIC OR FLAMMABLEGAS INTO POWER BLOCK REF EAL TABLE 7 with AFFECTED AREA HOUSES SAFETY RELATED EQUIPMENT with GAS IS FLAMMABLEOR LACK OF ACCESS IS A SAFETY PROBLEM with PLANT IN MODE 1, 2, 3 OR 4 EXPLANATION:

a8b. If an aircraft crash, missile impact or an unplanned explosion inside of the Protected Area (PA) occurs which affects safety related equipment, and the plant is not in cold shutdown, a Site Area Emergency (EAL 10-2-3) is declared.

c. Uncontrolled or unplanned release of Toxic (or Asphyxiant) or Flammable gas into the Power Block, if the release endangers personnel or equipment, or if access to equipment required to operate the plant is impeded, and with the plant not in cold shutdown results in declaration of a Site Area Emergency (EAL 10-3-3).

NUREG.0654, Att. 2, ExampIe IC Cross Reference to HNP EALs, Rev. 99-1 Pago 64 of 75

Attachment 3 Sheet17of18 (

SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 17 Other plant conditions exist that warrant activation of emergency centers and monitoring teams or a precautionary notification to the public near the site.

HNP EAL Initiatin Condition: EAL Reference No. 11-1-3 11-1-3 ANY PLANT CONDITION THAT IN THE JUDGEMENT OF THE SUPERINTENDENT-SHIFT OPERATIONS OR SITE EMERGENCY COORDINATOR WARRANTS A SITE AREA EMERGENCY DECLARATION.

EXPLANATION:

The Site Emergency Coordinator may declare a Site Area Emergency (EAL 11-1-3) when he feels that it is warranted, based on his judgment which is consistent with the NUREG.

NOTE:

An uncontrolled Boron dilution assessment had previously been included in the HNP EALs because it was an early indication of the potential toss of plant shutdown margin which could result in an unplanned criticality.

The EAL was removed in Revision 99-1, and was replaced with an inadvertent criticality assessment (EAL 8-2-1) using terminology contained in the second issuance of NEI 97-03, Final Draft, Rev. 3 (October 1998), Item SU8.

Inadvertent Boron dilution events are ANS Condition II events and are "self-limiting." As such, event classifications above an Unusual Event would not be appropriate.

NUREG.0654, Att. 2, Exampfo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 65 of 75

Attachment 3 Sheet 18 of 18 SITE AREA EMERGENCY CROSS REFERENCE NUREG-0654 EAL: SITE AREA EMERGENCY Example IC Item No.: 18 Evacuation of Control Room and control of shutdown systems not established from local stations in 15 minutes.

HNP EAL Initiatln Condition: EAL Reference No. 10-4-3 10<-3 CONTROL ROOM EVAC REQUIRED OR ANTICIPATED and the Auxiliary Control Panel (ACP) is not operational with the Control Room evacuated for >15 minutes EXPLANATION:

This item is as described in the NUREG.

If the Control Room must be evacuated, control is shifted to the Auxiliary Control panel (ACP) which contains all of the controls needed to maintain the plant in Hot Shutdown or to conduct a controlled cooldown to Cold Shutdown. If the evacuation is required or anticipated and the Auxiliary Control (Shutdown) Panel is not operational with the Control Room evacuated for >15 minutes, a Site Area Emergency (EAL 10-4-3) is declared.

NUREG.0654, Att. 2, ExampIe IC Cross Reference to HNP EALs, Rev. 99.1 Pago 66 ot 75

Attachment 4 Sheet 1 of 8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 1

a. Effluent monitors detect levels corresponding to 1 Rem/HR W.B. or 5 Rem/HR thyroid at the site boundary under actual meteorolo ical conditions.

b.'These dose rates are projected based on other plant parameters (e.g., radiation levels in containment with leak rate appropriate for existing containment pressure with some confirmation from effluent monitors) or are measured in environs.

HNP EAL Initfatln Condition: EAL Reference Nos. 1-1-4 and 1-2-4 1-1-4 PROJECTED DOSE >1000 MREM TEDE AT OR BEYOND THE SITE BOUNDARY, or PROJECTED DOSE >5000 MREM THYROID CDE AT OR BEYOND THE SITE BOUNDARY 1-2-4 MEASURED DOSE RATE >1000 MREM/HR AT OR BEYOND THE SITE BOUNDARY, or MEASURED I -131 EQUIV CONC >3.9 E-6 pCi/cc AT OR BEYOND THE SITE BOUNDARY EXPLANATION:

a. Projected Dose Calculations are performed using effluent rad monitors in accordance with Emergency Plan implementing procedures. If the levels of 1000 mREM TEDE or 5000 MREM Thyroid CDE are exceeded, a General Emergency (EAL 1-1-4) is declared.

The revised EPA (EPA-400) Protective Action Guidelines are expressed in terms of total effective dose equivalent, which includes the 50 year internal dose commitment from inhalation of the plume.

The internal dose commitment cannot be expressed as a dose rate without the period of time of exposure (inhalation) to the plume being stated for the point of interest. To convert the NUREG-0654 example EAL to a dose rate of 1 Rem/hour (TEDE) in a meaningful way, the assumed duration of inhalation must be one hour. I Similarly, a "Committed Effective Dose Rate" to the thyroid of 5 Rem/hour must assume an inhalation period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Any other periods of time for inhalation result in rates different than these which could, or could not, exceed the EPA 400 Protective Action Guidelines for total exposure, and would not be consistent with the definition of a General Emergency. I Hence the use of a separate Emergency Action Level expressed as a dose rate is redundant to the Emergency Action Level expressed as dose.

b. Projected Dose Calculations performed as addressed in Item a, also include projections based on Containment radiation levels when pressure is above 0 psig and SG radiation level based on Main Steam Line radiation. If levels of 1000 MREM TEDE or 5000 MREM Thyroid CDE are exceeded, a General Emergency (EAL 1-1-4) is declared.

A measured dose rate >1000 mRem/HR at or beyond the site boundary or l-131 equivalent concentration >3.9 E-6 pCi/cc at or beyond the site boundary will result in a General Emergency (EAL 1-2-4).

A measurement of 3.9 E-6 pCi/cc at/or beyond the site boundary is the equivalent of a dose rate to the thyroid of 5000 MREM/HR. The information is provided in this manner because it does not require conversion and can quickly be reported to the Site Emergency Coordinator, in this form.

The above values comply with NUREG-0654 as modified for use of TEDE and CDE values per EPA-400 and NRC EPPOS ¹1.

NUREG.0654, An. 2, Examplo IC Cross Roforonco to HNP EAl.s, Rov. 99-1 Pago 67 of 75

Attachment 4 Sheet 2of8 )

GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Exam pie IC Item No.: 2 Loss of 2 or 3 fission product barners with a potential loss of 3rd barner (e.g., loss of primary coolant boundary, clad failure, and high potential for loss of containment).

HNP EAL Initiatin Condition: EAL Reference No. 2-1-4 2-1-4 3 FPBs BREACHED/JEOPARDIZED EXPLANATION:

Refer to the analysis of the Fission Product Barriers for a detailed analysis of each barrier.

The NUREG requires that if two barriers are breached and a potential exists for a loss of the third fission product barrier, a General Emergency should be declared. The plant has taken the position that if all three barriers are intact, but the potential exists for all of them to breach, a General Emergency Classification (EAL 2-1-4) is warranted. Therefore, HNP considers a jeopardized (not yet breached) barrier as being in the same category as a breached barrier for emergency classification purposes.

NUREG-0654, Att. 2, Exampfa IC Cross Reference to HNP EALs, Rev. 99.1 Pago 68 of 75

Attachment 4 Sheet 3of8 (

GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 3 Loss of physical control of the facility HNP EAL Initiatin Condition: EAL Reference No. 7-1-4 7-1-4 A HOSTILE FORCE HAS TAKEN CONTROL OF PLANT EQUIPMENT AND/OR VITAL AREA(S) SUCH THAT PLANT PERSONNEL ARE UNABLE TO OPERATE EQUIPMENT REQUIRED TO MAINTAINSAFETY FUNCTIONS EXPLANATION:

This item complies with NUREG-0654.

HNP EAL Revision 99-1 updated the terminology used for assessment of this EAL condition.

The revised criteria for General Emergency conditions continue to correspond to the NUREG-0654 condition of" Loss of physical control of the facility " through descriptive terminology to include maintenance of "Safety Functions" as the definition of "control of the facility."

The revised terminology is based on the guidance provided in NUMARC/NESP-007, General Emergency HG1, as outlined in NRC EPPOS ff1, with updated guidance to reflect improvements similar to those in the second issuance of NEI 97-03, Final Draft Rev. 3 (October 1998).

NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 69 of 75

Attachment 4 Sheet 4of8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 4 Other plant conditions exist, from whatever source, that make release of large amounts of radioactivity in a short period possible, e.g., any core melt situation. See the specific PWR and BWR sequences below.

Notes:

a. For core melt sequences where significant releases from containment are not yet taking place and large amounts of fission products are not yet in the containment atmosphere, consider 2 mile precautionary evacuation. Consider 5 mile downwind evacuation (45'o 90'ector) if large amounts of fission products (greater than gap activity) are in the containment atmosphere.

Recommend sheltering in other parts of the plume exposure Emergency Planning Zone under this circumstance.

b. For core melt sequences where significant releases from containment are not yet taking place and containment failure leading to a direct atmospheric release is likely in the sequence but not imminent and large amounts of fission products in addition to noble gases are in the containment atmosphere, consider precautionary evacuation to 5 miles and 10 mile downwind (45'o 90 sector).

c. For core melt sequences where large amounts of fission products other than noble gases are in the containment atmosphere and containment failure is judged imminent, recommend shelter for those areas where evacuation cannot be completed before transport of activity to that location.

d. As release information becomes available, adjust these actions in accordance with dose projections, time available to evacuate and estimated evacuation times given current conditions.

HNP EAL Initlatin Condition:

N/A EXPLANATION:

Notes a, b, and c refer to core melt situations (Generic). Note d is general information and applies to any classification.

The notes are addressed by the Analysis of Fission Product Barriers. If a core melt situation exists, it will be addressed by the Fission Product Barrier analysis.

In the Emergency Action Level Network, a potential loss of any of the fission product barriers (jeopardized) is treated the same as a breached fission product barrier. For the purposes of Emergency Classification, this is an appropriate and conservative treatment of fission product barrier breaches. These situations would result in a General Emergency and the details at the NUREG would be addressed through the Protective Action Recommendation Emergency Plan Implementing Procedure.

NUREG 0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Pago 70 of 75

Attachment 4 Sheet 5of8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 5 Example PWR Sequences: (page 1 of 2)

a. Small and large LOCA's with failure of ECCS to perform leading to severe core degradation or melt in from minutes to hours. Ultimate failure of containment likely for melt sequences.

(Several hours likely to be available to complete protective actions unless containment is not isolated).

b. Transient initiated by loss of feedwater and condensate systems (principal heat removal system) followed by failure of emergency feedwater system for extended period. Core melting possible in several hours. Ultimate failure of containment likely if core melts.

c. Transient requiring operation of shutdown systems with failure to scram which results in core damage or additional failure of core cooling and makeup systems (which could lead to core melt).

HNP EAL lnitlatin Conditions: EAL Reference Nos. 2-1-4, 5-1-4 8 8-1-4 2-1-4 3 FPB's BREACHED/JEOPARDIZED 5-1-4 1A-SA OR 1B-SB NOT ENERGIZED with RCS PRESSURE >360 PSIG with >222.5 KPPH FEED FLOW NOT AVAILABLEwith FULL RANGE RVLIS LEVEL NOT >62%

8-1-4 ATWS WHILE IN MODE 1 OR 2 with MCB MANUALREACTOR TRIP NOT SUCCESSFUL (EITHER SWITCH) with FUEL FPB BREACHED EXPLANATION:

a. The Fission Product Barrier Analysis explains the reasoning used to determine if a breach or jeopardy of one or more fission product barriers is indicated.

The RCS would indicate breached from RCS leakage >50 gpm or CNMT Leak Detection Radiation Monitor Noble Gas Channel >8E-3 ijCI/cc or CNMT Hi Range Accident Monitor >17.5 R/HR.

For either a small or large size LOCA the Containment pressure would rapidly rise to >3 psig resulting in classifying the Containment fission product barrier as being in jeopardy. I A precursor to severe core degradation or melt would be indicated by CNMT High Range Monitor >17.5 R/HR, Gross Failed Detector or RCS activity measurement exceeding the breach criteria, or the Core Cooling Critical Safety function would indicate "Red" for thermocouple temperatures >1200 F or Full Range RVLIS >39% and thermocoupie temperatures >730 'F and would result in classifying the fuel as in Jeopardy.

As a result, there would be early indication of precursors to the NUREG condition which would result in declaration of a General Emergency (EAL 2-1-4).

b. The plausible conditions associated with a complete and sustained loss of all available feedwater flow would be initiated by a loss of electrical power. Refer to item 5.d on the next page

c. At HNP this condition is termed an Anticipated Transient Without Shutdown (ATWS).

If an ATWS event has occurred and the manual reactor trip from the Main Control, using either switch, was not successful an assessment of core status is performed. If the Fuel FPB is breached (core damage) a General Emergency (EAL 8-1-4) is declared.

The HNP Main Control Board has 2 independent Manual Reactor Trip switches. Operation of either switch accomplishes a "scram." I NUREG.0654, An. 2, Exampto IC Cross Reference to HNP EALs, Rov. 99-1 Pago 71 of 75

Attachment 4 Sheet 6of8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 5 Example PWR Sequences: (page 2 of 2)

d. Failure of offsite and onsite power along with total loss of emergency feedwater makeup capability for several hours. Would lead to eventual core melt and likely failure of containment.

e. Small LOCA and initially successful ECCS. Subsequent failure of containment heat removal systems over several hours could lead to core melt and likely failure of containment.

HNP EAL Initlatin Conditions: EAL Reference Nos. 5-1-4 8 2-1-4 5-1-4 1A-SA OR 1B-SB NOT ENERGIZED with RCS PRESSURE >360 PSIG with >222.5 KPPH FEED FLOW NOT AVAILABLEwith FULL RANGE RVLIS LEVEL NOT >62%

2-1-4 3 FPB'S BREACHED/JEOPARDIZED EXPLANATION:

d. If AC power is not available, (1A-SA and 1B-SB Emergency Busses are deenergized) then the possibility of a total loss of feedwater exists.

If the plant is in a condition where Feedwater flow, or secondary heat sink, is required (i.e., "RCS PRESSURE >360 PSIG") (reference EAL Table 3) and 222.5 KPPH OF FEED FLOW is not AVAILABLE'Lossof Steam Driven AFW pump) then core inventory is evaluated..

A General Emergency (EAL 5-1-4) is declared when RVLIS indicates <62% Full Range level.

The 62% Valve is conservative in that it is much higher than the 39% level at which, concurrent with an elevated thermocouple temperature of 730 F, the fuel fission product barrier would become jeopardized.

e. The fission product barrier analysis explains the reasoning used to determine if a breach or jeopardy of one or more of the fission product barriers is indicated.

The RCS would indicate breached from RCS leakage >50 gpm.

The loss of Containment heat removal would result in Containment pressure increasing to or remaining above 3 psig resulting in classifying the Containment fission product barrier as being in jeopardy.

A precursor to severe core degradation or melt would be indicated by CNMT High Range Monitor >17.5 R/HR, Gross Failed Detector or RCS activity measurement exceeding the breach criteria, or the Core Cooling Critical Safety function would indicate "Red" for thermocouple temperatures >1200 F or Full Range RVLIS >39% and thermocouple temperatures >730 'F and would result in classifying the fuel as in jeopardy.

Entry Points T, U, and V are EOP Network entry points. If plant conditions degrade during an off-normal event, the EOP Network directs entry into the EAL Network to reevaluate the current Emergency Classification. This is done regardless of the initiating event or the initial performance of ECCS. By integrating the EOP's in this fashion, a slow degradation of the Fission Product Barriers can be anticipated resulting in a new evaluation of the Emergency Action Level.

Therefore, an initially successful ECCS performance and subsequent loss of control of the event would cause a reevaluation of the Fission Product Barrier status as well as the rest of the EAL Network.

Several symptoms are present which would provide early indication of precursors to the NUREG condition and would result in declaration of a General Emergency (EAL 2-1-4). I The EOP Setpoint Study has calculated (under the guidance of the Westinghouse Owners Group) that this is sufficient flow to ensure that a heat sink exists.

NUREG.0654, Att. 2, CxampIo IC Cross Roforonco to HNP EALs, Rov. 99-1 Pago 72 of 75

Attachment 4 Sheet 7oi8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 6 Example BWR Sequences:

HNP EAL Initiatin Condition:

N/A EXPLANATION:

HNP is a PWR and this item is applicable only to BWR plants NUREG.0654, Att. 2, Example IC Cross Reference to HNP EALs, Rev. 99-1 Page 73 of 75

Attachment 4 Sheet Sof8 GENERAL EMERGENCY CROSS REFERENCE NUREG-0654 EAL: GENERAL EMERGENCY Example IC Item No.: 7 Any major internal or external events (e. g., fires, earthquakes, substantially beyond design basis) which could cause massive common damage to plant systems resulting in any of the above.

HNP EAL Initiatin Condition: EAL Reference No. 11-1-4 11-1-4 ANY RADIOLOGICALCONDITION WARRANTING RECOMMENDATIONTO EVACUATE OR SHELTER THE PUBLIC EXPLANATION:

The HNP EALs utilize an analysis of each of the fission product barriers. This philosophy, in combination with the assessment of the other conditions listed on the EAL Flowpath, provides adequate and anticipatory identification of effects on the plant from unpostulated internal or external events when adding the judgment item. I In the HNP EALs the Site Emergency Coordinator can declare a General Emergency if he feels that it is warranted based on radiological conditions (EAL 11-1-4). This option is provided to give the Site I Emergency Coordinator the flexibilityto declare a General Emergency and issue Protective Action Recommendations for the public in the event that some potential initiating condition was not previously addressed.

NUREG.0654, Att. 2, Example IC Cross Roforenco to HNP EALs, Rev. 99-1 Pago 74 of 75

DOCUMENT CHANGE

SUMMARY

Pa es Chan e Descri tion:

Various EAL Reference numbers have been added to EALs in place of the grid coordinates contained in the previsous revision.

"Site Emergency" classification title was updated to "Site Area Emergency."

Appearance/format of this document has been revised as a result of updates to computer word processing software.

In many instances, the HNP EAL Initiating Condition listed only the "Entry" decision block.

This revision added more complete text which more completely reflects to logic from the flow paths.

Many of the explanations have been re-worded to clarify the discussion, or be more consistent with the descriptions associated with similar EALs, without change in criteria (other than those listed below).

Indicated major aspect of this revision.

Added a discussion of the EAL numbering scheme and the linkage to the EAL Reference Manual for use by offsite authorities.

Deleted reference to EAL Flow Path coordinates (this was also done through out the remainder of the document).

11 -28 Added Unusual Event category numbering (1 - 11) to all items in Attachment 1. These correspond to the EAL numbering scheme.

12 Improved semantics of parenthetical information without change in intent.

13 Allowed T.S. corrective actions have been clarified.

14, 18, 19 Unusual Event Category name changed from "Inability to comply with Tech Spec shutdown requirements" to "Other plant or equipment problems." The new name provides a better description of vents in this category, due to addition of EAL 8-2-1 (Inadvertent Criticality) and relocation of Turbine Rotating Component Failure from category 10 to this category.

Allowed T.S. corrective actions have been clarified. This including listing RCS Pressure Boundary leakage as a separate Initiating Condition (IC) from the remainder of T.S.

leakage.

20 Renamed UE Category 10 from "Other Plant Hazards or Events" to "Other Hazards."

21 Changed lettered elements of these ICs to bullets for consistency within the EALs.

22 Replaced previous "Security Alert" terminology associated with this Unusual Event to be more descriptive.

23 Provided definition of conditions equivalent to a Hurricane and re-worded Tornado aspect.

24 Renamed UE Category 10 from "Other Plant Hazards or Events" to "Other Hazards."

Relocated turbine rotating component failure (EAL 8-3-1) from UE category 10 to category 8 (Other Plant or Equipment Problems).

26 Added an inadvertent criticality Unusual Event which replaced previous boron dilution related ICs.

28 Note format of break clarification was changed to a parenthetical phrase to be consistent with other and similar Unusual Event clarifying information.

39, 57 Added reference to use of either MCB Trip switch satisfied the criteria.

44, 62, 69 Replaced previous "Security Emergency" terminology to be more descriptive and less confusing, through use of NUMARC/NESP-007 EAL Methodology.

47, 65 Uncontrolled Boron Dilution was deleted. It has been replaced by Unusual Event EAL 8-2-1.

52, 53 Cross reference information was inadvertently omitted from previous revisions of this document. These have been restored.

NUREG-0654, Att. 2, Example JC Cross Reference to HNP EALs, Rev. 99.1 Pago 75 of 75

BAW-2355 October 1999 Analysis of Capsule X Carolina Power & Light Company Shearon Harris Nuclear Power Plant Reactor Vessel Material Surveillance Program by M. J. DeVan S. Q. King FTI Document No. 77-2355-00 (See Section 10 for document signatures.)

Prepared for Carolina Power & Light Company Prepared by Framatome Technologies, Inc.

3315 Old Forest Road P. O. Box 10935 Lynchburg, Virginia 24506-0935 F RAMATOME

Executive Summar This report describes the results of the examination of the third capsule (Capsule X) of the Carolina Power & Light Company (CP&L) Shearon Harris Nuclear Power Plant (HNP) as part of their reactor vessel surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of the reactor vessel materials by testing and evaluation of tension test and Charpy V-notch impact specimens. The program was designed in accordance with the requirements of Code of Federal Regulations, Title 10, Part 50, (10 CFR 50) Appendix H and ASTM Standard E 185-82.

Capsule X was removed from the HNP reactor vessel at the end-of-cycle 8 (EOC-8) for testing and evaluation. The. capsule received an average fast fluence of 3.25 x 10" n/cm'E ) 1.0 MeV). Based on the calculated eight-cycle-average full 'power flux and a 90% capacity factor, the projected end-of-life (36 EFPY) peak fast fluence at the base metal-clad interface of the HNP reactor vessel is 4.55 x 10" n/cm'.

The results of the tension tests indicated that the HNP surveillance materials exhibited normal behavior relative to the neutron fluence exposure. The Charpy impact data results for the HNP surveillance materials exhibited the characteristic behavior of transition temperature shifting to a higher temperature as a result of neutron fluence damage and a decrease in upper-shelf energy.

In accordance with Code of Federal Regulations, Title 10, Part 50.61, (10 CFR 50.61), the HNP reactor vessel beltline materials will not exceed the PTS screening criteria before end-of-life (36 EFPY). In addition, the upper-shelf energies of the HNP reactor vessel beltline materials are not predicted to fall below 50 ft-lb at end-of-life (36 EFPY) in accordance with Regulatory Guide 1.99, Revision 2.

F RAMATOME

Acknowledgement The author acknowledges the efforts of Kevin Hour of the B&W Services, Inc. Lynchburg Technology Center. His expertise in specimen testing contributed greatly to the success of this project.

F RAMATOM E n1

Table of Contents

1. Introduction.

2. Background 2-1

3. Surveillance Program Description 3-1

4. Tests of Unirradiated Material .. 4-1

5. Post-Irradiation Testing. . 5-1 5.1. Capsule Disassembly and Inventory. .. 5-1 5.2. Thermal Monitors . 5-1 5.3. Tension Test Results............................ .. 5-1 5.4. Charpy V-Notch Impact Test Results. .. 5-2 5.5. Compact Fracture Toughness and Bend Bar Specimens. .. 5-2

6. Neutron Fluence ...6-1 6.1. Objectives and Background . 6-1 6.2. Results .. 6-2

7. Discussion of Capsule Results .. 7-1 7.1. Unirradiated Material Property Data. 7-1 7.2. Irradiated Property Data. . 7-1 7.2.1. Tensile Properties .. . 7-1 7.2.2. Impact Properties. .. 7-2 7.3. Reactor Vessel Fracture Toughness .. 7-2 7.3.1. Adjusted Reference Temperature Evaluation. .. 7-2 7.3.2. Decrease in Upper-Shelf Energy Evaluation . .. 7-3 7.3.3. Pressurized Thermal Shock Evaluation. .. 7-3

8. Summary of Results 8-1

9. Surveillance Capsule Removal Schedule ..... 9-1

10. Certification .. 10-1

11. References .

lv F RAMATOME 55CHHO5OOIC5

endices ~Pa e A. Reactor Vessel Surveillance Program Background Data and Information... ...A-1 B. Unirradiated and Irradiated Tensile Data for the HNP RVSP Materials... .. B-l Unirradiated and Irradiated Charpy V-Notch Impact Surveillance Data for the HNP RVSP Materials Using Hyperbolic Tangent Curve-Fitting Method. C-1 D. Charpy V-Notch Shift Comparison: Hand-Drawn Curve Fitting vs. Hyperbolic Tangent Curve Fitting . D-1 Fluence Analysis Methodology List of Tables Table ~Pa e 3-1. Test Specimens Contained in HNP Capsule X ... 3-3 3-2 Chemical Composition of HNP Capsule X Surveillance Materials. 3-4 3-3 Chemical Analysis Results of Selected Charpy Specimens from HNP Capsule U for Copper and Nickel Contents 3-5

-4 Heat Treatment of HNP Capsule X Surveillance Materials. 3-6 Tensile Properties of HNP Capsule X Reactor Vessel Surveillance Materials, Irradiated to 3.25 x 10'~ n/cm'E > 1.0 MeV) . 5-3 5-2 Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat No.

B4197-2, Irradiated to 3.25 x 10" n/cm'E > 1.0 MeV) Longitudinal (LT)

Orientation 5-4 5-3 Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat No.

B4197-2, Irradiated to 3.25 x 10" n/cm'E > 1.0 MeV) Transverse (TL)

Orientation 5-5 Charpy V-Notch Impact Results for HNP Capsule X Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Irradiated to 3.25 x 10'/cm'E> 1.0 MeV) ............. 5-6 5-5 Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat-Affected-Zone, Irradiated to 3.25 x 10" n/cm'E> 1.0 MeV). 5-7 5-6 Hyperbolic Tangent Curve Fit Coefficients for the HNP Capsule X Surveillance Materials . 5-8 6-1 C/M Ratios for HNP Capsule X 6-3 6-2 Fast Neutron Fluence (E > 1 MeV) for the HNP Reactor Vessel "Wetted" Inside Surface .. 6-4 Fast Neutron Fluence (E > 1 MeV) for the HNP Reactor Vessel Clad-Base Metal Interface .............................................................................. ............ 6-5 F RAMATOM E

List of Tables cont.

Table ~Pa e 6-4. Peak Fast Fluence Locations of HNP Reactor Vessel Beltline Region Base Metals and Weld Metals 6-6 6-5 HNP Capsule X Fast Neutron Flux (E ) 1 MeV), Fast Neutron Fluence (E ) 1 MeV), and Lead Factors at EOC 8. 6-7 7-1 Summary of HNP Reactor Vessel Surveillance Capsules Tensile Test Results......... 7-4 7-2 Measured vs. Predicted 30 ft-lb Transition Temperature Changes for HNP Capsule X Surveillance Materials 3.25 x 10" n/cm'. 7-6 7-3 Measured vs. Predicte'd Upper-Shelf Energy Decreases for HNP Capsule X Surveillance Materials 3.25 x 10'/cm 7-7 7-4 Summary of HNP Reactor Vessel Surveillance Capsules Charpy Impact Test Results. 7-8 7-5 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 11 EFPY 7-9 7-6 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 12 EFPY 7-10 7-7 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Vessel Applicable to 14 EFPY 7-11 7-8 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 16 EFPY 7-12 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 18 EFPY .......................... 7-13 7-10 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 20 EFPY 7-14 7-11 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 23 EFPY 7-15 7-12 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 25 EFPY 7-16 7-13 Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 36 EFPY 7-17 7-14 Evaluation of Upper-Shelf energy Decreases for the HNP Reactor Vessel Applicable to 36 EFPY 7-18 7-15 Evaluation of Pressurized Thermal Shock Reference Temperatures for the HNP Reactor Vessel Applicable to 36 EFPY 7-19 A-1 HNP Surveillance Capsule Identifications, Original Locations, and Design Lead Factors ....... ................................. .......... .......... A-3 A-2 A-3 Description of the HNP Reactor Vessel Beltline Region Materials.

Heat Treatment of the HNP Reactor Vessel Beltline Region Materials.................. -0 F RAMATOME VI

List of Tables cont.

Table ~Pa e 8-1 Unirradiated Surveillance Tensile Properties of HNP Base Metal Plate, Heat No. 84197-2, Longitudinal Orientation . 8-2 8-2 Unirradiated Surveillance Tensile Properties of HNP Base Metal Plate, Heat No. 84197-2, Transverse Orientation. 8-2 8-3 Unirradiated Surveillance Tensile Properties of HNP Weld Metal, Wire Heat 5P6771 / Flux Lot 0342 8-3 8-4 HNP Capsule U Surveillance Tensile Properties of Base Metal Plate, Heat No.

84197-2, Irradiated to 5.52 x 10" n/cm'E > 1.0 MeV) Longitudinal Orientation ... 8-4 8-5 HNP Capsule U Surveillance Tensile Properties of Base Metal Plate, Heat No.

84197-2, Irradiated to 5.52 x 10" n/cm'E > 1.0 MeV) Transverse Orientation...... 8-4 8-6 HNP Capsule U Surveillance Tensile Properties of Weld Metal, Wire Heat 5P6771 /

Flux Lot 0342, Irradiated to 5.52 x 10'/cm (E > 1.0 MeV).. . 8-4 8-7 HNP Capsule V Surveillance Tensile Properties of Base Metal Plate, Heat No.

84197-2, Irradiated to 1.32 x 10" n/cm'E > 1.0 MeV) Longitudinal Orientation ... 8-5 8-8 HNP Capsule V Surveillance Tensile Properties of Base Metal Plate, Heat No.

84197-2, Irradiated to 1.32 x 10'/cm (E > 1.0 MeV) Transverse Orientation...... 8-5 HNP Capsule V Surveillance Tensile Properties of Weld Metal, Wire Heat 5P6771 /

Flux Lot 0342, Irradiated to 1.32 x 10" n/cm'E > 1.0 MeV)...;.... 8-5 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. 84197-2, Longitudinal (LT) Orientation.

C-2 HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

84197-2, Irradiated to 5.52 x 10" n/cm'E > 1.0 MeV) Longitudinal (LT)

Orientation ... .......................................................................... C-3 C-3 HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

84197-2, Irradiated to 1.32 x 10'/cm'E > 1.0 MeV) Longitudinal (LT)

Orientation ........................................... C-4 C-4 Hyperbolic Tangent Curve Fit Coefficients for HNP Base Metal Plate, Heat No.

84197-2, Longitudinal (LT) Orientation C-5 C-5 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. 84197-2, Transverse (TL) Orientation C-6 C-6 HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

84197-2, Irradiated to 5.52 x 10" n/cm'E > 1.0 MeV) Transverse (TL)

Orientation ..... ................................. .............................. C-7

~

C-7 HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

84197-2, Irradiated to 1.32 x 10" n/cm'E> 1.0 MeV) Transverse (TL)

Orientation F RAMATOME vll

List of Tables cont.

Table ~Pa e C-8. Hyperbolic Tangent Curve Fit Coefficients for HNP Base Metal Plate, Heat No.

B4197-2, Transverse (TL) Orientation C-9 C-9 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Weld Metal, Wire Heat SP6771 / Flux Lot 0342 C-10 C-10 HNP Capsule U Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342), Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV) ............ C-11 C-11 HNP Capsule V Surveillance Charpy Impact Data for Weld Metal (Wire Heat SP6771 / Flux Lot 0342), Irradiated to 1.32 x 10" n/cm (E) 1.0 MeV) ............ C-12 C-12 Hyperbolic Tangent Curve Fit Coefficients for HNP Weld Metal (Wire Heat SP6771 / Flux Lot 0342)

C-13 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Heat-Affected-Zone Material C-14 C-14 HNP Capsule U Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 5.52 x 10" n/cm (E) 1.0 MeV).

C-15 HNP Capsule V Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 1.32 x 10" n/cm'E) 1.0 MeV). C-16 C-16. Hyperbolic Tangent Curve Fit Coefficients for HNP Heat-Affected-Zone Material C-17 D-1 Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Base Metal Plate Heat No. B4197-2, Longitudinal (LT) Orientation........D-2 D-2 Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Base Metal Plate Heat No. B4197-2, Transverse (TL) Orientation..........D-3 D-3 Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Weld Metal (Wire Heat'5P6771 / Flux Lot 0342)

D-4 Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Heat-Affected-Zone.................................................................D-5 List of Fi ures

~Fi ere ~Pa e 3-1. Reactor Vessel Cross Section Showing Location of RVSP Capsules in HNP Reactor Vessel 3-7 3-2. Surveillance Capsule Assembly Showing Location of Specimens and Monitors for HNP Capsule X 3-8 F RAMATOME viii

List of Fi ures cont.

~Fi ure ~Pa e 5-1. Photographs of Thermal Monitors Removed from HNP RVSP Capsule X............ . 5-9 5-2 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL11, Tested at 70'F..................... 5-10 5-3 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL12, Tested at 300'F 5-10 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL10, Tested at 550'F 5-11 5-5 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Transverse Orientation, Specimen No. QT11, Tested at 70'F .... 5-11 5-6 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Transverse Orientation, Specimen No. QT12, Tested at 300'F 5-12 5-7 Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Transverse Orientation, Specimen No. QT10, Tested at 550'F 5-12 5-8 Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Specimen No. QW11, Tested at 70'F. 5-13 Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Specimen No. QW10, Tested at 300'F............................"........... 5-13 Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Specimen No. QW12, Tested at 550'F ......................................... 5-14 5-11 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces Base Metal Plate, Heat No. B4197-2 Longitudinal Orientation ............5-15 5-12 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces Base Metal Plate, Heat No. B4197-2 Transverse Orientation ..............5-16 5-13 Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces Weld Metal, Wire Heat 5P6771 / Flux Lot 0342 .........

5-14 Charpy Impact Data for Irradiated Base Metal Plate Heat No. B4197-2 Longitudinal (LT) Orientation............................... 5-18 5-15 Charpy Impact Data for Irradiated Base Metal Plate Heat No. B4197-2 Transverse (TL) Orientation. ..... 5-19 5-16 Charpy Impact Data for Irradiated Weld Metal Wire Heat 5P6771 / Flux Lot 0342 .......................................  ;....... ............. 5-20 5-17 Charpy Impact Data for Irradiated Base Metal Plate Heat-Affected-Zone ........ 5-21 5-18 Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate Heat No. B4197-2, Longitudinal (LT) Orientation 5-22 Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate 0 Heat No. B4197-2, Transverse (TL) Orientation 5-24 F RAMATOME 1x

~Fi ere 5-20.

Listof Fi ures cont.

Photographs of Charpy Impact Specimen Fracture Surfaces, Weld Heat 5P6771 /

Flux Lot 0342

~Pa

........... 5-26 e

I 5-21 Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate, Heat-Affected-Zone ...................................................................... 5-28 7-1 Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate, Heat No. B4197-2, Longitudinal'(LT) Orientation ..................7-20 7-2 Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation..................... 7-21 7-3 Comparison of Unirradiated and Irradiated Charpy Impact Data Cur'ves for 7-4 Weld Metal (Wire Heat 5P6771 / Flux Lot 0342).......7-22 Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Heat-Affected-Zone Material. .... 7-23 A-1 Locations and Identifications of Materials Used in the Fabrication of the HNP Reactor Pressure Vessel ...~............................................... A-6 A-2 Locations of Longitudinal Welds in HNP Reactor Vessel Upper and Lower Shell Courses . ................................................ .............. ...... .............A-7 A-3 Locations of Surveillance Capsule Irradiation Sites in the HNP Reactor Vessel.......A-8 C-1 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -.........~..... C-18 C-2 HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

B4197-2, Longitudinal (LT) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -................... C-19 C-3 HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

B4197-2, Longitudinal (LT) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -.................... C-20 C-4 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -. .............. C-21 C-5 HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

B4197-2, Transverse (TL) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -................... C-22 C-6 HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No.

B4197-2, Transverse (TL) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -.

.... C-23 C-7 Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method - . C-24 I F RAMATOME

List of Fi ures cont.

~Fi ure ~Pa e C-8. HNP Capsule U Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -. .. C-25 C-9. HNP Capsule V Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

Refitted Using Hyperbolic Tangent Curve-Fitting Method -. .. C-26 C-10. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method- .. C-27 C-11. HNP Capsule U Surveillance Charpy Impact Data for Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -.......... .... C-28 C-12. HNP Capsule V Surveillance Charpy Impact Data for Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method -..

.... C-29 E-1. Fluence Analysis Methodology . .. E-10 xi F RAMATOME

1. Introduction This report presents the examination results of the third reactor vessel surveillance" capsule (Capsule X) removed from the Carolina Power and Light Company's (CPS') Shearon Harris Nuclear Power Plant (HNP) reactor vessel. The capsule was removed and the contents evaluated after being irradiated in the HNP reactor vessel as part of the reactor vessel surveillance program (RVSP) s documented in WCAP-10502."'his report describes the testing and the post-irradiation data obtained from the HNP Capsule X after receiving an average fluence of 3.25 x 10" n/cm'E ) 1.0 MeV). The data are compared to previous HNP RVSP results from Capsule U"'nd Capsule V.'"

The objective of the program is to monitor the effects of neutron irradiation on the mechanical properties of reactor vessel materials under actual plant operating conditions. The program was lanned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-ear design life of the reactor pressure vessel. The HNP RVSP was designed and furnished by Westinghouse Electric Corporation and was based on American Society for Testing and Materials (ASTM) Standard E 185-82"'nd is in compliance with the Code of Federal Regulations, Title 10, Part 50, (10 CFR 50) Appendix G"'nd Appendix H."'

RAMATOME

2. Background

The ability of the reactor vessel to resist fracture is a primary factor in ensuring the safety of the primary system in light water-cooled reactors. The reactor vessel beltline region is the most critical region of the vessel because it is exposed to the highest level of neutron irradiation. The general effects of fast neutron irradiation on the mechanical properties of low-alloy ferritic steels used in the fabrication of reactor vessels are well characterized and documented. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. The most significant mechanical property change in reactor vessel steels is the increase in the ductile-to-brittle transition temperature accompanied by a reduction in the Charpy upper-shelf energy (CUSE) value.

10 CFR 50 Appendix G, "Fracture Toughness Requirements, " specifies minimum fracture oughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of light water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations for operation of the RCPB. The fracture toughness and operational requirements are specified to provide adequate safety margins during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of 10 CFR 50, Appendix G, became effective on August 16, 1973, the requirements are applicable to all boiling and pressurized water-cooled nuclear power reactors, including those under construction or in operation on the effective date.

10 CFR 50, Appendix H, "Reactor Vessel Materials Surveillance Program Requirements, "

defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of water-cooled reactors resulting from exposure to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens contained in capsules that are periodically withdrawn from the reactor vessel. These data permit determination of the conditions under which the vessel can be operated with adequate safety margins against non-ductile fracture iroughout its service life.

2-1 F RAMATOME

A method for guarding against non-ductile fracture in reactor vessels is described in Appendix G to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BgcPV)

Code, Section, III, "Nuclear Power Plant Components"'" and Section XI, "Rules for Inservice Inspection. '"" This method uses fracture mechanics concepts and the reference nil-ductility temperature, RTgpT which is defined as the greater of the drop weight nil-ductility transition temperature (in accordance with ASTM E 208-81') or the temperature that is 60'F below that at which the material exhibits 50 ft-lbs and 35 mils lateral expansion. The RT>>T of a given material is used to index that material to a reference stress intensity factor curve (K>> curve), which appears in Appendix G of ASME BEcPV Code Section HI and Section XI. The K>> curve is a lower bound of dynamic and crack arrest fracture toughness data obtained from several heats of pressure vessel steel. %hen a given material is indexed to the K>> curve, allowable stress intensity factors can be obtained for the material as a function of temperature. The operating limits can then be determined using these allowable stress intensity factors.

The RT>>T and, in turn, the operating limits of a nuclear power plant, are, adjusted to account for the effects of irradiation on the fracture toughness of the reactor vessel materials. The irradiation embrittlement and the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by a surveillance program in which surveillance capsules containing prepared specimens of the reactor vessel materials are periodically removed from the operating nuclear reactor and the specimens are tested. The increase in the Charpy V-notch 30 ft-lb temperature is added to the original RTgpT to adjust it for irradiation embrittlement. The adjusted RTgpT is used to index the material to the K>> curve which, in turn, is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

10 CFR 50, Appendix G, also requires a minimum initial CUSE of 75 ft-lbs for all beltline region materials unless it is demonstrated that lower values of upper-shelf fracture energy will provide an adequate margin of safety against fracture equivalent to those required by ASME Section XI, Appendix G. No action is required for a material that does not meet the initial 75 ft-Ibs requirement provided that the irradiation embrittlement does not cause the CUSE to drop below 50 ft-lbs. The regulations specify that if the CUSE drops below 50 ft-lbs, it must be demonstrated, in a manner approved by the Office of Nuclear Reactor Regulation, that the lower values will provide adequate margins of safety.

F RAMATOME 2-2

3. Surveillance Program Description The reactor vessel surveillance program for HNP includes six capsules designed to monitor the effects of neutron and thermal environment on the materials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned inside the reactor vessel between the neutron shielding pads and the vessel wall at the locations shown in Figure 3-1. The six capsules, designed to be placed in holders attached to the neutron shield pads, are positioned near the peak axial and azimuthal neutron flux.

WCAP-10502 includes a full description of the capsule locations and design. Capsule X was irradiated in the 287'osition during the time of irradiation in the reactor vessel (cycles 1 through 8).

Capsule X was removed during the eighth refueling shutdown of the HNP plant. The capsule contained Charpy V-notch (CVN) impact test specimens fabricated from one base metal plate SA-533, Grade B, Class 1), heat-affected-zone (HAZ) material, and a weld metal representative of the HNP reactor vessel beltline region intermediate to lower shell circumferential weld. The tensile test specimens were fabricated from the same base metal plate and weld metal. In addition, specimens were included for determining the fracture toughness of both the base metal plate and the weld metal. The number of specimens of each material contained in Capsule X is described in Table 3-1, and the location of the individual specimens within the capsule is shown in Figure 3-2. The chemical compositions of the surveillance materials in Capsule X, obtained from the original surveillance program report,"'re described in Table 3-2. In addition, six irradiated base metal and weld metal Charpy specimens from Capsule U were analyzed by emission spectrograph to determine the copper and nickel contents."'he results of these analyses are presented in Table 3-3. The heat treatment of the surveillance materials in Capsule X is presented in Table 3-4.

All base metal specimens were machined from the ~/4-thickness (~/4 T) location of the plate material after stress relieving. The base metal, HAZ material, and weld metal specimens were oriented such that the longitudinal axis of the specimen was either parallel or perpendicular to the principal working direction of the plate.

3-1' RAMAT0M E

Capsule X contained dosimeter wires of copper, iron, nickel, and aluminum-0.15 weight percent cobalt (cadmium-shielded and unshielded) and cadmium-shielded neptunium-237 P'Np) and uranium-238 P'U). The location of these dosimeters within Capsule X is shown in Figure 3-2.

Thermal monitors fabricated from two low-melting alloys were included in the capsule. The thermal monitors were sealed in Pyrex tubes and inserted in spacers located in Figure 3-2. The eutectic alloys and their melting points are listed below:

2.5% Ag, 97.5% Pb Melting Point 579'F 1.5% Ag, 1.0% Sn, 97.5% Pb Melting Point 590'F 3-2 F RAMAYOME

Table 3-1; Test Specimens Contained in HNP Capsule X Number of Test Specimens Material Description Tension CVN Impact Compact Bend Bar Base Metal Plate (Heat No. B4197-2)

Longitudinal 15 Transverse 15 HAZ Metal 15 Weld Metal 15 (Wire Ht. 5P6771 /

Flux Lot 0342)

Total 60 12 3-3

Table 3-2. Chemical Composition of HNP Capsule X Surveillance Materials Chemical Composition, wt%'"

Base Metal Plate Weld Metal Element Heat No. B4197-2 (Wire Ht. 5P6771 /

Flux Lot 0342) 0.22 0.04 Mn 1.34 1.23 P 0.014 0.006 0.016 0.006 Si 0.37 0.41 Ni 0.49 0.87 Mo 0.50 0.48 Cr 0.095 0.068 Cu 0.073 0.023 Al 0.043 0.018 Co 0.011 0.009

. Pb 0.001 <0.005 Not Detected '.02 0.004 <0.001 Not Detected <0.001 Not Detected <0.008 Sn 0.006 0.09 As 0.007 <0.005 Cb Not Detected <0.005 N~ 0.007 0.005 B <0.001 0.011 3-4 F RAMATOME

Table 3-3. Chemical Analysis Results of Selected Charpy Specimens from HNP Capsule V for Copper and Nickel Contents"'aterial Charpy Copper, Nickel, S ecimen wt%" wt%"

Base Metal Plate, QL1 0.090 0.52 B 4197-2 QL8 0.095 0.51 QL12 0.088 0.50 Weld Metal QW1 0.026 0.94 (Wire Ht. 5P6771 /

Flux Lot 0342) QW12 0.019 1.07 QW13 0.029 0.95

'*'nalysis performed with an emission spectrometer.

3-5 F RAMATOM E

Table 3-4. Heat Treatment of HNP Capsule X Surveillance Materials Material Heat Treatment" Base Metal Plate 1575-1625'F for 4 hrs., water quenched Heat No. B4197-2 1225-1275'F for 4 hrs., air cooled 1050'F for 4 hrs., air cooled 1100-1175'F for 35~/4 hrs., furnace cooled Weld Metal 1100-1175'F for 10 i/4 hrs., furnace cooled (Wire Ht. 5P6771 i Flux Lot 0342)

" Austenitizing, tempering, and stress relief times are from Lukens Steel Company Test Certificates. The post weld heat treatment times are from Chicago Bridge A. Iron Company thermal history records. The stress relief and post weld heat treatment times and temperatures were established to simulate that to which the materials within the HNP reactor vessel were exposed.

3-6 F RAMATOME

q y Figure 3-1. Reactor Vessel Cross Section Showing Location of RVSP Capsules in HNP Reactor Vessel CAPSULES po U (343 ) REACTOR VESSEL (340') Z CORE BARREL CAPSULES (290') Y (287 NEUTRON

) X PAD 270'0'APSULES V (107')

W (110')

180'LAN VIEW F RAMAYOME 3-7

Figure 3-2. Surveillance Capsule Assembly Showing Location of Specimens and Monitors for HNP Capsule X Tension Test Compact Charpy Specimens Fracture Specimens Specimens Dosimeter and Dosimeter and C Thermal Monitor Thermal Monitor IC Block Block 0 0 O O 0

Compact Charpy Tension Test Charpy Specimens Compact Tension Test Fracture Specimens Specimens Fracture Specimens Specimens Specimens Dosimeter Block Op o+

oO

<<m

4. Tests of Unirradiated Material Unirradiated material was evaluated for two purposes: (1) to establish baseline data to which irradiated properties data could be compared; and (2) to determine those material properties as required for compliance with 10 CFR 50, Appendices G and H.

Westinghouse Electric Corporation, as part of the development of the HNP RVSP, performed the testing of the unirradiated surveillance material. The details of the testing procedures are described in Westinghouse Electric Corporation Report WCAP-10502."'he unirradiated mechanical properties for the HNP RVSP materials are summarized in Appendices B and C of this report.

The original unirradiated Charpy V-notch impact data were evaluated based on hand-fit Charpy curves generated using engineering judgment. These data were re-evaluated herein using a yperbolic tangent curve-fitting program, and the results of the re-evaluation are presented in Appendix C. In addition, Appendix D contains a comparison of the Charpy V-notch shift results for each surveillance material, hand-fit versus hyperbolic tangent curve-fit.

4-1 F RAMAYOM E

5. Post-Irradiation Testing The post-irradiation testing of the tension test specimens, the Charpy V-notch impact specimens, thermal monitors, and dosimeters for the HNP Capsule X was performed at the B&W Services Inc. (BWSI) Lynchburg Technology Center (LTC)."

'.1.

Capsule Disassembly and Inventory After capsule disassembly, the contents of Capsule X were inventoried and found to be consistent with the surveillance program report inventory (WCAP-10502). The capsule contained a total of 60 standard Charpy V-notch specimens, nine (9) tensile specimens, twelve (12) 0.5T compact fracture specimens, one (1) bend bar specimen, four (4) dosimetry blocks, and three (3) temperature monitors.

~

5.2. Thermal Monitors

~ ~

he low-melting point (579'F and 590'F) eutectic alloys contained in Capsule X were x-rayed to reveal the shape of the monitors and examined for evidence of melting. No indication of melting

~

was observed (see Figure 5-1). Therefore, based on this examination, the maximum temperature that the capsule test specimens were exposed to was less than 579;F.

5.3. Tension Test Results The results of the post-irradiation tension test are presented in Table 5-1, and the stress-strain curves are presented in Figures 5-2 through 5-10. Tests were performed at room temperature (70'F), 300'F, and the HNP operating temperature (550'F). The tests were performed using a MTS servohydraulic test machine. All tension tests were run using stroke control with an initial actuator travel rate of 0.0075 inch per minute. Following specimen yielding, an actuator speed of 0.03 inch per minute was used. The tension testing was performed in accordance with the applicable requirements of ASTM Standard E 8-96a"" and ASTM Standard E 21-92.""

Photographs of the tension test specimen fractured surfaces are shown in Figures 5-11 through 5-13.

5-1 F RA MAT0 M E

5.4. Charpy V-Notch Impact Test Results The Charpy V-notch impact testing was performed in accordance with the applicable requirements of ASTM Standard E 23-91."" Impact energy, lateral expansion, and percent

-

shear fracture were measured at numerous test temperatures and recorded for each specimen.

The impact energy was measured using a certified Satec S1-1K Impact tester (traceable to NIST Standard) with a striker velocity of 16.90 ftlsec and 240 ft-lb of available energy. The lateral expansion was measured using a certified dial indicator. The specimen percent shear was estimated by video examination and comparison with the visual standards presented in ASTM Standard E 23-91.

The results of the Charpy V-notch impact testing are shown in Tables 5-2 through 5-5 and Figures 5-14 through 5-17. The curves were generated using a hyperbolic tangent curve-fitting program to produce the best-fit curve through the data. The hyperbolic tangent (TANH) function (test response, i.e., absorbed energy, lateral expansion, and percent shear fracture, "R," as a function of test temperature, "T") used to evaluate the surveillance data is as follows:

8 = A + 8 "'anh (T To)

C The Charpy V-notch data was entered, and the coefficients A, 8, To, and C are determined by the program minimizing the sum of the errors squared (least-squares fit) of the data points about the fitted curve. Using these coefficients and the above TANH function, a smooth curve is generated through the data for interpretation of the material transition region behavior. The coefficients determined for irradiated materials in Capsule X are shown in Table 5-6.

Photographs of the Charpy V-notch specimen fracture surfaces are presented in Figures 5-18 through 5-21.

5.5. Compact Fracture Toughness and Bend Bar Specimens The 0.5T compact fracture toughness specimens and the bend bar specimen were not tested at the request of Carolina Power 8c Light Company. The specimens are to be stored at the BWSI LTC facility for possible future testing.

F RAMATOME 5-2

Table 5-1. Tensile Properties of HNP Capsule X Reactor Vessel Surveillance Materials, Irradiated to 3.25 x 10" n/cm'E >1.0 MeV)

Stren th Fracture Pro erties Elon ation Reduction Specimen Test Temp. Yield Ultimate Load Stress Strength Uniform Total in Area Material No. ('F) ksi) (ksi) (jb) (ksi) (ksi) (% (%)

Base Metal Plate QL11 70 79.6 101.2 3611 185 73.6 10.2 22.9 Heat No. B4197-2 Longitudinal QL12 71.7 92.6 3332 171 67.9 9.18 20.3 QL10 550 72.8 99.0 4078 148 83.1 8.68 17.0 43.8 Base Metal Plate 70 QT11 80.0 101.0 3741 191 76.2 9.80 22.5 Heat No. B4197-2 Transverse QT12 300 72.6 92.8 3535 162 72.0 9.33 19.4 55.6 QT10 550 70.4 95.8 4032 176 82.1 9.87 17.8 53.2 Weld Metal QW11 70 82.8 95.4 3216 177 65.5 10.2 24.8 63.0 (Wire Heat 5P6771/

Flux Lot 0342) QW10 76.0 88.2 3040 174 61.9 9.96 22.3 QW12 550 75.6 93.0 3268 171 66.6 9.40 21.1 61.1 zg Op 0

oO

Table 5-2. Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat No. B4197-2, Irradiated to 3.25 x 10" n/cm'E >1.0 MeV)

Longitudinal (LT) Orientation Test Impact Shear Specimen Temperature, Energy, Expansion, Fracture, ID oF ft-lbs mil QL49 74 13.5 9 10 QL50 104 15.5 12 30 QL51 129 21.0 18 50 QL47 154 28.5 22 55 QL48 154 30.5 23 55 QL46 179 42.5 36 65 QL56 179 47.5 39 65 QL60 204 63.5'8.5'ateral 61.0 48 90 QL52 229 53.0 45 85 QL57 254 67.5* 57 100 QL58 254 68.5 53 100 QL54 304 56 100 64.5'1.0 QL55 304 53 100 QL53 354 59 100 QL59 404 59 100

~ Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.

5-4

Table 5-3. Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat

~

No. B4197-2, Irradiated,to 3.25 x 10" n/cm'E > 1.0 MeV)

~

~ ~ ~

~ ~

Transverse (TL) Orientation Test Impact Shear Specimen Temperature, Energy, Expansion, Fracture, ID oF ft-lbs mil QT56 74 13.5 8 10 QT55 104 19.5 16 30 QT50 129 20.5 16 40 QT49 154 24.5 20 50 QT52 154 32.0 25 60 QT57 179 43.5 37 50 QT60 179 36.5 34 65 QT58 204 47.5 39 95 QT46 229 42.5 38 70 55.0'0.0'0.5'4.5'6.5'ateral QT47 254 43.5 40 85 QT54 254 53 100 QT48 304 47 100 QT51 304 55 100 QT53 354 51 100 QT59 404 53 100

":Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.

5-5 F RAMAYO ME

Table 5-4. Charpy V-Notch Impact Results for HNP Capsule X Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Irradiated to 3.25 x 10" n/cm'E) 1.0 MeV)

Test Impact Shear Specimen Temperature, Energy, Expansion, Fracture, ID oF ft-Ibs mil QW59 -76 4.5 5 0 QW56 -36 11.0 6 5 QW58 4 11.5 7 25 QW49 24 21.5 17'4 45 QW54 44 31.0 50 QW52 74 43.5 33 70 QW48 104 52.0 45 80 QW53 104 50.5 41 75 QW55 129 70.0'5.5'4.0'ateral 58.5 51 85 QW57 154 63.5 52 95 QW51 204 61 100 67.0'8.5*

QW47 254 63 100 QW46 304 65 100 QW50 304 61 100 QW60 404 59 100

• Value used to determine upper-shelf energy (USE) in accordance .,

with ASTM Standard E 185-82.

5-6 F RA M AT 0 M E

Table 5-5. Charpy V-Notch Impact Results for HNP Capsule X Base Metal Plate, Heat-Affected-Zone, Irradiated to 3.25 x 10" n/cm'E) 1.0 MeV)

Test Impact Shear Specimen Temperature, Energy, Expansion, Fracture, ID oF ft-lbs mil QH57 -76 2.5 3 0 QH59 -36 16.5 8 10 QH55 4 22.5 14 40 QH56 24 30.5 19 50 QH51 44 40.0 28 60 QH48 74 67.0'ateral 46.5 32 60 QH50 104 59.5 45 100 QH58 104 41.5 32 50 QH52 129 53.5 43 85 QH60 129 76.5'6.0 56 100 QH54 154 43 95 QH49 204 67.5'1.5 49 100 QH47 254 46 100 QH46 304 62.5 54 100 QH53 404 55 100

":Value used to determine upper-shelf energy (USE) in accordance with ASTM Standard E 185-82.

5-7 F RAMATOME

Table 5-6. Hyperbolic Tangent Curve I~'it Coefficients for the HNP Capsule X Surveillance Materials Material Hyperbolic Tangent Curve Fit Coefficients Description Absorbed Energy Lateral Expansion Percent Shear Fracture Base Metal Plate A: 34.6 A: 29.6 A: 500 B4197-2 (LT) B: 324 B: 28 6 B: 500 C: 70.4 C: 76.6 C: 79.3 TO: 155.1 TO: 162.0 TO: 143.0 Base Metal Plate A: 29.3 A: 26.9 A: 500 B4197-2 (TL) B: 27.1 B: 25.9 B: 500 C: 99 9 C: 95.1 C: 900 TO: 148.8 TO: 160.5 TO: 150.1 Weld Metal A: 348 A: 31.6 A: 500 (Wire Ht. 5P6771 / B: 32.6 B: 306 B: 500 Flux Lot 0342) C:, 77.7 C: 84.0 C: 81.1 TO: 56.4 TO: 70.4 TO: 45.1 HAZ Metal A: 33.5 A: 26 3 A: 500 B: 31.3 B:. 25.3 B: 500 C: 85.7 C: 87.5 C: 92.6 TO: 33.4 TO: 48.1 TO: 36.0 5-8 F RAMAYO ME

t Figure 5-1. Photographs of Thermal Monitors Removed from the HNP RVSP Capsule X THERMAL SHEAIWN ~8N goN(~E~S TQP NG 84T zg Op 0

oo0 v Pl

Figure 5-2. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL11, Tested at 10'F 3

Fll~ t I...

OLII 1999 Spec ieeni OLI I 'lect teen.s 70 F I 2I CI R Stren9th lietdt 79600, UISt 101190.

e e

e

~ e pS 2'd d d

0. OO 0. Oti O. OS 0. 12 0. IS 0. 20 0. PI O. 28 0. 32 CnOlneerln9 Strain Dotted line indicates 0.2% offset.

Figure 5-3. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL12, Tested at 300'F 8 Ita... 1999 Filet OL12 Snecieeni OL12 lect I ~ .e 300 F I Ia8 CI Strength 1 t rid t 7I679.

UI St 92839.

pS

~

5 ee t p Pde w p d d 0.00 O,OI 0,08 0.12 0.16 0.20 0.21 0.28 0.32 Enyl nrrrl n9 Strain Dotted line indicates 0.2% offset.

F RAMATOME 0 00 I e 5-10 a a c ee et L a

Figure 5-4. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation, Specimen No. QL10, Tested at 550'F 3 h I 999 fll ~s

~

Otig Snecteens OLIO Test Ies . ~ 550 T I 28T CI R Strength I I e I 8 t 72826.

UT5s 98950.

ts st Q sn s te ts ss Z tr

~ ss cs cs

0. CO 0. OI 0. 08 0. 12 0. I 6 0. 20 0. Fi 0. 28 0. 32 Cnglneering Strain Dotted line indicates 0.2% offset.

Figure 5-5. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Transverse Orientation,,

Specimen No. QT11, Tested at 70'F 3 h . I999 fil ~s gill Scectsens OTII Test Tern.s TO CI 2I CI Strength

'Ilelgi 80000.

VTSs l00965.

e s

pS ~

go c) re ss cl

0. 00 0. OI 0. 08 0. I 2 0. I 6 0. 20 0. 2g 0. 28 0. 32 Cngine<<ring Strain Dotted line indicates 0.2% offset.

5-11 F s a RAMATO c ss ss L ME o o o Ie s

Figure 5-6. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. 84197-2, Transverse Orientation, Specimen No. QT12, Tested at 300'F 8 tte.. I999 Filet Oli2 Seeclnenl OII2 Teet Tenn.t 300 F( 198 CI Strength 1 tel dl 12559.

01St 92158.

j O L

pS O g tt tr o O d 0.00 O.OI 0.08 0.12 0, I6 0.20 ~ 0.29 0.28 0.32 Cngtneerlng Streln Dotted line indicates 0.2% offset.

Figure 5-7. Tension Test Stress-Strain Curve for Base Metal Plate, Heat No. B4197-2, Transverse Orientation, Specimen No. QT10, Tested at 550'F 3 h . I999 F I let OTIO Sneelnenl OTIO Teet Tern.l 550 F I 281 CI R Strength TI ~ Idl TTNOO.

IIIS: 95156.

Vl g

nl

~

V n L

L trS P

gO 2 tr 8

0.00 O.TTI 0.08 0.12 0.16 0.20 0.2tt 0 28 0.32 Cng Inc<<tng Stre ln Dotted line indicates 0.2% offset.

F RAMATOME TecHHotooI55 5-12

Ftigure 5-8. Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Specimen No. QW11, Tested at 70'F 3 h... 1999 F ll~ i Ottl I elean: 13tll lect Iree.s 70 FI 21 Cl Strength 1 Ie Id > 82800.

VISi SSg19 e

L trS L

FO O

0. 00 0. OI 0. 08 0. 12 0. 16 0. 20 0. 2g 0. 28 0. 32 Engineer lng Strain Dotted line indicates 0.2% offset.

Figure 5-9. Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Ftlux Lot 0342, Specimen No. QW10, Tested at 300'F 8 Ite.. 1999 F tier Ottlig Sneeteen> 0410 Feet I~ .> 300 FI 198 Cl Strength VI ~ tdi FSSSI.

UI3 l 88181.

n

~I L

pa ~

L e

Te d 4J O O

0. 00 O. OI 0. 08 0. 12 0. 16 0. 20 0. 2g 0. 28 0. 32 Engineer lng Strain Dotted line indicates 0.2% offset.

5-13 F RAMATOME aacHHoaoolae

Figure 5-10. Tension Test Stress-Strain Curve for Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Specimen No. QW12, Tested at 550'F 3 t( . ~ I 999 6ll~ SII2 s

Ssesteens OII12 'fest Teen.s 550 F( 281 Cl Stren0th iield: 356sl3.

015'XN6.

si o pS s

F L

e ts osr es R

ss O

0. 00 0. tN 0. 08 0. 12 0. l 6 0. 20 0. Pt 0. 28 0. 32 En9lneerln9 Streln Dotted line indicates 0.2% offset.

5-14 F

'I E RAMATO c H ss ME o s o o Is s

~

Figure 5-11. Photographs of Tested Tension Test Specimens and Corresponding

~

Fracture Surfaces Base Metal Plate, Heat No. 84197-2 Longitudinal Orientation h

'*~( ~'+/'/~+ ~Q w+'j'J, MQP'P+Q spy~

I kl la g~ ~ 6 ~~,gQ+ ~ ~~ptg q~@)p h g,a~a~

qI -12 300oF 4 g' . 't.4, QL-10 550oF 5-15 F RAMATO ME TCCHN&lOOIC$

Figure 5-12. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces Base Metal Plate, Heat No. 84197-2 Transverse Orientation I

'<< <<)~i,<<,+s

• > <<. -n. <<~~4>o ~ >~~.

8R ~

QT-11 70 F QT-12 300'I'

~a

<<5'Pv k44~4t~<<~~~V@~HiA&

QT-IO 550'F 5-16 F RAMATOME TCCHHOIOOIC5

t Figure 5-13. Photographs of Tested Tension Test Specimens and Corresponding Fracture Surfaces Weld Metal, Wire Heat 5P6771 / Flux Lot 0342 l

,-. ~~%PA@.

t QW-11 70'F QW-10 300oF QW-I2 550'F>>

F RAMATOME TCCHHOlOOIE5 5-17

Figure 5-14. Charpy Impact Data for Irradiated Base Metal Plate Heat No. B4197-2 Longitudinal (LT) Orientation 100 75 50 u

I o 25 CO

-100 100 200 300 400 500 600 Tomporaturo, F 100 80 a

0 60

~ ~

x 40 uj 20 0

-100 100 200 300 400 500 600 Tomporaturo, F 120 3SMLe: + 7

'40:

Tao: +145 F 100 CvUSE: 66 ft4b 80 P ~ ~

CI 60 tu Q.

E 40 Material: SA-533. Gr. B. Cl. 1 20 3,25x10'/cm~

"' "' """LE Fluence:

-100 100 200 300 400 500 600 Tomporaturo, F

~ /'

5-18 RAMATOME TCCHHOLOOICS

~

5'igure 5-15. Charpy Impact Data for Irradiated Base Metal Plate Heat No. B4197-2

~

Transverse (TL) Orientation 100

~e 75 L

50 u

25 co

-100 100 200 300 400 500 600 Tomporaturo, F 100 80 L

0

'III 60 C ~ ~

x 40 s ~

20 0

-100 100 200 300 400 500 600 Tomporaturo, F 120 55MEE 50 Too: +151 F 100 CvUSE: 55 ft4b 80 pc 60 uj O

CL E ~ ~ ~

40 Material: SA-533. Gr. B, Cl. 1 20 F(ueAoe: 3,25x10 h/em

-100 100 200 300 400 500 600 Tomporaturo, F F RAMATO ME 5-19 'I C C H H O 1 O O I E S

Figure 5-16. Charpy Impact Data for Irradiated Weld Metal Wire Heat 5P6771 / Flux I ot 0342 100

~4 75 50 IL 25 V)

-200 -100 0 100 200 300 400 500 600 Tomperaturo, F 100 80 C

O

'cn 60 x

uj 40 lo 20 0

-200 -100 0 100 200 300 '00 500 600 Tomporaturo, F 120 Ts5ul e'+80 F 100 SO Tgg.'+45 F CvUSE: 67 ft4b 80 4 ~

~

p 60 W

O CL E

40 Material: Weld Metal-Linde 124 20 Fluence: 3 25x10" n/cm~

~ ~ Heat Number: 5P6771 /0342

-200 <<100 0 100 200 300 400 500 600 Tomporaturo, F F RAMATOME 5-20 'V 0 C H 8 O LO 0 I @ S

Figure 5-17. Charpy Impact Data Irradiated Base Metal Plate Heat-Affected-Zone 100

~e 75 E

~ ~

50 u.

I 25 CO

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 100 E

80 C

0 60 x 40 ~ ~

uj 20 0

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 120 T55M<e +79 F 50 00 100 CvUSE: 66ft.lb 80 C7 4

Q) 60 ~ ~

UJ O

cL.

E 40 20 Material: SA-533, Gr. 8, Cl. 1 Fluence: 3,25x10'0 n/cm~

0

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 5-21 F RAMATOME TCCHHOLOOIC5

Figure 5-1S. Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate Heat No. B4197-2, Longitudinal (LT) Orientation 0

<<<<':

m))

t.<<<<r

)

Specimen No. QL49, Specimen No. QL48,

'<<"

) ++~+)r+

)'<<@~<

.'P~~ePr.

Specimen No. QI,50, Specimen No. QIANA>,

<<. 4<<

fr4,

,)Q)!<<re' P<<

Specinren No. QL51, Specimen No. QL56, mir<<rr

's<<'<<<<

<< 'ii I h

)'

<<h Specimen No. QIA7, Specimen No. QL60, 5-22 F 1

RAMATOME C C <<<<H O <<O 0 I C 5

4 1l p, 5-18 (Cont'd). Photographs of Charpy Impact Specimen Fracture Surfaces, J'igure Base Metal Plate Heat No. B4197-2, Longitudinal (LT) Orientation aqua Wi v

Specimen No. QI.52, Specimen No. QLSS, t '," <<

gl

-8"e -,

Specimen No. QI,57, Specimen No. QI,53, k~ -.<<'i .Qi Specimen No. QL58, Specimen No. QL59, Specimen No. QL54, 5-23 F

1 RAMATOME C C H HO<<OO I 4 $

Figure 5-19. Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate Heat No. B4197-2, Transverse (TL) Orientation A~I@ V 4 Specimen No. QT56, Specimen No. QT52,

$, .

~j Specimen No. QT55, Specimen No. QT57, t.

~ ~

Spcclttlctl Noe QT50, Speci nten No. QT60,

<'V< ' Ãw"<,<i '<'<4 Ew P

4w At

'I "g

Speci men No. QT49, Specltnen No. QT58, 5-24 F RAMATOME

<CCHHOIOOICS

Figure 5-19 (Cont'd). Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate Heat No. B4197-2, Transverse (TL) Orientation

'f

~ ~

~g"+jr~.pcf ~ I>>f5 I" Yr y $ f>> I

'<< '}I>> ..

, IHI*>>

Specimen No. QT4G, Specimen No. QT51, I

yaw

~ 'I<>>f. t

~Et I>>f>>p~ rfr

>>I I~

f'gg'pecimen No. Q'I'47, Speclnlell No pent.J

g4

~ki~wi Specimen No. QT54, Specimen No. QT59, Ir>>~f H+

II5My&'pecimen No. QT48, F RAMATOME 55CHHDLDOIE5 5-25

Figure 5-20. Photographs of Charpy Impact Specimen Fracture Surfaces, Weld Heat 5P6771 / Flux Lot 0342

<<*'.t~,6'$Q, ) '.1, ~,

Specimen No QW59, Specimen No. QW54,

) tt'f5. <<tj>~i

,p)

Specimen No. QW565 Specimen No. QWS2, r',Z 5r '5 II ~

tgt 1<< L" jt <<t .t Specimen No. QWS8, Specimen No. QW48, Specimen No. QW49, Specimen No. QW53, 5-26 F RAMATOME 55CHHOLOOI55

Figure 5-20 (Cont'd). Photographs of Charpy Impact Sp'ecimen Fracture Surfaces, Weld Heat 5P6771 / Flux Lot 0342 i(~.,-~B.i ~,e-i Specimen No. QW55, Specimen No. QW46, el-"~ .,

gA~i +m e!j>>

's

• e>-,5g,i

~(

e ge Specinxn No. QW57, Specimen No. QW50,

~+pi~

'(g

~;:tegj w,

~pP-.

>>Wee'.'4',

eA i~ft

~4- e Specinmn No. QW51, Specimen No. QW60,

~

gg;. ~ e y Specimen No. QW47, 5-27 F 1

RAMATOME C C H H O L O O'I 5 C

Figure 5-21. Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate, Heat-Affected-Zone Specimen No. QH51, Specimen No. QH51, f<p%. 1 gXj4',~@~

Spcclmen No. QH59, Specimen No. QH48, Specimen No. QH55, Specimen No. QH50, dyed+

Specimen No. QH56, Specimen No. QH58, F RAMATOME TCCHHOIOO 5-28 ~ ~ 5

Figure 5-21 (Cont'd). Photographs of Charpy Impact Specimen Fracture Surfaces, Base Metal Plate, Heat-Affected-Zone

- 5" Q.I Specin~n No. QH52, S~Ikcimen No. QH4V, 5 k- ~ I: -.~k=Qf 5

Specinmn No. QHCi0, Specimen No. QH46, Ii I I .&'+.I" .-ja Q Specinmn No. QHS4, Spcvinmn No. Qif53,

. ~ t *QW

+ ~

Specimen No. QH49, F RAMATO ME 5-29 55CHHOQOOI55

6. Neutron Fluence 6.1. Objectives and Background Over the last fourteen years, Framatome Technologies, Inc. (FTI) has developed a calculational-based fluence analysis methodology that can be used to accurately predict the fast neutron fluence in the reactor vessel using surveillance capsule dosimetry or cavity dosimetry (or both) to verify the fluence predictions."" The methodology was developed through a full-scale benchmark experiment that was performed at the Davis-Besse Unit 1 reactor."" The results of the benchmark experiment demonstrated that the accuracy of a fluence analysis that employs the FTI methodology would be unbiased and have a precision well within the NRC-suggested limit of 20%%u

""""

The FTI methodology was used to calculate the neutron fluence exposure to the pressure vessel, certain vessel welds in the beltline region, and surveillance Capsule X of the HNP reactor vessel.

he fast neutron fluences (E ) 1 MeV) at those points were calculated in accordance with the requirements of the U.S. NRC Draft Regulatory Guide DG-1053,"" as described in detail in the FTI fluence topical report, BAW-2241P, Revision 1.""

The energy-dependent flux at the capsule was used to determine the calculated activity of each dosimeter. The calculated activities were adjusted to account for known biases (photoflssion, plutonium build-in, and U-235 impurity in the U-238), and compared directly to the measured activities. It is noted that the measurements are not used in any way to determine the magnitude of the flux or the fluence. The measurements are used only to show that the calculational results are reasonable and to show that the HNP results are consistent with the FTI benchmark database

'f uncertainties.

F RAMATO ME

Explicit values of the fast fluence were computed for the following locations:

~ Surveillance Capsule

~ "Wetted" Surface for the HNP Reactor Vessel Beltline Materials

~ Clad-Base Metal Interface for the HNP Reactor Vessel Beltline Materials

~ Clad-Base Metal Interface Maximum Peak Location

~ ~/4 T, 'hT, ~/~ T, and Outside Surface at Maximum Peak Location where T is the reactor vessel thickness.

The multi-cycle-average full power flux at each of these locations was calculated for cycles 1 through 8, and the corresponding fluence at each location was then calculated by computing the product of each flux by the appropriate effective full power time, in seconds.

The calculated eight-cycle-average full power flux was used as the "extrapolation flux, " which is the flux used to project fluences for times beyond end-of-cycle 8 (EOC 8), reported in Tables 6-2 and 6-3.

6.2. Results The numerical and graphical results are presented in the following Tables and Figures:

~ Calculation to Measurement Ratios in Table 6-1,

~ Flux and fluence results at all points of interest in Tables 6-2 and 6-3, Locations of flux peaks on the various welds and plates in Table 6-4, '

Capsule Flux, Fluence, and Lead Factors in Table 6-5.

It was noted that all three dosimeter types showed an increasing trend in the measured activities from bottom to top. This trend was not observed in the calculated activities however, as they showed a peak in the middle for all dosimeters. Resolution of this discrepancy is not necessary since the uncertainty for each dosimeter, except for the Cu (top) is well within the limits established for the FTI fluence analysis methodologies."" Even though the validity of the Cu (top) dosimeter is in question, it is still included in the uncertainty evaluation for Capsule X.

F RAMATOME 6-2

Table 6-1. C/M Ratios for HNP Cap's'ule X Dosimeter C (pCi/g) M (pCi/g) C/M Fe (Top) 1429.2 1584 0.90 Fe (Mid) 1602.78 1528 1.05 Fe (Bot) 1566.57 1529 1.03 Ni (Top) 2003.12 2233 0.90 Ni (Mid) 2234.81 2122 1.05 Ni (bot) 2186.08 2071 1.06 Cu (Top) 8.36 11.11 0.75 Cu (mid) 9.472 10.65 0.89 Cu (Bot) 9.232 10.33 0.89 U-238 34.29 37.45 0.92 Avg. C/M = 0.944 6-3 F RA MAT0 M E

Table 6-2. Fast Neutron Fluence (E > 1 MeV) for the HNP Reactor Vessel "Wetted" Inside Surface Data for "Wetted" Inside Surface FLUENCE (n/cm') (Extrapolation Flux = EOC-8 flux)

Peak Flux Location Cyl-8 Avg EOC 8 ll EFPY 12 EFPY 14 EFPY 16 EFPY 18 EFPY (n/cm -s)

Weld AB 3.89148E+ 10 1.15979E+ 19 1.35086E+ 19 1.47367E+ 19 1.71928E+19 1.96489E+19 2.21050E+19 Weld BC 1.47860E+ 10 4.40673E+18 5.13271E+18 5.59932E+18 6.53254E+18 7.46576E+18 8.39899E+18 Weld BD 1.47860E+ 10 4.40673E+18 5.13271E+18 5.59932E+18 6.53254E+18 7.46576E+18 8.39899E+18 Weld BA 1.44110E+ 10 4.29497E+18 5.00254E+18 5.45731E+18 6.36687E+18 7.27642E+18 8.18597E+18 Weld BB 1.44110E+ 10 4.29497E+18 5.00254E+18 5.45731E+18 6.36687E+18 7.27642E+18 8.18597E+18 lnt Shell 4.05834E+ 10 1.20952E+19 1.40878E+19 1.53686E+19 1.79300E+19 2.04914E+19 2.30528E+19 Low Shell 3.96097E+ 10 1.18050E+19 1.37498E+19 1.49998E+19 1.74998E+19 1.99998E+19 2.24997E+19 FLUENCE (n/cm') (Extrapolation Flux = EOC-8 flux)

Peak Flux Location Cyl-8 Avg 20 EFPY 23 EFPY 25 EFPY 32 EFPY 36 EFPY (n/cm -s)

Weld AB 3.89148E+ 10 2.45611E+ 19 2.82453E+ 19 3.07014E+ 19 3.92978E+ 19 4.42101E+ 19 Weld BC 1.47860E+ 10 9.33221E+ 18 1.07320E+19 1.16653E+ 19 1.49315E+ 19 1.67980E+19 Weld BD 1.47860E+10 9.33221E+18 1.07320E+ 19 1.16653E+ 19 1.49315E+ 19 "'.67980E+19

.

Weld BA 1.44110E+ 10 9.09553E+ 18 1.04599E+ 19 1.13694E+19 1.45528E+ 19 1.63720E+ 19 Weld BB 1.44110E+ 10 9.09553E+ 18 1.04599E+ 19 1.13694E+ 19 1.45528E+ 19 1.63720E+ 19 Int Shell 4.05834E+10 2.56143E+ 19 2.94564E+ 19 3.20178E+ 19 4.09829E+ 19 4.61057E+ 19 Low Shell 3.96097E+10 2.49997E+ 19 2.87497E+ 19 3.12497E+19 3.99996E+19 19 I'.49995E+

6-4 F RAMATOME

Table 6-3. Fast Neutron Fluence (E ) I MeV) for the HNP Reactor Vessel Clad-Base Metal interface Data for Clad-Base Metal Interface FLUENCE (n/cd) (Extrapolation Flux = EOC-8 flux)

Peak Flux Cyl-8 Avg EOC 8 11 EFPY 12 EFPY 14 EFPY 16 EFPY 18EFI Y Location n/cm'-s Weld AB 3.83953E+ 10 1.14431E+19 1.33283E+19 1.45399E+19 1.69633E+19 1.93866E+ 19 2.18099E+ 19 Weld BC 1.46510E+10 4.36650E+18 5.08585E+18 5.54820E+18 6.47290E+18 7.39760E+ 18 8.32230E+ 18 Weld BD 1.46510E+10 4.36650E+18 5.08585E+18 5.54820E+ 18 6.47290E+18 7.39760E+18 8.32230E+ 18 Weld BA 1.43025E+ 10 4.26263E+18 4.96487E+18 5.41622E+ 18 6.31893E+ 18 7.22163E+ 18 8.12434E+18 Weld BB 1.43025E+10 4.26263E+18 4.96487E+18 5.41622E+ 18 6.31893E+ 18 7.22163E+ 18 8.12434E+ 18 Int Shell 4.00619E+10 1.19398E+19 1.39068E+19 1.51711E+19 1.76996E+19 2.02281E+ 19 2.27566E+ 19 Low Shell 3.90935E+ 10 1.16512E+19 1.35707E+19 1.48043E+19 1.72717E+19 1.97391E+ 19 2.22065E+ 19 I.S. Max 4.00619E+10 1.19398E+19 1.39068E+19 1.51711E+19 1.76996B+19 2.02281E+ 19 2.27566E+ 19 1/4T 2.32122B+10 6.91803E+18 8.05772E+18 8.79025E+18 1.02553E+19 1. 17203E+ 19 1.31854E+ 19 1/2T 1.18190E+10 3.52247E+18 4.10277E+18 4.47575E+18 5.22170E+18 5.96766E+ 18 6.71362E+ 18 3/4T 5.75046E+09 1.71383E+18 1.99618E+18 2.17765E+18 2.54059E+ 18 2.90353E+ 18 3.26647E+ 18 Outside 2.49588E+09 7.43858E+17 8.66403E+17 9.45167E+17 1.10269E+18 1.26022E+ 18 1.41775E+ 18 Surf FLUENCE (n/cm') (Extrapolation Flux = EOC-8 flux)

Peak Flux Cyl-8 Avg 20 EFPY 23 EFPY 25 EFPY 32 EFPY 36 EFPY Location n/an'-s)

Weld AB 3.83953E+10 2.42333E+ 19 2.78682E+19 3.02916E+19 3.87732E+19 4.36199E+19 Weld BC 1.465 10E+ 10 9.24700E+ 18 1.06341E+19 1.15588E+19 1.47952E+19 1.66446E+19 Weld BD 1A6510E+10 9.24700E+18 1.06341E+ 19 1.15588E+ 19 1.47952E+ 19 1.66446E+ 19 Weld BA 1.43025E+ 10 9.02705E+ 18 1.03811E+19 1. 12838E+19 1.44433E+19 1.62487E+19 Weld BB 1.43025E+ 10 9.02705E+ 18 1.03811E+19 1.12838E+19 1.44433E+19 1.62487E+19 Int Shell 4.00619E+ 10 2.52851E+ 19 2.90779B+ 19 3.16064B+19 4.04562E+19 4.55132E+19 Low Shell 3.90935E+ 10 2.46739E+ 19 2.83750B+19 3.08424B+19 3.94783E+19 4.44131E+19 I.S. Max 4.00619E+ 10 2.52851E+ 19 2.90779E+19 3.16064E+19 4.04562E+19'.55132E+19 1/4T 2.32122E+ 10 1.46504E+ 19 1.68480E+19 1.83130E+19 2.34407E+19 2.63708E+19 1/2T 1.18190E+ 10 7.45958E+ 18 8.57852E+ 18 9.32448E+ 18 1.19353E+19 1.34272E+19 3/4T 5.75046E+09 3.62941E+ 18 4.17382E+ 18 4.53677E+ 18 5.80706E+18 6.53294E+18 Outside 2.49588E+09 1.57528E+ 18 1.81157E+18 1.96910E+18 2.52045E+18 2.83550E+18 Surf

"'t should be noted that the EOL (32 EFPY) fluence calculated in this analysis is higher than the fluence calculated for the HNP Capsule V analysis."'his increase in fluence is due to the change in fluence alculational methodologies from the previous Capsule V analysis to the current analysis.

6-5 F RAMATOM E

Table 6-4. Peak Fast Fluence Locations of HNP Reactor Vessel Beltline Region Base Metals and Weld Metals Peak Location (cm)

Radial Azllllutllal Axial" Weld or Plate (cm) (degrees) (cnl) 199.7075 177.3238 BC & BD 199.7075 45 188.9515 BA & BB 199.7075 45 145.4585 Intermediate Plate 199.7075 188.9515 Lower Plate 199.7075 145.4585

'" Relative to the Lower Active Fuel elevation 6-6 F RAMATOME

Table 6-5. HNP Ca'psule X Fast Neutron Flux (E ) 1 MeV), Fast Neutron Fluence (E > 1 MeV), and Lead Factors at EOC 8 Exposure at EOC 8 3449.47 EFPD Capsule Average Neutron Flux (E > 1.0 MeV) 1.09 x 10" n/cm'-s Capsule Average Neutron Fluence (E > 1.0 MeV) 3.25x10" n/cm'.68t*>

Lead Factor (Capsules U, V, X to I.S. Maximum Location)

Lead Factor (Capsules U, V, X to ~/4T Location) 4.69t>

Lead Factor (Capsules W, Y, Z to I.S. Maximum Location) 2.38 "The lead factors changed from those calculated in the HNP Capsule V analysis due to the change in power distribution and minor changes in spectrum between the Capsule V analysis (cycles 2 - 3) and this analysis (cycles 1 - 8).

6-7 F RAMATOM f

7. Discussion of Capsule Results 7.1. Unirradiated Material Property Data The base metal and weld metal were selected for inclusion in the HNP surveillance program in accordance with the criteria in effect at the time the program was designed. The applicable selection criterion was based on the unirradiated properties only. A review of the original unirradiated material properties of the reactor vessel core beltline region materials indicated no significant deviation from expected properties except in the case of the upper-shelf energy properties of the base metal in the transverse (TL) orientation which was below the required minimum initial Charpy upper-shelf energy (CvUSE) of 75 ft-lbs. Based on the predicted end-of-service peak neutron fluence value at the '/4-thickness (i/4T) vessel wall location and the copper content of the base metal, it was predicted that the end-of-service CvUSE would not be below 50 ft-lbs.

he unirradiated mechanical properties for the HNP RVSP materials are summarized in Appendices B and C of this report.

7.2. Irradiated Property Data 7.2.1. Tensile Pro erties Table 7-1 compares the irradiated and unirradiated tensile properties. Review of the surveillance tensile test data indicates that the ultimate strength and yield strength changes in the base metal plate as a result of irradiation and the corresponding changes in ductility are within the ranges observed for similar irradiated materials. The changes in tensile properties for the surveillance weld metal, as a result of irradiation, are also within the observed ranges for similar irradiated materials.

The general behavior of the tensile properties as a function of neutron irradiation is an increase in both ultimate and yield strength and a decrease in ductility as measured by both total elongation and reduction in area. The most significant observation from these data is that the base metal with the transverse orientation exhibited slightly greater sensitivity to neutron irradiation than the weld metal. Conversely, the base metal properties in the longitudinal orientation were similar to those the weld metal.

F RAMATOME 7-1

7.2.2. Im act Pro erties Tables 7-2 and 7-3 compare the measured changes in irradiated Charpy V-notch impact properties from Capsule X with the predicted changes in accordance with Regulatory Guide 1.99, Revision n6]

2 The measured 30 ft-lb transition temperature shifts for the surveillance materials are greater than the values predicted using Regulatory Guide 1.99, Revision 2 except for the heat-affect-zone material. When the margin (2a) is added to these predicted shifts, the predicted 30 ft-lb transition temperature shift values are conservative for all the surveillance materials (see Table 7-2).

The measured percent decrease in CUSE for the measured surveillance base metal in the longitudinal (LT) orientation and the surveillance weld metal are slightly greater than the values predicted using Regulatory Guide 1.99, Revision 2 (0.7% and 5.5% respectively). However, the measured irradiated CvUSE values for these materials remains above the required 50 ft-lb limit.

The percent reduction in CUSE for the measured base metal surveillance data in the transverse (TL) orientation and the heat-affected-zone material are less than the Regulatory Guide 1.99, Revision 2 predictions.

The radiation-induced changes in toughness of all the HNP surveillance materials are summarized in Table 7-4. The original unirradiated and Capsule U and V Charpy impact data were evaluated based on hand-fit Charpy curves generated using engineering judgement. These data were re-plotted and re-evaluated herein using a hyperbolic tangent curve-fitting program to be consistent with the Capsule X Charpy curves and evaluations. The results of the re-evaluations are presented in Appendix C. In addition, Appendix D contains a comparison of the Charpy V-notch transition shift results for each surveillance material, hand-fit versus hyperbolic tangent curve-fit.

Comparisons of the unirradiated and irradiated Charpy V-notch impact curves are presented in Figures 7-1 through 7-4.

7.3. Reactor Vessel Fracture Toughness 7.3.1. Ad'usted Reference Tem erature Evaluation The adjusted reference temperatures for the HNP reactor vessel beltline region materials were calculated in accordance with Regulatory Guide 1.99, Revision 2 applicable to 11, 12, 14, 16, 18, 20, 23, 25, and 36 effective full power years (EFPY), and the results are presented in Tables 7-5 through 7-13. The evaluations were performed at the '/4-thickness ('/4 T) and '/4-thickness ('/4 T) wall location of each beltline material with chemistry factors determined from F RAMATOME 7-2

Tables 1 and 2 in Regulatory Guide 1.99, Revision 2, and determined from surveillance data if available. Based on these results, the controlling beltline material for the HNP reactor vessel is the intermediate shell plate, heat no. B4197-2.

7.3.2. Decrease in U er-Shelf Ener Evaluation An evaluation of the reactor vessel end-of-life (36 EFPY) upper-shelf energy at the ~/~ T wall location for the HNP reactor vessel beltline materials is presented in Table 7-14. The radiation induced reduction in upper-shelf energy was performed using the guidelines in Regulatory Guide 1.99, Revision 2. The HNP reactor vessel beltline material with the lowest predicted upper-shelf energy is the intermediate shell plate, heat no. B4197-2, however, the predicted value for this material will not fall below the required 50 ft-lb limit.

7.3.3. Pressurized Thermal Shock Evaluation A pressurized thermal shock (PTS) evaluation for the HNP reactor vessel beltline materials was performed in accordance with Code of Federal Regulation, Title 10, Part 50.61 (10 CFR 50.61),"" and the results are shown in Table 7-15. The results of the PTS evaluation demonstrate that the HNP reactor vessel beltline materials will not exceed the PTS screening criteria before end-of-life (36 EFPY). The controlling beltline material for the HNP reactor essel with respect to PTS is the intermediate shell plate, heat no. B4197-2 with a RTPTs value of 196.1'F which is well below the PTS screening criterion of 270'F.

7-3 F RAMATOME

Table 7-1. Summary Of HNP Reactor Vessel Surveillance Capsules Tensile Test Results Stren th, ksi Ductilit, %

Fluence, Test Total Reduction Material 10)~ g/cm~ Tem .,F Ultimate  %(4 Yield Elon . of Area  %(e Base Metal Plate, 0.00 75 94 0>> 7S.S>> 26.O>> 61.S>>

B4197-2 300 87.5>> 74.5>> 2LS>> 63.5>>

(Longitudinal) 600 88.0>> 73.S>> 19.o>> 54.0>>

0.552 70 96.7>> +2.9 72.4>> -4.1 23.3>> -10.4 62.6") +1.8 550 93.6 +6.4 65.7 + 10.6 19.3 +1.6 52.7 -2.4 1.32 70 97.7>> +3.9 73.9>> -2.1 22.1>> -15.0 62.5>> +1.6 550 93.8 +6.6 69.0 -6.1 15.9 -16.3 42.4 -21.5 70 101.2 +7.7 79.6 +5.4 22.9 -11.9 60.2 -2.1 300 +5.8 71.7 -3.8 20.3 -5.6 60.3 -5.0 550 99.0 +12.5 72.8 -1.0 17.0 -10.5 43.8 -18.9 Base Metal Plate, 0.00 75 91.0>> 68.5>> 26.0>> 61.S>>

B4197-2 300 83.O>> 61.0>> 23.O>> 58.5")

(Trans vetse) 600 86.5>> 60.0>> 2O.S>> s4.5>>

0.552 70 96.6>> +6.2 72.3>> +5.5 23.7>> -8.8 S7.S>> -6.5 550 93.7 +8.3 69.3 +15.5 17.4 -15.1 50.4 -7.5 1.32 70 97.4>> +7.0 76.0>> + 10.9 21.7>> -16.5 S8.4>> -5.0 550 92.7 +7.2 66.6 +11.0 17.3 -15.6 44.2 -18.9 70 101.0 +11.0 80.0 +16.8 22.5 -13.5 60.1 -2.3 300 92.8 +11.8 72.6 +19.0 19.4 -15.7 55.6 -5.0 550 95.8 + 10.8 70.4 +17.3 17.8 -13.2 53.2 -2.4

" Change relative to unirradiated material property.

>> Mean value of available test data.

z3 Op 0

o0 3

Table 7-1. (cont.) Summary of HNP Reactor essel Surveillance Capsules Tensile Test Results Stren th, ksi Ductilit, %

Fluence, Test Total Reduction Material 1O" n/cm'.00 Tem .,F Ultimate Yield  %(0 Elon .  %(a) of Area  %(4 Weld Metal 75 89.0>> 74.O>> 27.5(') 67.5>>

(Wire Ht. 5P6771 / 300 Sl.O>> 6s.o>> 23.5>> 66.O(')

Flux Lot 0342) 600 S3.5>> 63.0>> 22.5(') 64.5>>

0.552 70 91.1>> +2.4 76.3>> +3.1 25.8>> -6.2 67.1>> 4.6 550 87.5 +4.8 69.0 +9.5 20.9 -7.1 62.5 -3.1 1.32 70 91.8>> +3.1 76.0>> +2.7 22.5>> -18.2 61.9") -8.3 550 88.3 +5.7 70.3 +11.6 18.6 -17.3 61.6 -4.5 70 95.4 +7.2 82.8 +11.9 24.8 -9.8 63.0 -6.7 300 '8.2

+8.9 76.0 +16.9 22.3 -5.1 64.5 -2.3 550 93.0 +11.4 75.6 +20.0 21.1 -6.2 61.1 -5.3

'" Change relative to unirradiated material property.

>> Mean value of available test data.

zg Op o>>

o0

<<Pl

Table 7-2. Measured vs. Predicted 30 ft-lb Transition Temperature Changes for HNP Capsule X Surveillance Materials 3.25 x 10" n/cm Measured 30 ft-lb Transition 30 ft-lb Transition Temperature Shift Predicted in Tem rature, F Accordance With Re lato Guide 1.99, Revision2 Chemistry Material Unirradiated Irradiated Difference Factor" Margin (2') AT~ + 2cr Base Metal Plate, B4197-2, 51 145 94 55.9+'3.2 34 107.2 Longitudinal (L'I)

Base Metal Plate, B4197-2, 72 151 79 73.2 34 107.2 Transverse (TL)

Weld Metal 34 45 79 32.6>> 42.7 56 98.7 (Wire Ht. 5P6771 /

Flux Lot 0342)

Base Metal Plate, B4197-2, 24 68 73.2 34 107.2 Heat-Affect-Zone Material

'*'hemistry factor determined using Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

"'hemistry factor based on mean copper and nickel contents calculated using data in Tables 3-2 and 3-3.

zg Op 0

o0

Table 7-3. Measured vs. Predicted Upper-Shelf Energy Decreases for HNP Capsule X Surveillance Materials 3.25 x 10"

% Decrease Predicted Measured U r-Shelf Ener, ft-lb n/cm'aterial In Accordance With Regulatory Guide 1.99, Rev. 2 Unirradiated Irradiated  % Decrease Fi ure2 Base Metal Plate, B4197-2, 87 66 24.1 23 4<<

Longitudinal (LT)

Base Metal Plate, B4197-2, 70 55 21.4 23.4<<

Transverse (TL)

Weld Metal 67 27.2 21 7<<

(Wire Ht. 5P6771 /

Flux Lot 0342)

Base Metal Plate, B4197-2, 85 22.4 23 4<<

Heat-Affect-Zone Material

'" Based on mean copper content calculated using data in Tables 3-2 and 3-3.

zg Op 0

Oo0

Table 7-4. Summary of HNP Reactor Vessel Surveillance Capsules Charpy Impact Test Results Measured Measured Transition Tem rature U r-Shelf Fluence, DCv30, hCv50, Energy, Material Ca sule F F ft-Ib  % Decrease Base Metal Plate, B4197-2, Baseline n/cm'.52E+18 87 Longitudinal (LT) 30 37 -5.7 1.32E+19 43 53 3.4 3.25E+19 93 66 24.1 Base Metal Plate, B4197-2, Baseline 70 Transverse (TL)

U 5.52E+18 38 33 2.9 1.32E+19 35 47 65 7.1 X 3.25E+19 79 55 21.4 OO Weld Metal Baseline (Wire Ht. SP6771 /

5.52E+18 20 29 83 9.8 Flux Lot 0342) 1.32E+19 18 27 10.9 3.25E+19 79 94 67 27.2 Base Metal Plate, B4197-2, Baseline 85 Heat-Affected-Zone Material U 5.52E+18 59 76 10.6 1.32E+19 53 57 80 5.9 3.25E+19 68 87 66 22.4 zg Oy 0

oo

Table 7-5. Evaluation of Adjusted eference Temperatures for the HNP Reactor Vessel Applicable to 11 EFPY

)daterial Deter(prior/"

Chc mimi Composition'" I I EFPY Fluence, n/cm'RTsur, at II F

EFPY hlsrgin ART, F at I I EFPY Rmctor Vessel Mad. Heat lnithl Imide T/4 3/4T T/4 T/4 3/4T T/4 3/4T Beldinc Region Location ident. Number Type RTsurm Surface Location Location~ lstcathn Location Istcation Location Location Reguhtory Guide 1.99. Rcvirion 2, PcsiYion I. I intermediate Shell Phte A91$ 3-I A9153.1 S*-$33 Gr. Bl 0.09 0.46 +60 58.0 I.4 IE+19 8.59E+ IS 3.39E+18 55.6 40.7 34.0 34.0 149.6 134.7 Intermediate Shell Phtc B4 197-2 B4197.2 SA-$ 33 Gr. Bl 0.09 0.$ 0 +91 58.0 1.4 I E+19 8.59E+ IS 339E+ I 8 5$ .6 40.7 34.0 34.0 180.6 165.7 Loner Sbcll Phtc C9924-I C9924.1 SA433 Gr. Bl 0.08 0.47 +54 51.0 I.37E+19 S.35 E+ IS 3.29E+18 48.4 35.4 34.0 34.0 136.4 123.4 Loner Sbeil Phte C9924.2 C9924 2 SA.533 Gr. Bl O.OS 0.47 +S7 51.0 1.37E+19 8.35E+18 3.29E+18 48.4 34.0 34.0 139.4 126.4 IS Longit. W<<lds (Both 100%) BC/BD 4P4784 ASA/Linda 124 0.0$ 0.91 .20 68.0 $ .13E+ IS 3.13E+18 1.23E+ IS 46.3 31.3 463 31.3 72.6 42.6 IS m LS Circ. Weld (100%) AB 5P6771 ASA/Lindc !24 0.03 0.94 -20 41.0 1.35E+19 8.23E+ 18 3.25E+ 18 38.7 28.3 38.7 28.3 57.4 36.6 -,

LS Longe, WeMs (Bodt (00%) BA/BB 4P4784 ASA/Unde 124 0.0$ 0.91 -20 68.0 $ .00E+ IS 3 05 a+18 1.20E+18 45.8 30.9 45.8 71.6 41.8

/ r+

Reguhtory Guide 1.99. Revhbn 2. Position 2.1 Intcrmednse Shell Phte B4197-2 B4197-2 SA-533 Gr. BI 0.09 030 +91 51.4 1.4 IE+19 839E+18 3.39E+18 49.2 tl6l.lt

( 174.21 IS to LS Grc. WeM (100%) AB 5P6771 AS A/Unde 124 0.03 0.94 -20 49.1 1.35E+19 8.23E+18 3.25 E+18 46.4 33.9 28.0 '28.0 41.9 s

"See Appendix A.

+'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The 'AT and 3AT location fluence values are determined by calculating the '/4T and 3AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for 'AT = 2.0625 in. and "x" for 3AT = 5.9375 in.).

" Since two of the six surveillance data points are not credible, a full margin value is used to calculate the '/4 T and 3AT ART values.

[] - Controlling values of the adjusted reference temperatures.

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Table 7-6. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 12 EFPY Chemical dRTsur. F ART. F huter ial Descr iptiorP Compos itice/u 12 EFPY Fluence, n/em at 12 EFPY htarSin a[12 EFPY Rmetor Vessel htatt. I frat lnivial Che many lrnidc T/4 3/4T T/4 3/4T T/4 3/4T T/4 Belt! ine ReSion Location ident. Ãumhcr Type RTsurw Factor Sur/ace Location Locatioa Locathut Location Location Location Location RcSutatory Guide 1.99. Revision 2, Position I. I

!mcnnedia>> Shell Plate A91$ 3.1 A91S3 I SA-S33 Gr. Bl 0.09 0.46 +60 58.0 I.S4E+19 9.39E+18 3.1OE+ I 8 57.0 42.1 34.0 34.0 1$ 1.0 136.1 intermediate Shell Plate B4 197-2 B4 197-2 SA433 Gr. Bl 0.09 0.50 +91 58.0 1.54Ef 19 9.39E+18 3.70E+18 57.0 42.1 34.0 34.0 182.0 167.1 Lower Shell Phte C9924.1 C9924-I SA-$ 33 Gr. Bl 0.08 0.47 +54 51.0 I.SOE+19 9.14E+18 3.6IE+ IS 49:I 36.6 34.0 34.0 137.7 124.6 Leam Shell Phte C9924 2 C9924.2 SA-$ 33 Gr. Bl 0.08 0.47 +57 51.0 I.SOE+19 9.14E+18 3.61E+ IS 49.1 36.6 34.0 34.0 140.1 127.6 IS LonSit. Wchh (Both 100%) BC/BD 4 F4784 ASA/Lindc 124 0.0$ 0.91 68.0 5.60E+18 3.41E+18 1.3$ E+ IS 47.9 32.6 47.9 32.6 75.8 4$ .2 IS to LS Circ. Weld (100%) AB 5P6771 ASA/Liude 124 0.03 0.94 41.0 IA7E+ 19 S.96E+18 3.$ 4E+18 39.1 29.2 39.1 29.2 $ 9.4 38A LS LonSit. Welds (Both 100%) BA/BB 4P4784 AS A/Linde 124 0.0$ 0.91 68.0 SA6E+18 3.33E+18 1.3IE+18 47.4 32.2 47.4 32.2 74.8 44.4 ReSulatoty Guide 1.99. Revirion 2. Pcchion 2.1 lmermediatc Shell Plate B4197-2 B4197-2 SA433 Gr. Bl 0.09 0.50 +91 51.4 ICE+19 9.39E+ IS 3.70E+18 37.3 (175.5) tl62.3)

IS m LS Cire. WcM(100%) AB SP6771 ASA/Unde 124 0.03 0.94 -20 49.1 1.47E+19 8.96E+ IS 3.54E+18 47.6 35.0 2tLO 28.0 5$ .6 43.0

(" See Appendix A.

+) Calculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface flucnce is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The 'AT and SAT location fluence values are determined by calculating the IA4 T and AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for 'AT = 2.0625 in. and "x" for 3AT = 5.9375 in.).

"Since two of the six surveillance data points are not credible, a full margin value is used to calculate the IAT and 3AT ART values.

[ ] - Controlling values of the adjusted reference temperatures.

ug op o0

Table 7-7. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 14 EFPY Chetnhni bRTvrn, F ART, F huterial Description'e Composition'~ 14 EFPY Fluence n/etn at 14 EFPY htargin at 14 6'FPY Reaaor Vessel htad. Iieet Ni initial Chetnhny Inside T/4 T/4 3/4T T/4 3/4T Betdinc Region Location ldete. Iturnher Type RT Factor Sor/xe Lecaten Locathsn Location Location Location Regulatory Guide 1.99, Revhion 2, Penhion I. I Iraerrnediatc Shell Plate A9153-I A9153-I SA-533 Gt. Bl OA8 0.46 +60 58.0 1.79E+19 I ASE+ 19 4.316+18 59.4 44.4 34.0 34.0 153.4 138.4 Itnerntediate Shell Plate B4 197-2 B4197-2 SA433 Gr. Bl 0.09 0.50 +91 58.0 1.79E+19 1.096+19 43 IE+18 59.4 44.4 34.0 34.0 184.4 169.4 Ltraer Shel I Phtc C9924 I C9924.1 SA433 Gr. Bl 0.08 0.47 +$ 4 51.0 1.75E+19 1.076+19 4.216+18 51.9 38.8 34.0 34.0 139.9 126.8 Lower Shell Plate C9924 2 C9924-2 SA433 Gr. Bl O.OS 0.47 +57 SI.O 1.75E+19 I.DIE+19 4.216+18 51.9 38.8 34.0 34.0 142.9 129.8 IS Longit. WcMs (Bah 100%) BC/BD 4P4784 ASA/Lindc 124 O.OS 0.91 6S.O 6.53E+18 3.98E+18 I.S76+ lg S0.7 34.9 S0.7 34.9 81.4 49.8 IS to LS Cire. WcM (100%) AB SP6771 AS A/Lindc 124 0.03 0.94 41.0 1.72E+19 1.056+19 4.146+18 41.5 31.0 41.5 31.0 63.0 ~42.0 LS Longit. Welds (Both 100%) BA/BB 4P4784 ASA/Unde 124 0.05 0.91 6S.O 6376+ IS 3.8SE+18 I.S3E+18 50.2 348 50.2 349 80.4 49.0 Reguiaasty GuMe I 99. Revision 2. Position 2.1 Itrmntediate Shell Plate B4197-2 B4197-2 SA433 Gr. Bl OA8 0.50 +91 51,4 1.79E+19 I ASE+19 4.31E+18 S2.6 39.4 34.0'o [177.6} tl64.4)

IS to LS Circ. Weld (100%) AB SP6771 ASA/Lingo 124 0.03 0.94 49.1 1.72E+19 I.OSE+19 4.14E+ IS 49.7 37.1 28.0 28.0 57.7 45.1

"'ee Appendix A.

+) Calculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The IAT and 3AT location fluence values are determined by calculating the I/4T and SAT depth into the vessel and adding the minimum cladding thick)ess (i.e., "x" for /4T = 2.0625 in. and "x" for 3AT = 5.9375 in.).

" Since two of the six surveillance data points are not credible, a full margin value is used to calculate the I/4T and 3/4 T ART values.

[ ] - Controlling values of the adjusted reference temperatures.

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Table 7-8. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 16 EFPY hlaterial Descdptior/w 16 EFPY Floenee.

n/cm'RTMvr. F at 16 EFPY hfargin ART, F at 16 EFPY Reactor Vessel hhd. liest Ni Inithl Chetnistry hnide T/4 3/4T TI4 3/4T TI4 T/4 Bcldinc Region Lccation Ident. Nnnther wt% RTMvr'w Factor Sor/ace~ Locatior/" Ltcatiod Location Location Locatiea Location Regohtoty GoMc 1.99. Rcvhioa 2. PosiYion I. I Inter mediate Shdl Phte A9153.1 A9 153-I SA433 Gr. Bl 0.09 0.46 +60 58.0 2 05E+19 1.25E+19 4.93E+18 61.6 46.6 34.0 34.0 155.6 140.6 Intcttnedhte Shell Phte B4191-2 B4197.2 SAS33 Gr. Bl 0.09 0.50 +91 58.0 2.05E+19 1.25E+19 4.50E+ IS 61.6 46.6 34.0 34.0 186.6 171.6 Lower Shell Phte C9924.1 C9924.1 SA433 Gr. BI 0.08 0.47 +54 51.0 2.COB+19 1.22E+19 4.8 1 E+18 53.8 40,6 34.0 34.0 141.8 128.6 Los er Shdl Phte C9924.2 C9924.2 SA 533 Gr. Bl 0.08 0.47 +5 I 51.0 2.00E+19 1.22E+19 4.8 IE+18 53.8 40.6 34.0 34.0 144.8 131.6 IS Longh. WeMs (Both 100%) BC/BD 4P4784 ASA/Linda 124 0.05 0.91 68.0 7.41E+18 4.55E+Ig 1.80E+18 53.1 37.0 53.1 31.0 86.2 54.0 IS to LS Circ. WeM II00%) AB SP67/I ASA/Unde 124 0.03 0.94 41.0 1.96E+19 1.19E+19 4.1IE+18 43.1 32.4 43.1 32A 66.2 44.8 LS Longe. Welds (Both 100%) BA/BB 4P4784 AS A/Unde 124 0.05 0.91 6S.O 7.28E+ IS 4A4E+ lg 1.7SE+18 52.6 36,6 52.6 36.6 85.2 53.2 Regnhtory Goide 1.99, R<<v Yiioa 2. Poshiat 2.1 lntcrrnediate Shell Plate B4197-2 Bt 1 97-2 SA433 Gr. Bl 0.09 0.50 t91 51.4 2.05E+19 1.25E+19 4.93E+18 41.3 t 179.6) t166.3)

IS to LS Circ. WeM I100%) AB 5P6771 ASA/Unde 124 0.03 0.94 -20 1.96E+19 1.19E+ 19 4.71E+18 51.6 38,8 28.0 28.0 59.6 46.8 t'ee Appendix A.

"'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The I/4 T and 3AT location fluence values are determined by calculating the AT and 3AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for 'AT = 2.0625 in. and "x" for 3/4 T = 5.9375 in.).

"'ince two of the six surveillance data points are not credible, a full margin value is used to calculate the I/4 T and 3/4 T ART values.

[] - Controlling values of the adjusted reference temperatures.

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oo

Table 7-9. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 1S EFPY h'Iaterial Descriptiot(

Chemical Composutorf IS EFPY Flounce.

n/cm'Taur, F at 18 EFPY htargin ART. F at 18 EFPY Reactor Vessel hla. Heat Ni Inhial Inside T/4 3/4T T/4 3/4T T/4 T/4 3/4T Be! dine Region Location Ident. Number Type ws% RT Surface~ Locational Location~ Location Location Location Location Location Rcgulamry Guide 1.99. Rev Yiion 2, Fruition I. I Itzertncdiate Shell Piste A91$ 3 I A9153-I SA.533 Gr. Bl 0.09 0.46 +60 58.0 2.31E+19 1.4 I E+19 S.$ 6E+ IS 63.$ 48.5 34.0 34.0 151.5 142.$

Intermediate Shell Plate B4 197-2 B4197-2 S*.S33 Gr. Bl 0.09 OSO +91 58.0 2.3 IE+19 1.4 I 6+19 596E+18 63.5 48.5 34.0 34.0 ISS.S 173.$

Lower Shell Plate C9924 I C99ht-I SA.533 Gr. BI 0.47 +54 2.25E+19 O.OS $ 1.0 1.37E+19 SA IE+ I 8 $ $ .$ 42.2 34.0 34.0 143.$ 130.2 Lower Shell Plate C9924.2 C9924-2 SA $ 33 Gr. Bl 0.47 +57 O.OS 51.0 2.25E+19 1.37E+19 S.4 IE+ IS 55.5 42.2 34.0 34.0 146.$ 133.2 IS Longit. Welds (Both 100%) BCIBD 4P47SI ASA/Linde 124 0.0$ 0.91 -20 68.0 8.40E+18 S.12E+18 2.02E+IS 55.3 38.9 $ 5.3 38.9 90.6 51.8 IS m LS Cire. WcM (100%) AB 5P6771 ASA/Linda 124 0.03 0.94 -20 41.0 2.2 IE+19 1.35E+19 S.32E+18 44.4 33.1 44A 33.7 68.8 47.4 LS Longn. WcMs (Both 100%) BA/BB 4P4784 AS A/Linda 124 0.0$ 0.91 -20 68.0 8.19E+18 4.99E+18 1.97E+ I 8 54.8 38.$ 54.8 383 89.6 w51.0 Regulatory Guide 1.99. ReviYion 2, Poshion 2.1

,YF Intcrrnediate Shell Plate B4197.2 B4197.2 SA.533 Gr. B I 0.09 +91 51.4 2.31E+19 1.4 IE+19 S.56E+18 $ 6.3 43.0 t181.3I [168.0}

IS to LS Cite. WeM (100%) AB 5P6771 ASA/Unde 124 0.03 -20 49.1 2.21E+19 1.3$ E+19 S.32E+18 53.2 40A 28.0 28.0 61.2 48.4 "See Appendix A.

"'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The 'AT and 3AT location fluence values are determined by calculating the AT and 3AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for IAT = 2.0625 in. and "x" for 3AT = 5.9375 in.).

" Since two of the six surveillance data points are not credible, a full margin value is used to calculate the IA4 T and SAT ART values.

[ ] - Controlling values of the adjusted reference temperatures.

ve zg Op o~

oO v Itt

Table 7-10. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Uessel Applicable to 20 EFPY Cbetuical ARTsur. F ART. F huterial Descriptusn'o Cotuposition'e 20 EFP Y flutte n/catt at 20 EFPY h(ar8in at 20 EFPY Reaetee Vessel Marl. liest inside T/4 3/4T T/4 3/4T T/4 3/4T T/4 Be!dine Re8bsn Location Ident. Nutnber Type Sor/ace~ Locatiod Location~ Loeatusn Location Location Locatgn Location Re8utatoty GuMc 1.99. Revisbsn 2. PosiYion I. I Iotertoediate Shell Plate A91$ 3.1 A91$ 3.1 SA-533 Gr. Bl 0.09 0.46 +60 58.0 2.56E+19 I86E+19 6.16E+18 65.1 $ 0.1 34.0 34.0 159.1 144.1 lutertuedhte Shell Plate B4 197-2 B4 197-2 SA 533 Gr. Bl 0.09 0.50 +91 SS.O 2.56E+ 19 1.56E+19 6. 16E+18 6$ .1 $ 0.1 34.0 34.0 190.1 17$ .1 Loner Shell Plate C9924.1 C9924 I SA433 Gr. Bl 0.08 0.47 +54 $ 1.0 2.50E+ 19 1.$ 2E+19 6.0 IE+18 $ 7.0 43.8 34.0 34.0 14$ .0 131.8 Lean Sbeii Ph:e C9924 2 C9924.2 SA.533 Gr. Bl O.QS 0.47 +57 $ 1.0 2.50E+19 1.52E+19 6.0 IE+ IS 57.0 43.8 34.0 34.0 148.0 O4.8 IS Loe8h. WeMs(Both 100%) BC/BD 4P4784 ASA/Unde 124 0.0$ 0.91 68.0 9.33E+18 5.69E+18 2.24E+18 $ 7.3 40.6 56.0 40.6 93.3 61.2 IS to LS Circ. WeM (100%) AB SP6771 ASA/Linda 124 0.03 0.94 41.0 2.46E+19 I.SOE+19 5.92E+I 8 4$ .6 3$ .0 45.6 3$ .0 71.2 $ 0.0 LS Lurch WeMs (Both 100%) BA/BB 4P4784 ASA/Linda 124 0.05 0.91 68.0 9.10E+18 SDSE+ IS 2.19E+ IS $ 6.8 40.2 56.0 40.2 92.8 60.4 ReSuiasety GuMe 1.99. Revaion 2, position 2. i lntertnediasc Shell Plate B4197-2 B4197.2 SA433 Gr. Bl 0.09 0.50 +91 514 2.56E+ 19 1.56E+19 6.16E+18 57.7 44.4 (182.7} ( 169.4)

IS to LS Cire. WeM (100%) AB 5P6771 AS A/Linda 124 0.03 0.94 49.1 2.46E+19 1.50E+19 $ .92E+ IS 41.9 28.0 2RO 49.9

(') See Appendix A.

+) Calculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The IA4T and 3AT location fluence values are deterlnined by calculating the IAT and 3AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for IA4T = 2.0625 in. and "x" for SA4 T = 5.9375 in.).

') Since two of the six surveillance data points are not credible, a full margin value is used to calculate the IAT and 3AT ART values.

f ] - Controlling values of the adjusted reference temperatures.

Table 7-11. Evaluation of Adjuste eference Temperatures for the HNP Reactor Vessel Applicable to 23 EFPY Chemical bRTaut, F ART, F Material Dcseriptiouf ComposiYoty 23 EFPY Fluence, n/cmt at 23 EFPY h'll/8ln at 23 EFPY Reactor Vessel Matt. ucm Ni IniYial Inside T/4 T/4 3/4T T/4 3/4T T/4 3/4T Bc!dine Rc8ion Location Ident. Number Type wt% RTsuim Surfxe Location Location Location Location Location Location Location 44 Re8uhtory Guide 1.99. Revision 2, PosiYion I. I intermediate Shell Phtc A9 153-I A9153.1 SA.$ 33 Gr. Bl OA8 0.46 +60 58.0 2.95E+19 I.SOE+19 1.09E+18 67.3 52.4 34.0 34.0 161.3 146.4 lmermediate Sbcil Phtc B4197.2 B4 197-2 SA633 Gr. Bl OA8 0.50 +91 $ 8.0 2.95E+19 I.SOE+19 7.09E+18 67.3 $ 2.4 34.0 34.0 192.3 I77.4 Lower Shell Phte C9924.1 C9924.1 SA433 Gr. Bl 0.08 0.47 +$ 4 51.0 2.87E+19 1.75E+19 6.90E+18 $ 8.9 45.7 34.0 34.0 146.9 133.1 Lower Sbell Phte C9924-2 C9924-2 SA.$ 33 Gr. Bl 0.08 0.47 +$ 7 $ 1.0 2.87E+19 1.75E+19 6.90E+18 $ 8.9 45.1 34.0 34.0 149.9 136.7 IS Lon8it. WcMs (Both 100%) BC/BD 4P4784 ASA/Lindc 124 0.0$ 0.91 68.0 1.07E+19 6S2E+ IS 2.57E+18 $ 9.8 42.9 $ 6.0 42.9 9$ .8 6$ .8 IS to LS Circ. WeM (100%) AB SP6771 ASA/Unde 124 OAD 0.94 41.0 2.82Ee 19 I 72E+19 6.78E+18 47.1 36.5 47.1 36.5 74.2 53.0 IS Lon8it. WeMs (Botb 100%) BA/BB 4P4784 ASA/Lhde 124 0.0$ 0.91 6S.O 1.05 a+19 6.40E+18 2S3E+ IS $ 9.$ 42.6 $ 6.0 42.6 95S 6$ .2 Reguhtory Guide 1.99. ReviYion 2. Position 2.1 J y

Intermediate Shell Phtc B4197.2 Be(97-2 SA-$ 33 Gr. Bl 0.09 +91 $ 1.4 2.95E+19 I.SOE+19 7.09E+ IS 59.7 46S 34.tyu [184.7) [171.5)

IS m LS Circ. WcM (I00%) AB SP6711 ASA/Unde 124 0.94 -20 49.1 2.&2E+19 1.12E+19 6.78E+18 56.4 43.7 28.0 28.0 51.7 "See Appendix A.

~'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The 'AT and 3AT location fluence values are determined by calculating the I/4 T and 3AT depth into the vessel and adding the minimum cladding thickness (i.e., "x" for '/4 T = 2.0625 in. and "x" for 3AT = 5.9375 in.).

') Since two of the six surveillance data points are not credible, a full margin value is used to calculate the 'AT and 'AT ART values

[] - Controlling values of the adjusted reference temperatures.

z3 op o+

oo

Table 7-12. Evaluation of Adjusted Reference Temperatures for the HNP Reactor Vessel Applicable to 25 EFPY Chemical SRTsur. F ART, F Material Description'" Compos hhut 2S EFPY Fluence, n/emt at 25 EFPY Margin at 2S EFPY Ractcc Vessel bud. Heat Initbl (snide T/4 3/4T T/4 T/4 3/4T T/4 3/4T Behiioc Rcghut Location Ment. Number Type Sur/ace~

RYE~ Locatioa Loation Locathut Location Location Location Location Rcgubtory Guide 1.99, Revtshsn2. Position I.I Intermediate Shell Plam A9 IS 3-I A9153-I SA-533 Gr. BI 0.09 0.46 +60 58.0 3.206+19 Shell Phtc 1.95E+19 1.706+18 6L6 53.8 34.0 34.0 162.6 147.8 Intermediate B4197-2 Bc(97.2 SA.533 Gr. Bl 0.09 0.50 +91 58.0 3.206+19 1.956+19 7.706+18 68.6 53.8 34.0 34.0 193.6 178.8 Lower Shell Plate C9924.1 C9924 I SA433 Gr. Bl 0.08 0.47 +54 51.0 3.12E+19 1.906+19 WOE+ IS 60.0 46.9 34.0 34.0 148.0 134.9 Lower Shell Phtc C9924.2 C9924-2 SA-S33 Gr. Bl 0.08 0.47 +57 51.0 3.126+19 1.906+19 7.5OE+ IS 60.0 46.9 34.0 34.0 151.0 137.9 IS Lough. WcMs (Both 100') BC/BD 4P4784 ASA/(Jude 124 O.OS 0.91 20 68.0

~ 1.17E+19 7.13E+18 2.81E+18 61.5 44.S 56.0 44.S 97.5 69.0 IS to LS Circ. WeM (100%) AB SP6771 ASA/L(nde 124 0.03 0.94 -20 41.0 3.076+19 1.87E+ 19 7.386+18 48.1 37.5 48.1 37.S 76.2 55.0 LS Loogit. We1ds (Bah 100%) BA/BB 4P47SI ASA/Lindc 124 0.05 0.91 .20 6S.O 1.14E+19 6.95E+ I 8 2.74E+ IS 61.1 44.0 56.0 44.0 97.1 68.0 Regulatory Guide 1.99. ReviYion 2. Position 2.1 lmermediate Shell P(ate B4197.2 B4197-2 SA-533 Gr. Bl 0.09 0.50 +91 SI 4 3.206+19 1.95E+19 7.706+18 60.8 47.6 34.(yu (185.8) (172.6)

IS to LS Cire. WcM (l00%) AB 5P6771 ASA/Linde 124 0.03 0.94 49.1 3.076+19 1.876+ 19 7.386+ 18 573 44.9 28.0 28.0 52.9 "See Appendix A.

"'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The I/4 T and 3/4 T location fluence values are determined by calculating the I/4 T and 3/4 T depth into the vessel and adding the minimum cladding thickness (i.e., "x" for '/4 T = 2.0625 in. and "x" for 3/4 T = 5.9375 in.).

" Since two of the six surveillance data points are not credible, a full margin value is used to calculate the I/4T and 3/4 T ART values f ] - Controlling values of the adjusted reference temperatures.

zg op OO

Table 7-13. Evaluation of Adjuste eference Temperatures for the HNP Reactor Vessel Applicable to 36 EFPY Chemical dRTsur. F ART. F Mataial Descripten'" Compos(tet(u 36 EFPY F/uence n/cm at 36 EFPY hlargln at 36 EFPY Rcaaor Vessel hied. neat IniYial inside TI4 3/4T T/4 3/4T 3/4T T/4 Be(dine Region Location Idem. Numhet Type RTsue Sur/acne Locatet/ Locaten>> Lccat'en lncat'en Locat'en Locatiee Regulatory Guide I 99, Rev Yiion 2. Pmition I. I intermediate Shell Plate A9153.1 A9153.1 SA433 Gr. Bl 0.09 0.46 +60 58.0 4.61E+19 2.8 1 E+19 I. I IE+19 74.0 59.7 34.0 34.0 168.0 IS3.7 Intermediate Shell Plate B4197.2 B4 197.2 SA433 Gr. Bl 0.09 0.50 +91 58.0 4.61E+19 2.8 1 E+19 I.IIE+19 74.0 59.7 34.0 34.0 199.0 184.7 Lower Shell Plate C9924.1 C9924-I Sh-533 Gr. Bl 0.0$ 0.47 +S4 51.0 4.50E+19 2.74E+19 I.OSE+19 64.7 52.1 34.0 34.0 152.7 140.1 Lower Shell Plate C9924-2 C9924-2 SA-533 Gr. Bl 0.0$ 0.47 +57 51.0 4.50E+19 2.74E 419 I.OSE+19 52.'I 34.0 34.0 155.7 143.1 IS Longit. WeMs (Beth 100%) BC/BD 4P4784 Ash/Unde 124 0.05 0.91 6$ .0 1.68E+19 1.02E+19 4.04E+18 50.9 56.0 50.9 6S.S 104.S 81.8 IS to LS Circ. Weld (100%) AB 5P67/I ASA/Linda 124 0.03 0.94 41.0 4.42E+19 2.69E+19 1.06E+19 51.9 41.7 51.9 41.7 83.8 63.4 LS Longit. Wclds (Both (00%) BA/BB 4P47S4 ASA/Linda 124 0.05 0.91 68.0 I.64E+19 I.OOE+19 3.94E+ IS 6$ .0 503 56.0 50.$ 104.0 81.0 Regutaery Guide 1.99. RcviYen 2, Position 2.1 Intermediate Shel I Plate B4 197-2 B4197-2 SA.533 Gr. B I 0.09 0.50 +91 51.4 4.61E+19 2.8 IE+19 I.IIE+19 52.9 34.0'o ( 1903) (177.9J IS to LS Circ. Wekl ((00%) SP6771 ASA/Lindc 124 0.03 49.1 4.42E+19 2.69E+19 1,06E+19 49.9 62.1 28.0 28.0 57.9

'*'ee Appendix A.

"'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The I/4T and 3AT location fluence values are determined by calculating the I/4T and 34T depth into the vessel and adding the minimum cladding thickness (i.e., "x" for I/4 T = 2.0625 in. and "x" for SAT = 5.9375 in.).

"'ince two of the six surveillance data points are not credible, a full margin value is used to calculate the I/4 T and 3AT ART values

[] - Controlling values of the adjusted reference temperatures.

zg op 0

oO

Table 7-14. Evaluation of Upper-Shelf Energy Decreases for the HNP Reactor Vessel Applicable to 36 EFPY

'l~ T Predicted CUSE Material Descri Fluence+'x Per R.G. 1.99/2 Reactor Vessel Beltline tion'*'aterial Heat Cu n/cm'nitial 10") CUS Et'i CvUSE wt% ft-lbs ft-lbs Decrease Re ion Location Identification Number Intermediate Shell Plate (IS) A9153-1 A9153-1 SA-533 Gr. Bl 0.09 2.81 83 63.9 23.0 Intermediate Shell Plate (IS) B4 197-2 B4197-2 SA-533 Gr. Bl 0.09 2.81 71 54 23 3"'1.6 5'6.8 Lower Shell Plate (LS) C9924-1 C9924-1 SA-533 Gr. Bl 0.08 2.74 98 Lower Shell Plate (LS) C 9924-2 C9924-2 SA-533 Gr. B 1 0.08 2.74 88 69.0 21.6 IS Longit. Weld (Both 100%) BC/BD 4P4784 ASA/Linde 124 0.05 1.02 94 76.1 19.1 IS to LS Circ. Weld (100%) AB SP6771 ASA/Linde 124 0.03 2.69 80 59.2<'> 26.0<>

LS Longit. Weld (Both 100%) BA/BB 4P4784 ASA/Linde 124 0.05 94 76.1 19.0

" See Appendix A.

"'alculated based on guidelines in Regulatory Guide 1.99, Revision 2. The inside surface fluence is the calculated value at the "wetted" surface of the reactor vessel (Table 6-2). The i/a T location fluence value is determined by calculating the i/~ T depth into the vessel and adding the minimum cladding thickness (i.e., "x" for i!~ T = 2.0625 in.).

'alculated using surveillance data in accordance with Regulatory Guide 1.99, Revision 2, Position 2.2 (i.e., fitting the surveillance data with a line drawn parallel to the existing lines in Figure 2 as the upper bound of all the data).

Zg Op oO 3

Table 7-15. Evaluation of Pressurized Thermal Shock Reference Temperatures for the HNP Reactor Vessel Applicable to 36 EFPY Chemical 36 EFPY Fluence Material Descri tion" Com sitionto Initial at Clad-Base Reactor Vessel Matt. Heat Cu Ni Chem. RTNor." Metal Interface,o'/cm'luence /tRTrts, Margin, RTrrs. Screening Bcltline Re ion Matl. Ident. Number Ty wt'Fo wt% Factor F Factor F F F Criteria RTrrs Calculation Per 10 CFR 50.61 Using Tables Intermediate Shell (IS) A9153-I A9153-I SA-533 Gr. Bl 0.09 0.46 58.0 4.55E+19 1.383 80.2 34.0 174.2 270 Plate Intermediate Shell (IS) B4197-2 B4197-2 SA-533 Gr. Bl 0.09 0.50 -58.0 91 4.55E+19 1.383 80.2 34.0 205.2 270 Plate Lower Shell (LS) C99%-I C9924-I SA-533 Gr. Bl 0.08 0.47 "

51.0 54 4A4E+19 1.378 70.3 34.0 158.3 270 Plate Lower Shell (LS) C9924-2 C~-2 SA-533 GI'. Bl 0.08 0.47 51.0 57 4.44E+19 1.378 70.3 34.0 161.3 270 Plate IS Longit. Welds BC/BD 4P4784 Linde 124 0.05 0.91 68.0 -20 1.66E+19 1.140 56.0 "113.5 270 (Both 100Fo)

IS to LS Circ. Weld AB 5P6771 Linde 124 0.03 0.94 41.0 -20 4.36E+19 1.375 56.4 56.0 IOOFo)

LS Longit. Wclds BA/BB 4P4784 Linde 124 0.05 0.91 68.0 -20 1.62E+19 1.133 77.0 56.0 113.0 270 (Both 100'Yo)

RTrts Calculation Per 10 CFR 50.61 Using Surveillance Data Intermediate Shell (IS) B4197-2 B4197-2 SA-533 Gr. Bl 0.09 0.50 51.4 91 4.55E+19 1.383 71 I 34 0'c) [196.11 270 Plate IS to LS Circ. Weld AB 5P6771 Linde 124 0.03 0.94 49.1 -20 4.36E+19 1.375 67.5 28.0 75.5 (IOOFo t'ee Appendix A.

"'he inside surface fluence is the calculated value at the clad - base metal interface of the reactor vessel; attenuation through the cladding is based on deterministic methods (Table 6-3).

'ince two of the six surveillance data points are not credible, a full margin value is used to calculate the RTFrs value.

[] Limiting reactor vessel beltline material in accordance with 10 CFR 50.61.;.,

zg 0 '.

0 oo

Figure 7-1. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation 100

~4 Capsule U 75 e

Capsule X 50 Baseline 25 Capsule V co

-100 100 200 300 400 500 600 Tomporaturo, F 100 80 Baseline O

60 Capsule U Capsule X 40 20 Capsule V 8

p

-100 100 200 300 400 500 600 Temperature, F 120 Capsule U 100 80' Baseline Capsule V c 60 ru E Capsule X 4p 20 Material: SA-533 Gr. B Cl.1

-100 100 200 300 400 500 600 Tomporaturo, F F RAMATOME CHHOIOOIC5 7-20 7 ~

0 Figure 7-2. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Base Metal Plate, Heat No. B4197-2, .

Transverse (TL) Orientation 100

~e Capsule U 75 Capsule X 5 50 u Baseline 25 Capsule V co

-100 100 200 300 400 500 600 Tomporaturo, F 100 Baseline 60 C

0 6p Capsule U x 40 Capsule X 20 Capsule V p

-100 100 200 300 400 500 600 Temporaturo, F 120 100 Capsule U 80 g7 Baseline p

60 ill Capsule V EJ Q.

E 40 Capsule X 20 Material: SA-533 Gr. 8 Cl. 1 Heat Number. 84197-2 (TL)

-100 100 200 300 400 500 600 Tomporaturo, F F RAMATOM E 4 544 I 5 5 7-21 5 5 C H H

Figure 7-3. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342) 100

~O Capsule U 75 B Capsule X 50 Baseline u

25 Capsule V CO

-200 -100 100 200 300 400 500 600 Tomporaturo, F 100 Capsule U 80 Baseline C

0 60 Capsule V Capsule X 40 20 0

-200 -100 100 200 300 400 500 600 Tomporaturo, F 120 Capsule U 100 80 Baseline C7 Capsule V p

c 60 Iu 44 O

CL E Capsule X 40 20 Material: Weld Metal-Linde 124 Heat Number: 5P6771/0342

-200 -100 100 200 300 400 500 600 Tomporaturo, F 7-22 F T IRAMATO ME C H H O LOO I E 5

~ ~ ~ ~

Figure 7-4. Comparison of Unirradiated and Irradiated Charpy Impact Data Curves

~

for Heat-Affect-Zone Material

~O

~s u.

100 75 50 25 Capsule U Baseline

~ Capsule X Capsule V co

-200 -100 0 100 200 300 400 500 600 Tomperaturo, F 100 80 Baseline g 60 Capsule U gc 40 Capsule X 20 Capsule V 0

-200 -100 0 100 200 300 400 500 600 F 'omporaturo, 120 100 Capsule U Baseline

4. Capsule V op

~c 60 Lu 55 CL E

40 Capsule X 20 Material: SA-533 Gr. 8 Cl. 1 0

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 7-23 F 5 5 RAMATO C H H XP 5 XP ME O I 5 5

S. Summary of Results The analysis of the reactor vessel material contained in the third surveillance capsule, Capsule X, removed for evaluation as part of the HNP Reactor Vessel Surveillance Program, led to the following conclusions:

1. The capsule received an average fast neutron fluence of 3.25 x 10" n/cm'E > 1.0 MeV).

2. Based on the calculated eight-cycle-average full power flux, the projected end-of-life (36 EFPY) peak fast fluence at the base metal-clad interface of the HNP reactor vessel is 4.55 x 10'/cm (E > 1.0 MeV). The corresponding fluences at the /4T, '/zT, 3/4 T, and outside surface vessel wall in this peak location are 2.64 x 10", 1.34 x 10",

6.53 x 10", and 2.84 x 10" n/cm'E > 1.0 MeV) respectively.

3. The 30 ft-lb transition temperature for the base metal plate, heat no. B4197-2, longitudinal (LT) orientation, increased 94'F after the irradiation to 3.25 x 10" n/cm'E

> 1.0 MeV). In addition, the CUSE for this material decreased 24.1%.

4. The 30 ft-lb transition temperature for the base metal plate, heat no. B4197-2, transverse (TL) orientation, increased 79'F after the irradiation to 3.25 x 10'/cm (E > 1.0 MeV). In addition, the CUSE for this material decreased 21.4%.

5. The 30 ft-lb transition temperature for the weld metal, weld wire heat 5P6771 / flux lot 0342, increased 79'F after the irradiation to 3.25 x 10" n/cm'E > 1.0 MeV). In addition, the CUSE for this material decreased 27.2%.

6. The measured 30 ft-lb transition temperature shifts for the surveillance materials are greater than the values predicted using Regulatory Guide 1.99, Revision 2 except for the heat-affected-zone material. When the margin (2') is added to the predicted shifts, the predicted 30 ft-lb transition temperature shift values are conservative for all the surveillance materials.

7. The measured percent decrease in CUSE for the measured surveillance base metal in the longitudinal (LT) orientation and the weld metal are slightly greater than the values predicted using Regulatory Guide 1.99, Revision 2 (0.7% and 5.5% respectively).

However, these values remain above the required 50 ft-lb limit.

F RAMATOME 8-1

8. The measured percent decrease in CUSE for the measured surveillance base metal in the transverse (TL) orientation was less than the Regulatory Guide 1.99, Revision 2 prediction.

9. In accordance with Regulatory Guide 1.99, Revision 2, the CvUSE values for the HNP reactor vessel beltline materials are not predicted to fall below 50 ft-lb at end-of-life (36 EFPY).

10. In accordance with 10 CFR 50.61, the HNP reactor vessel beltline materials will not exceed the PTS screening criteria before end-of-life (36 EFPY).

8-2 F RAMATOME

9. Surveillance Capsule Removal Schedule Based on the post-irradiation test results of Capsule X, the following schedule is recommended for the examination of the remaining capsules in the HNP reactor vessel surveillance program:

Withdrawal / Evaluation Schedule" Capsule Location of Removal Expected Capsule Identification Ca sules+'ead Factor" Time Fluence (n/cm')"'

23 6.826 x 10" '~

110'90'40'.38 EFPY'tandby 2.38 2.38 Standby (a) In accordance with ASTM Standard E 185-82.

(b) Reference reactor vessel irradiation locations, Figure 3-1.

(c) The factor by which the capsule fluence leads the vessels maximum inner wall fluence.

(d) Based on current capsule analysis.

(e) Approximate fluence not less than peak EOL vessel fluence (4.551 x 10" n/cm') or greater than twice the peak EOL vessel fluence (9.102 x 10" n/cm'). Therefore, actual capsule removal times can range from 15.13 EFPY to 30.25 EFPY. This capsule may be held without testing following withdrawal.

(f) The specified fluence represents the peak inside surface vessel fluence at the clad base metal interface after 60 calendar year (54 EFPY) of operation based on the current fluence estimates for plant license renewal consideration.

F RAMATO ME 9-1

10. Certification The specimens obtained from the third CP&L HNP reactor vessel surveillance capsule (Capsule X) were tested and evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50, Appendices G and H.

M. J. DeVan (Material Analysis)

/o7y Date Materials & Structural Analysis Unit S. Q. King (Fluence Analysis) Date Performance Analysis Unit his report has been reviewed for technical content and accuracy..

1~imp W. A. Pavinich, (Material Analysis) ate Materials & Structural Analysis Unit

. Giave oni (Fluence Analysis) Date Performance Analysis Unit Verification of independent review.

. E. Moore, Manager Date Materials & Structural Analysis Unit This report is approved for release.

I~/(y Vp D. L. Howell Date Program Manager 10-1 F RAMATOME

11. References

1. L. R. Singer, "Carolina Power & Light Company Shearon Harris Unit No. I Reactor

"

Vessel Radiation Surveillance Program, WCAP-10502, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, May 1984.

2. A. L. Lowe, Jr., et al., "Analysis of Capsule U Carolina Power & Light Company Shearon Harris Unit No. Reactor Vessel Material Surveillance Program ," BAW-2083, I

Babcock 8c Wilcox, Lynchburg, Virginia, August 1989.

3. A. L. Lowe, Jr., et al., "Analysis of Capsule V Carolina Power & Light Company Shearon Harris Unit No. Reactor Vessel Material Surveillance Program ," BAW-2154, BOW I

Nuclear Technologies, Lynchburg, Virginia, March 1992.

4. ASTM Standard E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E 706 (IF)" American Society for Testing and Materials, Philadelphia, Pennsylvania.

5.'ode of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities, " A endix G Fracture Tou hness Re uirements, Effective Date:

January 18, 1996.

6. Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and R~.

Utilization Facilities, " A endix Hff D: T H'eactor Vessel Material Surveillance Pro ram y 18. 1996.

7. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, "Nitclear Power Plant Components, " A endix G Protection A ainst Nonductile Failure, 1989 Edition.

8. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for Inservi ce Inspection of Nuclear Power Plant Components, "

A endix G Fracture Tou hness Criteria for Protection A ainst Failure, 1989 Edition.

9. ASTM Standard E 208-81, "Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, " American Society for Testing and Materials, Philadelphia, Pennsylvania.

• - Available from Framatome Technologies Inc., Lynchburg, Virginia.

F RAMATO ME

10. K. Y. Hour, "Evaluation of Carolina Power dc Light Company Shearon Harris Capsule

"

X, 00:475-0188-01:02 FTG Document No. 31-1083271-01, BOW Services, Inc.,

Lynchburg, Virginia, May 1999.

11. ASTM Standard E 8-96a, "Standard Test Methods for Tension Testing of Metallic Materials, " American Society for Testing and Materials, Philadelphia, Pennsylvania.

12. ASTM Standard E 21-92, "Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, " American Society for Testing and Materials, Philadelphia, Pennsylvania.

13. ASTM Standard E 23-91, "Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, "American Society for Testing and Materials, Philadelphia, Pennsylvania.

14.J. R. Worsham III, "Fluence and Uncertainity Methodologies, "BAW-2241P Revision 1, Framatome Technologies, Inc., Lynchburg, Virginia, April 1999.

15. U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, " Draft Re ulator Guide DG-1053, June 1996.

16. U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials, "Re ulato Guide 1.99 Revision 2, May 1998.

17. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities, " Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, " Effective Date: August 28, 1996.

11-2 F RAMAYOME

APPENDIX A Reactor Vessel Surveillance Program Background Data and Information F RAMATO ME A-1

A.l. Capsule Identification The capsules used in the HNP reactor vessel surveillance program are identified in Table A-1 by identification, location, and design lead factors. The capsule locations within the HNP reactor vessel are shown in Figure A-3.

A.2. HNP Reactor Pressure Vessel The HNP reactor pressure vessel was fabricated by the Chicago Bridge Ec Iron Company. The HNP reactor vessel beltline region consists of two shells, containing four heats of base metal plate, four longitudinal weld seams, and one circumferential weld seam. Table A-2 presents a description of the HNP reactor vessel beltline materials including their copper and nickel chemical contents and their unirradiated mechanical properties. The heat treatments of the beltline materials are presented in Table A-3. The locations of the materials within the reactor vessel beltline region are shown in Figures A-1 and A-2.

A.3. Surveillance Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with ASTM Standard E '185-82, are shown in Table A-2. Based on the initial RTgpT value, copper and nickel contents and the projected end-of-life fluences, the reactor vessel intermediate shell plate B4197-2 is considered to be the limiting reactor vessel beltline material and has been used in the HNP reactor vessel surveillance program (RVSP). The surveillance weld used in the HNP RVSP was fabricated using the wire heat 5P6771 and Linde 124 flux lot 0342 which is identical to the intermediate to lower shell circumferential weld in the HNP reactor vessel.

F RAMATOME A-2

Table A-1. HNP Surveillance Capsule Identifications, Original Locations, and Design Lead Factors Capsule Capsule Design Lead Identification Location" Factor+'43'.12 V 107'.12 X 287o 3 12

~

W 110'.7 290o 2.7 340'.7 (a) Reference irradiation capsule locations as shown in Figure A-3.

(b) The factor by which the capsule fluence leads the vessels maximum inner wall fluence.

F RAMATOME A-3

Table A-2. Description of the HNP Reactor Vessel Beltline Region Materials'""'""

Chemical Com sition Unirradiated Tou hness Pro rties Material Material Beltline Cu, Ni, 30 ft-lb, 50 ft-lb, 35 MLE, CyUSE, TNDT~ RTNin, Heat No. T~e Region Location wt% wt% F F F ft-Ibs F F A9153-1 SA-533 Gr. Bl Intermediate Shell 0.09 '.46 83 -10 B4197-2 SA-533 Gr. Bl Intermediate Shell 0.09 0.50 71 . -10 91 C9924-1 SA-533 Gr. Bl Lower Shell 0.08 0.47 98 -10 54 C 9924-2 SA-533 Gr. Bl Lower Shell 0.08 0.47 88 -20 57 4P4784 / 3930'" ASA Weld/ Intermediate Shell 0.05 0.91 94 -20 -20 Linde 124 Longitudinal Welds 5P6771 / 0342" ASA Weld/ Intermediate to Lower 0.03 0.94 80 -80 -20 Linde 124 Shell Circ. Weld 4P4784 / 3930'*'SA Weld/ Lower Shell 0.05 0.91 94 -20 -20 Linde 124 Longitudinal Welds

"'eld wire heat number and flux lot identifiers.

z3 Op o~

o0 I

Table A-3. Heat Treatment of the HNP Reactor Vessel Beltline Region Materials Material Heat Treatment'ntermediate Shell Plates Austenitizing: 1600+ 25'F for 4 hrs., water quenched A9153-1 and B4197-2 Tempered: 1250+ 25'F for 4 hrs., air cooled (Core Ring No. 2) Stress Relief: 1050'F for 4 hrs., air cooled Post Weld: 1150 + 25'F 50'F for 35~/4 hrs., furnace cooled Lower Shell Plates Austenitizing: 1600+ 25'F for 4 hrs., water quenched C9924-1 and C9924-2 Tempered: 1250+ 25'F for 4 hrs., air cooled (Core Ring No. 1) Stress Relief: 1075'F for 4 hrs., air cooled Post Weld: 1150 + 25'F

- 50'F for 43 hrs., furnace cooled Intermediate Shell Post Weld: 1150 + 25'F Longitudinal Seam Welds 50'F for 333/4 hrs., furnace cooled'"

(Wire Heat 4P4784 /

Flux Lot 3930)

Lower Shell Post Weld: 1150 + 25'F Longitudinal Seam Welds 50'F for 36/4 hrs., furnace cooled'"

(Wire Heat 4P4784 /

Flux Lot 3930)

Intermediate to Lower Local Shell Girth Seam Weld Heat Treat: 1150 + 25'F (Wire Heat 5P6771 / 50'F for 10'/4 hrs., furnace cooled Flux Lot 0342)

'*'ustenitizing, tempering, and stress relief times are from Lukens Steel Company Test Certificates. The post weld heat treatment times are from Chicago Bridge & Iron Company thermal history records.

+'he post weld heat treatment includes the PWHT for the Intermediate to Lower Shell Girth Seam Weld.

A-5 F RAMATOME

Figure A-1. Locations and Identifications of Materials Used in the Fabrication of the HNP Reactor Pressure Vessel I

I Wf I

I Welds, lntermed Shell A9153-1 BC/BD'100%)

Weld AB(100%)

Weldsi BA/BB C 9924-1 (100%) Lower Shell A-6 F RAMATOME TICNHOLOOIEP

Figure A-2. Locations of Longitudinal Welds in HNP Reactor Vessel Upper and Lower Shell Courses 00 B 4197-2 BC 45'700 90'D A9153-1 180'B C 9924-1 45'0 270'0'-

BA C9924-2 F RAMATNME A-7

Figure A-3. Locations of Surveillance Capsule Irradiation Sites in the HNP Reactor Vessel

=

CAPSULES 0o U (343') REACTOR VESSEL (340') Z CORE BARREL 270'0 CAPSULES (290') Y (287') X NEUTRON PAD CAPSULES V (107')

W (110')

180O PIAN VIEW REACTOR VESSEL VESSEL WALL CAPSULE ASSEMBLY CORE CORE MIOPLANE NEUTRON PAO CORE BARREL ELEVATION VIEW F RAMATO ME A-8

.4.

~ ~ References A-1. Letter from W. R. Robinson (CP&L) to U.S. Nuclear Regulatory Commission, Attention: Document Control Desk, "Shearon Harris Nuclear Power Plant Docket No.

50-400/License No. NPF-63 Request for License Amendment RCS Pressure/Temperature Limits, "Serial: HNP-96-206, dated December 30, 1996.

A-2. Letter from James Scarola (CP&L) to U.S. Nuclear Regulatory Commission, Attention:

Document Control Desk, "Shearon Harris Nuclear Power Plant Docket No. 50-400/License No. NPF-63 Response to Request for Additional Information Regarding Generic Letter 92-01, Revision I, Supplement 1, " Serial: HNP-98-129, dated September 16, 1998.

A-9 F RAMATOM E

APPENDIX B Unirradiated and Irradiated Tensile Data for the HNP RVSP Materials B-1 F RAMATOME

Table B-l. Unirradiated Surveillance Tensile Properties of HNP Base Metal Plate, Heat No. B4197-2, Longitudinal Orientation Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform of Area,  %

Total 75 71.0 94.0 8 25 61 75 80.0 94.0 8 27 62 300 74.0 88.0 '6 22 64 300 75.0 87.0 6 21 63 550 74.0 88.0 6 19 54 550 73.0 88.0 19 54 Table B-2. Unirradiated Surveillance Tensile Properties of HNP Base Metal Plate, Heat No. B4197-2, Transverse Orientation Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield of Area, %

Ultimate Uniform Total 75 69.0 91.0 10 26 63 75 68.0 91.0 26 60 300 61.0 83.0 10 23 59 300 61.0 83.0 23 58 550 60.0 87.0 21 54 550 60.0 86.0 20 .55 B-2 F RAMATOME

Table B-3. Unirradiated Surveillance Tensile Properties of HNP Weld Metal, Wire Heat 5P6771 / Flux Lot 0342 Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform Total of Area, %

75 74.0 89.0 10 29 68 75 74.0 89.0 10 26 67 300 66.0 82.0 65 300 64.0 80.0 24 67 550 63.0 84.0 23 65 550 63.0 83.0 22 64 B-3 r/ F RAMATOME

Table B-4. HNP Capsule U Surveillance Tensile Properties of Base Metal Plate, Heat No. B4197-2, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Longitudinal Orientation Specimen Test Stren th; ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform of Area, %

Total QL2 70 72.4 97.0 10.1 23.7 63.0 QL3 70 72.3 96.4 10.4 22.8 62.1 QL1 550 65.7 93.6 9.6 19.3 . 52.7 Table B-5. HNP Capsule U Surveillance Tensile Properties of Base Metal Plate, Heat No. B4197-2, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Transverse Orientation Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield of Area, %

Ultimate Uniform Total

'T2 70 72.2 97.3 11.1 23.2 55.9 QT3 70 72.3 95.8 11.3 24.1 59.0 QT1 550 69.3 93.7 8.5 17.4 50.4 Table B-6. HNP Capsule U Surveillance Tensile Properties of Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield of Area, %

Ultimate Uniform Total QW3 70 76.3 91.0 11.2 25.6 66.8 QW1 70 76.3 91.1 11.3 25.9 67.3 QW2 550 69.0 87.5 9.2 20.9 62.5 F RAMATOME

N Table B-7. HNP Capsule V Surveillance Tensile Properties of Base Metal Plate, Heat No. B4197-2, Irradiated to 1.32 x 10" n/cm'E > 1.0 MeV)

Longitudinal Orientation Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform Total of Area, %

QL4 70 73.6 97.4 10.1 23.2 62.4 QL5 70 74.3 98.0 8.7 21.0 62.5 QL6 550 69.0 93.8 8.1 15.9 42.4 Table B-8. HNP Capsule V Surveillance Tensile Properties of Base Metal Plate, Heat No. B4197-2, Irradiated to 1.32 x 10" n/cm'E > 1.0 MeV)

Transverse Orientation Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform Total of Area, %

0 QT4 70 73.7 96.2 9.6 22.9 58.5 QT5 70 78.3 98.5 8.2 20.5 58.3 QT6 550 66.6 92.7 8.8 17.3 44.2 Table B-9. HNP Capsule V Surveillance Tensile Properties of Weld Metal, Wire Heat 5P6771 / Flux Lot 0342, Irradiated to 1.32 x 10" n/cm'E > 1.0 MeV)

Specimen Test Stren th, ksi Elon ation, % Reduction No. Temp. (F) Yield Ultimate Uniform of Area, %

Total QW4 70 75.3 91.6 9.3 22.4 64.0 QW5 70 76.6 91.9 9.4 22.6 59.8 QW6 550 70.3 88.3 7.6 18.6 61.6 F RAMATOME B-5

APPENDIX C Unirradiated and Irradiated Charpy V-Notch Impact Surveillance Data for the HNP RVSP Materials Using Hyperbolic Tangent Curve-Fitting Method F RAMATOM E C-1

Table C-1. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (mils) (%)

-60 5.0

-60 8.0 2.0 14.0 8.0 15 18.0 10.0 15 30 19.0 12.0 25 30 27.0 18.0 20 30 38.0 30.0 30 75 31.0 23.0 30 75 36.0 29.0 30 75 37.0 . 29.0 33 120 52.0 45.0 45 120 52.0 43.0 50 120 55.0 46.0 55 160 78.0 64.0 90 160 83.0 68.0 90 200 84.0 67.0 95 200 85.0 75.0 95 250 92.0 75.0 100 250 94.0 71.0 100 350 74.0 65.0 100 C-2 F RAMATO ME

Table C-2. HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Longitudinal (LT) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (ft-lb) (in.) (%)

QL1 11.5 0.008 QL3 40 21.0 0.017 20 QL10 70 22.5 0.021 30 QL11 90 40.0 0.045 30 QL12 110 37.0 0.030 30 QL15 125 43.5 0.038 50 QL6 140 49.0 0.043 60 QL14 175 66.0 0.049 85 QL2 225 80.5 0.064 100 QL4 275 89.5 0.067 100 QL8 325 90.0 0.072 100 QL9 375 99.0 0.069 100 QL5 425 89.0 0.071 100 QL7 475 92.5 0.069 100 QL13 550 100.0 0.079 100 C-3 F RAMATOME

Table C-3. HNP Capsule V Surveillance Charpy'Impact Data for Base Metal Plate, Heat No. B4197-2, Irradiated to 1.32 x 10" n/cm'E) 1.0 MeV)

Longitudinal (LT) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (in.) (%)

QL30 40 17.0 0.014 QL28 70 22.0 0.017 20 QL26 90 32.0 0.025 30 QL18 110 32.0 0.026 45 QL29 140 45.0 0.038 50 QL22 170 55.0 0.046 50 QL23 200 66.5 0.057 95 QL21 225 72.0 0.059 95 QL20 275 0;066 100 QL27 275. 76.0 0.063 100 QL19 325 87.5 0.070 100 QL16 375 87.0 0.073 100 QL24 425 84.0 0.071 100 QL25 475 84.0 0.071 100.

QL17 550 86.0 0.076 100 F RAMATOME

Table C-4. Hyperbolic Tangent Curve Fit Coefficients for HNP Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation H erbolic Tan ent Curve Fit Coefficients Absorbed Energy Lateral Expansion Percent Shear Fracture Unirradiated A: 47.0 A: 377 A: 500 B: 44.8 B: 36 7 B: 500 C: 102.3 C: 88 9 C: 91.9 TO: 92.3 TO: 93.9 TO: 102.3 Capsule U A: 49.1 A: 370 A: 500 B: 46 9 B: 360 B: 500 C: 120.5 C: 129.7 C: 78.1 TO: 133.3 TO: 117.0 TO: 122.4 Capsule V A: 44.3 A: 37:1 A: 500 B: 42.1 B: 36.1 B: 500 C: 118.1 C: 120.6 C: 85.6 TO: 136.0 TO: 138.5 TO: 133.8 F RAMATOME C-5

Table C-5. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (mils) (%)

-60 4.0

-60 4.0 12.0 7.0 10 14.0 8.0 15 40 23.0 19.0 30 40 27.0 22.0 25 75 '9.0 22.0 35 75 29.0 26.0 30 75 37.0 28.0 30 120 39.0 35.0 40 120 42.0 36.0 55 120 48.0 43.0 50 160 56.0 52.0 90 160 61.0 52.0 90 200 63.0 ~

56.0 100 200 65.0 58.0 100 250 67.0 60.0 100 250 69.0 60.0 100 350 65.0 62.0 100 350 93.0 82.0 100 C-6 F RAMATOM E

Table C-6. HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Transverse (TL) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (in.) (%)

QT10 70 17.0 0.018 20 QT15 70 15.0 0.015 20 QT9 90 29.5 0.025 35 QT6 110 29.5 0.026 40 QT1 140 47.0 0.046 100 QT14 140 32.0 0.030 60 QT5 175 51.5 0.047 80 QT8 175 50.0 0.045 85 QT11 225 62.0 0.056 100 QT13 275 66.0 0.057 100 QT12 325 70.0 0.062 100 QT2 375 68.0 0.059 100 QT3 425 68.0 0.061 100 QT4 475 71.5 0.064 100 QT7 550 68.5 0.059 C-7 F RAMATO ME

Table C-7. HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Irradiated to 1.32 x 10" n/cm'E) 1.0 MeV)

Transverse (TI ) Orientation Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (in.) (%)

QT22 70 18.0 0.016 10 QT29 90 27.0 0.024 40 QT27 110 32.5 0.031 50 QT26 140 38.0 0.036 50 QT21 170 47.5 0.040 70 QT28 170 48.0 0.040 50 QT18 200 50.0 0.045 75 QT23 200 49.5 0.044 50 QT30 225 60.0 0.052 50 QT19 275 60.0 0.053 55 QTA 325 63.5 0.058 100 QT20 375 60.0 0.056 60 QT25 425 66.0 0.060 100 QT17 475 61.5 100 QT16 550 67.0'.059 0.063 100 C-8 F RAMATOME

Table C-S. Hyperbolic Tangent Curve Fit Coefficients for HNP Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation H erbolic Tan ent Curve Fit Coefficients Absorbed Energy Lateral Expansion Percent Shear Fracture Unirradiated A: 40.1 A: 35.1 A: 500 B: 37 9 B: 34.1 B: 500 C: 125.2 C: 113.4 C: 83.1 TO: 105.9 TO: 110.0 TO: 104.3 Capsule U A: 35.8 A: 31.0 A: 500 B: 33.6 B: 300 B'00 C: 98 2 C: 97 9 C: 594 TO: 127.5 TO: 119.2 TO: 111.2 Capsule V A: 33.1 A: 306 A: 500 B: 309 B: 29.6 B: 500 C: 110.9 C: 135.2 C: 118.4 TO: 118.6 TO: 123.4 TO: 136.6 F RAMAYOM E C-9

Table C-9. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Weld Metal, Wire Heat 5P6771 / Flux Lot 0342 Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (ft-lb) (mils) (%)

-110 7.0 1.0 15

-110 8.0 2.0 15

-60 21.0 16.0 35

-60 25.0 18.0 40

-30 28.0 20.0 55

-30 31.0 26.0 45

-30 33.0 26.0 45 42.0 36.0 50 46.0 41.0 55 52.0 45.0 65 40 65.0 58.0 80 40 75.0 60.0 85 75 78.0 64.0 90 75 89.0 73.0 97 75 90.0 72.0 95 160 91.0 78.0 100 160 92.0 84.0 100 250 92.0 79.0 100 250 96.0 75.0 100 350 97.0 81.0 100 C-10 F RAMATOME

Table C-10. HNP Capsule U Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342),

Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (ft-lb) (in.) (%)

QW6 -80 7.5 0.005 20 QW2 40 19.0 0.017 40 QW12 -20 28.0 0.025 30 QW11 42.0 0.035 50 QW5 20 40.0 0.035 60 QW15 40 64.0 0.051 75 QW3 70 62.0 0.052 70 QW8 70 60.5 0.053 85 QW9 110 74.0 0.063 90 QW10 140 - 81.0 0.069 100 QW13 175 80.5 0.070 100

'00 QW4 225 83.0 0.074 QW1 275 80.0 0.073 100 QW7 375 83.0 0.071 100 QW14 550 89.0 0.078 100 F RAMATOME C-11

Table C-11. HNP Capsule V Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342),

Irradiated to 1.32 x 10" n/cm'E) 1.0 MeV)

Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. (F) (ft-lb) (in.) (%)

QW23 -80 12.0 0.010 15 QW28 -40 23.0 0.021 30 QW27 34.0 0.031 50 QW30 20 42.0 0.039 50 QW19 40 58.5 0.050 65 QW25 70 70.0 0.060 100 QW16 110 74.0 0.056 100 QW24 110 74.0 0.066 95 QW18 140 80.0 0.071 100 QW22 170 85.0 0.073 100 QW21 225 88.0 0.078 100 QW29 275 82.0 0.072 100 QW26 325 81.5 0.074 100 QW17 375 90.0 0.081 100 QW20 550 90.0 0.081 100 C-12 F RAMATOME

Table C-12. Hyperbolic Tangent Curve Pit Coefticienta for HNP Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

H erbolic Tan ent Curve Fit Coefficients Absorbed Energy Lateral Expansion Percent Shear Fracture Unirradiated A: 48.8 A: 402 A: 500 B: 466 B: 39.2 B: 500 C: 79 7 C: 74.2 C: 992 TO: 0.1 TO: 0.8 TO: 0.0 Capsule U A: 42.8 A: 37 3 A: 500 B: 40.6 B: 36 3 B: 500 C: 883 C: 96 8 C: 100.7 TO: 15.2 TO: 19.3 TO: 00 Capsule V A: 44.3 A: 392 A: 500 B: 42.1 . B: 382 B: 50 0 C: 93.1 C: 109.4 C: 77.1 TO: 16.9 TO: 19.7 TO: 26 F RAMATO ME C-13

Table C-13. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Heat-Affected-Zone Material Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. ('F) (ft-lb) (mils) (%)

-160 7.0 2.0

-160 13.0 9.0 25

-120 12.0 4.0 10

-120 16.0 5.0 10

-90 9.0 2.0 15

-90 11.0 5.0 13

-60 23.0 16.0 37

-60 24.0 16.0 35

-60 30.0 18.0 30

-30 '30.0 21.0 40

-30 34.0 26.0 40

-30 39.0 30.0 40 0 36.0 26.0 45 48.0 33.0 55 52.0 38.0 55 40 63.0 46.0 80 40 76.0 56.0 75 75 83.0 65.0 100 75 84.0 70.0 100 75 94.0 67.0 100 140 75.0 58.0 100 140 78.0 57.0 100 220 79.0 63.0 100 220 87.0 63.0 100 300 97.0 69.0 100 C-14 F RAMAT0M E

P Table C-14. HNP Capsule U Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 5.52 x 10" n/cm'E) 1.0 MeV)

Impact Lateral Shear Test'emp.

Specimen Energy Expansion Fracture No. ('F) (ft-lb) (in.) (%)

QH10 -80 10.0 0.006 10 QH7 -40 19.0 0.014 35 QH12 -40 18.5 0.013 30 QH5 31.5 0.024 50 QH14 27.5 0.024 40 QH11 40 41.0 0.034 65 QH3 70 54.0 0.044 50 QH13 70 65.0 0.050 100 QH15 110 63.0 0.048 100 QH2 175 72.5 0.055 100 QH4 225 78.0 0.058 100 QH6 275 75.0 0.061 100 QH8 375 78.0 0.060 100 QH1 450 84.0 0.064 100 QH9 550 90.5 0.070 100 F RA M AT 0 M E C-15

Table C-15. HNP Capsule V Surveillance Charpy Impact Data for Heat-Affected-Zone Material, Irradiated to 1.32 x 10" n/cm'E) 1.0 MeV)

Test Impact Lateral Shear Specimen Temp. Energy Expansion Fracture No. ('F) (ft-lb) (in.) (%)

QH29 -40 11.0 0.008 20 QH22 26.0 0.022 40 QH27 40 50.0 0.039 45 QH26 70 58.0 0.042 80 QH19 110 74.0 0.056 100 QH20 110 65.0 0.050 70 QH17 140 70.0 0.055 100 QH23 170 78.0 0.065 100 QH16 225 66.0 0.047 100 QH18 275 85.0 0.064 100 QH28 275 68.0 0.053 80 QH24 325 87.0 0.061 100 QH30 375 80.0 0.062 100 QH21 475 83.0 0.061 100 QH25 550 100.0 0.070 100 F RAMATO ME C-16

Table C-16. Hyperbolic Tangent Curve Fit Coefficients for HNP Heat-Affected-Zone Material H erbolic Tan ent Curve Fit Coefficients Weld Metal Absorbed Energy Lateral Expansion Percent Shear Fracture Unirradiated A: 45.5 A: 33.4 A: 500 B: 43.3 B: 324 B: 500 C: 864 C: 73 0 C: 92.3 TO: -11.7 TO: -9.3 TO: 0.0 Capsule U A: 41.7 A: 31.6 A: 500 B: 39 5 B: 306 B: 500 C: 107.9 C: 102.9 C: 97.4 TO: 33.0 TO: 28.3 TO: 64 Capsule V A: 41.8 A: 307 A: 500 B: 39 6 B: 29.7 B: 500 C: 884 C: 81.8 C: 883 TO: 35.7 TO: 28.6 TO: 27.0 C-17 F RAMATOME

Figure C-1. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation

~Q

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100 I

75 es 50 25 In

-100 100 200 300 400 500 600 Temperature, F 100 80 C

60 40 LU 20 0

-100 100 200 300 400 500 600 Temperature, F 120

$ 5AIIt.f Tre'. +99 F Tgo'. +51 F 100 CvUSE: 87 ft.lb 80 uI 60 C

uj CL E

40 20 ~ Material: SA-533. Gr. B, Cl. 1

~

Fluonco: None Heat Number: B4197-2 (LT)

-100 100 200 300 400 500 600 Temperature, F F RAMATOME rccHHoLooIcs C-18

Figure C-2. HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation

- Refitted Using Hyperbolic Tangent Curve-Flitting Method-100

~4 75 p 50 u

5

~ ~

25 Ul

-100 100 200 300 400 500 600 Temperaturo, F y> 100 80 O

60 x 40 20 0

-100 100 200, 300 400 500 600 Temporaturo, F 120 Tygule.'+110 F

$0 T5o'. +81 F 100 CvUSE: 92 ft4b

~ ~

80 4

Cl 60 Iu O

C5.

E 40

~ Material: SA-533. Gr. B. Cl. 1 20 ~

-100 100 200 300 400 500 600 C-19 FRAMATDME 5 5 C N H 55 5 O 0 I C 5

Figure C-3. HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Longitudinal (LT) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100

~ ~

75 I5 50 u.

I 25 Co

-100 0 100 200 300 400 500 600 Temperaturo, F 100 E eo C

O co 60 40 I5 20 0

-1 00 100 200 300 400 500 600 Temporature, F 120 T55me: +132 F Tso. +152 F T5o:

100 CvUSE: 64 ft.lb

~ ~

eo Ol 4'

C 60 Lu II Ilj 51.

E 40

~ ~

Material: SA-533, Gr. B, Cl. 1 20 Fluenco: 1.32x10'/cm

-100 100 200 300 400 500 600 Temperature, F C-20 F RAMATO ME 55CHHO5OOI55

~ ~ ~

Figure C-4. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP

~

Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation

~

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-

~

100

~4 75 50 u

25 CO

-100 100 200 300 400 500 600 Temperature, F 100 80 C

O 60 40 ll5 O 20 0

-100 100 200 300 400 500 600 ~

Tomporaturo, F 120

$ 5MLe 50 30 100 CvUSE: 70Nlb 80 60 tu CL E

40 20 Material: SA-533, Gr. B, Cl. 1 Fluenco: None Heat Number. 84197-2 (TL)

-100 100 200 300 400 500 600 Tomporaturo, F C-21 F RAMATO ME TECHHOLOOICS

Figure C-5. HNP Capsule U Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100

~4 75 I

50 u

I 25 co

-100 100 200 300 400 500 600 Tomporaturo, F 100 g

80 O

g 60 40 20 0

-100 100 200 ,'00 400 500 600 Tomporaturo, F 120 55MLB 50 Tig +110 F 100 CvUSE: 68 fl.tb 80 4-0) 60 LU E

40 Material: SA-533. Gr. B, Cl. 1 20

-100 100 200 300 400 500 600 C-22 F RAMATOME TCCNHOLOOIC5

1 Figure C-6. HNP Capsule V Surveillance Charpy Impact Data for Base Metal Plate, Heat No. B4197-2, Transverse (TL) Orientation

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100 RU 75 50 ~ ~ ~

IL 25 co

-100 100 200 300 400 500 600 Tomporaturo, F 100 80 C

O 60 40 I5 20 0

-100 "

0 100 200 300 400 500 600 Tomporaturo,'

120 75 IE 50 TRI: +167 6 100 CRU65: 6511.16 80 f7 OT o 60 Lu CL E

40 20 Material: SA-533, Gr. 8, Ct. 1

-100 100 200 300 400 500 600 Tomporaturo, F C-23 F RAMATO 5 4 C 56 56 ME6 O 6 O O I C

Figure C-7. Unirradiated Surveillance Charpy V-Notch Impact Data for HNP Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100

~O 75

~ ~

50 u

5 25 co

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 100

'. 80 C

O 60 40 20 0

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 120 35MLK 50 30 100 CvUSE: 92 ft-Ib I ~

0I 80 4-OI o 60 tu O

tL E

40 Material: Weld Metal-Linde 124 20 Fluence: None Heat Number. 5P6771/0342

-200 -100 0 100 200 300 400 500 600 F RAMATO ME 55CHHOIOOI55 C-24

Figure C-8. HNP Capsule V Surveillance Charpy Iinpact Data for Weld Metal (Wire Heat 5P6771 / Flux Lot 0342)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100 75 L

50 25 ce

-200 -100 0 100 200 300 400 500 600 Temperature, F 100 80 C

O 60 x 40 ~ ~ ~ ~

20 0

-200 -100 0 100 200 300 400 500 600 Temperature, F 120

$$ IA(.E'0 80 100 CvUSE: 83fl.lb 80 4

60 Ill CL E

40 Material: Weld Metal-Linde 124 20 F(uenco: 2x1 "n/

Heat Number. SP6771 /0342

-200 -100 0 100 200 300 400 500 600 C-25 F RAMATO TCC ME H HOIOO IC5

Figure C-9. HNP Capsule V Surveillance Charpy Impact Data for Weld Metal (Wire Heat 5P6771 / Flux I.ot 0342)

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-

<a 75 50 IP.

25 co

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 100 80 O

g 60 40 20 0

-200 -100 0 100 200 300 400 500 600 Tomporaturo, F 120 3$ MLe 50 00 100 CvUSE: 82fl-Ib 80 ~ ~

4 60 W

E 40 20 Material: Weld Metal-Linda 124 Fluence: 1.32x10'0 n/cm~

Heat Number. 5P6771/0342

-200 -100 0 100 200 300 400 500 600 Temperature, F C-26 F RAMATOME ICCHNOLOOIC5

Figure C-10. Unirradiated Surveillance Charpy V-Notch Impact. Data for HNP Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100

~O 75 50 U. ~

~

25 V)

-200 -100 100 200 300 400 500 600 Temperature, F 100 E 80 C

0 co 60 C

40 l5 20 0

-200 -1 00 0 100 200 300 400 500 600 Temperature, F 120 55ula Tg)', QF SO.

100 CrUSE: 85 fl.lb 80 Cl C

60 lu CL E

40

~ W ~

20 Materiat: SA.533. Gr. B. CI. 1

~ Fluence: None

~

~

! Heat Number: B4197-2 (HAZ)

-200 -100 0 100 200 300 400 500 600 Temperature, F C-27 F

'r RAMATO K C ME5 H H O L O 0 I C

Figure C-'11. HNP Capsule U Surveillance Charpy Impact Data for Heat-Affected-Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100 75 E

50 u.

L 25 Vl

-200 -100 0 100 200 300 400 500 600 Temperature, F 100 E 80 C

O 60 C

~ ~

Fc 40 E 20 0

-200 -100 0 100 200 300 400 500 600 ~

Temperature, F 120 05MLa 50 50 100 CvUSE: 76 ft-Ib 80 O

60 lu O

IC Q.

E 40 20 Material: SA-533, Gr. 8, Cl. 1 Fluence: 5.52x10'0 n/cm~

-200 -100 0 100 200 300 400 500 600 Temperature, F F RAMATO ME HHOCOOICC C-28 CCC k

Figure C-12. HNP Capsule V Surveillance Cha'rpy'mpact Data for Heat-Affected Zone Material

- Refitted Using Hyperbolic Tangent Curve-Fitting Method-100 75 50 IL I

25 co

-200 -100 0 100 200 300 400 500 600 Temperature, F 100 80 C

0 60 40 20 0

-200 -100 0 100 200 300 400 500 600 Temperature, F 120 Tggyga'. +41 F Tao:

100 T5o'9 F CvUSE: 80 ft.lb 80 Zl CB 60 C

Lu O

K E

40

. 20 Material: SA-533. Gr. 8, CI. 1 Fiuonco: 1.32x10'/cm

-200 -100 0 100 200 300 400 500 600 Temperature, F F RAMATO ME CCCHHOCOOI55 C-29

APPENDIX D Charpy V-Notch Shift Comparison:

Hand-Drawn Curve Fitting vs. Hyperbolic Tangent Curve Fitting D-1 F RAMATOME

Table D-1. Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Base Metal Plate Heat No. B4197-2, Longitudinal (LT) Orientation 30 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E> 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 60 51 U 0.552 85 25 81 30 1.32 101 41 94 43 50 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E>1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 110 99 0.552 141 31 136 37 V 1.32 153 43 152 53 35 MLE Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E> 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 90 87 U 0.552 123 33 110 23 1.32 132 42 132 45 D-2 F RAMATOME

Table D-2. Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Base Metal Plate Heat No. B4197-2, Transverse (TL) Orientation 30 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10>> ibtcm')

Capsule (E) 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 65 72 0.552 110 45 110 38 V 1.32 102 37 107 35 50 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E) 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 130 139 0.552 173 43 172 33 1.32 188 58 186 47 35 MLE Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" ecm')

Capsule (E> 1.0 MeV) Avg., 'F Shift oF Avg., 'F Shift, 'F Unirradiated 105 110 0.552 131 26 132 22 1.32 133 28 144 34 D-3 F RAMATOME

Table D-3. Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Weld Metal (Wire Heat 5P6771 / Flux Lot 0342) 30 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E) 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated -35 -34 0.552 24 -14 20 1.32 -10 25 18 50 ft-Ib Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E) 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 0.552 42 42 31 29 V 1.32 34 29 27 35 MLE Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10'9 n/cm')

(E) 1.0 MeV) Avg., 'F 'F 'F 'F

'apsule Shift, Avg., Shift, Unirradiated -15 -9 U 0.552 18 13 22 1.32 22 17 F RAMAYOME D4

Table D-4. Comparison of Curve Fit Transition Temperature Shifts for HNP Surveillance Material, Heat-Affected-Zone 30 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E> 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated 0.552 45 44 1.32 10 55 53 50 ft-lb Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10" n/cm')

Capsule (E> 1.0 MeV) Avg., 'F Shift, 'F Avg., 'F Shift, 'F Unirradiated U 0.552 57 57 56 59 V 1.32 47 47 54 57 35 MLE Transition Temperature Fluence Hand-Drawn Curve Fit Hyperbolic Tangent Curve Fit (x10'9 n/cm')

(E>1.0 MeV) Avg., 'F Shift, 'F Avg., 'F 'apsule Shift, 'F Unirradiated -6 U 0.552 39 40 46 V 1.32 40 40 41 47 F RAMATOME D-5

APPENDIX E Fluence Analysis Methodology F RAMATO ME

The primary tool used in the determination of the flux and fluence exposure to the welds, plates, and surveillance capsule specimens is the two-dimensional discrete ordinates transport code, DORT.~~n E<-1. Cycle 1 Through 8 Flux Calculational Procedures The standard Framatome Technologies Inc. (FTI) fluence analysis procedure was used to determine the fluence accumulated in the HNP Capsule X and on the HNP reactor vessel beltline region plates and welds for cycles 1 through 8.

Figure E-1 depicts the analytical procedure that is used to determine the incremental fluence accumulated over cycles 1 through 8. As shown in the figure, the analysis is divided into seven tasks:

(1) generation of the neutron source, (2) development of the DORT geometry models, (3) calculation of the macroscopic material cross sections, (4) synthesis of the results, (5) estimation of the calculational bias, (6) the calculational uncertainty, and (7) the final fluence.

Each of these tasks is discussed below.

E-2. Generation of Neutron Source The time-average space- and energy-dependent neutron source for cycles 1 through 8 was calculated using the SORREL code.' The effects of burnup on the spatial distribution of the neutron source were accounted for by calculating the cycle average fission spectrum for each fissile isotope on an assembly-by-assembly basis, and by determining the cycle-average specific neutron emission rate. These data were then used with the normalized time-weighted-average pin-by-pin relative power density (RPD) distribution to determine the space- and energy-dependent neutron source. The azimuthal-average, time average axial power shape in the peripheral assemblies was used with the fission spectrum of the peripheral assemblies to determine the neutron source for the axial DORT run. These two neutron source distributions were input to DORT as indicated in Figure E-1.

0 F RAMATOME E-2

~

E-3. DORT Analysis

~

The cross sections, geometry, and appropriate source were combined to create a set of DORT (R-8 and R-Z) for the cycle 1 through 8 analysis. Each DORT run utilized a cross 'odels section Legendre expansion of three (P,), a minimum of forty-eight directions (S,), and the appropriate boundary conditions. All outer boundaries employed vacuum boundary conditions.

(Note that when vacuum boundary conditions are used, the location of the vacuum boundary with respect to the location of the boundary flux was checked to ensure that the boundary source is being written sufficiently far into the inner model to ensure that the boundary location does not perturb the fiux significantly at the boundary flux location.) A theta-weighted flux extrapolation model was used, and all other requirements of Draft Regulatory Guide DG-1053' that relate to the various DORT parameters were met or exceeded for all DORT runs.

E<-4. Synthesized Three-Dimensional Results The DORT analyses produced two sets of two-dimensional flux distributions, one for a vertical cylinder and one for the radial plane. The vertical cylinder, which will be referred to as the R-Z plane, is defined as the plane bounded axially by the upper and lower grid plates and adially by the center of the core and a vertical line located 20 cm into the water biological shield. The horizontal plane, referred to as the R-8 plane, is defined as the plane bounded radially by the center of the core and a point located approximately two feet into the concrete of the primary biological shield, azimuthally by the major axis, and the adjacent 45'zimuth.

The vessel flux, however, varies significantly in all three cylindrical-coordinate directions (R, 0, Z). This means that if a point of interest is outside the planes of both the R-Z DORT and the R-0 DORT, the true flux cannot be determined from either DORT run. Under the assumption that the three-dimensional flux is a separable function, the two two-dimensional data sets were mathematically combined to estimate the flux at all three-dimensional points (R, 8, Z) of interest. The synthesis procedure outlined in Draft Regulatory Guide DG-1053 forms the basis for the FTI flux-synthesis process.

E-5. Development of the Geometrical Models The system geometry models for the mid-plane (R-0) DORT were developed using standard FTI interval size and configuration guidelines. The R-8 model for the cycle 1 through 8 analysis extended radially from the center of the core to a point approximately two feet into the concrete of the primary biological shield, and azimuthally from the major axis to 45'. The urveillance capsule was modeled explicitly in the R-0 model. The axial (R-Z) DORT E-3 FRAMATOME

geometry model was developed using FTI procedures for the radial part, and used the appropriate interval structure in the axial direction. The axial model extended from core plate to core plate. The geometrical models meet or exceed all guidance criteria concerning interval size that are provided in Draft Regulatory Guide DG-1053. In all cases, cold dimensions were used. The geometry models were input to the DORT code as indicated in Figure E-1. These models will be used in all subsequent Code of Federal Regulation, Title 10, Part 50 (10 CFR 50), Appendix H' and pressure-temperature (P-T) curve analyses.

E-6. Calculation of Macroscopic Material Cross Sections In accordance with Draft Regulatory Guide DG-1053, the BVGLE-93'~5i cross section library was used. The GIP code' was used to calculate the macroscopic energy-dependent cross sections for all materials used in the analysis, from the core out through the cavity and into the concrete and from core plate to core plate. The ENDF/B6 dosimeter reaction cross sections were used to generate the response functions that were used to calculate the DORT-calculated saturated specific activities.

E-7. Calculated Activities and Measured Activities The calculated activities for each dosimeter type "d" were determined using the following equation:

c= ÃSF~ *CF*gPRF.

where: NSFq Non-saturation factor for dosimeter type CFd factors for dosimeter type "d" "d'orrection dg Neutron flux for energy group function (dosimeter response / unit flux) for "g*'esponse RF~g .

group "g" and dosimeter type "d" It is noted that the measurements are not involved in any way in determining the calculated activities.

The measured activities are listed in Table 6-1.'~"

E-8. C/M Ratios Frequently, the terms "calculations" and "measurements" are used in the fluence analysis area, but adequate definitions of the terms are seldom provided. The following discussion will F RAMATO ME

clarify the meanings of the terms, "Measurements" (M) and "Calculations" (C) as used by FTI.

~ Measurements (M): The meaning of the term "measurements" as used by FTI is the measurement of the physical quantity of the dosimeter (specific activity) that responded to the neutron fluence, not to "measured fluence." For example, for an iron dosimeter, reference to the measurements means the specific activity of "Mn, which is the product isotope of the dosimeter reaction:

'Fe(n,p) Mn, in Ci('4Mn) / g(~Fe)

~ Calculations (C): The calculational methodology produces two primary results: tlie calculated dosimeter activities and the neutron flux at all points of interest. The meaning of the term "calculations," as used by FTI in the present context, is the calculated dosimeter activity. Calculated activities are determined in such a way that they are directly comparable to M, without recourse to the measurements (e.g. C is calculated in pCi("Mn)/g("Fe), and is directly comparable to M). ENDF/B6-based dosimeter reaction cross sections,. (i.e., response functions) were used in determining C for each individual dosimeter. It should be noted that in the FTI approach, the calculated activity is totally independent of the measurements.

~ C / M Benchmarks: The C / M benchmark is calculated for each dosimeter, using the C's and M's discussed above and in Section E-7.

E-9. Estimation of the Best-Estimate Flux The flux in the reactor vessel beltline region is determined by best-estimate calculations, which are, by definition, the DORT results corrected for the generic energy-dependent bias removal function, h,. The FTI cavity dosimetry database, which was developed in the cavity dosimetry benchmark experiment, determined that there is a slight bias in the calculations. The energy-dependent bias removal function was developed to remove biases from the DORT results in order to provide best-estimate calculational results.

As discussed in the uncertainty analysis, there is no significant bias associated with this analysis beyond that identified in the Cavity Dosimetry Program. Accordingly, the energy-dependent benchmark bias function was used with the DORT-calculated flux to determine the best-estimate E-5 F RAMATOM E

flux at each point of interest in the reactor vessel in accordance with the procedures discussed in the Fluence and Uncertainty Topical Report, BAW-2241P, Revision 1.'~"

E-10. Extrapolation to the End-of-Life (EOL)

By necessity, extrapolation of neutron fluence to points in the future is an inexact and approximate process. It is impossible to know with certainty the character of future core operations or to accurately estimate the effect of any given core operation on the fluence at any given location, before the fact. It is possible, however, to make reasonable estimates of the inside surface maximum flux using near-future fuel cycle design trends.

The "extrapolation flux" is defined as the constant flux used to determine the fluence at points in the future. In the FTI methodology, extrapolation flux is based on the DORT-calculated flux determined in the just-completed fluence analysis. Since it is the stated intention of Carolina Power & Light Company (CP&L) to continue operation into the indefinite future with loadings similar to those used in the later cycles of the HNP cycle 1 through 8 operations, and since the post cycle 8 operations will have vessel fluxes that are lower than the cycle 1 through 8 average fluxes, the extrapolation fluxes reported herein are appropriate and conservative. It is recognized that these extrapolation fluxes are greater than those used in the previous HNP Capsule V analysis; this is a consequence of the differences in power distribution and neutron spectra between the Capsule V analysis and this analysis.

It is emphasized that with proper monitoring, the magnitude of the fluence at end-of-life (EOL) and the uncertainty in the EOL fluence will be maintained such that the material properties will not exceed their lawful limits.

E-11. Uncertainty The HNP reactor vessel fluence predictions are based on the methodology described in the FTI Fluence and Uncertainty Topical Report.'~" The time-averaged fluxes, and thereby the fluences throughout the reactor, and vessel are calculated with the DORT discrete ordinates computer code using three-dimensional synthesis methods. The basic theory for synthesis is described in Section 3.0 of BAW-2241P, Revision 1 and in the previous Sections of this Appendix. The DORT three-dimensional synthesis results are the bases for the fluence predictions using the FTI "Semi-Analytical" (calculational) methodology. As noted in Sections 6.0 and 7.0 of BAW-2241P, Revision 1, the best-estimate fluence predictions are determined by removing any bias from the calculated fluence results. The bias removal function is dependent on the DORT solution procedures, the BUGLE-93 cross sections, and E-6 F RAMATOME

he FTI dosimetry benchmarks. It is independent of the HNP:flu'e'nce predictions and any plant-specific comparisons of dosimetry calculations to measurements.

The uncertainties in the HNP reactor vessel best-estimate fluence values have been evaluated to ensure that the greater than 1.0 MeV calculated values are accurate (with no discernible bias) and have a mean standard deviation that is consistent with the FTI benchmark database of uncertainties. Consistency between the fluence uncertainties in the updated calculations for HNP, cycles 1 through 8, and those in the FTI benchmark database ensures that the best-estimate vessel fluence predictions are consistent with the Code of Federal Regulations, Title 10, Part 50.61 (10 CFR 50.61),'~" Pressurized Thermal Shock (PTS) screening criteria and the Regulatory Guide 1.99, Revision 2'" embrittlement evaluations.

The verification of the best-estimate fluence uncertainty for HNP includes:

(a) estimating the uncertainties in the cycles 1 through 8 dosimetry measurements, (b) estimating the uncertainties in the cycles 1 through 8 benchmark comparison of calculations to measurements, and (c) determining if the specific measurement and benchmark uncertainties for cycles 1 through 8 are consistent with the FTI database of generic uncertainties in the measurements and calculations.

The embrittlement evaluations in Regulatory Guide 1.99, Revision 2 and those in 10 CFR 50.61 for the PTS screening criteria apply a margin term to the reference temperatures. The margin term includes the product of a confidence factor of 2.0 and the mean embrittlement standard deviation. The factor of 2.0 implies a very high level of confidence in the fluence uncertainty as well as the uncertainty in the other variables contributing to the embrittlement. The ten dosimeter measurements from HNP, cycles 1 through 8 would not directly support this high level of confidence. However, since one capsule from HNP is already included in the FTI database, and these results are consistent with the FTI database uncertainties, the calculational uncertainties in the updated fluence predictions for HNP are supported by 728 additional dosimeter measurements and 39 benchmark comparisons of calculations to measurements as shown in Appendix A of BAW-2241P, Revision 1. The calculational uncertainties are also supported by the fluence sensitivity evaluation of the uncertainties in the physical and operational parameters, which are included in the vessel fluence uncertainty.'E" E-7 F RAMATOME

The FTI generic uncertainty in the capsule dosimetry measurements has been determined to be unbiased and has a FTI proprietary standard deviation for the qualified set of dosimeters. The HNP cycle 1 through 8 dosimetry measurement uncertainties were evaluated to determine if any biases were evident and to estimate the standard deviation. The dosimetry measurements were found to be appropriately calibrated to standards traceable to the National Institute of Standards and Technology (NIST) and are thereby unbiased. The mean measurement uncertainties associated with cycles 1 through 8 are as follows:

Mean Measurement Uncertainty Cycle o'w%

1 8 5.173 These values were determined from Equation 7.6 in BAW-2241P, Revision 1 and indicate that there is consistency with the FTI database. Consequently, when the FTI database is updated, the HNP cycle 1 through 8 dosimetry measurement uncertainties may be combined with the other dosimeters. Since the cycles 1 through 8 measurements are consistent with the FTI database, it is estimated that the HNP Capsule X dosimeter measurement uncertainty may be represented by the FTI proprietary database standard deviation.

The weighted mean values of the ratio of calculated dosimeter activities to measurements (C / M) for cycles 1 through 8 have been statistically evaluated using Equation 7.15 from BAW-2241P, Revision 1. The standard deviations in the benchmark comparisons are as follows:

C / M Standard Deviation cr<<,

Cycle <<<m%

1-8 6.14 This standard deviation indicates that the benchmark comparisons are consistent with the FTI database. Consequently, when the FTI database is updated, the cycle 1 through 8 benchmark uncertainties may be included with the other 39 benchmark uncertainties in BAW-2241P, Revision 1. The consistency between the cycle 1 through 8 benchmark uncertainties and those in the FTI database indicates that the HNP fluence calculations for cycles 1 through 8 have no discernible bias in the greater than 1.0 MeV values.

E-8 F RAMATOME

Since dosimetry locations at various plants are both inside and outside of the vessel, it is very probable that the best-estimate fluence can be represented by an FTI proprietary standard deviation anywhere throughout the vessel. However, as discu'ssed'in BAW-2241P, Revision 1 to maintain a high level of confidence that the vessel fluence uncertainty is well defined, an analytical uncertainty evaluation was performed. The analytical uncertainty evaluation was performed in a manner that was statistically consistent with the benchmark comparisons and consistent with the high levels of confidence that the probability of vessel failure due to embrittlement is insignificant.

The combination of benchmark and analytical uncertainties discussed in BAW-2241P, Revision 1 indicates that the vessel fluence uncertainty includes an FTI proprietary standard deviation, and a 95 percent probability that the true vessel fluence will be within the best-estimate fluence values in Tables 6-2 and 6-3. The vessel fluence uncertainty values are presented in Table E-2 of BAW-2241P, Revision 1.

Based on the detailed discussion above, the uncertainties given in BAW-2241P, Revision 1 are applicable to the HNP fluence calculations. The uncertainties (U), that were determined in BAW-2241P, Revision 1 are as follows:

U(capsule) 7.94%

U(vessel, cycles 1 - 8) 12.36%

E-9 F RAMATOME

Figure E-1. Fluence Analysis Methodology ssem y x assem y fission spectrum pin x pin eactor eometry atena s o by fissile isotope distribution construction history 8 LE-93 mo es cross section co e library ime-average axial source Pco e ime-average neutron source osimetry So(E,R,S) ross sections counting 8 D R Ana ysis analysis RO and RZ (NES I) Data to calcu ate Power history absolute (saturation) ynt esize magnitude results 3-D results easure dosimetry activities a cu ate osimeter activities 88WOG benchmark tatistical analysis analysis Ca culational bias bias on irm plant-specific esse uence uncertanties E-10 F RA M AT 0 M E

I--)9

'-12D References E-1 Ed. M. A. Rutherford, N. M. Hassan, et al., "DORT, Two Dimensional Discrete Ordinates Transport Code, "BWNT-TM-107, Framatome Technologies, Inc.,

Lynchburg, Virginia, May 1995.

t E-2 L. A. Hassler and N. M. Hassan, "SORREL, DOT Input Generation Code User's Manual, " NPGD-TM-427, Revision 8, Framatome Technologies, Inc., Lynchburg, Virginia, July 1992.

E-3 U.S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, " Draft Re ulato Guide DG-1053, June 1996.

Code of Federal Regulation, Title 10, Part 50, "Domestic Licensing of Production and Utilization Facilities, " A endix H Reactor Vessel Material Surveillance Pro ram R~i.Eff ~:1 1 D 18, 1996.

E-5 D. T. Ingersoll, et al., "BUGLE-93, Production and Testing of the VITAMIN-B6Fine Group and the BUGLE-93 Broad Group Neutron/Photon Cross-Section Libraries Derived from ENDFIB-VINuclear Data, " ORNL-DLC-175, Radiation Safety Information Computational Center, Oak Ridge National Laboratory, Oak Ridge, 0 E-6.

Tennessee, April 1994.

L. A. Hassler and N. M. Hassan, "GIP Users Manual for B&W Version, Group Organized Cross Section Input Program, " NPGD-TM-456 Revision 11, Framatome Technologies, Inc., Lynchburg, Virginia, August 1994.

E-7 K. Y. Hour, "Evaluation of Carolina Power & Light Company Shearon Harris Capsule

"

X, 00:475-0188-01:02 FTG Document No. 31-1083271-01, B&W Services, Inc.,

Lynchburg, Virginia, May 1999.

J. R. Worsham III, "Fluence and Uncertainity Methodologies, " BAW-2241P Revision 1, Framatome Technologies, Inc., Lynchburg, Virginia, April 1999.

E-9. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities, " Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock, " Effective Date: August 28, 1996.

E-10 U.S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials, "Re ulato Guide 1.99 Revision 2, May 1998.

E-11 F RAMATOMF-

EAL STATUS BOARD FPB STATUS BOARD EAL REF. NO(S).:

GENERAL EMERGENCY FUEL RCS CONTAINMENT DE 1 INTACT FPB INTACT FPB INTACT FPB SITE AREA EMERGENCY ALERT UNUSUAL EVENT JEOPARDIZED JEOPARDIZED JEOPARDIZED CONTINUING ACTION TIME BREACHED BREACHED BREACHED BEGAN:

LOSS OF REQUIRED DURATION (MIN)

NP ALL ONSITE ESF DC BUSSES (125VDC IASA EXPIRES AND IBSB)

INDICATE FUEL INDICATE CNMT BREACHED ON FPB EAL TABLE 1 CURRENT CLASSIFICATION BREACHED ON FPB STATUS BOARD STATUS BOARD RADIATION MONITOR TAG NUMBERS NEW OR 'POTENTIAL'LASSIFICATION MONITOR DESCRIPTION INSTRUMENT RM-11 CHANNEL ERFIS POINT ALL INDI CA TE R CS IAG NO ID NO. ID NO. ON SITE ESF DC NO BREACHED ON FPB LOST FOR STATUS BOARD CNMT LEAK DETECTION REM-ILT-3502ASA 2GS303 RLT35028 > 15 MINs MONITOR ALERT EXCEEDED CNMT HI RANGE ACCIDENT RM-ICR-3589-SA 2AX162 RCR3589A MONITORS RM-ICR-3590-58 2AX163 RCR3590A CONTINUING ACTION:

SITE AREA EMERGENCY PLANT VENT STACK ITI RM-21 AV3509 ISA IEX817 RAV3509H IF CONDITION CON11NUES WRGM RlLJLk4 EXCEEDED F'R > 15 MINs, THEN SITE EFFLUENT CHANNEL AREA EMERG. EAL EXCEEDED INITIATE MONITORING OF CRI1ICAL SAFETY FUNCTION TURBINE BUILDING STACK 3A RM-I TV-3536-1 2EX827 RTV3536D STATUS TREES WRGM EFFLUENT CHANNEL NOTE MODES I 4, TOTAL IF ALBs = 30 MAIN STEAM LINE MONITORS RM IMS-3591 SB SAX164 RMS3591A NO ANY RAD MON RM-IMS-3592 SB 5AX165 RMS3592A MODES 5-6. TOTAL IF ALBs 20 IN EAL TABLE I, RM IMS-3593 SB 5AX166 RMS3593A (ALB I, 2, 4-13, 15, 22, 23, IN HIGH ALARM EITHER 24 OR 25 (BASED ON EDG FUEL BREACH AREA RAO. MONITORS: OPERABILITY), 26-28, 8 30)

VOLUME CONTROL TANK RM RM -I R R- 3595 IAA056 RRR3595A CHARGING PUMP IA ROOM RM-IRR-3599A IAA060 RRR3599A CHARGING PUMP 18 ROOM RM IRR-35998 IAA061 RRR35998 CHARGING PUMP IC ROOM RM IRR-3599C IAA062 RRR3599C LOSS OF PLANT VENT RECYCLE EVAP. VLV GAL RM-IRR-3600 IAA063 RRR3600A STACK tFI WRGM YES RM I RR-3601 1AA064 RRR3601A NO >75K DF MCB LETDOWN HX VALVE GAL ANNVNCIATORS EFFLUENT CHNL > 3.6E5 MODERATING HX VLV. GAL. RM I RR 3602 IAA065 RRR3602A uCI/SEC (ALBs)

STACK &5 MONITOR RM I WV3546-1 4EX837 RWV3546H WRGM EFFLUENT CHANNEL EITHER STACK IIISA MONITOR RM- IWJ-3547-1 4EX847 RWV3547H AFFECTED CNMT HI WRGM NO SYTEMS'RFIS RANGE ACCIDENT EFFLUENT CHANNEL DATA MON > 17.5 R/HR AVAILABLE TREATED LAUNDRY (c REM 1st 3540 4LL666 RWL3540A HOT SHOWER DISCHARGE YES NO SECONDARY WASTE SAMPLE REM-21WS-3542 4LL686 RWS3542A TANF DISCHARGE slsutFtGANT ANY EAL IRANSIENT (EOP TABLE 2 PARI I EN TERED.

MONITOR > IODO YES WASTE MONITOR TANK REM IWL-3541 4LL676 RWL3541A NO

>25K RUNBACK OR TIMES DISCHARGE >IGX POWER NORMAL osaLLAGGNS)

TURBINE BLDG. DRAIN REM -I LID3528 2LL276 RMD3528A SIGNIFICANT TRANSIENT (EGP NO YES PATH I ENTEREG, NO

>2SF RUNBACK OR

>10K POSER GSCIILATIGNS)

WAS YES ENTRY AT POINT EAL TABLE 2 LOSS OF ALBs T

FUEL BREACH AREA RAD MONITORS PREPLANNED NO ALBs NO

1. VOLIJME CONTROL TANK RM. LOST FOR NO INDICATE ALL 1HREE 2. CHARGING PUMP IA ROOM > 15 Mills FPB BREACHED ON 3 CHARGING PUMP 18 ROOM GFFD FPB STATUS BOARD 4. CHARGING PUMP IC ROOM INCREASEO > YES

5. RECYCLE EVAP. VLV. GAL.

I.OE5 CPM IN A(Bs NO YES

6. LEIDOWN HX VALVE GAL LOST FOR 30 MINs

7. MODERATING HX VLV. GAL > 15 MINs SITE AREA EMERGENCY

~A - EXCEEDED NO RCS ACDVITY (1-131 DOSE EQUIVALENT)

> 300 UCI/CC YES LAL TABLE 3 FUNCTIONS NEEDED TO ACHIEVE SHUTDOWN

~ ALERT EXCEEDED CONTINUING ACTION:

IF CONDITtON CONTINUES FOR > 15 MINs, THEN ALERT EAL EXCEEDED CONTINUING ACTION:

IF CONDITION CONTINUES FOR > 15 MINs, THEN SllE AREA EMERG. EAL EXCEEDED NO INDICATE FUEL MODE 4 5 MODE 3 BREACHED ON FPB STATUS BOARD IF RCS PRESS.

CAPABILITY FOR FW FLOW > 222.5 KPPH COMPLETE LOSS 1 360 PS(G ANY FUNCTION CORE COOLING CSF REO USTED ON EAL IF RCS PRESS. TABLE 3 INDICATE FUEL SECONDARY HEAT SINK CAPABILITY JEOPARDIZED ON 360 PS(G NO FPB STA1US BOARD RHR CAPABILITY INDICATE FUEL CCW CAPABIUTY INTACT ON FPB FUNCTION STATUS BOARD ESW CAPABILITY LOSS DUE CHARGING CAPABILITY TO LOSS OF ALL AC OR ALL STEAM GENERATOR SAFETY VALVES DC POWER BORATION CAPABILITY RCS PRESS. CONTROL CAPAE'ILITY NO ACCUM. ISOL. OR VENTING CAPABILITY

~

EOP PA1H-2

= REQUIRED LOST ENTERED X FUNCTION REQUIRED YES FOR MODE 3 ANY MAIN SITE AREA EMERGENCY NO STEAMLINE RAD YES EXCEEDED MON > 20 mR/HR NO CNMT LEAK DET RAD MON NOBLE YES LOST NO GAS CHNL > BE-3 3 FPBs BREACHED/ ES FVNCTION REQUIRED uCI/CC JEOPARDIZED FOR MODE 4 OR 5 GENERAL EMERGENCY

~

EXCEEDED NO NO ALERT DECLARE GENERAL EXCEEDEO WAS YES YES EMERGENCY ENTRY AT 2 FPBs BREACHED/

POINT JEOPARDIZED V

SITE AREA EMERGENCY NO BA~~ EXCEEDED MAKE NOTIFICATIONS NO A HOSTILE FORCE RE-EVALUATE EALS HAS TAKEN CONTROL INDICATE FUEL AND ND RCS LEAKAGE YES I FPB BREACHED/ AS CONDITIONS WARRANT OF PLANT EQUIPMENT AND/OR RCS BREACHED ON YES

> 50 GPM JEOPARDIZED VITAL AREA(S) SUCH THAT PLANT FPB STATUS BOARD PERSONNEL ARE UNIABLE TO OPERATE EQUIPMENT REQUIRED TO MAINTAIN GENERAL EMERGENCY INDICATE RCS SAFETY FUNCTIONS EXCEEDED BREACHED ON FPB STATUS BOARD ALERT

~A& EXCEEDED NO DECLARE GENERAL INTEGRITY CSF YES EMERGENCY MAGENTA OR RED INDICATE RCS NO JEOPARDIZED ON ATWS MAKE NOVFtCAHONS FPB STATUS BOARD WHILE IN EITHER OF THE FOLLOWING SECURITY INDICATE RCS MODE I OR 2 EVENTS WIIHIN A VITAL AREA:

INTACT ON FPB BOMB DISCOVERED IMTHIN A VITAL AREA POTENTIALLY STATUS BOARD AFFECTING SAFETY RELATED RE-EVALUATE EALS EQUIPMENT AS CONDITIONS WARRANT CONFIRMED INTRUSION INTO A VITAL AREA BY A HOSTILE FORCE MCB MANUAL REACTOR YES TRIP SUCCESSFUL 511E AREA EMERGENCY WAS (EtTHER SWITCFI) EXCEEDED ENTRY AT POINT V

~8 ALERT EXCEEDED NO NO NO Ntt)E EAL ALERT CONDIIION EXISTS ONLY FOR THE PERIOD EITHER OF THE FOLLOWIMG SECURITY IS WHILE THE RODS ARE OUT EVENTS WITHIN 1HE PA:

CNMT PHASE A BOMB DISCOVERED WITHIN 1HE YES OR VENT PA POTENTIALLY AFFECTING SOLATION REQUIRED' SAFETY RELATED EQUIPMENT IMMINENT THREAT OF. OR ACTUAL FUEL FPB INTRUSION INTO THE PA BY BREACHED NO

~4 GENERAL EMERGENCY A HOSTILE FORCE

~

EXCEEDED ALERT NO EXCEEDED BOTH FUEL AND RCS NO NO SITE AREA EMERGENCY INTACT ON FPB DECLARE GENERAL EXCEEDED STATUS BOARD EMERGENCY PAII-IWAY FOR SPENT FUEL POOL YES NO NO FISSION PRODUCTS TO NIIE LEVEL < I FT ABOVE ESCAPE CNMT EXISTS OTHER EAL EMERGENCY EXISTS MAKE NOTIFICATIONS TOP OF FUEL THAN SECONDARY SYSTEM ONLY FOR THE PERIOD (STEAM/FEED)

WHILE 11-IE RODS ARE OUT PRIMARY TO RE-EVALUATE EALS SITE AREA EMERGENCY NO SECONDARY LEAKAGE AS CONDITIONS WARRANT EXCEEDED IN ANY SG

> 10 GPM YES DAMAGE NO TO SPENT NO FIRE FUEL AFFECTED SG NO SAFETY VALVES SHUT YES FIRE MAY ANY SPENT FUEL NO AFFECT SAFETY NO POOL AREA AFFECTED SG NO RELATED (ESF) RAD MON PORV SHUT EQVIPMENT > t00 mR/HR CAN AFFECTED scs YES PDRV -OR- BLOCK VALVE BE SHUT COMPLETE ANY SPENT PJEL LOSS OF ANY POOL ARFA NO SAFETY RELATED RAD MON YES (ESF) FUNCTION > 700 mR/HR DUE TO FIRE ALERT ALERT 8&& TBJ-&EXCEEDED EXCEEDED N(ZTT'.

YES A BREAK IS A LEAK WHICH EXCEEDS THE OPERATORS ABILITY TO SITE AREA EMERGENCY SITE AREA EMERGENCY SHUTDOWN 1HE PLANT IN A CONTROLLED MANNER DR TO NOT EXCEEDED ~EXCEEDED EXCEED TECH SPEC COOLDOWN LIMIT'S NONISOLABLE STEAM AND/OR FEED YES TA-SA OR 18 ENERGIZED 58

) NO PLANT IS IN BREAK OUTSIDE CNMT MODE 6 IN AFFECTED SG NO NO INDICATE CNMT BREACHED ON FPB SG PRESS STATUS BOARD NO RCS PRESSURE 360 PSIG

) NO VALID CNMT VENT ISOL ACTUATION

> 1230 PSIG YES NO SG LEVEL NO

> 82.48

> 222.5 KPPH FEED ~I)0 FLOW AVAILABLE 8011-1 CNMT NO Hl RANGE ACCIDENT CNMT YES MON > 6.5 R/HR

> 3 PSIG FULL RANGE NO ALERT

~

YES EXCEEDED RVLIS LEVEL 622 INDICATE CNMT JEOPARDIZED ON GENERAL EMERGENCY NO FPB STATUS BOARD fkLJkhk EXCEEDED SITE AREA EMERGENCY INDICATE CNMT EXCEEDED INTACT ON FPB STATUS BOARD DECLARE GENERAL EMERGENCY IA-SA AND 18 SB m tlO LOST FOR

> 15 MINs ALERT MAKE NOTIFICATIONS B&L ILL&E'ICEEDED SITE AREA EMERGENCY CONTINUING ACTION: RE-EVALUATE EALS EXCEEDED IF COtlDITION CONTINUES AS CONDITIONS WARRANT FOR > 15 IJINs, THEN SITE HARRIS NUCLEAR PLANT AREA EMERG EAL EXCEEDED EMERGENCY ACTION LEVEL FLOW PATH SIDE 1 REVISION: 99 1

EAL STATUS BOARD EAL TABLE 4 EAL TABLE 5 EAL TABLE 6 EAL REF. NO(S).;

DOSE PROJECTION MONITORS EFFLUENT MONITORS SEISMIC EVENT SYMPTOMS

1. PLANT VENT STACK 91 WRGM EFFLUENT CHANNEL GENERAL EMERGENCY

1. CNMT HI RANGE ACCIDENT MONITORS 2 TVRBINE BUILDING STACK Jtsa WRGM EFFLUENT CHANNEL

1. SEISMIC MON SYS OBE EXCEEDED" ALARM (ALB ID-4-4)

2. PLANT VENT STACK gl WRGM EFFLUENT CHANNEL

3. MAIN STEAM LINE MONITDRS 3 STACK $ 5 WRGM EcFLUENT CHANNEL AND/OR 'ALARM" LIGHT ON THE SEISMIC SWITCH PANEL. SITE AREA EMERGENCY 4 STACK 95 MONITOR WRGM EFFLUENT CHANNEL 4. STACK gsa WRGM EFFLUENT CHANNEL 2. WHITE EVENT INDICATOR ON SMA CONTROL PANEL. ALERT

5. STACK PSA MONITOR WRGM EFFLUENT CHANNEL 5. IREATED LAUNDRY AND HOT SHOWER DISCHARGE MONITOR

6. SECONDARY WASTE SAMPLE TANK DISCHARGE MONITOR 3. ALARM AT TRIAXIAL RESPONSE SPECTRUM ANNUNCIATOR. UNUSUAL EVENT

7. WAS1E MONITOR TANK DISCHARGE MONITOR

4. NOWCEABLE TREMORS OR MBRATION 8 IURBINE BLDG. DRAIN MONITOR FROM CONTINUING ACTION TIME BEGAN:

REQUIRED DURATION (MIN)

EXPIRES:

ANY TWO CURRENT CLASSIFICATION INDICATE CURRENT EAL No INDICATIONS OF ON EAL STAlUS BOARD A SEISMIC EVENT NEW OR 'POTENTIAL'LASSIFICATION LISTED ON EAL TABLE 6 YES No ANY RAD MON ON EAL TABLE 4 IN HIGH ALARM ANY YELLOW No UGHT ON TRIAXIAL RESPONSE SPECTRUM EMERGENCY ANNUNCIATOR DOSE PROJECTIONS LIT No ANY OR ENVIRONMENTAL MONITORING RESULTS RADIOLOGICAL

~

AVAILABLE CONDITION WARRANTING YES RECOMMENDATION To EVACUATE OR SHELTER ANY RED lHE PUBLIC LIGHT ON TRIAXIAL GENERAL EMERGENCY RESPONSE SPECTRUM EXCEEDED ANNUNCIATOR LIT EMERGENCY ALERT No DOSE PROJECTION YES CALMEXCEEDED ANY DECLARE GENERAL RESULTS PLANT CONDITION EMERGENCY AVAILABLE SITE AREA EMERGENCY THAT IN THE JUDGEMENT PERFORM DOSE ILL)-4-8 EXCEEDED No OF THE SUPERINTENDENT SHIFT PRoJEcnoN FER OPERATIONS OR SITE EMERGENCY PEP-343 YES COORDINATOR WARRANTS A MAKE NOTIFICATIONS SITE AREA EMERGENCY oEcLananoN

~

PROJECTED DOSE >1000 MREM TEDE YES AT OR BEYOND SITE RE-EVALUATE EALS BOUNDARY No ADVERSE WEATHFR AS CONDITIONS WARRANT RE-EVALUATE EALS AT SITE AREA EMERGENCY ENTRY POINT Y WHEN DOSE EXCEEDED PROJECTIONS COMPLETED No YES PROJECTED DOSE >5000 MREM THYROID CDE AT OR BEYOND SITE BOUNDARY No YES

~

GENERAL EMERGENCY EXCEEDED DECLARE GENERAL TORNADO HAS HIT THE POWER BLOCK YES YES PROJEC1ED OR MEASURED SUSTAINED WIND SPEED AT 10 AIRBORNE RAD LEIIELS INDICATE SEVERE DEGRADATION IN RADIOACTIVE MATERIAL CONTROL No EMERGENCY PROJECTED DIOSE METERS > 90

> 50 MREM TEDE AT OR MPH ANY BEYOND SITE BOUNDARY SUSTAINED WIND No PLANT CONDIWON USING ADVERSE MET MAKE NOTIFICATIONS NP EXISTS THAT IN THE SPEEDS AT 10 METERS

~

DATA JUDGEMENT OF THE SUPERINTENDENT YES

> 100 MPH SHIFT OPERATIONS OR SITE EMERGENCY ALERT EXCEEDED COORDINATOR WARRANTS No RE-EVALUATE EALS AN ALERT AS CDNDITIONS WARRANT At.ERT DECLARATION EXCEEDED SITE AREA EMERGENCY PROJECTED DOSE EXCEEEED No

> 250 MREM THYROID CDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA HAS A ESTIMATED No YES SITE AREA EMERG OR DURATION OF No AIRCRAFT CRASH ALERT BEEN RELEASE > 30 MINs No MISSILE IMPACT OII EXCEEDED UNPLANNED EXPLOSION EVALUATE AGAINST INSIDE POWER UNUSUAL EVENT MATRIX BLOCK DECLARE HIGHEST YES EMERGENCY ON EAL STATUS BOARD PROJECTED DOSE ANY UNUSUAL EVENT No

> 500 MREM TEDE AT PLANT IN MODES MAKE CONDITION IDENTIFIEO OR BEYOND SITE BOUNDARY I, 2, 3, OR 4 NonFtcanons USING ADVERSE MET No EMERGENCY DATA DECLARATION YES RE-EVALUATE EALS AS CONDITIONS WARRANT DECLARE UNUSUAL EVENT CONSULT AP-617, ERC-003. PLP-500, PLP-502 SAFETY RELATED No AND SP 017 FOR OTHER PROJECTED DOSE EQUIPMENT OR REPORTABILITY MAKE

> 2500 MREM THYROID STRUCTURE REQVIRMENTS NOTIFICAWONS CQE AT CR BEYOND SITE BOUNDARY AFFECTED ALERT USING ADVERSE MET DATA ~EXCEEDED ESTIMATED RE-EVALUATE EALS

~

DURATION OF YES AS CONDITIONS WARRANT SITE AREA EMERGENCY No RELEASE > 2 MINs ~BEXCEEDED SITE AREA EMERGENCY EXCEEDED EAL DECLARATION TIME UNCONTROL'D OR UNPLANNED No RELEASE OF TOXIC OR FLAMMABLE GAS INTD PDWER BLOCK ENVIRONMENTAL REF EAL TABLE MONITORING RESULTS 7 AVAILABLE RE-EVALUATE EALS AT YES ENTRY POINT Y WHEN ENVIRONMENTAL MONITORING YES RESULTS AVAILABLE AFFECTED AREA HOUSES SAFETY RELATED MEASURED DOSE EQUIPMENT RAlE > 1000 MREM/HR AT OR BEYOND SITE BOUNDARY YES No GAS IS FLAMMABLE OR No LACK OF ACCESS

~

MEASURED 1-131 EQUIV CONC IS A SAFETY

> 39E-6 uclfcc PROBLEM AT OR BEYOND SITE BOUNDARY GENERAL EMERGENCY EXCEEDED No PLANT IN MODE DECLARE GENERAL I, 2, 3 OR 4 EAL TABLE 1 EMERGENCY

~EXCEEDED ALERT RADIATION MONITOR TAG NUMBERS MAKE NOTIFICATIONS MONITOR DESCRIPTION INSTRUMENT RM I'I CHANNEL ERFIS POINT TAG No. ID No. ID NO.

SITE AREA EMERGENCY MEASURED OSE RATE > 50 MREM/HR

~EXCEEDED CNMT LEAK DETECTION REM-ILT-3502A-SA 2GSJD3 RLT35028 AT OR BEYOND SITE RE-EVALUATE EALS MONITOR BoVNoaRY AS CONDITIONS WARRANT CNMT HI RANGE ACCIDENT RM-ICR-3589-SA 2AX162 RCR3589A MONITORS RM-ICR-3590-58 2AX163 RCR3590A No CONTROL PLANT VENT STACK gt RM-21AV-3509-ISA IEX817 RAV3509H MEASURED ROOM EVAC REQUIRED WRCM 1-131 EQUIV CONC OR ANTICIPATED EFFLUENT CHANNEL

> 1.9 6-7 <<CI/cc AT OR 4EYONG SITE BOUNDARY RM 1 TV-3536-1 2EX827 RTV3536D TURBINE BUILDING STACK 3A WRGM YES EFFLUENT CHANNEL No MEASURED YES LEVEL HAS EXISTED 5AX164 RMS3591A MAIN STEAM LINE MONITORS RM IMS 3591 SB

> 30 MINs AVX CONTROL RM-IMS-3592-58 5AX165 RMS3592A No PANEL (ACP) RM-IMS 3593 SB 5AX166 RMS3593A OPERATIONAL No FUEL BREACH AREA RAD. MONITORS.

VOLUME CONTROL TANK RM. RM-IRR3595 IA AD 56 RRR3595A CONTINUING ACTION: CONIROL IF CONDITION CONTINUES No CHARGING PUMP IA ROOM RM-IRR-3599A IAA060 RRR3599A ROOM EVACUATED FOR > 30 MINs THEN SITE FOR > 15 MINs CHARGING PUMP 18 ROOM RM-IRR35998 IAA061 RRR35998 AREA EMER. EAL EXCEEDE CHARGING PUMP IC ROOM RM-IRR-3599C IAA062 RRR3599C RECYCLE EVAP VLV GAL RM-IRR-3600 IAA063 RRR3600A LETDOWN HX VALVE GAL. RM IRR-3601 IAA064 RRR3601A MODERATING HX VLV GAL RM IRR-3602 IAA065 RRR3602A CONTINUING ACTION:

IF CONDITION CONTINUES STACK 95 MONITOR RM IWV-3546-1 4EX837 RWV3546H MEASURED FOR > 15 MINs THEN SITE YES WR0 M DOSE RATE > 500 MREM AREA EMER. EAL EXCEEDE EFFLUENT CHANNEL

/HR AT OR BEYOND stTE BoUNoaRY STACK 65A MONITOR RM-IWV-3547I 4EX847 RWV3547H WRGM No EFFLUENT CHANNEL MEASURED 1-131 EQVIV CONC

~4- ALERT EXCEEDED SITE AREA EMERGENCY 88LL~-4EXCEEDED LIQUID EFFLUENT MONITORS TREATED LAUNDRY 4 HOT SHOWER DISCHARGE REM IWL-3540 4LL666 RWL354DA

> t.s T.-e <<CVcc ac 09 Se<<O<<9 BOUNDARY SITE SECONDARY WASTE SAMPLE REM 21WS 3542 4LL686 RWS3542A TANK DISCHARGE MEASURED REM-'IWL3541 4LL676 RWL3541A WASTE MONITOR TANK LEVEL HAS EXISTED DISCHARGE FOR > 2 MINI SITE AREA EMERGENCY TURBINE BLDG. DRAIN REM-IMD-3528 2LL276 RMD3528A BARB EXCEEDED No MONITOR IN EAL TABLE 5 READING No EAL TABLE 7

> 10 TIMES THE HIGH ALARM TOXIC, FLAMMABLE, (Ic ASPHYXIANT GASES SET POINT

~ ALERT EXCEEDED ACETYLENE HYDROGEN GAS (I)

X C

X Y?+

8+

YY e"X AP-501-00610 AP-501-001 29 AP-501-00085 CHEMICAL FACT 5HEET NUMBER CARBON DIOXIDE AMMONIA X X AP-501 00637 P-10 (USED IN PORTAL MONITORS) AP-501-00591 OXYGEN (2) AP-501 OOD68 Ict >OGEN AP-5C/1 (tol 30 ARGON AP-501-00319 HELIUM X AP 501 00573 FREON X AP 501 00987 THIS LIST INCLUDES GASES STORED IN BULK ANO IS NOT INTENDED To BE ALL INCLUSIVE (I) VENTING OF H2 FROM THE H2 TANK VENT STACK OR H2 HEADER IN ACCORDANCE WITH DESIGN AND IS A PLANNED EVENT.

(2) OXYGEN ITSELF IS NOT FLAMMABLE, HOWEVER, ITS PRESENCE INCREASES THE FLAMMABILITYOF MATERIALS.

ASPHYXIANT GASES IN LARGE ENOUGH QUANDTIES CAN DISPLACE OXYGEN AND POSE A DANGER To PERSONNEL AND SHOULD BE CONSIDERED A TOXIC THE TYPE AND QUANTITY OF GAS RELEASED, VOLUME OF AREA, AND VENTILATION SYSTEMS IN SERVICE SHOULD BE EVALUATED To DEIERMINE WHETHER PERSONNEL OR EQUIPMENT COULD BE IN AN ADVERSE ENVIRONMENT, OR WHETHER ACCESS To EQUIPMENT REQUIRED To OPERATE THE PLANT IS IMPEDED.

UNUSUAL EVENT MATRIX 8)ODIER PLANT 9)NATURAL 10) OTHER HAZARDS 11) SITE EMERGENCY DEFINITIONS OF AREAS REFERRED To IN THESE EAL st POWER BLOCK INCLUDES 1?IE CONTAINMENT, REACTOR 3)LOSS OF 4)LOSS OF REACTOR 5)LOSS OF 6)LOSS OF MCB ANNUNCIATORS, 7) SECURITY THREAT OR I) GASEOUS OR LIQVID 2)LOSS OF SECONDARY COOLANT POWER ERFIS OR COMMUNICATIONS EQUIPMENT PHENOMENA COORDINATOR AUXIVARY BLDG., TURBINE BLDG, FUEL HANDLING EFFLUENT(S) SECONDARY PROBLEMS JUDGMENT COOLANT OR COOLANT OR CAPABILITY BLDG. (INCLUDING THE K AREA). WASTE PROCESSING EXCEEDING TECHNICAL COOLING COOLING BLDG., DIESEL GENERATOR BLDG., DIESEL FUEL OIL SPECIFICATIONS CAEMJ LOSS OF ALL UNPLANNED LOSS OF >75K OF MCB B!LEACH CONFIRMED SECURITY 88~

INAB'LITl'o REACH INDICATION Of'NY FIRE WITHIN THE PROTECTED AREA CALJbhl OTHER PLANT STORAGE. FIRE HOUSE, TANK AREA. INTAKE STUCTURES.

AND DUCT BANKS SERVING THESE AREAS.

VALID HIGH ALARM ON GROSS FAILED FUEL RAPID FAILURE OF A OFFSITE POWER ANNUNCIATORS (ALBS) FOR >15 EVENT WHICH RE')UIRED SHUTDOWN TWO VALID SEISMIC NOT EXTINGVISHED WITHIN 15 CONDITIONS EXIST 1HAT ANY OF THE MONITORS DE1ECTOR INDICATES AN DEPRESSURIZATION PRESSURIZER PROTECTED AREA INCLUDES 1HE PLANT AREA INSIDE THE MINUTES AS DEFINED BY: INDICATES A (MODE 3) CONDITION SYMPTOMS USTED ON MINUTES OF CONTROL ROOM WARRANT INCREASED IN EAL TABLE 5 INCREASE > 2E4 CPM OF 51EAM SAFETY OR PORV To

• MODES I 4 TOTAL 9 ALBS = 30 POTENTIAL WtltttN TECH SPEC EAL TABLE 6 NOWFICATION OR VERIFICATION OF A AWARENESS ON 1HE SECURITY FENCE INCLUDING THE INTAKE STRUCTURES.

WITHIN 30 MINUTES GENERATOR CLOSE FOLLOWING MODES 5 6 TOTAL 8 ALBS = 20 DEGRADADON IN THE DVE UMITS CONTROL ROOM ALARM PART OF THE PLANT SECONDARY SIDE REDUCTION OF AND ME RELEASE HAS NOT APPLICABLE PRESSURE LOSS OF BOTH EMERGENCY DIESEL (ALB 1,2, 4-13, 15, 22, 23, EITHER 24 OR 25 BASED ON EDG LEVEL OF SAFETY OF THE PLANT AS INDICATED BY: Ckt JHLJ

~4 SUSTAINED WIND (MIS DOES NOT INCLUDE FIRES 941HIN OFFICE AREAS, TRASH BIN OPERATION STAFF, CHATHAM COUNTY, HARNETT COUNTY LEE SITE BOUNDARY (SB) IS APPROXIMATELY A CIRCLE OF 2500 FT. RADIUS (0.47 MlLES)

BEEN TERMINATED RCS SPECIFIC ACTIVITY OPERABILITY, 26 28, dc 30) SPEED AT 10 METERS FIRES, H2 TANK VENT STACK FIRES COUNTY. WAKE COUNTY, GENERATORS <<VNAUTHORIZED IN/ 0 VER TENT EXCLU5ION AREA BOUNDARY (EAB) IS APPROXIMATELY (UNUSUAL EVENT EXISTS EXCEEDS TECHNICAL SPECIFICADON 3.4.8 FAILURE OF A SG SAFETY OR PORV To BLAH ALTERATION OR CRITICALITY OF 74 MPH OR GREATER EXDNGVISHED PER OP-152.02 OR OTHER SMALL FIRES OF No PLANT OR THE STATE OF NORTH CAROLINA A CIRCLE OF 7000 FT. RADIUS (1.3 MILES)

FULLY RESET AFTER ANY RCS PRESSURE TAMPERING HAS OR EXTENDED AND UNTIL EFFLUENT LIMITS FOR DOSE BOUNDARY LEAKAGE IS OCCURRING UNPLANNED SUSTAINED SAFETY CONSEQUENCE)

DISCHARGE IS EQUIVALENT 1-131 OR OPERATION INABILITY OF ERFIS To PERFORM ITS TERMINATED AND WHEN ALL REQUIRED GROSS RADIOACTIVTY INTENDED FUNCTION FOR A CONTINUOUS PERIOD OF 4 HOURS, AFFECTING SAFETY RELATED POSHIVE STARTUP RATF ~4A TORNADO REPORTED EAL REFERENCE NUMBERS (X-Y-2)'

EQUIPMENT (THIS DOES NOT NonocanoNs aRE (F'R DOSE EQUIVALENT ANY 01HER RCS OTHER THAN PREPLANNED REMOVAL

• HOSTAGE/EXTORTION WITHIN THE EAB AIRCRAFT, TRAIN OR OTHER VEHICLE 131 THE EAL IS NOT INCLUDE CRITICALITY MADE) I MAIN SIEAM LINE OR LEAKAGE IN EXCESS FROM SERVICE FOR MAINTENANCE OR SIIUATION THAT EARLIER THAN CRASH THAT MAY DAMACE PLANT CATEGORY (1-11)

EXCEEDED UNLESS THE FEEDWATER LINE MODIFICATION PURPOSES WHILE IN Y = IDENWFIER WITHIN CATEGORY OF lECHNICAL THREATENS To ESRMATED DURING STRUCTURES CONTAINING FUNCTIONS 48 HOUR TIME BREAK SPECIFICATION MODES 1,2,3 OR 4 AS DEFINED BY: OR SYSTEMS REQUIRED FOR SAFE CLASSIFICATION (1-4)

INTERRUPT NORMAL PLANNED REACTOR 2 INTERVAL, OR FIG. 3.4-1 FAILURE OF BOTH OPUS SHUTDOWN OF THE PLANT 3.4.6.2 WllH THE 4 PLANT OPERATIONS STARTUPS) I = UNUSUAL EVENT LIMITS ARE EXCEEDED) (A BREAK IS A LEAK HOUR CORRECTIVE FAILURE OF BOTH DATA

• CIVIL DISTURBANCE WHICN EXCEEDS lHE 2 ALERT Acnons Nof CONCENTRATORS ONGOING BETWEEN 3 SITE AREA EMERGENCY OPERATORS ABILITY SAWSFIED FAILURE OF BOM DATA DISCS THE SITE BALkLt 4 = GENERAL EMERGENCY To SHUTDOWN THE INABILITY To DISPLAY SPDS IN BOUNDARY AND TV<<BINE ROTAWNG UNPLANNED EXPLOSION WITHIN THE PLANT IN A THE CONTROL ROOM THE PROTECTED COMPONENT FAILURE PROTECTED AREA RESULTING IN CONTROLLED INABIUTY To UPDATE CURRENT AREA RE<<JLTING IN A IIISIBLE DAMAGE To PERMANENT MANNER OR To NOT DATA DISPLAYS IN THE CONTROL REA TOR TRIP, CASING STRUCTURES OR EQUIPMENT EXCEED TECH. SPEC ROOM PENETRATION OR COOLDOWN LIMITS) (THIS IS NOT To BE CONSTRUED AS SICHIFICANT DAMAGE A FAILURE OF A SINGLE VARIABLE OR TO THE MAIN SMALL DATA SUBSET) C4NERATOR SEALS UNPLANNED TOXIC OR FLAMMABLE GAS RELEASE WITH ME EAB STEAM GENERATOR BLOWDOWN LINE (REFERENCE EAL TABLE 7)

HARRIS NUCLEAR PLANT BREAK FAILURE OF BOTH SITE TELEPHONE (MODES 1,2 ANID 3) AND EMERGENCY (HEJ<<EC) TELEPHONE SWITCHES EMERGENCY ACTION LEVEL FLOW PATH SIDE 2 REVISION: 99-1

EAL STATUS BOARD FPB STATUS BOARD EAL REF. NO(S).:

FUEL RCS CONTAINMENT GENERAL EMERGENCY FPB FPB FPB SITE AREA EMERGENCY ALERT INTACT IN TACT INTACT UNUSUAL EVENT JEOPARDIZED JEOPARDIZED JEOPARDIZED CONTINUING ACTION: TIME BREACHED BREACHED BREACHED BEGAN:

LOSS OF REQUIRED DURATION (MIN):

ALL ONSITE ESF DC eussss (12svoc lass EXPIRES:

INDICATE FUEL INDICATE CNMT AND 1858)

BREACHED ON FPB BREACHED ON FPB EAL TABLE 1 CURRENT CLASSIFICATION STATUS BOARD STATUS BOARD RADIATION MONITOR TAG NUMBERS YES NEW OR POTENTIAL'LASSIFICATION INDICATE RCS MONITOR DESCRIPTION INSTRUMENT RM-11 CHANNEL ERFIS POINT ALL TAG No. ID No. ID No.

BREACHED ON FPB ON SITE ESF DC No STATUS BOARD LOST FOR CNMT LEAK DETECTION REM ILT-3502ASA 2GS303 RLT35028 > 15 MINs MONITOR ALERT EXCEEDED CNMT HI RANGE ACCIDENT RM-ICR-3589-SA 2AX162 RCR3589A MONITORS RM-ICR-3590-58 2AX163 RCR3590A CONllNVING ACTION:

PLANT VENT STACK 4( RM-21AV-3509-ISA 1EX817 RAV3509H SITE AREA EMERGENCY IF CONDITION CON1INUES WRGM EXCEEDED FOR > 15 MINs, THEN SITE EFFLUENT CHANNEL AREA EMERG. EAL EXCEEDED INITIATE MONITORING OF CRITICAL SAFETY FUNCTION STATUS TREES TURBINE BUILDING STACK 3A RM-IIV-3536I 2EX827 RTV3536D WR0 M EFFLUENT CHANNEL NOTE MAIN STEAM LINE MONITORS RM-HAS-3591 SB SAX164 MODES 1-4, TOTAL IT ALBs 30 RMS3591A ANY RAD MON NO RM-UAS-3592-58 5AX165 RMS3592A MODES 5-6. TOTAL 8 ALes 20 IN EAL TABLE I, RM HAS-3593 Se 5AX166 RMS3593A (ALB I, 2. 4 13, 15. 22. 23.

IN HIGH ALARM EITHER 24 OR 25 (BASED ON EDG FUEL BREACH AREA RAO. MONITORS: DPERABIUTY), 26-28, 8 30)

VOLUME CONTROL TANK RM. RM-IRR-3595 IAA056 RRR3595A CHARGING PUMP IA ROOM RM -IRR -3599 A laaoeo RRR3599A CHARGING PUMP 18 ROOM RM-IRR-35998 IAA061 RRR35998 PLANT VENT CHARGING PUMP IC ROOM RM-IRR-3599C IAA062 RRR35990 RECYCLE EVAP. VLV GAL. RM IRR-3600 IAA063 RRR3600A LOSS OF STACK gl WRGM YES >75K OF MCB LETDOWN HX VALVE GAL. RM-IRR-3601 IAA064 RRR3601A No EFFLUENT CHNL > 3.6E5 RM-IPR-3602 ANNVNCIATORS MODERATING HX VLV. GAL IAA065 RRR3602A uCI/SEC (ALes)

STACK $ 5 MONITOR RM I WV-3546-1 4EX837 RWV3546H NO WRGM EFFLUENT CHANNEL EIMER STACK $ 5A MONITOR RM-IWV-3547I 4EX847 RWV3547H CNMT HI AFFECTED WRGM RANGE ACCIDENT EFFLUENT CHANNEL SYTEMS'RFIS No MON > 17.5 R/HR DATA AVAILABLE LIOLI'D EFFLUENT MONITORS TREATED LAUNDRY 8c REM-;WL-3540 4LL666 RWL3540A No HOT SHOWER DISCHARGE SECONDARY WASTE SAMPLE REM-21WS-3542 4LL686 RWS3542A ANY EAL TAIAV. DISCHARGE SIGNIFICANT TABLE 2 WIAWSIENT (EOP YES MONITOR > 1000 WASTE MONITOR TANK REM-IWL-3541 4LL676 RWL3541A No PATH I ENTERED, TIMES DISCHARGE >25K RUN6ACK OR NORMAL >10K POWER OSCILLA h ON 9)

TURBINE BLDG. DRAIN REM-IMD-3528 2LL276 RMD3528A SIGNIFICANT NO TRANSIENT (EOP No PAIH I ENIERED.

>sss Ruueach oh

>IOX POWER WAS OSCILLATIONS)

YES ENTRY AT POINT T

EAL TABLE 2 LOSS OF aces FUEL BREACH AREA RAO MONITORS PREPLANNED No

1. VOLUME CONTROL TANK RM. ALBs No No LOST FOR INDICATE ALL THREE 2. CHARGING PUMP IA ROOM

> 15 MIN>

GFFD FPB BREACHED ON 3 CHARGING PUMP 18 ROOM INCREASEO > FPB STATUS BOARD 4 CHARGING PUMP lc ROOM I.OES CPM IN 5. RECYCLE EVAP. VLV. GAL.

ALes No 30 MINs 6 LETDOWN HX VALVE GAL. LOST FOR

7. MODERATING HX VLV. GAL. > 15 MINs SNE AREA EMERGENCY No gkL&hg EXCEEDED Rcs ach>4 TY (1-131 DOSE EQUIVALENT)

> 300 VCI/CC No YES EAL TABLE 3 FUNCTIONS NEEDED TO ACHIEVE SHUTDOWN

~ ALERT EXCEEDED CONTINUING ACTION; IF CONDITION CONTINUES FOR > 15 MINs, THEN ALERT EAL EXCEEDED CONTINUING ACTION:

IF CONDITION CONIINUES FOR > 15 MINs, THEN SITE AREA EMERG. EAL EXCEEDED INDICATE FUEL MODE 4-5 MODE 3 BREACHED ON FPB STATUS BOARD IF RCS PRESS.

CAPABILITY FOR FW FLOW > 222.5 KPPH COMPLElE LOSS CORE COOLING CSF

> 360 Pslo NO ANY FUNCTION REO LISTED ON EAL INDICATE FVEL SECONDARY HEAT SINK CAPABILITY IF Rcs PRESS TaeLE 3 JEOPARDIZED ON 360 Pslo No FPB STATUS BOARD RHR CAPABILITY INDICATE FVEL CCW CAPABIVTY INTACT ON FPB STATUS BOARD ESW CAPABILITY FUNCTION LOSS DVE CHARGING CAPABILITY To LOSS STEAM GENERATOR SAFETY VALVES OF ALL AC OR ALL DC POWER BORATION CAPABILITY RCS PRESS. CONTROL CAPABIUTY ACCUM, ISOL. OR VENTING CAPABX ITY EOP PATH-2

~

ENTERED X REQUIRED LOST FUNCTION REQUIRED FOR MODE 3 ANY MAIN No STEAMLINE RAO YES SITE AREA EMERGENCY MON > 20 mR/HR EXCEEDED No CNMT LEAK DET RAD MON NOBLE YES GAS CHNL > BE-3 YES LOST 3 FPBs BREACHED/

uCI/CC JEOPARDIZED FUNCTION REQUIRED FOR MODE 4 OR 5 GENERAL EMERGENCY

~

No EXCEEDED No YES ALERT WAS DECLARE GENERAL YES EXCEEDED ENTRY AT 2 FPes BREACHED/ YES EMERGENCY POINT JEOPARDIZED V

SITE AREA EMERGENCY No gaae ExcEEDED MaKE Nooncahous No A HOSTILE FORCE INDICATE FUEL AND RE-EVALUATE EALS HAS TAKEN CONTROL RCS LEAKAGE YES I FPB BREACHED/ as coNoroous waRRaNT RCS BREACHED ON OF PLANT EQUIPMENT AND/OR

> 50 GPM JEOPARDIZED FPB STATUS BOARD VITAL AREA(s) SUCH THAT PLANT PERSONNEL ARE UNABLE To OPERATE EQUIPMENT INDICATE Rcs YES REQUIRED To MAINTAIN GENERAL EMERGENCY No BREacHEo oN n e SAFETY FUNCTIONS EXCEEDED STAlVS BOARD ALERT EXCEEDED INTEGRITY CSF YES NO DECLARE GENERAL MAGENTA OR RED EMERGENCY INDICATE Rcs No JEOPARDIZED ON ATWS FPB STATUS BOARD YES WHILE IN MAKE NOTIFICATIONS INDICATE RCS EITHER OF THE FOLLOWING SECURITY MODE I OR 2 EVENTS WIMIN A VITAL AREA:

INTACT ON FPB BOMB DISCOVERED WITHIN A STATUS BOARD VITAL AREA POTENTIALLY AFFECTING SAFETY RELATED RE-EVALUATE EALS No EQUIPMENT AS CONDITIONS WARRANT CONFIRMED INTRUSION INTO A VITAL AREA BY A HOSTILE FORCE MCB

~

MANUAL REACTOR IRIP SUCCESSFUL WAS (EITHER SWITCH) SITE AREA EMERGENCY YES ENTRY AT ALERT EXCEEDED POINT EXCEEDED V

No No ~NT EAL ALERT CONDITION EXISTS ONLY FOR THE PERIOD IS EIMER OF ME FOLLOWING SECURITY CNMT PHASE A WHILE THE RODS ARE OVT EVENTS WITHIN 1HE PA:

ES OR VENT BOMB DISCOVERED WITHIN 1HE YES SOLA110N REQUIREDT PA POTENTIALLY AFFECTING

~

SAFETY RELATED EQUIPMENT FUEL FPB ~ IMMINENT THREAT OF. OR ACTUAL BREACHED INTRUSION INTO THE PA BY A HOSTILE FORCE GENERAL EMERGENCY

~

EXCEEDED ALERT eoM No EXCEEDED FUEL AND Rcs NO INTACT ON FPB SITE AREA EMERGENCY No EXCEEDED DECLARE GENERAL STATUS BOARD EMERGENCY PAMWAY FOR SPENT FUEL POOL YES FISSION PRODUCTS To N(LTE No LEVEL g 1 FT ABOVE ESCAPE CNMT EXISTS OTHER THAN SECONDARY SYSTEM EAL EMERGENCY EXISTS MAKE Nohncahous TOP OF FUEL ONLY FOR THE PERIOD

~

(STEAM/FEED)

WHILE THE RODS ARE OUT PRIMARY TO RE-EVALUATE EALS No SECONDARY LEAKAGE AS CONDITIONS WARRANT SITE AREA EMERGENCY IN ANY SG EXCEEDED

> 10 GPM DAMAGE NO No TO SPENT AFFECTED SG NO FIRE FUEL SAFETY VALVES SHUT YES FIRE MAY NO AFFECT SAFETY ANY SPENT FUEL AFFECTED SG No RELATED (ESF) NO POOL AREA PORV SHUT EQUIPMENT RAD MON 100 mR/HR CAIA AFFECTED poev -oh- ecocK No VALVE BE SHUT COMPLETE LOSS OF ANY ANY SPENT FUEL

~

No POOL AREA No SAFETY RELATED RAD MON (ESF) FUNCTION OVE To FIRE 700 mR/HR ALERT

~DEXCEEDED ALERT YES EXCEEDED A BREAK IS A LEAK WHICH EXCEEDS ME OPERATORS ABILITY TO SHUTDOWN THE PLANT IN A SITE AREA EMERGENCY CONTROLLED MANNER OR To NOT ~EXCEEDED J SITE AREA EMERGENCY

~EXCEEDED EXCEED TECH SPEC COOLDOWN LIMITS NONISOLABLE YEs IA-SA OR ENERGIZED IB-SB ')

STEAM AND/OR FEED YES BREAK OUTSIDE CNMT No PLANT IS IN IN AFFECTED SG MODE 6 No No INDICATE CNMT BREACHED ON FPB SG PRESS YES STATUS BOARD No Rcs PRESSURE 360 PSIG

) NO VALID CNMT VENT

> 1230 PSIG ISOL ACTUATION YES No SG LEVEL No

> 82.48

> 222.5 KPPH FLOW AVAILABLE FEED 'O 801H CNMT CNMT YES Hl RANGE ACCIDENT

> 3 PSIG MON > 6.5 R/HR

~

FULL RANCE No ALERT YES

~

RVLIS LEVEL EXCEEDED INDICATE CNMT > 622 No JEOPARDIZED ON GENERAL EMERGENCY FPB STA1VS BOARD EXCEEDED INDICATE CNMT SITE AREA EMERGENCY YES INTACT ON FPB EXCEEDED STAlus BOARD DECLARE GENERAL EMERGENCY IA-SA AND IB-SB No LOST FOR

> 15 MIN>

ALERT MaKE NOIIncaTIONS EXCEEDED SITE AREA EMERGENCY CONTINUING ACTION. RE-EVALUATE EALS EXCEEDED IF CONDITION CONTINUES AS CONDITIONS WARRANT FOR > 15 MINs, 1HEN SITE AREA EMERG. EAL EXCEEDED HARRIS NUCLEAR PLANT EMERGENCY ACTION LEVEL FLOW PATH SIDE 1 REVISION: 99-1 4

EAL STATUS BOARD EAL TABLE /I EAL TABLE 5 EAL TABLE 6 EAL REF. NO(S).:

DOSE PROJECTION MONITORS EFFLUENT MONITORS SEISMIC EVENT SYMPTOMS SDE 2 1. CNMT HI RANGE ACCIDENT MONITORS 2, PLANT VENT STACK II>t WRGM EFFLUENT CHANNEL

3. MAIN STEAM LINE MONITORS

4. STACK 175 MONITOR WRGIJ EFFLUENT CHANNEL

5. STACK 175A MONITOR WRGM EFFLUENT CHANNEL

1. PLANT VENT STACK gt WRGM EFFLUENT CHANNEI

2. TURBINE BUILDING STACK ttsa WRGM EFFLUENT CHANNEL

3. STACK 65 WRGM EFFLUENT CHANNEL

4. STACK t)SA WRGM EFFLUENT CHANNEL

5. TREATED LAUNDRY AND Hol SHOWER DISCHARGE MONITOR

1. SEISMIC MON SYS OBE EXCEEDED- aLaRM (aLB AND/OR "ALARM" VGHT ON THE SEISMIC SWITCH PANEL.

2. WHITE EVENT INDICATOR ON SMA CONTROL PANEL.

10-4 4)

GENERAL EMERGENCY SITE AREA EMERGENCY ALERT 6 SECONDARY WASTE 5AMPLE TANK DISCHARGE MONITOR 3. ALARM AT TRIAXIAL RESPONSE SPECTRUM ANNUNCIATOR. UNUSUAL EVENT

7. WASTE MONITOR TANK DISCHARGE MONITOR

8. TURBINE BLDG. DRAIN MONITOR 4. NOTICEABLE TREMORS OR VIBRATION.

FROM CONTINUING ACTION: TIME BEGAN:

REQUIRED DURATION (MIN)

EXPIRES:

ANY TWO CURRENT CLASSIFICATION INDICATE CURRENT EAL No INDICATIONS OF ON EAL STATUS BOARD A SEISMIC EVENT NEW OR POTENTIAL CLASSIFICATION LISTED ON EAL TABLE 6 No ANY RAD MON ON EAL TABLE 4 IN HIGH ALARM ANY YELLOW No UGHT ON TRIAXIAL RESPONSE SPECTRUM EMERGENCY ANNUNCIATOR DOSE PROJECTIONS YES No LIT OR ENVIRONMENTAL ANY MONITORING RESULTS RADIOLOGICAL

~

YES CONDITION WARRANTING AVAILABLE RECOMMENDATION TO EVACUATE OR SHELlER ANY RED THE PUBUC LIGHT ON TRIAXIAL No GENERAL EMERGENCY RESPONSE SPECTRUM EXCEEDED ANNUNCIATOR LIT NO ALERT EMERGENCY EXCEEDED DOSE PROJECTION ANY DECLARE GENERAL RESULTS PLANT CONDITION EMERGENCY AVAILABLE SITE AREA EMERGENCY THAT IN THE JUDGEMENT PERFORM DOSE 88LB t 8 ExCEEDED No OF THE SUPERINTENDENT SHIFT PROJECTION PER OPERATIONS OR SITE EMERGENCY PEP 343 YES COORDINATOR WARRANTS A MAKE NOTIJTCATIONS SITE AREA EMERGENCY DECLARATION

~

PROJECTED DOSE >IOOD MREM TEDE YES AT OR BEYOND SITE RE-EVALUATE EALS BOUNDARY NO ADVERSE WEATHER AS CONDITIONS WARRANT RE EVALUATE EALS AT SITE AREA EMERGENCY ENTRY POINT Y WHEN DOSE EXCEEDED PROJECTIONS COMPLETEO NO YES PROJECTED DOSE >5000 MREM THYRoto cos al oa esiouo SITE BOUNDARY NO YES

~GENERAL EMERGENCY EXCEEDED DECLARE GENERAL TORNADO HAS HIT THE POVJER BLOCK YES YES PROJECTED OR MEASURED SUSTAINED WIND SPEED AT 10 AIRBORNE RAD LEVELS INDICATE SEVERE DEGRADATION IN RADIOACTIVE MATERIAL CONTROL No EMERGENCY PROJECTED DOSE METERS > 90

> 50 MREM TEDE AT OR MPH ANY BEYOND SITE BOUNDARY PLANT CONDITION SUSTAINED WIND NO USING ADVERSE MET MAKE NOTIFICATIONS NQ SPEEDS AT 10 METERS EXISTS THAT IN THE

~

DATA JUDGEMENT OF THE SUPERINTENDENT YES

> 100 MPH SHIFT OPERATIONS OR SITE EMERGENCY ALERT EXCEEDED COORDINATOR WARRANTS No RE-EVALUATE EALS AN ALERT AS CDNDITIONS WARRANT ALERT DECLARATION EXCEEDED SITE AREA EMERGENCY PROJECTED DOSE EXCEEDED NO

> 250 MREM THYROID CDE AT OR BEYOND SITE BOUNDARY USING ADVERSE MET DATA HAS A ESTIMATED No YES SITE AREA EMERG OR DURAllON OF NO AIRCRAFT CRASH ALERT BEEN RELEASE > 30 MINs No MISSILE IMPACT PII EXCEEDED UNPLANNED EXPLOSION EVALUATE AGAINST INSIDE POWER UNUSUAL EVENT MATRIX BLOCK DECLARE HIGHEST YES EMERGENCY ON EAL STATUS BOARD PROJECTED DOSE ANY UNUSUAL EVENT NO

> 500 MREM TEDE AT YES PLANT IN MODES MAKE CONDITION IDENTIFIED OR BEYOND SITE BOUNDARY I, 2, 3, OR 4 NO11FICATIONS NO EMERGENCY USING ADVERSE MET DECLARATION DATA YES RE-EVALUAlE EALS AS CONDITIONS WARRANT DECLARE UNUSUAL No EVENT CONSULT AP 617 ERC-003. PIP-500, PLP-502 SAFETY RELATED NO AND SP-017 FOR OTHER PROJECTED DOSE EQUIPMENT OR MAKE REPORTABILITY

> 2500 MREM THYROID STRUCTURE REQUIRMEN Ts AFFECTED N ORE tea TION s CDE AT CR BEYOND SI1E BOUNDARY ALERT USING ADVERSE MET EXCEEDED DATA RE-EVALUATE EALS ESTIMATED

~

DURATION OF YES AS CONDITIONS WARRANT SITE AREA EMERGENCY NO RELEASE > 2 MINB ~BEXCEEDEO SITE AREA EMERGENCY EXCEEDED EAL DECLARATION No TIME UNCONTROL'D OR UNPLANNED NO RELEASE OF TOXIC OR FLAMMABLE GAS INTO POWER BLOCK EN VIR ON MENTAL REF EAL TABLE MONITORING RESULTS 7 AVAILABLE YES RE-EVALUATE EALS AT ENlRY POINT Y WHEN ENVIRONMENTAL MONITORING RESULTS AVAILABLE AFFECTED No AREA HOUSES SAFETY RELATED MEASURED DOSE EQUIP MEN 7 RATE > 1000 MREM/HR AT OR BEYOND SITE BOUNDARY NO GAS IS FLAMMABLE CR No MEASURED LACK OF ACCESS IS A SAFETY I131 EQUIV CONC

> BSE-6 Ci/c PROBLEM AT OR BEYOND SITE BOUNDARY GENERAL EMERGENCY 88S 4 BM EXCEEDED NO PLANT IN MODE No DECLARE GENERAL I, 2, 3 OR 4 EAL TABLE 1 EMERGENCY

~EXCEEDED ALERT RADIATION MONITOR TAG NUMBERS MAKE NOTIFICATIONS MONITOR DESCRIPTtON INSTRUMENT RM-11 CHANNEL ERFIS POINT TAG No. ID NO ID No.

SITE AREA EMERGENCY MEASURED OSE RATE > 50 MREM/HR

~A EXCEEDED CNMT LEAK DETECTION REM ILT3502 4- SA 2GS303 RLT35028 RE-EVALUATE EALS MONITOR AT OR BEYOND SITE BOUNDARY AS CONDITIONS WARRANT CNMT HI RANGE ACCIDENT RM-ICR-3589-SA 2AX162 RCR3589A MONITORS RM ICR-3590-SB 2AX163 RCR3590A NO CONTROL PLANT VENT STACK gt RM 21AV-3509-ISA IEX817 RAV3509H MEASURED ROOM EVAC REQUIRED WRGM I 131 EQUIV CONC EFFLUENT CHANNEL OR ANTICIPATED

> I.S 6-7 uot/cc AT OR BEYOND SITE BOUNDARY TURBINE BUILDING STACK 3A RM ITV-3536-1 2EX827 RTV3536D WRGM EFFLUENT CHANNEL No MEASURED YES LEVEL HAS EXISTED RM IMS 3591-SB 5AX164 RMS3591A MAIN SlEAM LINE MONITORS

> 30 MINs AVX CONTROL RM IMS-3592-58 5AX165 RMS3592A No PANEL (ACP) RM-I MS-3593-58 5AX166 RMS3593A OPERATIONAL No FUEL BREACH AREA RAD. MONITORS.

VOLUME CONTROL TANK RM RM IRR-3595 IAA056 RRR3595A CONTINUING CONTROL RM-IRR-3599A IAA060 RRR3599A No CHARGING PUMP IA ROOM ACTION'F CONDITION CONTINUES ROOM EVACUATED CHARGING PUMP 18 ROOM RM-IRR 35998 IAA061 RRR35998 FOR > 30 MINs THEN SITE FOR > 15 MINs AREA EMER. EAL EXCEEDE CHARGING PUMP 'IC ROOM RM-I RR-35990 IAA062 RRR3599C RECYCLE EVAP. VLV. GAL. RM-IRR-3600 IAA063 RRR3600A LETDOWN HX VALVE GAL. RM-IRR 3601 IAA064 RRR3601A MODERATING HX VLV. GAL RM-IRR3602 IAA065 RRR3602A CONTINUING ACTION:

IF CONDITION CONTINUES STACK 25 MONITOR RM IWV-3546 I 4EX837 RWV3546H MEASURED FOR > 15 MINs THEN SITE WRGM DOSE RATE > 500 MREM YES AREA EMER. EAL EXCEEDE EFFLUENT CHANNEL

/HR AT OR BEYOND SITE BOUNDARY STACK IISA MONITOR RM-IWV-3547-1 4EX647 RWV3547H WRGM No EFFLUENT CHANNEL MEASURED 1-131 EQVIV CONC YES

~4- ALERT EXCEEOEO SITE AREA EMERGENCY EXCEEDED LIQUID EFFLUENT MONITORS TREATED LAUNDRY Jc HOT SHOWER DISCHARGE REM-ISIL-3540 4LL666 RWL3540A

> I.s E-4 uol/cc AT OR BEYOND BOUNDARY SITE SECONDARY WASTE SAMPLE REM-2185-3542 4LL686 RWS3542A TANK DISCHARGE MEASURED YES WASTE MONITOR TANK REM IWL-3541 4LL676 RWL3541A LEVEL HAS EXISIEO DISCHARGE FOR > 2 MINs SITE AREA EMERGENCY TURBINE BLDG. DRAIN REM-I MD-3528 2LL276 RM03528A

~84- EXCEEDED MONITOR IN EAL TABLE 5 READING No EAL TABLE 7

> 10 TIMES THE HIGH ALARM TOXIC, FLAMMABLE, (Ic ASPHYXIANT GASES SET POINT

~ ALERT EXCEEDED ACETYLENE HYDROGEN CARBON DIOXIDE GAS (I) to4>O AP-501-00610 AP 501 001 29 AP-501-00085 CHEMICAL FACT SHEET NUMBER AMMONIA AP-501-00637 P-lo (USED IN PORTAL MONITORS) AP-501-00591 OXYGEN (2) AP-501-00068 NITROGEN AF'-501-VVI30 ARGON X AP-501 00319 HELIUM X AP 5III 00573 FREON AP-50 -00987 MIS UST INCLUDES GASES STORED IN BULK AND IS NOT INTENDED To 8 ALL INCLUSIVE.

(I) VENTING OF H2 FROM THE H2 TANK VENT STACK OR H2 HEADER IN A CORDANCE WIM DESIGN AND IS A PLANNED EVENT.

(2) OXYGEN ITSELF IS NOT FLAMMABLE, HOWEVER, ITS PRESENCE ltJCREASE THE FLAMMABILITYOF MATERIALS.

ASPHYXIANT GASES IN LARGE ENOUGH QUANIITIES CAN DISPLACE OXYGEN ND POSE A DANGER To PERSONNEL AND SHOULD BE CONSIDERED A TOXIC THE TYPE AND QUANTITY OF GAS RELEASED, VOLUME OF AREA, ANO VENTILATION SYSTEM5 IN SERVICE SHOULD BE EVALUATED To DETERMINE WHETHER PERSONNEL OR EQUIPMENT COULD BE IN AN ADVERSE ENVIRONMENT, OR WHETHER ACCESS To EQUIPMENT REQUIRED To OPERATE THE PLANT IS IMPEDED.

UNU SU AL EVEN T M ATRI X DEFINITIONS OF AREAS REFERRED TO IN THESE EAL's:

7) SECURITY THREAT 8)0'BIER PLANT OR 9)NATURAL ID) OTHER HAZARDS 11) SITE EMERGENCY POWER BLOCK INCLVDFS THE CONTAINMENT, REACTOR 3)LOSS OF 4)LOSS OF REACTOR 5)LOSS OF 6)LOSS OF MCB ANNUNCIATORS, I) GASEOUS OR LIQUID 2)LOSS OF COOLANT POWER ERFIS OR COMMUNICATIONS EQI >PMENT PHENOMENA COORDINATOR AUXIUARY BLDG, TURBINE BLDG, FUEL HANDLING SECONDARY SECONDARY EFFLUENT(s)

COOLANT OR CAPABILITY PROBLEMS JUDGMENT BLDG. (INCLUDING THE "K AREA), WASTE PROCE5SING EXCEEDING TECHNICAL COOLANT OR COOLING COOLING BLDG.. DIESEL GENERATOR BLDG., DIESEL FUEL OIL SPECIFICATIONS PLACE UNPLANNED LOSS OF >752 OF MCB SLM CONFIRMED SECURITY INDBILIIY TO REACH INDICATION OF ANY FIRE WITHIN THE PROTECTED AREA OTHER PLANT STORAGE, FIRE HOUSE, TANK AREA. INTAKE STUCTURES.

AND DUCT BANKS SERVING THESE AREAS.

GROSS FAILED FUEL RAPID FAILURE OF A LOSS OF ALL VALID HIGH ALARM ON EVENT WHICH REQUIRED SHUTDOWN TWO VALID SEISMIC NOT EXTINGUISHED >MTHIN IS CONDITIONS EXIST THAT DETECTOR INDICATES AN DEPRESSURIZATION PRESSURIZER OFFSITE POWER ANNUNCIATORS (ALBS) FOR >15 PROTECTED AREA INCLUDES THE PLANT AREA INSIDE THE ANY OF THE MONITORS MINUTES AS DEFINED BY: INDICATES A (MODt. 3) CONDITION SYMPTOMS USTED ON MINUTES OF CONTROL ROOM WARRANT INCREASED INCREASE > 2E4 CPM OF STEAM SAFETY OR PORV To IN EAL TABLE 5 MODES 1-4 TOTAL 6 ALBS = 30 POTENTIAL WITHBI TECH SPEC EAL TABLE 6 NOTIFICATION OR VERIFICATION OF A AWARENESS ON 1HE SECURITY FENCE INCLUDING THE IN7AKE STRUCTURES.

WITHIN 30 MINVTES GENERATOR CLOSE FOLLOWING MODES 5-6 TOTAL (t ALBS 20 DEGRADATION IN THE TIME LIMITS CONTROL ROOM ALARM PART OF THE PLANT SECONDARY SIDE REDUCTION OF AND LOSS OF BOTH LEVEL OF SAFETY OF OPERATION STAFF SITE BOUNDARY (SB) IS APPROXIMATELY A CIRCLE APPLICABLE (ALB 1,2, 4-13, 15, 22, 23, EITHER CHATHAM COUNTY, EMERGENCY THE PLANT AS (THIS DOES NOT INCLUDE FIRES THE RELEASE HAS NOT QLLL1 PRESSURE DIESEL 24 DR 25 BASED ON EDG OPERABILITY, 26-28, 8 30)

INDICATED BY: SUSTAINED WIND WITHIN OFFICE AREAS, TRASH BIN HARNETT COUNTY, LEE OF 2500 FT, RADIUS (0.47 MILES)

BEEN TERMINATED RCS SPECIFIC ACTIVITY GENERATORS c UNAUIHORIZED SPEED AT 10 METERS FIRES, H2 TANK VENT STACK FIRES COUNTY WAKE COUNTY FAILURE OF A SG IN D DYER TEN T EXCLUSION AREA BOUNDARY (EAB) 15 APPROXIMATELY EXCEEDS TECHNICAL ALTERATION OR CRITICALITY- OF 74 MPH OR EXTINGUISHED PER OP 152.02 OR OR THE STATE OF (UNUSUAL EVENT EXISTS SPECIFICATION 3.4.8 SAFETY OR PORV To GREAlER OTHER SMALL FIRES OF No PLANT NORTH CAROLINA A CIRCLE OF 7000 FT. RADIUS (1.3 MILES)

ANY RCS PRESSURE TAMPERING HAS OR EXTEIIDED AND UNTIL EFFLUENT LIMITS FOR DOSE FULLY RESET AFTER BOUNDARY LEAKAGE 8ALbL1 IS OCCURRING UNPLUJNEO SUSTAINED SAFETY CONSEQUENCE)

EQUIVALENT 1-131 OR OPERATION DISCHARGE IS TERMINATED AND WHEN ALL REQUIRED GRDSS RADIOACTIVTY

~B INABILITY OF ERFIS TO PERFORM ITS INTENDED FUNCTION FOR A CONTINUOUS PERIOD OF 4 HOURS, AFFECTING SAFETY RELATED EQUIPMENT POSII'VE STARTUP RA IE ~9 TORNADO REPORTED EAL REFERENCE NUMBERS (X-Y2):

(THIS DOES NOT NOTIFICATIONS ARE (FOR DOSE EQUIVALENT ANY OTHER RCS OTHER THAN PREPLANNED REMOVAL

• HOSTAGE/EXTORTION INCLL'DE CRITICALITY NITHIN THE EAB AIRCRAFT, 1RAIN OR OTHER VEHICLE MADE) 131 ME EAL 15 NOT MAIN STEAM LINE OR I ROM SERVICE FOR MAINTENANCE OR X CATEGORY (1-11)

I LEAKAGE IN EXCESS SITUATION THAT EARLIER THAN CRASH THAT MAY DAMAGE PLANT EXCEEDED UNLESS ME FEEDWATER LINE DF TECHNICAL MODIFICATION PURPOSES WHILE IN STRUCTURES CONTAINING FUNCTIONS Y = IDENTIFIER WITHIN CATEGORY THREATENS TO ESTIMATED DURING 48 HOUR TIME BREAK MODES 1,2,3 OR 4 AS DEFINED BY:

SPECIFICATION INTERRUPT NORMAL PLA>INEO REACTOR OR SYSTEMS REQUIRED FOR SAFE 2 CLASSIFICATION (1-4)

INTERVAL, OR FIG. 34-1 FAILURE OF BOTH OPUS SHUTDOWN OF THE PLANT 3.4.6.2 WIM THE 4 PLANT OPERATIONS STAR'UPS) I = UNUSUAL EVENT LIMITS ARE EXCEEDED) (A BREAK IS A LEAK HOUR CORRECTIVE FAILURE OF BOTH DATA CIVIL DISTURBANCE 2 = ALERT WHICH EXCEED5 THE ACTIONS NOT CONCENTRATORS ONGOING BETWEEN OPERATORS ABILITY SATISFIED FAILURE OF BOTH DATA DISCS 88LBW 3 = SITE AREA EMERGENCY THE SITE TO SHUTDOWN THE INABIVTY To DISPLAY SPDS IN BOUNDARY AND TURB'ttE ROTATING UNPLANNED EXPLOSION WITHIN THE 4 = GENERAL EMERGENCY PLANT IN A THE CONTFIOL ROOM THE PROTECTED COIJPONENT FAILURE PROTECTED AREA RESULTING IN CONTROLLED INABILITY To UPDATE CURRENT AREA RESUt TING IN A VISIBLE DAMAGE To PERMANENT MANNER OR TO NOT DATA DISPLAYS IN THE CONTROL REACTOR TRIP, CASING STRUCTURES OR EQUIPMENT EXCEED TECH SPEC ROOM PEIJETRATION OR COOLDOWN LIMITS) (THIS 15 NOT To BE CONSTRUED AS SIGNIFICANT DAMAGE A FAILURE OF A SINGLE VARIABLE OR TO Tt!c MAIN SMALL DATA SUBSET) GEIJERATOR 5EALS UNPLANNED TOXIC OR FLAMMABLE EAEbkf GAS RELEASE WITH 1HE EAB STEAM GENERATOR BLOWDOWN LINE ~LO (REFERENCE EAL TABLE 7)

HARRIS NUCLEAR PLANT BREAK FAILURE OF BOTH SITE TELEPHDNE (MODES 1,2 ANID 3) AND EMERGENCY (HEJcEC) TELEPHONE SIMTCHES EMERGENCY ACTION LEVFL FLOW PATH SIDE 2 REVISION: 99 1