ML20135F916

Submits Summary of 970226 Meeting Conducted in Region IV Re Programs,Initiatives & Actions Taken or Under Consideration for Issues Associated W/Operation of Plant

ML20135F916
Person / Time
Site:	Fort Calhoun
Issue date:	03/06/1997
From:	Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Gambhir S OMAHA PUBLIC POWER DISTRICT
References
NUDOCS 9703170116
Download: ML20135F916 (216)

Text

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[ 611 RY AN PL AZA DRIVE SUIT E 400 3, ,[ AR LlNGTON, TE XAS 76011 8064 MAR -6 1997 S.K. Gambhir, Division Manager Production Engineering Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 399 Hwy 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023-0399

SUMMARY

OF MEETING WITH OMAHA PUBLIC POWER DISTRICT MANAGEMENT REPRESENTATIVES

Dear Mr. Gambhir:

This refers to the meeting conducted in the Region IV office on February 26,1997. This meeting included your discussion of programs, initiatives and actions taken or under consideration for issues associated with the operation of the Fort Calhoun Station. During the meeting, representatives of the Omaha Public Power District presented information involving fuel handling, the 1996 Fort Calhoun Refueling Outage, configuration control, quarterly trends, and change management / coaching. In the area of configuration control, the presentation included recent issues, history, actions taken as a result of recent issues, I and planned actions for the future. You stated a goal of zero tolerance for unauthorized configuration changes.

Normally, a drop-in visit does not require a meeting notice, a meeting summary, or documentation, since a drop-in meeting consists of a general exchange of information.

However, the NRC staff concluded following the meeting that the visit was more than a general discussion, including a thoroughly prepared and specific presentation of the topics. l Therefore, in accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, ,

Title 10, Code of Federal Regulations, a copy of this letter and its enclosures (attendance I list and licensee presentation materials) will be placed in the NRC's Public Document Room.

Should you have any questions concerning this matter, we will be pleased to discuss them with you.

Sincerely,

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L h Arth r T. HowellIll, Acting Director Division of Reactor Projects 9703170116 970306 PDR ADOCK 05000285 p PDR M.Eh .. l 1

Omaha Public Power District Docket No.: 50-285 License No.: DPR-40

Enclosures:

1. Attendance List

2. Licensee Presentation cc w/ enclosures:

James W. Tills, Manager Nuclear Licensing Omaha Public Power District Fort Calhoun Station FC-2-4 Adm.

P.O. Box 399 Hwy. 75 - North of Fort Calhoun Fort Calhoun, Nebraska 68023-0399 l James W. Chase, Manager  :

Fort Calhoun Station P.O. Box 399 Fort Calhoun, Nebraska 68023 l Perry D. Robinson, Esq.

Winston & Strawn

, 1400 L. Street, N.W.

Washington, D.C. 20005-3502 Chairman Washington County Board of Supervisors Blair, Nebraska 68008 Cheryl Rogers, LLRW Program Manager Environmental Protection Section Nebraska Department of Health ,

301 Centennial Mall, South P.O. Box 95007  :

Lincoln, Nebraska 68509-5007 i

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Omaha Public Power District MAR -6 1997 bec to DCD (IE45) bec distrib. by RIV:

Regional Administrator Resident inspectc,r I DRP Director l

Branch Chief (DRP/B) DRS-PSB Project Engineer (DRP/B) MIS System '

Branch Chief (DRP/TSS) RIV File -

Leah Tremper (OC/LFDCB, MS: TWFN 9E10) K. Perkins, Director, WCFO B. Henderson, PAO C. Hackney, RSLO l

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To receive copy ofgument, indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy RIV:C:DRP/B 5 ADD #fiD j D:DRFC)# 1 l l WDJohnson:df// DDCHfibef9ain ATH6MH [f

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l bec to DCD (IE45) .

] l bec distrib. by RIV: 1 Regional Administrator Resident inspector DRP Director Branch Chief (DRP/B) DRS-PSB  ;

Project Engineer (DRP/B) MIS System Branch Chief (DRP/TSS) RIV File Leah Tremper (OC/LFDCB, MS: TWFN 9E10) K. Perkins, Director, WCFO '

B. Henderson, PAO C. Hackney, RSLO  ;

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To receive copy o@ument, indicate in box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy RIV:C:DRP/BM - ADD QBD j D:DRpt)# 1 l l WDJohnson:df // DDCHfilDef9ain ATH&XH [f

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ENCLOSURE 1 MEETING: DROP-IN MEETING

SUBJECT:

FORT CALHOUN OPERATIONAL ISSUES DATE: FEBRUARY 26, 1997 ATTENDANCE LIST .

NAME ORGANIZATION POSITION TITLE Sudesh Gambhir OPPD - FCS Division Manager -

Production Engineering Gary Gates OPPD - FCS Vice President Ron Short OPPD - FCS Manager Operations James Tills OPPD - FCS Manager - Nuclear Licensing J. E. Dyer NRC Acting Regional Administrator T. P. Gwynn NRC Acting Deputy Regional Administrator  !

A. T. Howel' III NRC Acting Director of Division of Reactor <

Projects l W. D. Johnson NRC Chief. Branch B.

Division of Reactor  !

Projects i

K. E. Brockman NRC Acting Director of Division of Reactor 4 Safety D. N. Graves NRC Senior Project Engineer.

Branch B. Division of Reactor Projects

ENCLOSURE 2 J

Exnmples of Conservative Fuel Handling t .

An Assembly would not fully insert during the Cycle 17 core reload.

Successfully used adjustment allowance within procedure.

After sectnd bundle wouldn't fully insert, the System Engineer was contacted and he diagnosed the problem to the setting of the reference point for the indexing i system. Once reset, the remainder of the fuelinsertions were completed without i any further difficulties.

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As a result of a misplaced fuel assembly in Cycle 16, which was discovered before completion of the core reload, several Humsn Factor enhancements were incorporated into our fuel handling practices:

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The Trolley coordinates were adjusted to ensure that the bridge and trolley 1 1

coordinates could not be transposed.

The format of the fuel movement sequence sheets was revised to be consistent with the bridge and trolley displays on FH-1.

Independent verification of fuel movement is now required.

The bridge and trolley displays on FH-12 were moved to be consistent with FH-1 and minimize the potential for transposition errors.

FH-1 has both an indexing system and an above water coordinate system. I l

FH-1 was modified to easily handle ~ e overload / underload settings to handle the various insert weights (CEAs, FLPRs, Sources) and different fuel vendors (different fuel l assembly weights) to ensure that fuel damage potential is minimized.

No misplaced fuel assembly or insert was found during the core verification performed at the completion of the reload.

Core Alignment checks are performed to ensure that fuel damage from the placement of the UGS does not occur.

ECN-96-411 was approved by a NPRC to reformat the control console of FH-1 for human factors concerns.

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8kuna Omaha PublicPowerDistrict 444 South 16th Street Mall Omaha NE 68102-2247 moH 2.1 January X. 1997 LIC-96-0199 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-137 Washington. DC 20555

References:

1. Docket No. 50-285

2. Letter from NRC (T.P. Gwynn) to OPPD (T. L. Patterson) dated December 4. 1996

SUBJECT:

NRC Inspection Report No. 50-285/96-12. Reply to a Notice of Violation (NOV)

The subject report transmitted a Notice of Violation resulting from an NRC inspection conducted October 15 through November 1,1996 at the Fort Calhoun Station (FCS). Attached is the Omaha Public Power District (0 PPD) response to this NOV.

In addition, the subject inspection report discusses several concerns or issues related to refueling operations during the Fort Calhoun Station (FCS) 1996 Refueling Outage. These concerns / issues involve: 1) poor reactor-side cavity water clarity. 2) apparent lack of concern by operators related to monitoring neutron count rates 3) lack of utilization of TV camera aids, and 4) OP-11.

" Reactor Core Refueling" procedure not explicitly stating a safe condition for a fuel assembly if' refueling operations are unexpectedly suspended. A brief response to each of these items is provided below:

1) Poor Reactor-side Cavity Water Clarity During fuel reload in the 1996 Refueling Outage, the reactor cavity side water clarity was murky due to the inability of an improperly operating Tri-Nuc filter in the cavity to clean up the water. OPPD's standard will be raised for all subsequent refueling outages to ensure good reactor-side cavity water clarity that will permit visual observation of possible '

obstructions and debris which could impede safe fuel movement. Poor clarity did not jeopardize the 1996 fuel reload because: 1) the integrity of each fuel assembly was verified during core off-load by means of fuel sipping and only new, repaired or non-damaged fuel bundles were reloaded. 2) the fuel

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. LIC-96-0199 Page 2

, is moved and placed in accordance with precise Fuel Handling Machine'(FH-1)

! bridge and trolley coordinates for which there are two independent position i indicating systems. The primary system is a digital encoder which is calibrated to a fixed benchmark location within 14 days of fuel movement.

i The backup position indicating system is a permanently mounted, above water

n alpha-numeric indexing system. A procedurally required cross-check between g the two indexing systems is performed each time FH-1 indexes over the e

reactor core. 3) The FH-1 operators closely monitor fuel assembly weights y 3 for potential under/ overload conditions; which is a good indication of fuel $

handling problems' which could lead to fuel bundle damepe, 4) independent g 1 verification of the correct fuel bundle being moved eitha in the Spent Fuel 3

, Pool (SFP) or Containment is completed by the Refueling Crew Control Room $

Coordinator, and 5) FCS utilized the fuel handling machine (FH-1) mast y camera to perform a complete and thorough core alignment check (0P-11, g Appendix L) and fuel loading verification check (0P-11, Appendix M) after y l fuel reload was completed.

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c To enhance cavity water clarity for future refueling outaces, the following R changes will be made prior to the 1998 refueling outage: $ l m \

a) Radiation Protection will take ownership to implement an action I plan to maintain clarity of the reactor-side cavity and Spent Fuel Pool, b) OP-11 will be revised to include a sign-off step by the Fuel Handling Coordinator (normally a Shift Supervisor) or Control

• Room Shift Supervisor that the clarity standard of the spent @

fuel pool, fuel transfer canal, and reactor-side cavity are met =

to permit visual observation of possible obstructions and 8

debris which could impede fuel loading or jeopardize safe fuel 3 movement, "

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2) Apparent Lack of Operator Concern on Monitoring Neutron Count Rates 0 FCS Operations takes the movement of fuel and associated monitoring of plant g parameters very seriously. In fact, for both fuel off-load and reload, only m

, two (2) wide range neutron flux channels were to be operable and the 5

. Operations Staff insisted that at least three (3) channels be operable to E provide additional indication.

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Additional improvements are discussed in the enclosed attachment.

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December 17,1996 LIC-96-0199 Page 3

3) Lack of Utilization of TV Cameras.

FCS utilized the fuel handling machine (FH-1) mast camera to perform a complete and thorough core alignment check (OP-11 Appendix L) and fuel -

l loading verification check (OP-11. Appendix M) after fuel reload was

• completed. These checks ensure the fuel is properly aligned and all fuel

assemblies. CEAs and sources are properly located and oriented, including verifying fuel ' assembly serial numbers. Due to an equipment limitation resulting from a modification to the refueling machine mast for in-mast sipping, little useful information could be obtained in using the camera during withdrawal and insertion of a fuel assembly. The current camera configuration points straight down and is only capable of viewing one adjacent fuel assembly. Viewing of a grappled fuel assembly was not possible with the mast camera, Modification of the current camera configuration is under consideration.

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4) Procedure OP-11 does not explicitly state a safe fuel assembly position.

OP-11 was written to allow the Control Room Shift Supervisor or Refueling Crew Coordinator to determine the proper, safe configuration to leave a fuel assembly depending on the reason for the suspension of fuel movement. OPPD will review OP-11 to determine if additional guidance should be provided.

If necessary, OP-11 will be revised prior to the start of the 1998 refueling outage.

If you should have any questions, please contact me.

Sip erely, I

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Sude h K. Gambhir Division Manager Production Engineering SKG/ddd I Attachment c: Winston and Strawn I L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector  ;

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Attachment LIC 96-0199 Page 1 REPLY TO A NOTICE OF VIOLATION Omaha Public Power District Ibcket: 50-285 Fort Calhoun Station Liamse: DR-40 l

During an NRC inwticx1 conducted on October 15 through Nuvh 1,1996, one violation of NRC requirements was identified. In accordance with the " General Statenent of Policy and Procedure for NRC Enforcement Actions," NUREG 1600, the J violation is listed below: 1 Tae4mical Specification 2.8(4) requires neutron flux to be con *4mnaly  ;

nonitored by at least two source range neutron nonitors whenever core gecmetry is being changed, with each nonitor providing ccx1&4m = m visual indication in the control room.

F Contrary to the abcae, en October 28, 1996, seven fuel assemblies were loaded into the reactor vessel near inoperable Wide Range Logarithnic Power Channel D (a source range neutron nonitor) while the operable channels in other quadrants of the core were unable to provide the required continuous visual indication due to their distance from the assemblies. i This is a Severity Level IV Violation. (Supplement I) (285/9612-01)  !

i DPPD Resnonse OPPD agrees that' placing the initial seven fuel assemblies near an inoperable excore detector was a violation of the intent of Technical Specification 2.8(4) but it did not impact nuclear safety.

1. The Reason for the Violation Fuel reload is controlled under procedure OP-11. " Reactor Core Refueling".

This procedure does not contain guidance concerning the requirement to place the initial fuel assemblies near operable excore detectors. OP-11 was also deficient in providing guidance to address contingency actions when an excore detector is inoperable prior to, or fails during, core reload. As a result, the fuel reload sequence was not revised and the first seven fuel assemblies loaded into the reactor vessel were placed

I Attachment LIC 96-0199 l

Page 2 1

near an inoperable excore detector.

2. Corrective-Steps Which Have Been Taken and the Results Achieved l

OPPD has reviewed our practices associated with fuel reload and will not a initially load fuel assemblies near inoperable excore detectors. o 8 -

3. Corrective Steps Which Will Be Taken to Avoid Further Violations e I R l OP-11 will be revised prior to the 1998 Refueling Outage to require 8 i reloading the fuel initially near operable detectors. A reload $;

sequence will be developed for two preferred operable excore y detectors and contingency actions will be added to OP-11 should m these detectors become inoperable prior to or during the reload process. This will ensure that the intent of Technical o Specification 2.8(4) will be met. E  ;

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4. Date When Full Compliance Will be Achieved S I 1

OPPD is currently in full compliance. l l

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Gunoa CEP7blicP7werDE 444 South 16th StreetMall Omaha NE68102-2247 January 21, 1997 LIC-97-0003 1

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555

Reference:

Docket No. 50-285 l l

Subject:

Licensee Event Report 96-015 Revision 0 for the Fort Calhoun Station Please find attached Licensee Event Report 96-015 Revision 0 dated January 21, 1997. This report is being submitted pursuant to l 10 CFR 50.73(a)(2)(i)(B). If you should have any questions, please contact me.

l Sincerely, 1

S 'G Division Manager Production Engineering l DDD/ddd l Attachment c: Winston and Strawn L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector INP0 Records Center wau

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (445)

APPROVED BY OMB NO. 3150-0104 EXPIRES 4/30/98

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- IOFOP.MATI COL Cl N EO E :60.0 R EP R L $ $L LICENSEE EVENT REPORT (LER) Ay,Ej,NegggagN c gy,EggEgsggcggNyg,egxg  ;

(See reverse for required number of RE U T COMM$10 WASH , 05 5 , No digits / characters for each block) e[c'cNAE8Tofo"No#'[3y0m s mcE OF MANAGEMENT ANo

, FAC3JTY NAME(1) DOCKET NUMSER (2) PAGE(3) 1 Fort Calhoun Station Unit Tu.1 05000285 10F 5 1

TlVLE (4)

Improper Fuel Loading next to an Inoperable Wide Range Nuclear Instrument

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1 EVENT DATE(5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) l FACILTTY NAME DOCKET NUMBER MONTH DAY YEAR YEAR MONW DAY YEAR NUMBE N 8 Q$QQQ FACillTY NAME DOCKET NUM8ER 3 10 28 96 96 - 015 -

00 01 21 97 05000  !

OPERATING THIS REPORT IS SUBMITTED PURSUANTTOTHE REoutREMENTS OF 10 CFRG (Check one or more)(11) 5 MODE (9) 20.2201(d) 6 20.2203(ax2)(v) x 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 0 20.2203(a)(2)(i) 20.2203(a)(3)(iii 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below l or k'n NRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) l LICENSEE CONTACT FOR THIS LER (12)

MME TELEPHONE NUMBER (induce Area Code) l Carl Stafford - Principle Reactor Engineer (402) 533-6670 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) i CcUSE SYSTEM COMPONENT MANUFACTURER T LE L pp CAUSE SYSTEM COMPONENT MANUFACTURER O NPR $

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONm DAY YEAR VES SUBMISSloN pt yes, complete EXPECTED SUBMISSION DATE) )( NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e.. approximately 15 single-spaced typewritten lines) (16)

On October 28,1996, 'during cycle 17 fuel reload, seven fuel assemblies were loaded into the reactor vessel near inoperable Wide Range Power Channel D (source range) while the operable channels in other quadrants of the core were unable to provide visual indication due to their distance from the assemblies. Technical Specification 2.8(4) requires neutron flux to be continuously monitored by at least two source range neutron monitors whenever core geometry is being changed, with each monitor providing continuous visual indication in the control room.

OP-11, " Reactor Core Refueling" did not contain guidance concerning the requirement to place the initial fuel assemblies near operable detectors and did not provide guidance to address contingency actions when an excore detector is inoperable. As a result, the fuel reload sequence was not revised to load near an operable detector.

l The Omaha Public Power District has reviewed the practices associated with fuel reload 4

and will not initially load fuel assemblies near inoptrable detectors. OP-11 will be revised before the 1998 refueling outage to require reloading the fuel initially near operable detectors. A reload sequence will be developed for initially loading fuel adjacent to two operable excore detectors and contingency actions will be added to OP-11 should these detectors become inoperable prior to or during the reload process.

NRC FORM 366 (4-95)

RC FORM 366A U.S. NUCLEAR REGULATORY CoMMISSloN LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER :6) PAGE (3) g SEQUENTLAL REVISION Fort Calhoun Station Unit No.1 05000285 2 OF 5 96 - 015 - 00 7 EXT (dsnore space is requeed, ne acthbonalcopies of NRC Fam 366N (11)

BACKGROUND The nuclear instrumentation external of the core at Fort Calhoun Station includes four l wide (source) range logarithmic channels (A, B, C and D), four power range safety l channels (A, B, C, and D) and two power range control channels (A and B). The detectors for these channels are located in instrument thimbles in the biological shield around the reactor core.

The four wide range nuclear instrumentation channels monitor and indicate greater than ten decades of neutron flux. Each wide range channel generates a rate-of-change of power signal for indication, control and protection. A signal is also provided by the l wide range channels to enable and/or inhibit reactor protection functions, and provide an audible indication of neutron flux. Each wide range detector assembly contains dual high sensitivity fission chambers surrounded by an aluminum housing. l When the reactor is shutdown, both fission chambers of the wide range detector are ,

used to provide additional sensitivity for measuring neutron leakage in counts per i second (cps). Once power reaches approximately 4.0 E-05 percent, approximately 1000 cps, the unshielded fission chamber is electronically cut-off since adequate counts are available from one detector. During fuel movement, low power physics testing and plant start-ups, external counters are connected to the wide range channels to support l generation of count rate ratio plots (also referred to as inverse multiplication or 1/M plots). These plots are used for predicting if or when criticality will occur.

Technical Specification 2.8(4) requires, in part, that whenever core geometry is being changed, neutron flux shall be continuously monitored by at least two source range neutron monitors, with each monitor providing continuous visual indication to the control room.

EVENT DESCRIPTION On October 28, 1996, Fort Calhoun Nuclear Station (FCS) was in Mode 5 (Refueling Shutdown) for a refueling outage. A plant modification was in progress to change out the four wide range nuclear instrument drawers in the control room. This modification was required to replace the existing equipment which had become obsolete, making replacement parts difficult to obtain.

A requirement of the modification was to have two wide range neutron flux detector instrumentation channels operable including source range and start up rate indication during fuel movement. An additional more restrictive' requirement of the modification requested by operations was to have three channels operable prior to start of the fuel reload activities. Two of those channels requested by operations were "A" and "D". The PCC FORM 36sA(445)

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N FORM 366A U.S. NUCLEAR REGULATORY COMMISSlo[

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

, l FACILITY NAME (1) DOCKET LER NUMBER 6) PAGE (3)

YEAR SEQUENTIAL REVISION Fort Calhoun Station Unit No.1 05000285 3 or 5 ,

96 - 015 - 00 TEXT (e more space is reqwed, use adnonalcones of NRC Form 366A)(n) requirement to have channels "A" and "D" operable was based on past fuel reload practices and experience.

During the Cycle 17 reload (as in at least the past three cycles) the fuel assemblies containing the neutron sources were placed in the reactor early in the fuel reload

sequence to provide a background count rate for monitoring purposes. This was accomplished by placing four fuel assemblies against the East wall of the core shroud i in locations T6, T8, T10, and T12 to provide support for the fuel assembly with the 4 neutron source which is placed one row in from the periphery of the core shroud. Three additional fuel assemblies were placed in core locations S11, S9, and S7, with the l fuel assembly containing the first neutron source being placed in location S9. This

, arrangement provides support on three sides of the fuel assembly which contains the l 1

4 source. The process was then repeated on the West wall of the core shroud to support i the fuel assembly with the second neutron source.

When fuel reload activities were scheduled to begin, wide range channel "D" was not available. The delay in the completion of channel "D" of the modification was caused by hardware and testing problems. A field design change request (FDCR) was written to i remove the requirement of having channels "A" and "D" operable for fuel reload ,

activities. This change was made to keep fuel reload activities on schedule. 1 This change resulted in reducing the requirement of having channels "A" and "D" operable, to having a minimum of any two channels of the wide range nuclear instrumentation system operable during fuel reload, without any preference as to which channels were available.

The FDCR and its poss.ible impacts on the fuel reload were reviewed by Operations, Construction Management and Engineering prior to approval. Nuclear Engineering evaluated the change from a safety standpoint and concluded that the fuel reload could j proceed as scheduled.

1 The decision to proceed with the fuel reload with wide range channel "D" inoperable i was based on the following four points: 1) Three of the wide range system channels were operable, which exceeded the requirement of Technical Specification 2.8(4) of having two channels operable, 2) the boron concentration of the reactor coolant system was greater than the required Cycle 17 refueling boron concentration of 2100 parts per million (ppm), 3) without including the control element assemblies which were inserted in the fuel assemblies prior to reload, the shutdown margin was much greater than 5%,

and 4) any criticality concerns would be detected by the three operable wide range channels during fuel reload activities.

After all of the initial condition requirements of Operating Procedure OP-11, " Reactor Core Refueling" were met, fuel reload activities commenced at 0442 hours0.00512 days <br />0.123 hours <br />7.308201e-4 weeks <br />1.68181e-4 months <br /> on NRC FORM 346A (4-95)

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i NRC FORM 366A im U.S. NUCLEAR REGULATORY COMMISSION l LICENSEE EVENT REPORT (LER)  !

TEXT CONTINUATION '

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FACILITY NAME (1) DOCKET LER NUMBER C6) PAGE (3)

Fort Calhoun Station Unit No. 1 03000285 4 OF 5 l 96 - 015 -

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TEXT gincre space W- M. use m.;comes orNRC rom M64 (17)

October 28, 1996. l During fuel reload activities an inverse count rate ratio,1/(Co/Ci), plot is maintained for at least two wide range system channels throughout core loading. The control room operator completes this calculation for each fuel assembly. Following the loading of each fuel assembly, the control room operator ensures that the reactor will not go critical and allows the next fuel assembly to be moved.

While placing fuel assemblies against the East wall of the core shroud in the vicinity of the inoperable channel "D" detectors, the value for the base count detected on channels "A"' and "C" remained at zero during the first five fuel assembly moves, and the count rate ratio value remained at one. During the sixth movement of fuel assembly.

T028, which contained the first neutron source, the base count value showed an increase to 0.010 cps on channel "A" and 0.008 cps on channel "C".

The next series of fuel assembly moves were on the West side of the reactor adjacent to the operable channel "A". These moves were to place the second source in the reactor using the same process to build out to the second row from the shroud to provide support for fuel assembly T027, which contained the second neutron source.

During the next seven moves, the count rate as well as the count rate ratio values changed for each fuel assembly move.

During refueling an NRC inspector visited the control room on October 28, 1996 and witnessed the initial fuel reload activities as previously described. The inspector noted a lack of audible counts during loading of the initial seven fuel assemblies in the core, including the one which contained the neutron source. The inspector verified that the first seven fuel assemblies were loaded adjacent to channel "D", which was inoperable.

The inspector then observed the loading of the first fuel assembly in the West quadrant, in the vicinity of operable channel "A". The inspector noted that in addition to the audible count rate, a substantial increase in neutron flux indication was observed.

When the monitored neutron flux indication from the operable ciannels were compared for the two series of moves, they were quite different, even though the characteristics of the fuel assemblies placed in the two areas were similar.

A corrective action document (Condition Report 199601336) was written to document the event. On December 4,1996, a Notice of Violation was received from the NRC. On December 20, 1996, the FCS Plant Manager determined that a violation of Technical Specification 2.8(4) had occured; in that, seven fuel assemblies were loaded into the NRC FORM M4A(4-95)

l NR FORM 366A U.S. NUCLEAR REGULATORY CoMMISSN

. LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION

} FACILITY NAME (1) DOCKET LER NUMBER L6) PAGE (3) ~

l Fort Calhoun Station Unit No.1

• *Ee7 EsTE 05000285 5 OF 5 96 - 015 -

00 TEXT (#more spece is required, use adcitortaf copes etMC Form 364%17) reactor vessel near an inoperable Wide Range Logarithmic Power Channel D (a source range neutron monitor) while the operable channels in other quadrants of the core were unable to provide the required continuous visual indication due to their distance from the assemblies. This report is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B) .

CONCLUSIONS Fuel reload is controlled under procedur OP-11. This procedure did not contain guidance concerning the requirement to ; 4 ace the initial fuel assemblies near operable excore detectors. OP-11 was deficient in providing guidance to address contingency actions when an excore detector is icperable prior to, or fails during, core reload.

As a result, the fuel reload sequence was not revised and the first seven fuel assemblies loaded into the vessel were placed near an inoperable excore detector.

SAFETY ASSESSMENT i

Review of the significance of this event indicates that there was no significance with  ;

respect to nuclear safety. At all times during the fuel reload, three of the four channels of the wide range nuclear instrumentation system were operable. In addition, 1 the reactor coolant system was borated to greater than the Cycle 17 refueling boron I concentration of 2100 ppm which provided a minimum shutdown margin of 5%. Additional safety margin against the reactor reaching criticality existed as the fuel assemblies were loaded with control element assemblies inserted. I CORRECTIVE ACTIONS OP-11 will be revised prior to the next refueling outage to require reloading the fuel initially near operable detectors. Reload sequence will be developed for initally loading next to two operable excore detectors and contingency actions will be added to OP-11 should either or both of these detectors become inoperable prior to or during the reload process. This will ensure that the intent of Technical Specification 2.8(4) is met.

PREVIOUS SIMILAR EVENTS There have been no previous similar events at Fort Calhoun Station.

CID 970099/03 ASSIGNED TO RW SHORT

=j l NRC FORM 366A 44-95)

i' 1996 Outage Initiatives and Lessons Learned f.

1. The Reactor Coolant System (RCS) was not breached until all material was through the I equipment hatch and the equipment hatch was reinstalled. This was done to minimize the l

activity that would have been released due to OPPD's failed fuel. j i

! l

2. An RCS leak check was performed instead of a Hot Hydro. This is a safer option as it j

eliminates an evolution requiring the RCS to be in a solid condition for the length of the I

inspections.

4 i 3. Two weeks prior to the shutdown, reactor power was reduced to 70%. One week later power was reduced to 45 %. This greatly reduced RCS activity (most notably iodine) and containment  ;

radiation levels at the start of the outage. l l

? 4. In future outages, Control Element Drive Mechanism (CEDM) tool access flanges will be

{ removed prior to disassembling Heated Junction Thermo-Couples (HJTCs). In the 1996 i outage, reactor coolant leaked out of the disassembled HJTCS when the access flanges were

removed.

l 5. To minimize the possibili> of a Ventilation Isolation Actuation Signal (VIAS), the

{ containment purge will be secured prior to removing the steam generator manways.

I a

i 6. OPPD performed a temporary modification to put charcoal filters in-line with the containment l and auxiliary building ventilation exhaust to remove iodine, mimmizing the iodine released to i the public.

}

4

7. Substitute fuel bundles were used as much as possible to reduce the need for fuel reconstitution

and attempt to improve fuel reliability for the current cycle. The substituted bundles were all

, CE fuel.

i 1

f 8. OPPD took precaution during nozzle dam work and steam generator manway removal to avoid )

i personnel safety issues noted in the last outage which were cau::ed by hot (temperature) l j systems. {

I

9. The effectiveness of degassing the RCS, was improved by standardizing the use of the Volume  ;

Control Tank (VCT) for " burping" rather than floating on the vent header. l i

1 l

f

1996 REFUELING OUTAGE REPORT AND CRITIQUE Outage Accomplishments / Issues l Outage Completed in Record Time l

No Lost Time Injuries ,

I Better Control Of Work Hours 1

1 Control of Iodine Release by

+ Power Reduction 2 Weeks Prior to Outage

• Aggressive Cleanup Efforts

• Purification System Operation Addition of Charcoal Filters to Contaimnent and Auxiliary Building Vent Paths i

Steam Generator Tube Inspections Reactor Coelant Pump Motor RC-3B Replacement Lessons Learned VIAS when Steam Generator Manways are Opened Controls on Containment Closure Fuel Load Pattern based on Operable Wide Range Nuclear Instruments Cooldown During Steam Generator Sweeps l 1

l l

l 1

-- . - . _ - . -- . - _ . _ . - - = ~ - . . - - --

MONTHLY HUMAN PERFORMANCE INDICATOR Description Recent interest in providing a means of trending overall human performance improvements or declines has resulted in the development of the Monthly Human Performance Indicator shown above. This indicator is based on plotting the number of human performance related CRs divided by a measure of the amount of work being performed during a month. Currently, work hours are combined from Maintenance, Operations, Chemistry, Radiation Protection, and Security as these groups work hours tend to fluctuate dependent upon the overall amount of work being performed throughout the plant. The amount of work performed is assumed to reflect a measure of the amount of risk or potential for performance errors. A four month moving average and 1a standard deviation is plotted to assist in understanding how a given month compares to previous monthly trends.

Observations / Analysis The monthly trend indicates that the months of June and Octot'er,1996 saw an increase in the number of human performance related errors beyond a level of assumed, normal fluctuation. Review of csuse code information for these time periods indicates that an increase in the number of conditions having resulted from errors in the WORK PRACTICE, PROCEDURES and SUPERVISORY METHODS categories were the primary contributors.

The negative trend in WORK PRACTICE during these months appears to be co mentrated in the area of

' Skill Based Slips

• and

• Failure to Perfurm Self-CheckingNerifk.atien*

The negative trend in the area of PROCEDURES is primarily due to an increase in the number of identified

• Drawing Errors

• and ' Situations not adequately covered by procedures

• The negative trend in SUPERVISORY METHODS is not as significant as the other two categories however, these periods indicate that an increase in ' inadequate supervision" (in the field) and " inadequate scheduling" are the primary contributors.

The fact that the 1996 refueling outage and the unplanned outage related to the reactor coolant pump l

ARD failure coincide with these time periods may indicate that outage related activities resulted in these increases.

d Human P@rformance Indicator Human Performance Indicator Dec-95 Jan-96 Fetr96 Mar-96 Apr-96 May-96 Aug-96 Sep-96 Oct-96 Jun-96 Jul-96 Nov-96 Dec-96

1. HP Related CRs 127 123 124 133 100 121 148 98 106 110 171 159 Ill Hours Worked 59991 49(75 49776 58479 62535 49347 52885 55807 47139 58908 64144 65839 50429

1. CRs/100 Hours Worted 0.212 0.248 0.249 0.227 0.160 0.245 0.280 0.176 0.225 0.187 0.267 0.241 0.220 4 Month Moving Average 0.237 0.234 0.221 0.220 0.228 0.215 0.231 0.217 0.213 0.230 4 Month Stanadard Deviation 0.015 0.015 0.036 0.036 0.044 0.049 0.038 0.041 0.036 0.029 Upper Error Level (+ ls) 0.252 0.249 0.257 0.256 0.272 0.264 0.269 0.258 0.249 0.259 Lower Error Level (-Is) 0.222 0.219 0.185 0.185 0.184 0.166 0.194 0.176 0.178 0.201

. _ ~ . _ _ _ . .

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E 4

4 REVIEW of OUTAGE CONDITION REPORTS

\

On October 24. 1996, a presentation was made to the PRC comparing the number i

' of significant irs from the 1995 RF0 to the level 1. 2 & 3 CRs occurring in '

the 1996 RFO. Significant increases were evident in the event areas as I described in the attached memorandum (FC-CAG-96-061) and as listed below. l The following actions were taken as a result of this comparison:

The PRC agreed that department briefings would be held on the outage

. goals not being met, i

The message on outage goal status was conveyed on 3N network.

The Operations Manager and Operations Supervisor were transferred from i the outage organization back to operations to maintain a full time

management oversight.

The last column of the table shows a marked improvement in the event category 1 performance for the last half of the 1996 RF0. after the above actions were implemented.

EVENT Entire 1st Half 2nd Half DESCRIPTION 1995 RF0 1996 RF0 1996 RF0 Equipment Actuation Events 0 4 0 Equipment Operational Control 1 4 1 (Valve Misalignment)

Equipment Operability Concern 0 3 4 (Technical Specifications)

Equipment Related Events 4 5 2 (Malfunction. Damage, or Degradation of Equipment)

Tagging Events 0 4 0 Radioactive Releases 0 4 0

= _-

. =Elam RYYEl MEMORANDUM l i

c

~

FC-CAG 96-061 4

DATE: November 1,1996 I FROM: M. A.Tesar l TO: Distribution I

SUBJECT:

Outage Status of Level 1,2, and 3 Condition Reports l

This report is provided to assess the event status of the Outage from the start date of October 5,1996 to October 31,1996. A comparison between the 1996 Refueling Outage and the 1995 Refueling Outage is provided on Attachment 1. Also, an assessment of the j

^

events by comparing the first two (2) weeks of the Refueling Outage to the next two (2) weeks up to October 31,1996 is provided so that managers can monitor more recent events compared to early outage events.

1 Attachment 1 is a comparison between the 1996 Refueling Outage and 1995 Refueling Outage.

1996 RFO 1995 RFO DESCRIPTION 4 0 l Equipment Actuation Events l

- Equipment Operational Control 4 1 i (Valve Misalignment) i 3 0 Equipment Operability Concerns (Technical Specifications) 5 4 Equipment Related Events (Malfunction, Damage, or Degradation of Equipment) 4 0 Tagging Events 4 0 Radioactive Releases

- During the 1995 Refueling Outage, the Significant problems were in Industrial Safety and Maintenance.

As-$101

. _ . . . * . . _ , - - m. . - _ . . _ . _ _ - _ - . . _ . _ . ... ~ . - . . - , - . . _ . . . . _ - . .

a i

i Distribution i Page 2 l

FC-dAG-96-061 -

l I

- Attachment 2 compares the events by two (2) week periods.

. Engineering

- Engineering problems have surfaced these last two (2) weeks.

1 l

= Equipment Operational Control (Valve Misalignment) l l

\

I' - This continues to be a concern, at the same rate as the first two (2) week period.  ;

. Equipment Malfunction, Degradation, or Damage

- Continues to increase, surpassing 1995. ,

. Tagging Incident

- One (1) additional tagging incident occurred during the second l

two (2) week period, totaling four (4) events compared to zero (0) during the 1995 Refueling Outage.

' . Radioactive Releases i

j - One (1) additional event involving radioactive release totaling four (4) events for this outage compared to zero (0) during the 1995 Refueling Outage.  ;

- Attachment 3 is a description of selected Event Codes which will be helpful as i you review the graphs in Attachments 1 and 2. )

- Attachment 4 is a list of Level 1 and Level 2 Condition Reports written after October 4,1996 which require a Licensee Event Report (LER). l i

k l I

d l

f Distribution i Page 3 I FC-CAO-96-061

SUMMARY

f i The review of the Level 1,2, and 3 Condition Reports during the 1996 Refueling Outage indicates continual problems in areas related to Operations as compared to Industrial j

Safety and Maintenance during the 1995 Refueling Outage. A possible explanation for

' this could be directly related to the lack of system windows and scheduling problems this outage. Without system windows, an increased workload has been placed on Operations as there is an increase in equipment manipulations and more tagging required.

l 4 Operations is, therefore, more vulnerable to equipment operational events. It is'possible I

that we have helped Maintenance by eliminating system windows at the expense of an increased burden on Operations.

How this event status relates to safety t,ignificance can best be summarized by reviewing l the LER Status. Currently we have five (5) LERs this Outage, whereas, we had two (2) j LERs during the 1995 Refueling Outage. Four (4) out of the five (5) 1996 Refueling l

Outage LERs occurred during active manipulations ofinstalled equipment as opposed to an event that occurred when the system was out-of-service for testing or maintenance (see j

i Attachment 4).

a We are experiencing a negative trend compared to the 1995 Refueling Outage. We must focus on safety in every aspect of our jobs to help each other succeed.

1 If you have any questions or require additional information, please call me at extensio 7250.

j W

M. A. Tesar Manager Corrective Actions Group MAT:lah Attachments

M Distribution Page 4 FC-CAG-96-061 Distribution:

W. G. Gates 8W/EP 1 J. K. Gasper FC-2-4 R. L. Andrews FC-2-4 J. L. Skiles FC-1-3 S. K. Gambhir FC-2-4 H. J. Faulhaber FC-1-1 T. L. Patterson FC-2-4 M. R. Core FC-1-9 J. W. Chase FC-1-1 R. G. Haug FC-2-1 R. L. Phelps FC-1-1 R. L. Jaworski FC-2-4 '

O. J. Clayton FC-2-4 R. L. Wylie FC-1-8 W. J. Ponec FC-2-4 ' R. W. Short FC-1-1 H. J. Sefick FC-2-4 R. G. Conner FC-3-1 S. J. Willrect FC-1-5 C. J. Brunnert FC-2-2 D. R. Trausch FC-1-1 D. E. Spires FC-1-1 M. T. Frans FC-1-1 J. B. Herman FC-1-1 S. W. Gebers FC-1-1 T. J. Melvor FC-2-1 E. R. Lounsberry FC-2-2 F. C. Scofield FC-2-4

' O. C. Bishop FC-1-1 c: CAG Memo File FC-1-1

ATTAC. .4T 1 .

L I I I I I -l .__

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l. I I I e1996 RFO 01995 RFOl8 Event Codes on Level 1-3 CRs vs Significant irs __

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INS LBL MNT MOD MSC MWO NCR OA REG REL SAF ST TAG ADM C~lR CON DOC ENG EQA EQC EQO EQU FP 1995 RFO data is for entire RFO 1996 RFO data is for CRs written after 10/4/96 i in 7 irs was Significant 1 in 7 CRs made Level 1,2 or 3 Data Date: 10/31/96

i

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, 1 ,. ._ .. - '

tsum.1ois to 1ott9 a1o/20 to 1o/31 -+-- t ois to 10/31 i  ;

1996 RFO Level 1,2 & 3 CR Event Code Comparison ~

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NCR REG REL ST TAG EQO EQU INS LDL MSC ADP CON ENG EQA EOC ADE Data Date: 10/31/96

ATTACHMENT 3 EVENT CODE DESCRIPTION CODE ENG ENGINEERING:

- Events involving Design Basis concerns such as invalid assumptions, values or criteria contained in design basis documents. plant systems or components which do not meet design basis requirements or the identification of previously unanalyzed conditions.

EQA EQUIPMENT ACTUATION:

- Events involving unplanned or inadvertent equipment actuation. This event code would include events which occur during the course of operations, testing and maintenance and are not part of the normal expected sequence of events or system / component response.

EQC EQUIPMENT OPERATIONAL CONTROL:

- Events involving unauthorized manipulation of plant system or equipment, unauthorized performance of maintenance on equipment, or misalignment of valves or control switches.

EQO EQUIPMENT OPERABILITY:

- Events involving Technical Specification equipment operability concerns. This event code would include unexpected Tech Spec entry due to equipment malfunctions, exceeding Tech Spec allowable outage times for inoperable equipment or violating equipment related Technical Specifications.

EQU EQUIPMENT:

- Events involving equipment malfunction, damage or degradation.

LBL EQUIPMENT LABELING:

- Events such as incorrect or missing equipment labels.

MNT EQUIPMFNT MAINTENANCE:

- Events associated with preventative or corrective maintenance of plant equipment.

EVENT CODE DESCRIPTION CODE i

MWO MAINTENANCE WORK DOCUMENT:

1

- Events involving the planning, approval, performance or review of maintenance work requests, maintenance work order or other maintenance work documents.

]

NCR OPERATIONAL NONCONFORMANCE:

- Events involving deficiencies such as physical defects, test failures, inconect or inadequate documentation, or deviations from prescribed processing, inspection or test procedures, which render the quality of an 4

operable plant component (CQE, LCQE, FPM, Radioactive Material Packaging) unacceptable.

^

REG REGULATORY:

- Events associated with interactions or commitments to outside regulatory or governmental agencies.

REL RADIOACTIVE RELEASE:

4

- Events involving actual or potential radioactive releases not in l

' compliance with restrictions contained in the Off-site Dose Calculation

' Manual (ODCM).

SAF INDUSTRIAL SAFETY:

- Events involving industrial safety concerns, lost time accidents and housekeeping issues which represent a personnel safety issue.

TAG EQUIPMENT TAGGING:

- Events involving the plant equipment tagging process as established in Standing Order G-20A. These may be problems with the preparation and review of tagouts, or the hanging, removal or verification of tags. This also includes events associated with plant related Hold Order tags issued by System Operations.

1 I

i l

l l

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..- .... - . - . . . - - . - . -_. - . - . . - - - - ~ .. - -_-.- ... _ -

ATTACnMENT 4  :

LEVEL 1 AND LEVEL 2 CONDITION REPORTS WRITTEN AFTER OCTOBER 4,1996 ,

REQUIRING AN LICENSEE EVENT REPORT OWNER DESCRIPTION DATE LEVEL CONDITION REPORT NO.

OP-3 A directs to place the steam generator low pressure f October 5,1996 I J. D. Kecy CR 199601203 trip switches in bypass when the Reactor Protection System pre-trips are received. During the cooldown to 400*F, the LSO did not give timely direction to the ,

Reactor Operators to bypass the...

D. E. Spires After removing steam generator primary side manway October 12,1996 2 CR 199601238 radiation levels in Containment increased above the Containment Radiation digh Signal setpoint resulting in a Ventilation Isolation Actuation Signal (VIAS). The  ;

VIAS automatically secured the...

C. N. Bloyd As found set pressure for RC-142 was determined to be October 17,1996 2 CR 199601272 about 3% below the specified seapoint of 2485 psig. l Technical Specification 2.1.6 allows a tolerance of + or

- 1% on the set pressure of the pressurizer safety valves.

This testing was performed...

Containment closure was breached during fuel movement  ;

October 18,1996 I J. D. Kecy  ;

CR 199601279 due to Maintenance activities on HCV-1388A. Once the Control Room was notified, refueling operations were stopped.

J. D. Kecy Fuses were pulled on YCV-1045 to support various CR 199601346 October 30,1996 I Maintenance activities. When the fuses were pulled, YCV-1045A and YCV-1045B opened. YCV-1045A and YCV-1045B are on OI-CO-4 tables "D" and "F' to remain closed for containment closure.

i

_ __._. _____._________________.______._____________m _ _ _ _ _ _ . _ _ _ . _ _ . _ . _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ _ _ .- __ __

Configuration Control Recent Issues:

1. Improper authorization for removal of Post Accident Sampling System (PASS) equipment.

2. Changing of Like-for-Like components prior to authorization by engineering.

3. Lessons learned from Industry.

4. Insights from 50.54(f) project.

History:

1. FCS completed Design Basis Reconstitution project in 1990.

2. Several walkdowns were conducted to verify that the physical plant matched with the paper plant.

3. Modifications implemented prior to the DBD project were reviewed for adequacy of 50.59's.

4. 1770 open items were identified. Most (about 93 %) of them have been closed out.

5. Many new calculations have been generated as a result of the DBD project.

6. New modification control and 50.59 procedures were developed.

7. A USAR upgrade project was completed in 1995.

8. A Procedure upgrade project was completed in 1993.

9. A Vendor Manual upgrade project was completed in 1991.

Actions Taken as a Result of Recent Issues:

1. Training was conducted with the maintenance planners on the recent Like-for-Like replacement problems that have been noted.

2 Management expectation with regard to Configuration Control are being emphasized with j

maintenance personnel during the quarterly maintenance conferences that are currently being i conducted.

3. S0-M-101 has been modified to implement a new streamlined process for nonsignificant <

replacement issues. PED-QP-2 is being revised to match the SO-M-101.  !

4. A Causal Factor Analysis was completed for the issues identified up to the end of December 1996. The analysis is going to be updated for the most recently identified issues.

5. The issue of Configuration Control was discussed and management expectations reenforced during a recent site wide meeting.

6. Configuration Control issues have been discussed in the several recent Engineering Support Program (ESP) continuing training sessions. Topics have included Millstone problems and issues as well as the recent FCS problems previously noted. '

l GOAL: Zero tolerance for unauthorized Configuration Control changes. i Future actions:

1. Conduct an SSFI every other year to verify that procedures, DBDs USAR, Drawings, calculations etc., are being adequately maintained and are consistent with the approved design basis.

2. USAR verification will be conducted for all safety related and safety significant systems per NEI guidance.

3. Ongoing vigilance from System Engineers.

Updated through 1/31/97 Type of Discrepancy Plant Documents Physical Design Documents Discrepancy Cause (definitions on next page) TOTAL Cause Unknown 5 10 2 17 Error Updating Plant Documents for Configuration Change 16 16 Unauthorized Configuration Change 16 16 Discrepancy Due To Original Equipment 19 19 Old Modification Caused Discrepancy 3 18 6 27 Recent PE or ECN caused Discrepancy 3 3 Maintenance Activity Caused Discrepancy 18 18 Error in Original Documentation 1 11 12 Error in Updating Design Docunertation for 19 19 Configuration Change Original Building Docunents 4 4 Cause Not Detennined 5 TOTAL 24 84 43 157 The definitions of the causes are:

1. Cause Unknown - The investigation of the discrepancy was unable to identify a cause.

2. Error in Updating Plant Documents for Configuration Change - A plant docunent was not updated for a configuration change or an error was made when the docunent was revised.

The configuration charge was made using the OPPD's revised design engineering and configuration control processes implemented following the SSOMI inspection.

Training has been conductedfor Engineering Support Personnel emphasizing the importance of identifying all required document changes and the importance of all SAURT members identifying document changes during the SAMRTreview.

3. Unauthorized Configuration Change - A configuration change was made without the proper d6cumentation or an unauthorized change in the construction work order was made.

Training has been conductedfor Engineering Support Personnel emphasizing the importance of completing all required documentation and obtaining the necessary authorization prior to making any configuration changes.

4. Discrepancy Due To Original Equipment - A deviation between the original design basis and the as-built plant was identified and no configuration changes could be identified which would have modified the as-built plant.

5. Old Modification Caused Discrepancy - A modification was incorrectly installed using the design engineering and configuration control processes in place prior to the SSOMI inspection.

6. Recent MR or ECN Caused Discrepancy - A modification or configuration change was incorrectly installed using the design engineering and configuration control processes implemented following the SS0MI inspection.

1 l

/

\

l

' I Updated through 1/31/97 l

?. Maintenance Activity Caused Discrepancy - A configuration change was made using the Maintenance Work Order process. I

8. Error in Original Documentation - A deviation between the original design basis documents and the as-built plant was identified and no document changes could be identified which  !

would have modified the design basis documents

9. Error in Updating Design Documentation for Configuration Change - A design document was not updated for a configuration change or an error was made when the document was revised. The configuration change was made using the OPPD's revised design engineering and configuration control processes implemented follcwing the 550M1 inspection.

Training has been conductedfor Engineering Support Personnel emphasizing the importance of identifying all required document changes and the importance of all SMART members identifying document changes during the SMARTreview.

10. Original Building Doctsnents Errors or omissions are identified in documentation for non-safety structures (e.g. warehouse) constructed using comercial grade standards.

Effectiveness of Design Basis Reconstitution None of the CRs has identified any discrepancy in the Design Basis Documents. The design document discrepancies are associated with drawings which were not typically reviewed as part of the DBR Procedure Uonrade No CRs have identified discrepancies associated with the procedure upgrade project Labeline There are 19 CRs associated with label discrepancies. Of these 9 identify equipment labels that were not included in the scope of the labeling project. In eight cases the cause could not be identified and in two cases the labels were not properly specified in old modifications.

9

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08/2G 67 00:32 CIPD FCS PLR4T RDMIN + H DFW H NO.170 932 4

ASSESSMENT PLAN I. Assessment

Title:

Configuration Control II. Assessment Scope:

A. Prograannatic and Administrative Controls g B. Program FEectiveness i C. Training and Qualificadon Ill. Assessment Measures:

A. Programmatic and Administrative Controls Clear Policy Established which Defines Configuration Control and Its Requirements Formal Guidance for Implementing Changes in Accordance with

(' 10CFR50.59 Adequate Controls for Maintaining Configuration Control Integrated Review and Approval Process B. Program Effectiveness Condition Report Ristory Timely Implementation of Corrective Action Review of MWD's Implementing Configtvation Changes Field Observation of Maintenance / Modification Activities  ;

> Modifications I > Engineering Change Notices

> Temponuy Modifications

- Tagging /Equipraent Abandonment

' Interviews with Personnel Mamtaining Configuration Control Availability of Design Requirements in the Field. i Documentation is Maintained to Reflect Actual Plant Configuration / Current Design l' '

C. Training and Qualification Configuration Control Training for Engineering Configuration Control Training for Maintenance l

t ,

I IV. Conclusions of the effectiveness based on criteria in NRC Inspection ManualInspection i Procedures, INPO Perfonnance Objectives and Criteria and FCS Procedure Requirements, l

1

(DPGalle, RDMartin, JGKeppler, JHMacKinnon, JWShannon)--SARC Members 9 Omaha Pub!!cPowerDistrict RLJaworski, FC-2-4 RLPhelps, FC-1-1 TJMcIvor, FC-2-1 i

\

444 South 16th StreetMall SJWi!irett, FC-1-5 i Omaha NE66102-2247 RLWylie, FC-1-8 l JWChase, FC-1-1 l RWShort, FC-1-1 1 February 7, 1997 JBHerman, FC-1-1 ,

LIC-97-011 JBishop, FC-1-1 j HJFaulhaber, FC-1-1 '

U.S. Nuclear Regulatpry Commission SWGebers, FC-1-1 l

Attn: Document Control Desk DESpires, FC-1-1 Mail Station P1-137 )

i Washington, D.C. 20555-0001  !

References:

1. Docket No. 50-285

2. Letter from NRC (J. M. Taylor) to OPPD (F. M. Petersen), t

" Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy and Availability of Design Bases Information," dated October 9, 1996

SUBJECT:

Response to Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Adequacy and Availability of Design Bases Information On October 11, 1996, Omaha Public Power District (0 PPD) received the Reference 2 letter, which required submittal of a response under oath or affirmation within 120 days of receipt. Accordingly, this letter with its attachments constitutes the required response from OPPD; this information is intended to provide the U.

S. Nuclear Regulatory Commission (NRC) added confidence and assurance that Fort Calhoun Station is operated and maintained within the plant design bases, with a.y deviations identified and reconciled in a timely manner.

. The following information is provided to facilitate review of this submittal.

The response format consists of this transmittal letter with 3 attachments.

Attachment 1 is an affidavit; Attachment 2 includes the responses to the i

requested information items (A) through (E), including descriptions of design  !

review and reconstitution efforts. Attachment 3 is a listing of Acronym )

Definitions and Referenced Procedures. 1 1

The responses also contain references to important milestones in the history of Fort Calhoun Station (FCS). The Safety System Outage Modification Inspections (SSOMI) were conducted by the NRC in late 1985, as documented by Inspection i Reports 85-22 dated January 25, 1986 and 85-29 dated March 19, 1986. The subsequent SSOMI reinspection was documented by Inspection P.eport 88-200 dated September 16, 1988. These inspection reports identified several violations and other deficiencies associated with maintenance of the FCS design bases and processes used to design, install, and test modifications. These and other l problems led to FCS being placed on the list of plants requiring additional NRC j attention. OPPD commissioned an Independent Nuclear Assessment (INA) by an engineering consultant to identify areas needing improvements.

In order to best administer implementation of the numerous corrective actions resulting from the SS0MI and the INA, OPPD created the Safety Enhancement Program WGGates, GRWilliams, SKGambhir, JBKuhr, RGConner, DRPodoll, ALHale, ERLounsberry, JWTills, TCMatthews, FILE COPY (Pat)

t U. S. Nuclear Regulatory Commission LIC-97-011 4

Page 3 4

with its design bases, and that any deviations discovered are resolved in a j timely manner using the corrective action program.

I Please contact me if you have any questions.

i Sincerely,

} W. G. Gates

Vice President

Attachments

TCM/ tem c: Winston & Strawn F. J. Miraglia, Jr., Acting Director, Office of Nuclear Reactor Regulation L. J. Callan, NRC Regional Administrator, Region IV L. R. Wharton, NRC Project Manager W. C. Walker, NRC Senior Resident Inspector

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LIC 97 011 Attachment 2 Omaha Public Power District Responses to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Adequacy and Availability of Design Bases Information ,

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e LIC-97-011 Attachment 2 Page 1 Omaha Public Power District Response to NRC Request A l

(A) Description of engineering design and configuration control processes. I including those that implement 10 CFR 50.59, 10 CFR 50.71(e) and Appendix B to 10 CFR Part 50 I i

Executive Summary 1 This response provides a description of the Omaha Public Power District (OPPD) engineering design and configeration control processes, including those that implement 10 CFR 50.59,10 CFR 50.71(e> and Appendix B to 10 CFR 50. These cr] cesses are ,

designed to provide reasonable cssurance that Fort Calhoun Station (FCS) is operated l and maintained in accordance witt the design bases. The current programs and procedures were developed following the NPJ Safety System Outage Modification Inspection (SSOMI) in 1985. They implemented impro.ed eng4neering design and configuration control processes as Dart of the Safety Enhancement Program (SEP). in parallel with the Design Basis Reconstitution (DBR) Project. To ensure that activities such as modifications, procedure chaiges and safety evaluations could be conducted while the DBR Project was in progress, interim controls were established to ensure that design basis margins were not abrogated. These controls are discussed in the response to Request C.

Significant elements of the OPPD engineering design and configuration control processes are listed below.

1. Design work is reviewed for conformance with the design bases; if required, changes are made to the design bases. The Design Change Packages and supporting material (calculations, analyses, specifications, drawings, etc.) are developed using approved procedures and are maintained as controlled documents. (Section A.1)

2. System interaction analysis is used to assess the effects of a design change. The impact of a design change on plant procedures and training is assessed. Design verification is conducted in accordance with established procedures. Plant and Engineering management approval processes are defined. (Section A.1) 3.10 CFR 50.59 requirements are implemented. Proposed changes to procedures and the plant design, including the USAR, are evaluated for 10 CFR 50.59 applicability. If applicable. Unreviewed Safety Question determinations are performed. 'he review and approval process for 10 CFR 50.59 evaluations is defined. (Section A.4)

4. 10 CFR 50.71 requirements are implemented. Changes conducted in accordance with 10 i CFR 50.59 are reviewed to determine if USAR changes are required. NRC Safety j Evaluation Reports are reviewed and appropriately included in the USAR. (Section A.5) j

LIC-97-011 Attachment 2 Page 2

5. FCS personnel are trained and qualified to perform the processes and procedures for engineering design and configuration control (Section A.6).

6. To ensure conformance with the design bases for modifications designed and installed from initial commercial operation up to the Design Basis Reconstitution. OPPD conducted a safety evaluation check of Design Change packages (for safety related and nonsafety related modifications that had potential to impact safety systems).

Concerns identified through this review were dispositioned using the 080 open item process. (Section C.2)

7. Assessments conducted since the SSOMI provide assurance that effective processes and procedures have been adopted. These audits, surveillances and inspections also confirmed that deficiencies are promptly corrected. The audits of current programs and ongoing trending of configuration control discrepancies confirm that the current processes are effectively implemented and are producing quality products. These trends also confirm the adequacy of the Safety Evaluation checks performed for modifications designed and installed before the Design Basis Reconstitution.

A.1. Engineering Design & Configuration Control Processes I

The OPPD engineering design and configuration control processes for FCS are designed ,

to ensure configuration changes and document changes are conducted in accordance with l applicable rules and regulations including the requirements of 10 CFR 50 Appendix B l Criterion III. These processes are depicted on Figure 1.

The following processes are used to make configuration changes to FCS plant systems, structures and components:

1. Modification Request (MR). l

2. Facility Change Engineering Change Notices (FC-ECN).

3. Minor Configuration Change or Replacement (GEI-35), and

4. Temporary Modification (TM)

The Substitute Replacement Item Engineering Change Notice (SRI-ECN) and the Document Change Engineering Change Notice (DC-ECN) are processes used to revise design bases documents without changing the form. fit or function of the FCS systems. structures, and components. The procedures listed below and other documents provide detailed instructions for engineering design and configuration control at FCS.

1. Production Engineering Division Quality Procedure PED-0P-2. Configuration Change Control

2. Standing Order S0-G-21. Hodification Control

3. Production Engineering Division General Engineering Instruction. PED-GEI-3.

Preparation of Design Change Packages

i i LIC-97-011 Attachment 2 Page 3 4 Production Engineering Division General Engineering Instruction. PED-gel-29.

Facility Change Evaluation

5. Production Engineering Division General Engineering Instruction. PED-GEI-35, i Preparation of EARS for Minor Configuration Chsnges and Replacements

6. Standing order 50-0-25. Temporary Modification Control

7. Nuclear Operation Division Quality Procedure N00-0P-3.10 CFR 50.59 Safety l Evaluations

8. Nuclear Operation Division Quality Procedure. N00-0P-16. Updated Safety Analysis Report (USAR)

, 9. Production Engineering Division Quality Procedure PED-OP-3. Calculation Preparation. Review, and Approval

10. Production Engineering Division Quality Procedure PED-0P-5. Engineering Analysis Preparation. Review, and Approval d

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LIC-97-011 Attachment 2 Page 4 Figure 1 Engineering Design and Configuration Control Processes 10 CFR 50 App. B Criterion 111 lOPPD QA Plan l ANSI N45.2.11 Configuration Change

! Control PEO-QP-2 I

f f i f f 1 2 3 4 Temporary 6-Minor Faciuty Change Modification Subeutue Moddics5on Confoura5on ECN Recuest Replacement Cha W ltem Replacement SO-O-25 PED-gel-29 PED-gel-3 PEO-gel-60 PEMI-35 u ir 1r 1f. If if 6

5 9_ FocR Maintenance Document Work Document so o-21 Change PED-GE142 S0-M 101 PED-gel-51 " Y "

7 Construc6on Work Order SO-G-21 l

if lf k

l DesignBasesDocumentUpdate l I

1. Temporary Modification: Short term 5. Maintenance Work Document: Authorize, document alteration to the Plant that involve changes to ]

and control maintenance work activities i the design bases. 6. Field Design Change Request: Field change to an

2. Minor Configuration approved design change package or ECN Change / Replacement 7. ConstructionWork Order: Authorize, document and Simple changes that do not affect design control construction work activities.

pararneters, operaung conditions or functions 8. Substitute Replacemerit item: Authorize and

3. Facility Change ECN: Changes to non- document the equivalency for the replacement of a part or CQE systems, structures or components that component that does not have the same make, rnodel do not affect safety related design parameters, and/or part number as the origint.l.

operating conditions,or funcbons 9. Document Change: update of existing documents to

4. Modification Request: A change to the reRect as-built conditions.

form, fit, or function of a COE or Llrnited COE system, structure or component I

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LIC-97-011 l Attachment 2 Page 5 A,2, Configuration Changes The configuration change processes used at FCS (Modification Request (MR), Facility Change Engineering Change Notices (FC-ECN), Minor Configuration Change or Replacement (gel-35), and Temporary Moalfication (TM)) are described in the following sections.

Configuration Changes --Modification Requests Standing Order S0-G-21, Modification Control, establishes procedural guidance for initiating, reviewing, approving and canceling Modification Requests. Modification Requests generally involve changes to the station that are complex from a system interaction perspective and affect system design parameters as well as operating conditions. The Modification Request process must be used for any configuration change that involves:

1. a change, directly or indirectly, to the fit, form or function of a COE or a Limited COE system, structure, or component.

OR

2. a change to a Non-COE system, structure, or component that affects a COE or a Limited COE system, structure, or component.

COE designation corresponds to the more common industry designation " safety related."

Limited COE designation corresponds to the more common industry designation " Category II." The definitions of COE and Limited COE are:

Critical Quality Element (COE): Those structures, systems, components or items whose satisfactory performance is required to prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health  !

and safety of the public. ,

l Limited Critical Quality Element (Limited COE): Those structures, systems, components or items whose satisfactory performance is required to prevent or mitigate the failure of those structures, systems, components or items identified as Critical Quality Elements, Modification Requests are documented in a Design Change Package (DCP) in accordance with PED-gel-3. A DCP typically includes the following:

l 1. A description and evaluation of the problem addressed by the modification and the selected solution.

2. Any regulatory requirements associated with the modification. This includes the

! impact of the modification on the USAR, Technical Specification and licensing commitments.

3. Design Inputs

1 LIC-97-011 Attachment 2

, Page 6

4. Design Analysis

5. Systems Interaction Analysis

6. 10 CFR 50.59 Safety Evaluation

7. Operating Impact -

8. Installation and Testing Requirements

9. Drawing and Document Revisions DCP Deveinnmont -- Conformance with Design Rasis l

The information necessary to assess impact of a modification on the Design Bases is obtained from controlled documents including the DBDs, the USAR. drawings, j specifications and other controlled design documents. This information is assembled l in the Design Inputs section of the DCP. Using the Design Inputs and the guidance provided in PED-GEI-3, the conformance to the design bases of a modification is confirmed.

l DCP Deveinnmant -- Oncomentation of Sunnnrting Analysos Analyses and calculations to support a modification are included or referenced in the DCP. Cumulative changes are confirmed to be consistent with the design bases.

Calculations are detailed technical packages that include the equations, references.

data inputs, outputs. assumptions. and conclusions needed to substantiate engineering I decisions. Calculations are prepared reviewed and approved using PE0-QP-3. An l

Engineering Analysis (PEO-QP-5) may also be used to prepare supporting analyses. '

Engineering Analyses and calculations are reviewed and independently reviewed for analyses designated as COE. Limited COE and Fire Protection to provide design verification.

The review is performed by an individual technically qualified in the specific discipline or field of analysis. The analysis is reviewed for technical accuracy, completeness, and compliance with licensing commitments and procedure requirements.

Independent reviews may be performed by one of three methods.

1. Analysis review

2. Alternate calculation

3. Qualification testing DCP Opvolonment -- Systems interaction Analyses System Interaction Analyses are used to determine the impact of a modification on plant level design bases. (Plant level design bases documents are discussed in the' response to Request C.) Checklists are used to screen a modification for the following potential system and plant interactions.

r LIC-97-011 Page 7

1. Fire Protection

2. Special Service Engineering Programs

3. Motor Operated Valve Program

4. Electrical Equipment Qualification

5. High Energy Line Break

6. Seismic Interaction. Qualification, and Effect on the USI A-46 Program

7. Electrical System Analysis

8. Human Factors Review

9. Security System

10. Environmental and Radiological Release

11. Materials Compatibility

12. Containment Integrity

13. Control Room Habitability

14. Missile Protection

15. Structural Impact

16. Independence Criteria

17. Single Failure Criteria

18. Possibility of Operator Error

19. Heavy Loads

20. Impact on HVAC

21. Use of Vendor Procedures

22. Station Blackout This analysis confirms that the modification conforms to the design bases.

DCP Develonment.-- Safety Evaluation (10 CFR 50 59)

The modification DCP contains the 10 CFR 50.59 Safety' Evaluation. The Safety Evaluation addresses the three stages of the modification: Design. Installation and Testing. The safety evaluations are prepared and reviewed by engineers who are traired and qualified.

The Design Safety Evaluation ensures that the modification, as installed, will not introduce an unreviewed safety question. The safety evaluation also identifies the USAR changes that are needed as a result of the modification. (The USAR is later updated 1 in accordance with procedure N00-0P-16.)

The Installation Safety Evaluation addresses the specific configuration requirements, restrictions on interfacing equipment, systems and structures and the impact on the safety of the unit during construction.

The Testing Safety Evaluation assesses the modification testing requirements, including operational, functional and special tests, and associated test acceptance criteria.

Details of the 10 CFR 50.59 safety analysis process are provided in Section A.4.

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LIC-97-011 Attachment 2 Page 8 I 'DCP Opveinnmant -- Onoratina imnact

! The impact of a modification on As Low As Reasonably Achievable (ALARA) exposure goals.

Training. Testing. Operations and Maintenance is evaluated. The ALARA evaluation '

i describes the impact of the modification on overall worker exposure during construction. -

! and subsequent operation and maintenance of the components. The Training evaluation i documents training program changes required due to design considerations or equipment -

' selection. Impact on the Training Simulator is also addressed. . The Testing evaluation <

describes any new or revised' periodic testing requirements (e.g.. changes to Inservice Inspection programs, etc.) for the modification. The Operation evaluation identifies j changes to the operation of equipment including revision of operating procedures. The i . Maintenance evaluation describes any new or revised periodic maintenance requirements

[

due to installation of the modification.

i DCP Deveinnment -- Revisions to Proendures and Desian Rasis Dornments -

j The DCP contains draft additions or revisions to the FCS Operating Manual procedures i

and documents. Drafts of the additions or revisions to the USAR. DBDs and other design bases documentation are also provided in the DCP.

DCP Opveinnmant -- Installation and Testina i The DCP provides a sumary of the construction specifications that are required to

! install and test the modification. The package addresses the major construction aspects of the work and integration of plant outage and operational requirements _for construction. Testing requirements are identified, evaluated and sumarized. The test specifications include testing the component (s) and. system to ensure functionality in i

actual operation. This may include any unique or special criteria for interruption and continuation of a test requests for documenting, and/or review of unusual conditions.

The Construction Work Order (CWO) is used to authorize. document and describe c construction work activities, including required QC inspections. for the installation

, of modifications. The CWO work instructions are developed from the installation and

technical description sections of the DCP.

DCP Opveinnment -- Review Anoroval and Accentance  !

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The DCP is reviewed to ensure technical adequacy of the modification. An Independent '

Review is performed for DCPs affecting COE. Limited CQE. Fire Protection or Security systems or components. The Independent Review is performed by someone other than the j preparer who is technically competent in the specific discipline or field of analysis.

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LIC-97-011 Page 9 The DCP is reviewed and approved by the Station Modification Acceptance.and Review Team (SMART). a multi discipline team composed of personnel from Design and System Engineering. Operations and appropriate suppcrt groups.

Following SMART approval of the DCP the DCP is presented to the PRC for review and recommendation of approval. in accordance with Technical Specification 5.5.1.6. In parallel, the Nuclear Safety Review Group (NSRG). on behalf of the Safety Audit and Review Committee, reviews the DCP in accordance with Technical Specification 5.5.2.7 to verify that the modification does not constitute an unreviewed safety question. -

Final approval of the DCP is by the Manager - FCS.

Following installation of the DCP. the following steps are completed to return the new or modified equipment to service:

1. Completion of installation and testing of a modification is verified by Engineering and Quality Control personnel.

2. Engineering personnel prepare the as-built drawings panel schedules and electrical load distribution lists.

3. The appropriate System Engineer and Operations personnel update the Operating Manual procedures necessary for operation of the modified system. ,

Approval for return to service (Accepted for Operability) is authorized by the Manager-Operations. Manager-Maintenance. Manager-System Engineering, and the Shift Supervisor. ,

DCP Develonment - Field Changes A Field Design Change Request (FDCR)is used to change a DCP after the DCP has been approved for construction. The FDCR documents the reason for the change and the documents affected. The affected documents are revised and attached to the FDCR.

Engineering reviews the requested FDCR for conformance with scope. intent and design basis of the DCP. This includes a review of the 10 CFR 50.59 Safety Evaluation included ,

in the DCP. The FDCR is screened to determine if Plant Review Committee (PRC) review is required. The FDCR requires PRC review and recommendation of approval to the Manager-FCS if the change reduces the ability of a system, component or structure to perform its design function, affects the conclusion of the 10 CFR 50.59 Safety i Evaluation'or significantly changes the modification scope. An FDCR not requiring PRC review is approved by the appropriate Engineering department head. After approval of '

. the FDCR. the revised documents are inc'orporated into the original DCP. The

-Construction Work Orders are then revised to incorporate the FDCR.

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LIC-97-011 1 Attachment 2 1 Page 10

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DCP Develonmant - Modification close-nut Upon verification that all portions of a modification are completed, the Modification Completion Report is prepared to initiate the modification close-out process. This process includes the following steps listed.

1. Quality Control verifies that all construction and testing have been cocpleted.

Items verified are as follows:

a. All CW0s have been completed and approved.

b. All pre-operational testing has been completed and approved.

c. All FDCRs have been reviewed, approved, and implemented.

d. Independent Design Verification or Independent Review is complete.

e. Pre-operational Tests have been completed.

f. All changes to the construction drawings and drawing list have been documented by FDCRs.

g. Construction and testing documentation is complete.

h. All Condition Reports which could affect operability of the modification have been resolved.

2. The appropriate System Engineer verifies that the modification has been installed, tested, and Accepted for Operability in accordance with the approved DCP and any FDCRs.

3. The construction drawings and the plant drawings are verified to ensure they reflect the as-built condition.

4. Labels are installed for all equipment affected by the modification.

5. Any additional revisions to the Operating Manual are completed. The documentation revised includes applicable:

Abnormal Operating Procedures (A0P)

Emergency Operating Procedures (EOP)

Annunciator Response Procedures (ARP)

Operating Procedures (09)

Operating Instructions (01)

Calibration Procedures (CP)

Maintenance Procedures (MP)

Preventive Maintenance Procedures (PM)

Emergency Plan implementing Procedures (EPIP)

Radiological Emergency Response Procedures (RERP) i Security Procedures l Surveillance Test Procedures (ST)

Standing Orders (S0) 1 j

LIC-97-011 Attachment 2 Page 11

6. The Vendor Manuals are updated and changes to Preventive Maintenance practices are made.

7. The affected System Training Manuals are revised.

8. The EE0 Program is updated.

9. The recommended spare parts are identified.

10. The ISI Program Plan is updated to reflect changes made by the modification.

11. Training materials are updated and training is conducted or scheduled.

12. The Maintenance Rule scoping documents are updated.

13. A PRC subcomnittee reviews the DCP and close-out documentation to verify that the modification was correctly installed and tested, and that appropriate documentation is revised. Following completion of its review the PRC subcommittee recommends acceptance of the modification to the PRC.

14. The PRC accepts the modification.

15. Following acceptance, the PRC notifies the NSRG of their acceptance of the modification. The NSRG conducts a review of the Safety Evaluations.

16. The "as-built" DCP information is incorporated into the DBDs and other design documentation.

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17. The DCP is processed for permanent retention. j i

Facility Changel CN l I

The Facility Change ECN allows simple configuration changes to FCS equipment and l supporting facilities. These changes: )

1. Do not directly or indirectly affect COE or Limited COE systems, structures or l components:

2. Do not affect the Technical Specifications:

3. Do not result in an Unreviewed Safety Question: and i

4. Do not directly or indirectly affect Nuclear Safety.

Facility Change ECNs are documented in an ECN in accordance with PED-GEI-29. The FC-ECN provides the information necessary to construct and test the modification.

1 LIC-97-011 ,

Attachment 2 l Page 12 EC-FCN Opveinnment The processes used to evaluate the conformance with design bases, conduct the systems interaction analysis, determine the operating impact and determine installation and testing requirements for a FC-ECN are similar to those required to develop a MR. The processes are reduced in scope compared to the MR because a FC-ECN may not directly or indirectly affect nuclear safety.

FC-FCN Develonment -- Safety Fvaluation (10 CFR 50 59)

Facility Change ECNs require a documented safety evaluation to ensure the configuration change does not result in an unreviewed safety question or require a Technical '

Specification change. This evaluation is performed in accordance with N00-0P-3 and attached to the ECN. The safety evaluation considers all aspects of the Facility Change i including design, installation and testing phases. I FC-FCN Opvolonment -- installation and Testing i

FC-ECNs are installed and tested via the CWO process in a manner similar to that used '

for Modifications.

FC-FCN Opveinnment -- Review Annroval and Accentance FC-ECNs are reviewed for technical adequacy by someone other than the Preparer competent in the specific discipline, or field of analysis. An Independent Review is performed for all FC-ECNs affecting Fire Protection. Radioactive Waste Disposal or Security. The Independent Review is performed by someone other than the Preparer who is technically competent in the specific discipline or field of analysis. An Engineering Department Head ECN review is also conducted as an overview to ensure commitments are met and the package is complete.

l FC-ECNs are reviewed and approved by the Manager-FCS or Assistant Plant Manager-FCS prior to installation.

Completion of installation and testing of a FC-ECN is verified by Engineering. The l close-out process for FC-ECNs is the same as that used for MRs. I Minor Configuration Change ne Renlacement l

A Minor Configuration Change or Replacement is a change to the facility which does not alter the design, function. or method of performing the function of a component, system, or structure described in the Safety Analysis Report. A Minor Configuration Change or .

Replacement does not require detailed engineering and is simplistic in character.  !

LIC-97-011 Page 13 To demonstrate a minor configuration change or replacement does not alter the design, function, or method of performing the function of a component, system, or structure described in the Safety Analysis Report, a 10 CFR 50.59 screening evaluation per N00-0P-3 is required.

A Minor Configuration Change or Replacement is installed and tested using a Maintenance Work Document (MWD). Post Maintenance Testing requirements are specified within the MWD. The appropriate System Engineer reviews the MWD for adequate Post Maintenance Testing requirements. Documents that require updating as a result of the change (e.g.,

drawings, procedures, manuals, data base, etc.) are updated using a Document Change Engineering Change Notice.

Configuration Control During Maintenance Activities Standing Order 50-M-101. Maintenance Work Control, identifies the methods for control of maintenance activities and defines the different methods of configuration control for maintenance work.

1. MWDs are verified to ensure that no configuration changes are proposed unless properly authorized.

2. Craft Personnel are responsible for verifying those components being replaced do not cause an unauthorized configuration change. Prior to beginning work on the maintenance work package, the craft personnel are responsible for comparing the parts and materials against approved drawings and/or design change documents to ensure that configuration control will be maintained.

3. The appropriate System Engineer is responsible for performing the technical review of maintenance work documents and ensuring that no unauthorized configuration changes are performed. The System Engineer also ensures adequate post maintenance testing is identified.

4. The Shift Supervisor is responsible for verifying the tag out boundaries such that the plant is operated in accordance with the Technical Specifications.

Iemporarl Modifications Temporary Modifications (TM) are temporary minor alterations made to plant equipment that do not conform with original drawings or other design documents. These alterations are temporary in that they are expected to be installed for a short duration.

Temporary Modifmations receive a screening for applicability of 10 CFR 50.59 and an unreviewed safety question determination if 10 CFR 50.59 is applicable. During this process, the design, installation and testing phases of each temporary modification are

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l LIC-97-011 Attachment 2  !

Page 14 evaluated. The technical evaluation of a temporary modification *s impact is assessed with regard to items such as fire hazards. electrical loading, environmental qualifications, flooding, system interaction, and associated procedures.

Temporary Modifications are required to be reviewed and approved by the PRC prior to their installation. The Shift Supervisor may authorize installation prior to PRC approval in case of an emergency. Emergencies include situations when it is necessary to increase the margin of plant safety. prevent injury to personnel or to prevent damage to equipment. The Temporary Modification must be reviewed and approved by the PRC within two working days after installation. Temporary Modifications are installed using the MWO process. Temporary Modifications are identified on controlled drawings in both the Control Room and the Operations Control Center.

The use of a Temporary Modification for an extended period of time is minimized.

Installed temporary modifications are reviewed by the PRC on a six-month frequency.

Temporary Modifications that require an outage for removal are expected to be installed for only one operating cycle. Those that can be removed during plant operation are expected to be removed within six months of installation. Progress on removal of Temporary Modifications is tracked on a weekly basis through a report to management and monthly basis by a performance indicator. Operator awareness of installed temporary modifications is maintained by requiring *.he shift supervisor to review the TM log during shift turnover.

A.3 Design Document Changes Involving No Changes to Design Bases A DC-ECN is used to authorize the evaluation and update of existing documents to reflect as-built conditions. An SRI-ECN is used to ensure equivalency of the replacement for a part or component that does not have the same make. model and/or part number as the original.

Document Chanan Fnaineerino Chance Notice -- DC-FCN Document Change ECNs (DC-ECNs) are used when discrepancies are found between design and station documents or when field conditions do not match the design or station documents.

DC-ECNs may only be used when the proposed change is substantiated by existing design basis documents and does not change the current station configuration. When document discrepancies are identified, they are evaluated by Engineering to ensure that the existing design basis supports the As-Built condition. If, based on that evaluation. ,

the field condition is not substantiated by the existing design bases. the DC-ECN is I canceled and the condition is documented and resolved via a Condition Report in I accordance with Standing Order R-2.

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LIC-97-Oll Attachment 2 i Page 15 I l

DC-ECNs do not require .a safety evaluation because there are no proposed changes or tests to the facility that could impact the Technical Specifications, USAR,'or basis -

)

'thereof, and/or result in an unreviewed safety question.

A review of a DC-ECN is conducted by an engineer other than the preparer. The ,

l appropriate Engineering department head approves a DC-ECN. Upon approval the appropriate changes are made to the documentation.  ;

l Substitute _ Replacement item Engineering _ Change Notice -- SRI _ECN l

' An SRI-ECN is used to conduct and document an engineering equivalency evaluation of

replacement parts or components that do not have the same make, model and/or part number i

as the original. This evaluation ensures that interface, interchangeability, safety, fit and function requirements are maintained in accordance with applicable regulatory '

or . code requirements. The SRI-ECN Equivalency Evaluation performed per PED-GEI-60 l demonstrates that the Substitute Replacement Item is being procured.to an equivalent specification. If Equivalency cannot be demonstrated, then an SRI-ECN is not applicable.

Provided that equivalency of the SPI has been demonstrated, the component or part replacement will not result in a change to the facility as described in the USAR.

Therefore, a 10 CFR 50.59 safety evaluation is not required.

The SRI-ECN documents that the Substitute Replacement item is equivalent to the original item for all critical design characteristics. The SRI-ECN also documents that the design basis of the plant, system or component is not altered by the Substitute Replacement Item.

Each SRI-ECN is reviewed for technical adequacy by someone competent in the specific discipline or field of analysis and by someone other than the preparer. An Independent Review is' performed for all SRI ECNs affecting COE, LCOE, Fire Protection, Radioactive Waste Disposal or Security. The appropriate System Engineer reviews the SRI-ECN and performs an overall . check to ensure the SRI will accomplish its objective. The Engineering department head reviews and approves the SRI-ECN.

The part .or component approved by the SRI-ECN is installed using the MWO process discussed in Section A.2.

A.4 Safety Evaluations Implementing 10 CFR 50.59 10 CFR 50.59. Paragraph (a). Item 1, permits changes to FCS and its operation as  !

described in the USAR without prior NRC approval, provided a change in the Technical  !

Specifications is not involved or an Unreviewed Safety Question is not created.

Proposed changes that are determined to involve a change in the Technical Specifications 1

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! LIC-97 011

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i or an Unreviewed Safety Question must be submitted to the NRC in accordance with 10 CFR -

}- 50.90 and 50.92 and 10 CFR 51.21.

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Information from the NRC Inspection Manual. Part 9900. April 4.1996: NRC I&E Manual, l

! Part 9800. January 1.1984: NRC Inspection Manual. Inspection Procedure 37001: NSAC-125:

and NSAC-105 was used to develop the 10 CFR 50.59 Safety Evaluation program at FCS. The i program is defined by Nuclear Operation Division Quality Procedure N00-QP-3.10 CFR

. 50.59 Safety Evaluations.

i N00-09-3 is the only procedure used at FCS for the preparation of 10 CFR 50.59 safety j evaluations. The activities which must be evaluated for potential unreviewed safety

questions in accordance with this procedure are:

1. .All proposed procedures and proposed changes to procedures governing activities 1

at FCS that could affect nuclear safety.

1

. 2. All proposed tests or experiments that could affect nuclear safety.

[ 3. All proposed changes or modifications to plant systems or equipment that could

! affect nuclear safety.

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__N00-QP-3 is comprehensive in its guidance:

l 1. The procedure presents a series of questions for review, followed by c detailed i discussion of areas to be reviewed in responding to the question. The guidance 1

provided in this document is structured to provide assurance that relevant aspects of the activity are considered by the preparer and identified for evaluation by the reviewers.

2 2 The procedure specifies the qualification and training requirements which must j be satisfied by personnel preparing and reviewing nuclear safety evaluations.

3. The. procedure provid's e the necessary direction for the accumulation of the 1

information needed for a report to the NRC describing those activities performed 1

under the provisions of 10 CFR 50.59. Paragraph (a). Item 1.

4. The procedure contains definitions of key words and phrases pertaining to the 4

safety evaluation process to provide those individuals who are assigned preparation and review responsibilities with a consistent interpretation of their meaning.

The safety evaluation is a two step process. The first step is the 10 CFR 50.59 Applicability Screening: the second is the Unreviewed Safety Question Determination,

. . ~ _ _. .-._, . m. __ - . .

i LIC 97-011 AttacFment 2 Page 17 ScreeningELDetermine Comnliance witlLUSAILand_IechnicaLSpecifications i i

l A guide is provided for documenting the 10 CFR 50.59 applicability screening process.

This guide requires answers to the following series of questions.

1. What activity is being evaluated?

2. What is being done? l

3. Why is this being done? i

4. Does the activity involve a change to the Technical Specifications?  !

5. Does the activity being evaluated involve changes in the facility as described l in the USAR? (N00-0P-3 includes guidance on assessing interim changes not I included in the USAR.)

6. Does the activity involve proposed tests or experiments not described in the l USAR?

~7. Could the activity adversely affect nuclear safety?

The Applicability Screening is used to determine if the activity affects the USAR, Technical Specifications, or License Conditions for. FCS, When the responses on an Applicability Screening indicate that the change affects the USAR, an unreviewed safety )

question determination is initiated. If, however, the activity involves a. change to

'the License Conditions or Technical Specifications, other than the basis section, the preparer is directed to process the change in accordance with Nuclear Operation Division Quality Procedure N00-0P-7, Facfif ty Lfcense Changes (FLC). This screening ensures compliance with the requirements of both 10 CFR 50.59 and 10 CFR 50.92.

'Unreviewed Safety Question Determination The applicability screening determines which activities and modifications require evaluation of the unreviewed safety question criteria contained in 10 CFR 50.59.

Paragraph (a), item 2, The three criteria as stated in 10 CFR 50.59 have been expanded into several questions for clarity in responding to the requirement. The three criteria center on four main areas of evaluation:

1. Probability of an accident or equipment malfunction

2. Possibility of a previously unevaluated accident or equipment malfunction

3. Consequences of an accident or equipment malfunction

4. Reduction of the margin of safety A guideline is provided for documenting the development of background information and the unreviewed safety question determination for the proposed activity. Completion of this documentation results in a positive or negative determination. Any unreviewed safety question identified must be submitted to the NRC for approval as part of a license amendment application.

LIC-97 011 Attachment 2 Page 18 Review and Annrnvals Unreviewed safety question determinations are presented to and reviewed by the Plant Review Committee (PRC) and approved by the Manager-FCS. PRC makes a recomendation for approval or disapproval to the Manager-FCS.

All approved Safety Evaluations are reviewed by the Safety Audit Review Comittee (SARC) through the Nuclear Safety Review Group (NSRG), acting on behalf of the SARC. -The NSRG provides a bimonthly status report of this independent review and evaluation to the Safety Audit Review Comittee (SARC).

-A5 10 CFR 50.71(e) USAR Updating Process Procedure N00-QP-16 provides assurance of compliance to 10 CFR 50.71(e). Revisions to the USAR are identified and documented during the 10 CFR 50.59 safety evaluation process and reviews of NRC Safety Evaluation Reports.

identifying tlSAR Revisions for Iicense Amendments and New Safety issues The USAR update procedure. in conjunction with the procedure for license amendments, requires that a review of NRC Safety Evaluation Reports be conducted to determine any changes to the USAR and Design Basis Documents. This ensures that " safety evaluations l performed by the licensee in support of requested license amendments" are all reviewed and appropriately included in the USAR as required by 10 CFR 50.71(e).

Other SERs are reviewed for potential revisions to the USAR. This. ensures that "all analyses of new safety issues performed by or on behalf of the licensee at Comission request." are appropriately reviewed for inclusion into the USAR.

A.6 Training The Engineering Support Personnel (ESP) Training Program provides the training for those individuals who are engaged in engineering support functions at FCS.

The design and development of the instructional material used in the Engineering Support Training Program are based on National Academy for Nuclear Training document ACAD 91-017. Guide 11nes'for Training and Qualification of Engineering Support Personnel. The Instructional Systems Design (150) process for performance-based training was utilized to implement the ACAD guidelines. The training program consists of two components:

initial training, which includes orientation and position-specific training, and continuing training.

The objective of the Engineering Support Training Program is to provide engineers with job-related skills and knowledge to supplement their formal education. Initial training

LIC-97-011 Page 19 provides technical knowledge of station operation related to the individual's job responsibilities. It also provides participants "'ith the skills to perform assigned duties independently in a manner that promotes safe and reliable plant operations.

Continuing training helps participants maintain and enhance their job performance and develop a broader scope and depth of knowledge and skills.

Maintenance personnel re aive training on the requirements for configuration control as part of the training conducted on 50-M-101.

A.7 Audits and Inspections Internal Quality Assurance (0A) audits have evaluated the engineering design and configuration control processes both during normal (SSFI type Audit) operation and during refueling (SS0MI type Audit) operations. Additionally. Quality Assurance personnel perform additional audits of both the Engineering Configuration Management program and the Station Engineering program. Several NRC inspections have also assessed the FCS engineering design and configuration control processes. Discussions of these audits and inspections is included in the responses to Requests C and E.

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LIC-97-011 Attachment 2 Page 20 Omaha Public Power District Response to NRC Request B (B) Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures t Executive Summary Omaha Public Power District (OPPD) processes, procedures, and programs provide reasonable assurance that Fort Calhoun Station (FCS) design bases requirements have been properly translated into operating, maintenance and testing procedures. This assurance is based on:

1. Completion in 1991 of an upgrade of safety related operating, maintenance and testing procedures to industry standards (Section B.1)

2. Development and use of a procedure writer's guide with requirements for incorporation of design bases information into operating, maintenance and testing procedures (Section B.1 and B.2)

3. Requirements for. review of new procedures and procedure changes for conformance with the design bases (Section B.2)

4. Requirements for verification and validation of new procedures, along with selected verification and validation of procedure changes dependent on the complexity of the changes (Section B.2)

5. Access to and familiarity with the design bases by procedure writers and reviewers (Section B.2) 1

6. Overall positive experience with upgraded FCS procedures (Section B.3)

7. Lack of programatic deficiencies in the procedures, their implementation or I results, as identified by self assessment (Section B.3)  !

8. Results of inspections by the NRC of procedures, their implementation and results ,

(Section B.3)

9. Additional verification of the procedures for conformance with the USAR and design bases currently being conducted by 0 PPD. (Section B.3)~

J LIC-97-011

' Attachment 2 Page El B.1 Procedures Improvements Proceduces_.UWrade_ Project The primary objective of the Procedures Upgrade Project (PUP) was to upgrade operating, maintenance and testing procedures at Fort Calhoun Station (FCS) to current industry standards. Those standards included the following: l o 1. ANSI N18.7, American National Standard Administrative Controls for Nuclear Power

\

. . Plants l 2. Regulatory Guide 1.33 Quality Assurance Program Requirements (Operation) l

3. NUREGICR-1369. Procedure Evaluation Checklist for Maintenance. Test and Calibration i Procedures Used in Nuclear Power Plants j

4. INPO 85-026 Writing Guideline for Maintenance. Test, and Calibration Procedures

5. INPD 85-038. Guidelines for the Conduct of Maintenance at Nuclear Power Station The PUP satisfied corm 11tments associated with the FCS Safety Enhancement Program (SEP) l Item 48. The scope of the PUP was to completely rewrite / upgrade the operating procedures, maintenance procedures and surveillance tests at Fort Calhoun St e m . The l term " upgrade." as used for'the PUP. means'to correct technical content deficiencies 1

and human performance deficiencies, such that verbatim compliance with the procedure  !

can be achieved by the least experienced but qualified individual. Approximately 3000 safety related and non-safety related procedures were upgraded by the PUP. The i following procedures were specifically included:

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1. . 0perating Instructions

2. Operating Procedures

3. Annunciator Response Procedures l

4. Chemistry Procedures i 5. Mechanical Maintenance and Pressure Equipment Procedures

! 6. Instrument and Control Procedures

7. Surveillance Tests )

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Radiation Protection Procedures. Emergency Plan Procedures. Security ProcedJres. I Emergency Operating Procedures and Abnormal Operating Procedures were upgraded under  !

I separate projects.  !

i 1 l The PUP instituted administrative controls currently in use such as the procedure i writer's guide, now documented as Standing Order G-73. Fort Calhoun Station Writer's 1 Guide, The purpose of the Writer's Guide is to provide the necessary guidance for l writing FCS Operating Manual procedures. The intent of the Writer's Guide is to ensure  !

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LIC-97-011 Attachment 2 Page 22 station procedures are clear. concise, technically correct and written for the users with little or no margin for interpretation.

The PUP involved several phases. During data gathering and drafting, design bases requirements e e incorporated into the draft procedures through implementation of the Writer's Guide. Information sources used during the drafting phase included, but was not limited to. the following:

1. Technical Specifications

2. USAR

3. 0A Plan

4. Vendor manuals

5. OPPD and Vendor Drawings

6. Industry Standards (ANSI. ASME. IEEE. ASTM. etc.)

7. System Training Manual

8. NRC/INP0 commitments

9. FCS Operating Manual

10. Experienced plant personnel

11. Configuration Changes (Huds. ECNs)

12. Design Basis Documents 0 hen available)

After each procedure was drafted, a verification and validation was performed. The verification of procedures refers to the processes used to ensure that:

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1. Procedures are technically correct.

2. Correspondence exists between the procedures and the control room / plant hardware, i and

3. Language and nomenclature used in the procedures are consistent with the terms familiar to users.

The validation of procedures refers to a method for ensuring that procedures are usable by plant personnel and able to accomplish the intended objectives. Essentially. it is a simulation (or rehearsal) of the procedure, progressing through the steps in as

' realistic a manner as practical. After validation was performed, each procedure was sent to the Plant Review Cocnittee (PRC) for review and recommendation of approval to the Manager-FCS. After approval and any required training, each procedure was issued for use.

Design Rases / Safety Related Onprating Procedure Cnmnliance Proieri In 1990. OPPD completed the Design Bases / Safety Related Operating Procedure Compliance Project, a separate but related procedure upgrade project which satisfied the coamitments associated with SEP Item 5. The objective of this project was to confirm that, concurrent with the Design Basis Reconstitution project. Safety Related Operating Procedures. Emergency Operating Procedures (EOPs). Abnormal Operating Procedures (AOPs).

i LIC-97-011 Attachment 2 Page 23 and the Technical Data Book (TDB) were in compliance with the FCS Design Bases. The scope of Safety Related 'SR) Operating Procedures included SR Operating Instructions and SR Operating Procedures. At the time of the project a total of 75 Ols/0Ps were classified as safety related.

Verification that the 75 Ols/0Ps complied with the FCS Design Bases was accomplished by determining if the procedure complied with the requirements of the Design Basis Documents (DBDs). Because of the timing of issuance of DBDs and upgraded procedures, some upgraded procedures were verified after they had been issued. The verification included determination that the purpose or functions described in procedures were consistent with the DBDs, that setpoints or system requirements were appropriately called out in the procedures, and that there were no contradictions between functions and setpoints in the procedures and the DBDs.

The Technical Data Book was verified against applicable DBDs using a two-volume cross-reference matrix of sorted data. One volume presented TDB Figures in numerical order with the applicable DBDs listed beside them. The other matrix volume presented the DBDs in numerical order with the applicable TDB figures listed beside them. No discrepancies between the TDB and the DBD were identified during the review process.

The verification of E0P and A0P compliance with the FCS Design Bases was completed through development of E0P and A0P Technical Basis Documents. Each E0P Technical Basis Document notes the differences between the E0P and report CEN-152. Combustion Engineering Emergency Procedure Guidelines. which was used as the basis for the FCS E0Ps. These differences resulted from the incorporation of plant-specific technical information and operating philosophy. The FCS-specific information was obtained from design documents. The AOP Technical Basis Documents include reference information for these procedures, including appropriate design documents. The verification and i validation process for E0Ps and AOPs ensured the technical adequacy of the procedures. l I

B.2 Current Programs and Processes i l

The FCS Quality Assurance Plan. Section 2.0. establishes requirements for implementing procedures which prescribe activities affecting safety. Standing Orders G-73 and G-30.

Procedures Changes and Generat1on establish responsibilities and general requirements for procedure change and generation. The Standing Orders are responsive to ANSI N18.7 -

1972. Administrative Controls for Nuclear Power Plants; the FCS Updated Safety Analysis Report (USAR). Section 12; and FCS Technical Specification 5.8. Procedures. The ,

Standing Orders also incorporate commitments related to SEP ltem 46, which implemented I a procedures control and administration program.

Standing Order G-73 is the Writer's Guide for procedures, and is intended to provide the necessary guidance for writing of FCS Operating Manual procedures. The Writer's Guide provides consistency in matters of format. human factors considerations and technical content and implements the verification and validation processes, l I

4 LIC-97-011 i Attachment 2 Page 24 The effectiveness of the procedures is directly dependent upon the writer's ability to t obtain and apply information that exists relative to operating or maintaining the affected system or equipment. Information sources include, but are not limited to, the following:

1. Technical Specifications

2. USAR

3. Vendor Manuals

4. Industry Standards

5. Design Basis Documents

6. NRC/INP0 Commitments

7. Configuration Changes in accordance with Standing Order G-30. Procedure Changes and Generation, the writer '

is responsible for ensuring the administrative and technical accuracy of the proposed procedure / procedure change and performance of a 10 CFR 50.59 Safety Evaluation as required by N00-OP-3. including review of any applicable Modification Request or Engineering Change Notice Safety Evaluation for consistency. The writer optionally forwards the procedure / procedure change to the Engineering Department for preparation of the Safety Evaluation if the change is specifically based on calculations or engineering analyses. This process provides reasonable assurance that procedure writers incorporate the design basis into operating, maintenance and testing procedures.  ;

The writer then compiles a verification / validation review package. Verification and validation are defined and implemented consistent with the program initiated by the Procedures Upgrade Program (previously discussed in section B.1).

FCS Technical Specification 5.8 delineates the requirements for the review and approval of procedures that affect nuclear safety; these requirements are implemented by the Administrative Controls Standing Orders such as G-30. A procedure or procedure change ,

is reviewed by a Qualified Reviewer (OR) who is knowledgeable in the functional area l affected but is not the individual writer. The OR renders a determination in writing j on whether or not cross-disciplinary review of a procedure. or change thereto, is  !

necessary. If necessary. such review is performed by appropriate personnel.  ;

FCS Technical Specification 5.8.2.4 requires " Qualified Reviewers meet or exceed the respective qualifications for either Supervisors Requiring an AEC License. Professional- '

Technical Personnel, or Technical Support Per'sonnel, as specified in ANSI N18.1 - 1971.

Personnel recomended to be ors shall be reviewed by the PRC and approved and designated as such by the PRC Chairman."

A procedure, or change thereto, is reviewed by the Department Head designated by Administrative Control Standing Orders as the responsible Department Head for that procedure. The review includes a determination of whether or not a 10 CFR 50.59 safety evaluation is required. If a 10 CFR 50.59 safety evaluation is not required, the

LIC-97-011 Page 25 procedure, or change thereto, is approved by the responsible Department Head or the Manager-FCS prior to implementation.

If the responsible Department Head determines that a procedure, or change thereto, requires a 10 CFR 50.59 safety evaluation, the Department Head renders a determination in writing of whether or not the procedure, or change thereto, involves an Unreviewed Safety Question (US0) and forwards the procedure, or change thereto, with the associated safety evaluation to the Plant Review Committee (PRC) for review in accordance with Technical Specification 5.5.1.6.a. If a US0 is involved. NRC approval is required prior to implementation of the procedure, or change. The PRC recommends approval or disapproval to the Manager - FCS.

Procedure changes requiring training are reviewed and approved, with issuance held until the requirements of the Procedure Training Request are met. Completion of training before or after procedure approval may be specified. After review and prior to implementation, the responsible Department Head within the functional area or the Manager-FCS must approve the procedure or change in accordance with Standing Order G-30.

B.3 Assessments, Surveillances and NRC Inspections A review of Quality Assurance Surveillances (1991 to present) related to station procedure control did not find any programmatic deficiencies in the procedures their implementation or results. Since January 1995, the FCS Quality Assurance Department has ensured the acceptability of FCS procedures by implementing the Biennial Procedure Review Oversight Program. Quality Assurance Manual Procedure #14 (0AM-14) is used to administer the program. The sampling plan used in OAM-14 provides a 95% confidence level that there are no more than 5% nonconforming procedures in the total procedure population being reviewed. To date, the FCS procedures have been in coroliance with the criteria specified in 0AM-14. This indicates the overall effectiveness of the procedure improvements and current process. Significant discrepancies or deficiencies found in individual FCS procedures are identified and resolved through the corrective action program.

In Licensee Event Report (LER) 91-08. OPPD voluntarily reported the discovery of inappropriate Technical Specification (TS) surveillance requirements for Reactor l Protective System bistables, and inadequate documented verification of the incorporation ,

of all applicable TS surveillance requirements into FCS surveillance test procedures. l The corrective actions included implementation of a TS Verification Action Plan.

Surveillance procedures were compared to the TS to ensure each required surveillance has a corresponding surveillance procedure which meets the intent of the TS. To ensure l that the information gathered during this phase of the project was captured, a surveillance test / technical specification matrix was developed and installed in an on- l line, main frame computer system that can be readily accessed. This was completed in November 1992.

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LIC-97-011 Attachment 2 Page 26 However, a surveillance test to monitor Boric Acid Storage Tank level was not properly reviewed during implementation of the Action Plan due to administrative oversight. This ,

procedure did not satisfy TS requirements, a condition reported by LER 93-01.

In March 1993, the FCS staff began a detailed effort to ensure that the content of l individual station surveillance tests met the intent of the requirements of the technical specifications. A review of all surveillance tests assigned to system engineers was performed by the respective system engineer. In addition, an independent review was performed of approximately 10% of all the surveillance tests.

The results of these reviews uncovered a number of minor discrepancies, none of which I were contrary to the requirements of the technical specifications, but rather I enhancements to the test program, individual surveillance tests, or enhancement to the technical specifications. An action plan that included each discrepancy was developed with specific assignment to station staff. by primarily to system engineering.

Currently, the action plan is scheduled to be complete in April 1997.

NRC Inspection Report 91-01. Electrical Distribution System Functional Inspection.

concluded that the FCS DBDs were generally comprehensive, user friendly and provided l a good information source for the engineering staff. Design information was noted as I being readily retrievable and accurate. The procedures developed through the FCS procedure rewrite program were noted as appearing greatly improved over earlier revisions.

NRC Inspection Report 94-01 documented a station blackout team inspection. The report states that the team found the calculations and analyses to be technically sound and in accordance with approved guidelines (i.e., the design basis). The team also determined that the station blackout coping procedures adequately addressed control room l and local operator actions necessary to establish and maintain decay heat removal. l minimize RCS inventory loss, minimize DC loads and restore the availability of vital components.

NRC Inspection Report 96-01, a Resident Inspector report. included a review of activities relative to plant practices, procedures, and parameters detailed in the USAR (a portion of the plant's design basis). The inspectors verified that the USAR wording was consistent with the observed plant practices procedures, and parameters. NRC Inspection Report 96-03, another Resident Inspector report. stated that in the l Maintenance area good procedural compliance was noted.

B.4 Conclusion l FCS procedures were rewritten / upgraded in the late 1980s through the early 1990s.

Technical content and accuracy of the procedures continues to be verified including verification of correspondence between the procedures and plant hardware. Responses 1 to Requests C and E provide OPPD's rationale for concluding that the plant hardware is

LIC-97-011 Attachment 2 Page 27 consistent with the design basis. The verification process and the correspondence between the procedures and plant hardware provide OPPD with reasonable assurance that the FCS design bases are reflected in the operating, maintenance and testing procedures used since the implementation of these programs, projects, and processes.

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l LIC-97-011 Attachment 2 Page 28 1

Omaha Public Power District Response to NRC Request C  !

(C) Rationale for concluding that system. structure, and component configuration and performance are consistent with the design bases Executive Summary Otsha Public Power District (0 PPD) has reasonable assurance that the configuration and i performance of the Fort Calhoun Station (FCS) systems, structures, and components are consistent with the design bases. This reasonable assurance is based on the following  !

items:

1. Reconstitution of the FCS design bases (Section C.1)

2. The process used to disposition missing or conflicting design documents (Section C.2)

3. Verification that the reconstituted design bases and design documents are ,

accurate and complete (Section C.3) l

4. Verification that the FCS physical configuration is consistent with the reconstituted design basis (Section C.3)

5. The system oriented assessments conducted by OPPD, including audits based on Safety System Functional Inspections (Section C.3) 1

6. Inspections by the NRC of the reconstitution of the design basis and the processes, programs, and procedures for changing and controlling the  ;

configuration of the station (Section C.4) '

7. Verification that the USAR is consistent with the design bases (Section C.5)

8. Effective control of design changes made after the design bases reconstitution which ensures the plant continues to conform to the design basis (Section C.6) l

9. Additional verification of the USAR and design bases currently being conducted l by OPPD using Nuclear Energy Institute guidance (Section C.7) i l

Background

Following a Safety System Outage Modification Inspection (SSOMI) by the NRC in December 1985. OPPD agreed to organize the original Fort Calhoun Station (FCS) design basis information in a more manageable form to ensure that the original design margins are not unintentionally abrogated. The design bases were documented and reconstituted by I

LIC-97-011 Page 29 the Design Basis Reconstitution Project. Various aspects of this project were tracked by Safety Enhancement Program Items 4. 5. 6. and 7.

C.1 Design Basis Reconstitution The FCS design bases were reconstituted by the Design Basis Reconstitution Project which began in March 1987 and completed the development of the Design Basis Documents (DBDs) in April 1990. OPPD contracted with Combustion Engineering and Stone & Webster Engineering for the production of the DC0s.

One of the challenges to the Design Basis Reconstitution Project was the recovery of the reference design data necessary to develop the design basis documents. Because of the vintage of the plant, current record keeping and. storage requirements were not in effect at the time of original plant design. Therefore, the original design and construction records were not always properly stored. maintained, indexed or even identified as being required. OPPD also contacted the architect / engineer, reactor vendor, and major equipment suppliers for FCS design records they possessed. These searches yielded thousands of records containing design basis information.

The Design Basis Reconstitution Project developed 47 Design Basis Documents (DBDs). Of these documents. 33 were specific to selected systems and 14 were topical documents covering a subject that spanned many of the DBDs. The goal of the project was to develop system design basis documents which would reflect the current condition of the plant. Selection of systeme needing DBDs was dependent on the safety significance of the system. frequency at mnWnion, complexity of the system and importance of the system for sustained ;afe plM operition. The following is a list of the System level Design Basis Document: Miff'i t'idt were developed for FCS.

SDBD-AC-CCW-100 Component Cooling Water SDBD-AC-RW-101 Raw Water SDBD-AC-SFP-102 Spent Fuel Pool Cooling SDBD-CA-IA-105 Instrument Air i SDBD-CH-108 Chemical & Volume Control SDBD-CN-110 Plant Communications SDBD-DG-112 Emergency Diesel Generators SDBD-DW-113 Demineralized Water l SDBD-FP-115 Fire Protection SDBD-FW-116 Feedwater SDBD-FW-AFW-117 Auxiliary Feedwater SDBD-HG-122 Nitrogen & Hydrogen Gas SDBD-MS-125 Main Steam SDBD-RC-128 Reactor Coolant SDBD-SI-130 Shutdown Cooling SDBD-SI-CS-131 Containment Spray SDBD-SI-HP-132 High Pressure Safety Injection l

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LIC-97-011 Attachment 2 Page 30 SOBD-SI-LP-133 Low Pressure Safety Injection SDBD-SL-PAS-134 Post Accident Sampling SDBD-SL-135 Primary and Secondary Sampling SDBD-VA-AUX-138 Aux.111ary Building HVAC 1

SDBD VA-CON-139 Containment HVAC SD8D-VA-CR-140 Control Room Habitability SDBD-WD-144 Waste Disposal SDSD-EE-200 120 VAC Vital Distribution SOBD-EE-201 AC Electrical Distribution SDBD-EE-202 DC Electrical Distribution SDBD-EE-203 Cathodic Protection SDBD COMP-300 ERF Computer & OSPDS Computer SDBD-CONT-501 Containment SDBD-AUX-502 Auxiliary Building SDBD-STRUC-503 Intake Building SDBD-STRUC-504 Security Building Plant level design basis documents were developed when a specific topic covered more than one system. The following is a list of the Plant Level Design Basis Documents (PLDBDs) for FCS.

PLDBD-ME-10 Pipe Stress & Supports PLDBD-ME-11 Internal Missiles & HELB PLOBD-EE-21 Electrical Equipment PLDBD-IC-30 Instrumentation Installation PLDBD-IC-32 Instrumentation & Controls Systems i PLDBD-CS-50 External Missiles PLOBD-CS-51 Seismic Criteria PLDBD-CS-52 Heavy Loads i PLOBD-CS-54 Geotechnical l PLDBD-CS-55 Masonry Walls PLDBD-NU-61 Regulations. Codes and Standards l PLOBD-NU-63 Personnel Protection i PLDBD-EV-70 Site Meteorology Not all systems at FCS were candidates for design basis documents based on the selection l criteria above. The following is a listing of those systems. I I

Auxiliary Steam System Chemical Feed System -

Circulating Water System l Lubricating Oil System l Potable Water System Sanitary & Storm Drains Toxic Gas Monitoring System l

LIC-97-011 Attachment 2 Page 31 Turbine Supervisory System Heater Vents & Drains Vacuum Priming System To ensure conformance with the design basis for modifications designed and installed prior to the Design' Basis Reconstitution, OPPD conducted a safety evaluation check of Modification Design Change packages (for safety related and nonsafety related modifications that had potential to impact safety systems) that were installed since initial commercial operation. Concerns identified through this review were dispositioned using the DBD open item process (described in Section C.2).

To ensure that activities such as mocifications, procedure changes and safety evaluations could be conducted while the DBR Project was in progress, controls were established to ensure that desig., basis margins were not abrogated. These controls provided:

1. An interim position and guidance for selection of design inputs and re-creation of design bases on an as needed basis. and

2. A review of modifications by third parties, including 0 PPD Quality Assurance and ,

o tside Architect / Engineer firms. I Prior to starting the development of the design basis documents, writer's guides were written, reviewed and approved by 0 PPD staff. In order to maintain consistency between the documents, writer's guides for both the SDBDs and PLDBDs were created. During the development of the DBDs, it was important to clearly define the scope of each design I basis document. Without a clear definition of the scope, system boundaries or topics may have overlapped. To ensure that an overlap did not occur, system boundary drawings were developed first. These boundary drawings, similar to Piping & Instrumentation Diagrams (P&lDs) containing major components, aided the DBD preparer in clearly understanding the scope of the design basis documents during development. ,

l The initial steo in the preparation of each system DBD was the development of the requirements sec'. ion. The primary references for the requirements sections were:

l

1. The US;R tand FSAR),

The Commitment Tracking System.

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2.

3. System modification packages, and l

4. NRC Safety Evaluation Reports. l The following is a listing of additional resources used to assist in the development of the DBDs:

1. Technical Specifications

2. Equipment Specifications 1

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1 NEREHER

- PPR

- Performance Improvement Plan

- Development Action Plan l'

- Use of Performance Measures

- Conducting value added analysis

- Setting Performance Measures 1

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L i.IC-97-011 Attachment 2 Page 32

3. Calculations

. 4. Test Reports

5. Correspondence

6. Licensing Correspondence

7. Contract Documents

8. Drawings

9. Vendor Information

10. Design Changes (Modifications Packages)

C.2 Hissing Documentation

! During the process of reconstituting the design bases for FCS. information or references I

were not always available to fully define or support the design requirements or to verify the adequacy of the design bases. In many instances, because of the vintage of the plant, that documentation did not exist. There also were instances where controlled documentation contained conflicting data. In order to identify evaluate and classify the safety significance of these discrepancies. OPPD initiated a DBD Open Item Program.

These open items ranged from minor discrepancies having no impact on the safety of the plant to issues potentially requiring notification of the NRC. When a discrepancy was found in the design basis, the initiator identified it with an open item number. OPPD procedure PED-QP-29. Evaluating. Reconstituting, and Closing Design Basis Docunent Open Items, was esed to properly evaluate and classify the open items and prioritize them based on their safety significance. Each open item was assigned a category ranging from Category 1 (potentially reportable) to Category 6 (least serious).

Category 1 Open items which identify specific design deficiencies that can adversely affect safety-related systems, structures, components or equipment and are potentially reportable to the NRC Category 2 Open items which (1) address safety-related equipment or components that perform an active function in satisfying plant system operations requirements as specified in the Technical Specifications and do not meet the reportability criteria of 10CFR50.72.10CFR50.73 or 10CFR21: or (2) identify conflicting requirements or missing important documentation, or (3) identify missing 10CFR50.59 evaluations.

Category 3 Open Items which (1) address safety-related equipment or components that do not perform an active function in satisfying plant operational requirements as specified in the Technical Specifications and do not meet the reportability criteria of 10CFR50.72.10CFR50.73 or 10CFR50.21: or (2) identify missing documentation that is needed to confirm compliance with structural, seismic, pipe stress. pipe support, or HELB design requirements: (3) identify missing supporting documentation for 10CFR50.59 evaluations, or (4) identify unresolved code issues or missing

LIC 97-011 Attachment 2 Page 33 1

. calculations / analysis required to prove conformance with regulatory 4 commitments.

Category 4 Open Items which are being resolved on a generic basis (i.e. NRC's unresolved issues, or other issues identified by industry groups).

Category 5 Open Items which reflect minor discrepancies. They have no safety significance and are not necessary to improve the overall understanding of the design basis.

Category 6 Open Items for which the appropriate paperwork has been initiated to closecut/ resolve the condition.

During the course of reconstituting the design bases for FCS. a total of 1715 open items were identified, evaluated classified and dispositioned. A list of the category subtotals follows.

Category 1: 10 Cater.ory 2: 19]

Category 3: 839 Category 4: 13 Category 5: 586 Category 6: 76 Because the Category 1 open items were potentially reportable, they were dispositioned imediately after identification. The Category 2 open items were closed approximately twelve months (April 1991) after the completion of the reconstitution project. The remaining open items were evaluated and determined to have minimal impact on the quality of the DBDs and have been scheduled for closure based on resource availability. As of February 4. 1997, a total of 200 open items remain in the following categories.

l Category 1: 0 Category 2: 0 Category 3: 93 Category 4: 0 Category 5: 101 Category 6: 6 These open items are not safety significant and are being actively pursued: completion of closure is currently scheduled for June 1997.

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LIC-97-011  !

Attachment 2 Page 34 C.3 Design Basis Verification During reconstitution of the design basis for FCS, OPPD began to verify that the design documents were in agreement with the configuration of the plant. Verification during the DBD project was accomplished by physical walkdowns, procedure reviews, licensing commitment reviews, functional checks, safety evaluation checks, and system oriented assessments.

Walkdowns-Over the years. OPPD has conducted various walkdowns such as the Piping and Instrumentation Diagram (P&ID) drawing walkdown. Drawing Update Project walkdown,

, Computerized History and Maintenance Program (CHAMPS) walkdown, NRC Bulletins 79-02 and 79-14 walkdowns, and the Environmental Qualification Program walkdown. These walkdowns served to document the as-built condition of the plant which provided assurance during the Design Basis Reconstitution Project that the physical plant was correctly reflected in the DBDs. The results of these walkdowns were used to verify that the contents of the DBDs reflect the as-built configuration and regulatory requirements.

Procedure Revipws-The Operating Procedures. Emergency Operating Procedures. Abnormal Operating Procedures and Technical Data Book were verified to be in compliance with the design bases as part of the Design Bases / Safety Related Operating Procedure Compliance Project described in the response to Request B.

Functional Checks-This verification confirmed that safety systems and other selected systems would perform their intended functions. A review of the FCS operating experience was done to verify that systems performed their normal operating functions as designed.

Systems required to operate in the post-accident mode were reviewed to verify that the ability to perform their intended functions was not compromised as a result of the modification process. This verification included reviews of post modification functional testing, reviews of the surveillance tests, and review of operating experience.

Safety Fvaluation Checks-Concurrent with the reconstitution effort. OPPD conducted a review of documentation packages for safety related modifications (and non-safety related modifications which had a potential to impact safety systems) which were installed since coamercial operation in 1973. There were 1304 modifications reviewed for potential impact on safety related systems. Of these.1001 were identified for review of Unreviewed Safety

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L LIC-97-011  !

l Attachment 2 i Page 35 Question (US0) Evaluations-. The 10 CFR 50.59 evaluations associated with these j modifications were reviewed to insure that m unreviewed safety questions were )

4 overlooked. Fifty-nine modifications were dispositioned through on-going 0 PPD programs. i i

One hundred and ninety-seven modifications lacked sufficient documentation to adequately l

. determine whether a US0 existed and were dependent on the closure of pre-identified open l items. Closure of these open items resulted in no additional identification of modifications with Unreviewed Safety Questions. Based on the project's review,16 1 modifications were determined to contain US0s. They were assigned to a Category 1 open  ;

item and immediately resolved in accordance with PED-0P-29. Evaluating, Reconstitution, and Closing Design Basis Document Open Items. .l J

System Oriented Assessments-

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I Since 1987, OPPD has completed 9 (including one follow-up) audits based on NRC guidance for Safety System Functional Inspections (SSFIs). During these audits, vertical slice evaluations were performed on the following 8 systems: Auxiliary Feedwater. Instrument Air, 120 VAC Vital Distribution, Component Cooling Water, 125 VDC, Raw Water, Stator j Cooling, and Diesel Generator. The purpose of these audits was to verify the operational readiness of the systems (including system modifications, testing, maintenance and operation) to perform required safety functions in conformance with the ,

current licensing bases. This verification also served to validate the effectiveness of the Design Basis Reconstitution Program. ,

I The. follow-up SSFI audit was performed in 1991 to ensure that corrective actions were performed for deficiencies identified during the Instrument Air, Component Cooling Water, and Auxiliary Feedwater SSFI audits. This follow-up audit also included verifications that deficiencies identified by OPPD prior to the NRC Electrical Distribution System Functional Inspection (EDSFI) wcce corrected or on track for correction. During this audit, more than seventy corrective action documents were verified to have corrective actions taken. Corrective actions for the identified deficiencies included revisions to procedures, drawings, and various maintenance activities; establishment of the interim set-point control program; equipment label changes; and modification and design evaluations.

'The subsequent SSFI audits each generated approximately three concerns. These audits show a continuing trend of improved conformance of the plant to its design bases. The deficiencies dealt with items such ~as inconsistencies between USAR values and calculations, inadequate post-maintenance testing of equipment, and administrative deficiencies with Design Basis Documents, plant procedures, and drawings.

Review of corrective action documents generated as a result of these inspections noted that corrective actions such as updates to procedures, System Training Manuals and

-drawings, changes to operation and testing procedures, and corrections to the Computerized History and Maintenance Program System (CHAMPS) equipment database have taken place to corre:t the identified deficiencies.

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l LIC-97-011 i Page 36 1 During the months of October and November 1994. OPPD personnel conducted the Service Water System Operations Performance Inspection (SWSOPI). This assessment determined that the Raw Water Component Cooling Water and related supporting systems are in good condition. The assessment also concluded that the systems are operated within the I design criteria for FCS.

C.4 NRC Inspections l l

In cddition to assessments conducted by 0 PPD which evaluated effectiveness of the FCS Design Basis Reconstitution Project and configuration control process, the NRC has conducted several major inspections of OPPD's programs. In April 1990, the NRC conducted an Electrical Distribution System Functional Inspection (EDSFI), as documented by Inspection Report 91-01, dated May 20, 1991. The inspection team reviewed selected modifications, design calculations and design basis documents. Additionally, the inspection team conducted interviews and reviewed training records, corrective action l records, maintenance records and surveillance test records. The review focused on the l

off-site power supply grid and associated on site. safety-related busses and loads. l l

During the EOSFI. the team noted a number of strengths und relatively few weaknesses.

The team noted that the design basis documents were generally comprehensive, user i friendly, and provided a good information source to the engineering staff. Design information was readily retrievable and accurate. l I

An NRC/NRR Special Audit of Control Processes for Commitments and Current Licensing Basis was documented in a report dated October 19. 1993. The staff conducted audits I at a cross section of reactor plants to assess the processes used by licensees for controlling commitments that affect the plants' current licensing basis. including maintaining and updating the final safety analysis report. At FCS. the audit team found that commitments which affected the USAR were captured by the USAR update process. The team's review of several plant modifications to identify changes to plant systems and ,

verify incorporation of the changes in the USAR showed that all affected text I descriptions and system drawings in the USAR were properly revised to reflect the associated modifications.

C.5 USAR Verification Project The NRC SS0MI report and Inspection Report 91-19 also identified problems with the adequacy of information in the USAR. Specifically, certain design information and safety analyses sent to the NRC were not incorporated. and a radiation monitor was still described years after its removal from service. As corrective action. OPPD implemented '

the USAR Verification Project in January 1993 to accomplish 3 objectives:

1. Verification of the accuracy of information contained in the USAR utilizing l information (design basis documentation) gathered during the Design Basis l Reconstitution Project l 1

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LIC-97-011 Attachment 2 Page 37

2. Verification that a proper review and inclusion of information required by 10 CFR 50.71(e) had been completed 1

3. Assurance that processes for USAR update and review of procedures were adequate for maintaining information current and accurate 2

The scope of the USAR Verification Project was to review and verify the information in Sections 1-11 and 14. The remaining sections (12 and 13) and the appendices (A through.

N) were outside the scope of the Project because they were either archived, relocated out of the USAR, or had recently been verified by other means. Section 13 was archived.

+ Appendix A is updated annually in accordance with the OA Plan. Appendices B through 0 and G through L were archived. Appendix E was incorporated into USAR Section 12.

Conduct of Operations. Appendix F was updated by Design Engineering. Appendix M was updated under the normal USAR update process. Appendix N was updated in 1992.

The USAR review and update procedure (Quality Procedure N00 0P-16, Updated Safety Analysis Report) was reviewed and appropriately upgraded to ensure the USAR is maintained accurate and current. A USAR Writer's Guide was developed to provide guidelines in matters regarding content, format, style and detail. The Writer's Guide is an appendix to N00-0P-16.

Verification was accomplished by utilizing the informaticn gathered during the Design Basis Reconstitution Project. NRC Safety Evaluation Reports (SERs) NRC Generic Letters (GLs) and NRC Inspection and Enforcement Bulletins (IEBs) issued from 1973 through 1994 i were reviewed. If inaccuracies or omissions were discovered during the verification 1 process, torrections and additions were initiated along with a justification for each.

Each verified section, with the applicable changes / justifications, was sent to System Engineering, Design Engineering and Operations for an independent review. The System Engineer, Design Engineer and the Operations Engineer not only reviewed the changes for accuracy and concurrence, but reviewed the entire section for accuracy. Any necessary 4 additional changes, along with justifications, were incorporated before review and i approval by the Plant Review Committee and subsequent incorporation into the next )

scheduled USAR update.

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The USAR Verification Project was completed in July 1995. Recently completed and ongoing self assessments (described in Section C.7 below) have identified insignificant discrepancies between the USAR and the plant configuration. These discrepancies have been entered into the Corrective Action Program for resolution. j NRC inspection Reports 96-01 through 96-04 and 96-08 verified that, in the areas inspected, the USAR wording was consistent with the observed plant practices.

procedures, and/or parameters.  !

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l LIC-97-011 Attachment 2 Page 38 l

. C.6 Engineering Design and Configuration Control i

I Engineering Design and Configuration Control processes have been upgraded to ensure that the FCS design bases are available and maintained during design change development and installation of new designs. The Configuration Control process delineates responsibilities, interfaces, review, approval, and training requirements. The control processes ensure that only approved configuration changes are made in the station and  !

l engineering documents, and databases are accurately updated in a timely manner. Clear l guidance is provided on the various types of authorized configuration changes, l Procedure PED-0P-2. Configuration Contro?. assists the Design Engineer in determining j the type of configuration change to use. These processes are discussed in more detail l in the response to Request A.

C.7 Ongoing Verification Activities In order to provide additional assurance of conformance with the design bases. OPPD is currently completing assessments of selected FCS systems using the Puclear Energy 1 Institute (NEI) Guidelines for Assessing Programs for Maintaining the Licensing Basis (NEl 96-05). Deficiencies, found are entered into the Condition Report system for resolution. The assessment of the Chemical and Volume Control System is complete. and l assessments of the Safety Injection and Instrument Air systems are ongoing. OPPD will complete assessments of all remaining FCS safety related and safety significant systems no later than February 1.1999.

C.8 Design Basis Document Maintenance In order to maintain the dynamics of the Design Basis Reconstitution Program. OPPD  !

institutionalized the Design Basis program by conducting training for the users and by l issuing procedures that control the changes to the documents. OPPD procedure PEO-GEI-3.  !

Preparation of Design Change Packages, requires that the design engineer use the DBDs as a source for the design input of the design change package. Any changes in the design basis due to the modification result in revision to the appropriate DBDs. The actual update of the design basis documents is controlled by PED-0P-13. Design Basis Docament Contro?. This procedure establishes the requirements and responsibilities for control and approval of revisions to the DBDs. When proposed changes are submitted by other than the design basis document Sponsor an additional technical review is required per procedure GEI-12. Design Engineering Review of Design Basis Document Revisions. This procedure insures that the proposed configuration change is technicaliy correct.

substantiated by design documentation and is evaluated for impact on other Fort Calhoun Station documents.

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LIC-97-011 Page 39 C.9 NUMARC 90 12 The OPPD Design Basis Reconstitution Program, although completed before the issuance ,

of NUMARC 90-12 contained all the attributes necessary for a successful program as outlined in NUMARC 90-12. This document was assembled by a team representing 12 utilities, 4 NSSS vendors, and 4 A/E firms. Input to the guideline was also received from INP0 based on trips to utilities that had mature programs in place. Some of the lessons learned and reflected in NUMARC 90-12 are the result of evaluation of the FCS program.

C.10 Conclusion The design bases reconstitution verification. maintenance, and assessment activities described above provide OPPD with reasonable assurance that the configuration and performance of the FCS systems. structures and components are consistent with the design bases. Discrepancies identified during these activities have been and will continue to be resolved through appropriate corrective action processes. ,

LIC-97-011 Attachment 2 l Page 40 Omaha Public Power District Response to NRC Request D l (0) Processes for identification of problems and implementation of corrective -

l actions. inc'luding actions to determine the extent of problems, action to prevent '

recurrence, and reporting to NRC.  ;

Executive Summary The Corrective Action Program (CAP) is used for the identification, reporting. I documenting, controlling. tracking and trending of conditions adverse to quality at Fort Calhoun Station (FCS). This applies to conditions that represent a failure to meet requirements or expectations, such as malfunctions, deficiencies, defects, deviations. ,

non-conformances, or abnormal occurrences. The Corrective Action Program provides ,

reasonable assurance that problems are identified, required reports are ma:ie to the NRC, and effective corrective actions are implemented, all on a schedule cocensurate with the nuclear safety significance. This assurance is based on the effective implementation of the processes making up the Corrective Action Program, summarized as follows:

1. The Condition Reporting (CR) system, proceduralized in Standing Order R-2 4

implements the Corrective Action Program and provides the primary tool for identifying, categorizing, tracking and documenting correction of conditions adverse to quality. (Section 0.2)

2. Standing Order R-2 requires personnel to identify conditions adverse to quality

, including nonconformances in design basis documents, as-built plant and  ;

procedures. (Section 0.2) '

1

3. Design concerns are formally documented in the CR system and evaluated by both {

the Corrective Action Group and the Condition Review Grwp. (Section D.2 and 0.5) j

4. Identified conditions adverse to quality are evaluated to determine their impact, using the Operability Determination Program which provides a means of formal documentation for operability evaluations. (Section 0.3)

5. ' The Condition Review Group evaluates identified conditions adverse to quality and assigns a priority for timely corrective action based on their significance to safety. (Section 0.5) i

6. The CR system is used to track identified conditions adverse to quality through l completion of corrective actions and is also used to trend CRs to identify' l potentially significant adverse conditions. (Sections D.5. D.7 and 0.8)

7. The Root Cause Analysis Program provides a formal method for root cause determination of significant conditions adverse to quality and the identification l

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-Attachment 2 Page 41 i

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and implementation of actions to prevent recurrence and -correct any generic - ,

implications. (Section 0.6) i 8;

i If it is pecessary to operate FCS with a degraded condition that does not conform .,

j with the design basis. the condition is formally analyzed and documented using

i. the Safety Analysis for Operability process, which. includes the performance of '

[ a 10 CFR 50.59 safety evaluation. (Section D.3) i

.9. The Reportability Determination Program provides a method to ensure that i identified conditions adverse to. quality are evaluated for reportability to the

NRC under requirements including 10 CFR 50.72, 10 CFR 50.73, 10 CFR 50.9. 10 CFR i

{ Part 21, and the Technical Specifications. (Section 0.4)

10. . Audits, surveillances, and assessments of the corrective action program and its I

, implementation have been conducted by the Quality Assurance Department, the q Nuclear Safety Review Group, and independent third parties. These audits, a surveillancesi and assessments have identified no significant deficiencies in the

procedures, including implementation or results. However, an ongoing problem with timeliness of responses and deficiency correction has been noted. 'and is  ;

being addressed through increased management attention. (Section 0.9) l

j. 11. Inspections by the NRC of the current FCS Corrective Action Program have  !

identified no . significant or . programmatic deficiencies in the procedures, j including implementation or results. (Section D.9)

.D.1 Corrective Action Program Prior to the Condition Report System i

i Prior to implementation of the current Condition Report (CR) system at FCS. several i systems comprised the Corrective Action Program. The Operational Non-conformance (0NC).

Incident Report (IR), and Corrective Action Report (CAR). systems were the Corrective i Action Programs in general use up to September 1995. In addition, the Design Basis j Reconstitution program had its own Open Item identification and tracking process. .

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{ OpecationaLNon-conformance (ONC) System i^

The ONC System was used to identify and document resolution of nonconforming conditions found in any " operable" system, structure, component or software identified during plant operation, testing, inspection or maintenance. A nonconforming condition is defined as a deficiency or noncompliance in characteristic, documentation or procedure which renders the quality of an item unacceptable or indeterminate. The ONC was a separate screening process that had been operated by Quality Assurance. The ONC system was principally used by Engineering to disposition material conditions adverse to quality  ;

in operating ~ equipment.

LIC-97-011 Attachment 2 Page 42 incident Renort (IR) SysterD The IR system was the primary mechanism used to report operating anomalies and deficiencies affecting the design basis, licensing basis, or configuration of FCS.

Typically, when an employee identified an item of concern, it was discussed with the employee's supervisor before being considered for further reporting. When initiated, an IR was reviewed by the Shift Technical Advisor (STA) and Shift Supervisor for reportability, and an operability determination was made for equipment governed by Technical Specification Limiting Conditions for Operations. The STA would also perform a safety assessment for any event that placed the plant in an abnormal situation or any plant parameter that affected or reflected an abnormal indication of a safety related system.

The IR was then reviewed at the weekly Incident Evaluation Team (IET) meeting. The IET l was established as a multidisciplinary subcommittee of the Plant Review Comittee (PRC). '

For each IR, the IET:

1. Ensured reportability requirements were met,

2. Determined significance, i

3. Assigned appropriate responsibility, depth of investigation, and documentation I requirements,

4. Reviewed for operational nonconformances, and

5. Identified relevant industry experience. 1 As another multidisciplinary management review, the PRC then reviewed each IR with the assignments, recommendations and comments made by the IET and modified them as applicable.

The IR assignee was responsible for initiating, completing, and documenting corrective actions to ensure issues related to the IR were resolved. One of the actions may have been to perform the ONC screening and provide a basis for correcting, removing, or accepting the condition. ONCs were incorporated as part of the IR process which allowed i ONCs to be approved, closed out and trended through the IR system.

The IR system was replaced with the CR system in September of 1995. The irs with open corrective actions were left on the IR system for completion. Upon completion of corrective actions, all irs are reviewed by an IR coordinator to ensure all issues are resolved and documented. If the IR is considered significant, the IET and PRC will review the IR for closure, again ensuring both a multidisciplinary and managerial perspective of the completed corrective actions.

Corrective Action Renort (rAR) System The other system in cormion use was the Corrective Action Report (CAR). The CAR system was used primarily by the Quality Assurance organization to record conditions adverse

LIC-97-Oll Page 43 to quality identified during their audit process; however, it was available for use by all personnel. The corrective actions were reviewed and accepted by OA management or for a significant deficiency, by the Safety Audit and Review Committee (SARC)

Chairperson. The deficiencies identi fied by the CAR system tended to be more programmatic in nature.

DesjgrLBans_ Reconstitution _0 pen _Ltem Tracting During the process of reconstituting the design bases for FCS information or references were not always available to fully define or support the design requirements or to verify the adequacy of the design bases. In order to identify, evaluate and classify the safety significance of these discrepancies. OPPD initiated -a DBD Open Item Program.

(This program is also discussed in the response to Request C.) These open items ranged from minor discrepancies having no impact on the safety of the plant to issues potentially requiring notification to the NRC. When a discrepancy was found in the design basis, the initiator identified it with an open item number. These open items were not entered into the corrective action process, but they were tracked on a separate open action item list. OPPD procedure PED-0P-29. Evaluating. Reconstituting, and Closing Design Basis Document Open Items, was used to properly evaluate and classify the open items and prioritize them based on their safety significance. Each open item was assigned a category ranging from Category 1 (potentially reportable) to Category 6 (least serious). If an open item was found to be significant to plant safety or reportable to the NRC, the deficiency was then converted into the applicable corrective action process for resolution and tracking of corrective actions.

D.2 Transition / Upgrade - Condition Report System The corrective action programs under the Incident Report (IR) Corrective Action Report (CAR), and Operational Non-conformance (ONC) systems were considered sound but inefficient. There was also some confusion among personnel as to which system should be used to report a problem. Therefore, these processes were benchmarked by interviews, procedure reviews and direct observations. The corrective action processes at the leading industry performers were also reviewed. This information was assessed and the decision was made to develop an improved integrated corrective action program.

The new corrective action system, entitled the Condition Report (CR) system, was ,

implemented in September 1995. Plant personnel were trained on the new system. The CR system is procedurally controlled by Standing Order R-2. Condition Reporting and Corrective Action. The CR system was developed to provide a consistent approach to .

effect actions and management oversight, to properly characterize and prioritize l problems, and to provide for permanent fixes to problems. Benefits in combining these l systems were consistency in threshold, significance, status, trending, and closure of l 1ssues important to safe efficient, and reliable operation.

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LIC-97-011 Page 44 The Corrective Action Group (CAG) was established to provide an integrated l administration of the Corrective Action Program at FCS. The CAG includes experienced  !

personnel familiar with plant configuration, operation, and administrative controls; some of these personnel have held Senior Reactor Operator licenses and have served as '

Shift Supervisors'. As discussed later in Section 0.5. the CAG presents preliminary and background information on each CR to the management Condition Review Group. The CAG also includes a supervisor responsible for coordinating and administrating the Root Cause Analysis process. The CAG personnel can provide expert assistance to anyone using the CR system.

Standing Order R-2 requires personnel to use the CR system to identify conditions adverse to quality including nonconformances in design basis documents, as-built discrepancies in plant systems, and procedure deficiencies. The CR system is electronically available to everyone, on a Local Area Network; however, a paper form or CAG assistance is also available for initiation of CRs. In the event an individual discovers a condition requiring imediate response by Operations (e.g., fire, flooding.

medical emergency, possible equipment failures, etc.) the individual immediately contacts the Control Room. For events that require immediate response by Security (e.g., tampering, assault, bomb threats, unauthorized individuals, etc.) the individual will imediately contact the Central Alarm Station (CAS). The CR will be generated after the nonconforming condition or event has been stabilized.

The CR process was designed to encourage timely problem identification at all levels of the organization. No management approval is required for any employee to initiate a CR on any concern or deficiency. A low threshold of reporting is maintained, with the philosophy being "when in doubt report it." The expectation that conditions be immediately reported is supported by senior management, as demonstrated by the Vice President's request for all supervisors to make the expectation clear to their personnel. This expectation was included in General Employee Training which is required for all personnel having unescorted access to the protected area. Experience to date shows that this expectation is being met.

0.3 Operability Determination Program Formal operability evaluations at FCS are controlled by Nuclear Operations Division procedure N00-QP-31. Operability and Reportability Determinations. The N00-QP-31  ;

process has been designed to implement guidance on operability determinations provided i in NRC Generic Letter 91-18 and Inspection Manual Part 9900. N00-0P-31 applies to all l non-routine events and conditions discovered during activities associated with the i operation of FCS. Such activities may include plant operations, maintenance, testing.

analysis, surveillance or any other process which reveals the potential inoperability of specified structures, systems, or components (SSCs).

After initiating a CR. the originator ensures notification of the Shift Technical .

Advisor and/or the Shift Supervisor. The Shift Supervisor is responsible for completion ll

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l LIC-97-011 ) l Page 45 of an initial operability and reportability determination. Initial operability determinations are made as soon as reasonably possible following discovery of a 1 deficiency. Determinations are generally made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, even though complete information may not be available. Initial determinations are revised appropriately as new or additional infonnation becomes available. Most initial determinations are performed without completion of N00-0P-31 documentation.

For some conditions, the SSC is clearly inoperable. If the affected SSC is considered operable based on available information but further evaluation is required, the Shift Supervisor contacts appropriate management personnel to initiate a documented N00-QP-31 operability evaluation.

Positive determinations of operability must be Justified and include, as applicable.

a technical discussion of why the concern identified does not prevent the item from fulfilling its intended safety function (s). This demonstrates that the item is not exceeding its design basis specified in reference documents, or the item maintains the ability to provide minimum parameter values (e.g., flow, pressure, voltage), and does not create any inter-system effects. The discussion addresses issues such as common mode failure, seismic event. station blackout, and other design basis events.

Following preparation and review of an N00-0P-31 operability evaluation, it is forwarded to an appropriate level of management for concurrence. Following this concurrence, the Shift Supervisor takes appropriate actions based on the results of the operability determination. The Plant Review Committee (PRC) also reviews these evaluations.

Procedure N00-QP-22, Preparation and Approval of a Safety Analysis for Operability (SAO), defines a process at FCS similar to the industry Justification for Continued Operation (JCO). An SA0 provides technical justification for continued plant operation when unexpected events or conditions arise that are contrary to the established design basis for FCS. In the event that a component is determined to be inoperable, and there is reasonable assurance of safety, operation may be acceptable during the corrective action phase of the problem. An SA0 package includes documentation of the issue, including an evaluation in accordance with 10 CFR 50.59 requirements. and documentation of necessary compensatory and corrective actions. Following appropriate management review, an SA0 is reviewed by the PRC and approved by the PRC Chairman.

D.4 Reportability Determination Program The Condition Reporting system is used to initiate and coordinate the reportability process for FCS. Several procedures govern portions of the process. Assigned personnel provide support as required for research, analysis, and review associated with reportability determinations.

The Shift Supervisor is responsible for completing and making prompt verbal reportability determinations. The Shift Technical Advisor (STA) performs a safety

LIC-97-011 Attachment 2

Page 46 assessment for any event or condition that places the plant in an abnormal situation or for any plant parameter that affects or reflects an abnormal indication of a safety related system. This safety assessment is intended to help ensure abnormal plant conditions are properly assessed against reportability criteria. The STA assists the Shift Supervisor'in making prompt verbal reportability determinations. The Shift Supervisor determines whether the condition requires verbal reporting to a governmental agency using guidance in Standing Order R-11. Notiffcation of Signf ficant Events.

Standing Order R-11 is used to perform a 10 CFR 50.72.10 CFR 20.2201.10 CFR 20.2202 or 10 CFR 26.73 notification to the NRC. For events reportable under-10 CFR 73.71.

appropriate Security management personnel review the event in accordance with Security Administrative Procedure SAP-35. Reporting of Safeguards Events, and Standing UNer R-

12. Reportfng of Physfcal Securf ty Events, to determine reporuabi1itv If further evaluation is needed to determine reportability appropriate management personnel are asked for assistance. In general, procedure PED-0P-19. Evaluation of Potentfally Reportable Conditions. is used by Engineering in determining if a reportable condition exists. This procedure guides the user through an evaluation of the condition against reporting criteria of 10 CFR 50.72.10 CFR 50.73.10 CFR 21. and 10 CFR 50.9(b).

Completion of the PED-QP-19 documentation results in a recommendation from Engineering that a condition meets or does not meet any of the reporting criteria. This recommendation is then reviewed by the PRC which subsequently makes a reportability recommendation to the Manager - FCS. The Manager - FCS makes the final determination of reportability and ensures that required reports are made.

Written reports are prepared in accordance with N00-0P-35. Licensee Event and Special Reports. and Standing Order R-3, Reportable Occurrences. These procedures ensure that OPPD complies with the regulatory guidance on written reports promulgated in 10 CFR 20.

10 CFR 50.73.10 CFR 73.71.10 CFR 50.36 and NUREG-1022. The reporting of issues subject to 10 CFR 21 is addressed in N00-0P-12. Reporting of Defects and Noncortpliance to the Nuclear Regulatory Cortmissfon (10 CFR 21).

D.5 Condition Reports - Hanagement Review After the initial operability and reportability reviews are complete. the CR is prepared for management review by the Corrective Action Group (CAG). The CAG performs a preliminary investigation on the condition and determines whether the condition is repetitive. indicative of an adverse trend, or a generic concern. The CAG makes a preliminary determination of the significance level and assigns trend codes. This preparation enables management to complete a more informed, accurate, and timely review.

Management review of new reports in the CR process is performed during normal work day meetings by the Condition Review Group (CRG). The CRG is chaired by the Manager - FCS (or alternate) and includes department managers. Either the Vice President or one of the Division Managers (or designated alternate) normally attends the CRG meeting. The various department managers provide a multi disciplinary review team. The key purpose

LIC-97-011 Page 47 of the mdeting 1s to focus on safety, significance, operability, generic implications, reportability, and assignment of an owner.

e The CRG assigns to each Condition Report a significance level dependent on: the impact to quality, reliability or safety: the recurrence of the condition; and actions already taken to correct the condition.

If subsequent evaluation of the condition reveals new information that may affect reportability. operability, or safety significance, the Manager - FCS or Shift Supervisor is notified and the CR is again presented to management at the CRG. The CRG may reconsider the CR significance level recommended by the CAG or as previously set by the CRG.

In accordance with Standing Order R-2 assigned CR owners determine the cause of the condition, develop the corrective actions, and set the schedule for completion. Only l selected CRs are subject to a Root Cause Analysis. Any condition determined by the CRG to be significant must meet the requirements of 10 CFR 50 Appendix B. Criterion XVI (understanding the cause of the event and establishing corrective act;ons to prevent recurrence) through assignment of a Root Cause Analysis.

D.6 Root Cause Analysis Prior to 1989, a formal and procedurally driven process for the determination of causes of events did not exist at FCS. The completion of Safety Enhancement Program Items 2

10. and 15 resulted in Quality Procedures N00-0P-19. Root Cause Analysis, and N00-0P-

20. Human Performance Enhancement System.

The goal of the Root Ceuse Analysis (RCA) program implemented through NOP-0P-19 is to provide a methodology for evaluation and analysis of selected incidents or conditions to determine root cause(s) and generic implications. Currently. OPPD requires completion of RCA: for events or conditions considered significant.

Although the implementation of the RCA program under N00-0P-19 includes the application of Human Performance Enhancement System (HPES) process techniques, the HPES program can also be applied as a stand alone process to non-significant conditions or events through use of N00-0P-20.

The scope of RCA program includes guidance on analytical methods and techniques, corrective action development. report preparation and approval, trending. and qualification requirements for individuals performing RCA Evaluations. Analytical methods include the use of the various MORT analysis techniques, as well as Kepner-Tregoe problem solving techniques and HPES methodologies. The RCA process relies on the use of qualified individuals throughout the nuclear organization to perform cause and generic implication determinations.

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LIC-97-011 Attachment 2 Page 48 A Peer Review Team assesses each RCA report for adequacy in determining root, contributing causes and generic implications. Based on the identified cause(s), the Condition Report Daner then determines corrective actions and presents the RCA to the PRC for final approval. The PRC provides a final assessment of the adequacy of the '

causal determinations, generic implications and corrective actions for the condition under investigation. The causal determinations and corrective actions are then entered into the CR as a means of providing trending data and corrective action tracking.

The CRG may, at its discretion require a root cause analysis for nonsignificant conditions when a formal method of cause determination is desirable. Otherwise, the (apparent) cause of the nonsignificant events and appropriate corrective actions are ,

determined by the CR owner. I D.7 Corrective Actions j 1

1 The owner of a CR assigns action items to appropriate personnel as a tracking vehicle  !

for the corrective actions. Timeliness for implementation of corrective actions is maintained by raising the level of management approval for due date extensions. Manager

- FCS or Division Manager approval may be required for an extension, depending on the length of time for implementation and the priority level of the action item.

When the owner has confirmed completion of the corrective actions. the CAG will perform an administrative review for CR closure, ensuring that all required corrective action documentation is in the CR record file.

1 D.8 Trend and Search Capabilities l An added benefit of the CR system is the enhanced search and trending capabilities.

The low threshold of reporting allows for analysis and trending that can identify ,

precursors before serious problems occur. This provides valid information that allows for proper prioritization of resources. When the process measurement reveals ineffective corrective actions through repetitive problems, the recurrent issue can be given a higher priority or resources. I Available reports address topics such as whether the condition was self-identified, causal factors, work groups involved with the condition, and event symptoms involved.

The process measurement not only identifies effectiveness of corrective actions, but also can provide data for quarterly trend reports and self-assessment. Personnel now have the ability to electronically view trend information relevant to their functional area without the need for extensive manual tabulation.

The improved search capabilities enable the CAG to provide the CRG with an accurate presentation of condition trends and history, identifying recurring events that need further corrective actions directed at the cause. Trending reports are periodically presented to both the PRC and the Safety Audit and Review Cormittee.

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Page 49 W

The ability to provide timely data on reported conditions has proven valuable, allowing l 4

early compensation and correction of activities in the field. For example, significant

. and repetitive CRs during the 1996 refueling outage were compared with Incident Reports from the 1995 refueling outage. A prominent shift was seen from industrial safety and maintenance events to equipment operational events. Significant events in the 1996

refueling outage occurred during active manipulation of installed equipment instead of

during times a system was out of service for testing or maintenance. A possible explanation for this trend was scheduling problems, including a lack of system windows. ,

This shift in events prompted immediate concern by the Manager - FCS. At his direction.  ;

) a site-wide briefing was conducted to communicate this trend to the workforce. Although '

these problems could not be imediately corrected, the site wide briefing provided a l 3 heightened awareness and a sensitivity to safety in all aspects of the outage.

From September 1995, when the Condition Reporting System was initiated, through December 1996, approximately 2100 Condition Reports (CR) were initiated. Out of this total 126 have been classified as configuration discrepancies. These have been further classified as plant document, design document, and physical discrepancies. A plant document I discrepancy is a plant document, typically a procedure, not consistent with the design 1

or licensing basis. A design document discrepancy is a design document not consistent l with the design or licensing basis. A physical discrepancy is a plant system, structure

or component not conforming to the design basis. The following table shows a breakdown of the CRs by type and cause.

Type of Discrepancy 1 Plant Documents Physical Design Documents Discrepancy Cause (definitions o. next page) TOTAL Cause Unknown 4 9 2 15 i

Error Updating Plant Documents for Configuration Change 15 15

)

Unauthorized Configuration Crange 11 11 Discrepancy Due To Original Equipment 17 17 Old Modification Caused Discrepancy 3 13 5 21 )

Recent MR or ECN caused Discrepancy 3 3 l

Maintenance Activity Caused Discrepancy 12 12 Error in Original Documentation 11 11 Error in Updating Design Documentation for 17 17 Configuration Change Original Building Documents 4 4 TOTAL 22 65 39 126 I

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The definitions of the causes are:

l 1. Cause Unknown The investigation of the discrepancy wts unable to identify a cause.

2. Error in Updating Plant Documents for Configuration Change A plant document vias not updated for a configuration change or an error was made when the docucant was revised. Tne configuration change was made using the OPPD's revised design engineering and configuration control processes implemented following the SSOHI inspection.

. 3. Unauthorized Configuration Change - A configuration change was cade without the proper documentation or an unauthorized change in the construction work order was race.

4. Discrepancy Due To Original Equipment - A deviation between the original design basis and the as-built plant was identified and no configuration changes could be identified which would have modified the as built plant.

5. Old Hodification Caused Discrepancy - A modification wts incorrectly installed using the design engineering and configuration control procisses in place prior to the SSOHI inspection.

6. Recent HR or ECN Caused Discrepancy - A modification ce configuration chan;e was incorrectly installed using the design engineering and configuration control processes icplemented following the SSOHI inspection.

7. Maintenance Activity Caused Discrepancy - A cordiguration change was made usir.g the Haintenance Work Order process.

8. Error in Original Documentation - A deviation bitween the original design basis documents and the as built plant was identified and no document changes could 'oe identified which would have modified the design basis documents

9. Error in Updating Design Documentation for Canfiguration Change - A design document was not updated for a configuration change or an error wts trade when the document was revised. The configuration change was made using the OPPD's revised design engineering and configuration control processes implemented following the SSOH! inspection.

10. Original Building Documents - Errors or omissions are identified in documentation for non-safety structures (e.g., warehouse) constructed using corrercial grade standards. 1 D.9 Audits and Inspections of the CAP Quality Assurance Audit #45 of the CAP includes verification that conditions adverse to quality are effectively identified, documented and corrected in a timely manner.

The 1994 and 1995 audits determined that the CAP was effectively implemented but that there was an excessive number of overdue responses. The auditors also noted an attitude among personnel which encouraged rapid rather than complete corrective actions.

Operational non-conformance packages reviewed provided adequate technical resolution to the identified problems.

The 1996 audit determined that the CR system was exhibiting consistency and uniformity.

Although not all administrative problems were completely resolved, the system was verified to be effective in completing corrective actions. Strengths noted were effectiveness of the CRG in defining a problem, assigning ownership and determining significance level; excellent CAG presentations; and management's lower threshold of problem identification.

LIC-97-011 Attachment 2 Page 51 The 1996 FCS Joint Utility Management Audit report noted an ongoing problem with timeliness of responses and deficiency correction. The timeliriess of corrective action responses was found to need improvement. To improve responsiveness and compliance with due dates, a daily status report is available to the Vice-President and to managers responsible for condition reports. Improved compliance with commitment dates for response is being enforced by all levels of management.

The 1994 NRC Systematic Assessment of Licensee Performance (SALP) report for FCS (Inspection Report 94-99) noted a weakness in knowledge about the CAP on the part of station personnel. Personnel were confused as to which system was appropriate for use and which problems should be reported. These were key reasons for the upgrade of the CAP to "le CR system. In 1996. the NRC SALP Report 96-99 noted that personnel knowledge of the LAP had improved to a good level. The report also notes that plant personnel

, were using the CAP to document problems including minor configuration discrepancies.

Training programs had been updated to include methods for identifying problems.

Engineering was noted as being effective in identifying, resolving and preventing problems in the 1995 NRC Integrated Performance Assessment Process (IPAP) report for FCS (Inspection Report 95-18). The IPAP report also noted FCS was effective in identifying and resolving performance problems and cited strong problem ownership at FCS 6s a key to resolution of issues and correction of deficiencies.

D.10 Conclusion

. The current Corrective Action Program provides an effective process for problem identification, operability and reportability determination, implementation of corrective actions, and trending. Based on the increased number of CRs and the independent assessments the CAP is meeting management expectations of a low threshold for reporting of problems. However, cause determination and corrective action determination and completion for some CRs are not being completed in accordance with the due dates. OPPD has initiated corrective action to resolve this issue, including reinforcement of mar.dgement expectations and establishment of additional tracking to ensure completion of actions in accordance with the due dates.

LIC-97-011 Attachment 2 Page 52 Omaha Public Power District Response to NRC Request E (E) The overall effectiveness of your current processes and programs in concluding that the configuration of your plant is consistent with the design bases.

Executive Summary The Omaha Public Power District (0 PPD) processes, procedures and programs provide reasonable assurance that the Fort Calhoun Station (FCS) physical configuration is consistent with the design bases. This assurance is based on:

1. The reconstitution and verification of the FCS design bases completed in 1990 (Section E.1)

2. The confirmation that the plant physical configuration conformed to the design bases following completion of the reconstitution of the design bases (Section E.2)

3. The verification of the Updated Safety Analysis Report (USAR) to be consistent with the design bases (Section E.3)

4. The completion of the Procedures Upgrade Project and the associated Design Bases / Safety Related Operating Procedure Compliance Project (Section E.4)

5. The use of the Operating Experience Review Process and the Commitment Action Tracking System to ensure that OPPD complies with new NRC requirements (Section E.5)

6. The timely identification and resolution of configuration discrepancies using the Corrective Action Program (Section E.13)

7. The participation of the system engineers in the design change and configuration control process, ensuring effective implementation of configuration control procedures and compliance with the design basis (Section A.2)

8. The ready availability of the USAR, Technical Specifications. Design Basis Documents and other licensing documents (Section E.6)

9. The upgrade of the processes, programs and procedures for changing and controlling the configuration of the station as part of the Safety Enhancement Program in i conjunction with the Design Basis Reconstitution (Section E.9)

10. The completion of vertical slice SSFI and SS0MI type audits since the deployment of the upgraded procedures and Design Basis Reconstitution which have confirmed these processes were and are being effectively implemented (Sections E.7 and E.10) l

LIC-97-011 Page 53

11. The results of inspections by the NRC of the reconstitution of the design basis and the processes, programs and procedures for changing and controlling the configuration of the station (Sections E.8 and E.11)

12. The additional verification of the USAR and design bases currently being conaucted by OPPD which have identified areas for improvement, but have confirmed the overall effectiveness of processes, programs and procedures for maintaining conformance with reconstituted design bases (Section E,12)

E.1 Design Basis Reconstitution The FCS design bases were reconstituted and conformance of the as-built plant with the design bases was verified following the NRC Safety System Outage Modification Inspections in 1985. The Design Basis Reconstitution Project completed development of the Design Basis Documents in April 1990 as described in the response to Request C.

This reconstitution included an evaluation of modifications installed on safety related systems since the initial operation of the station. The Open Item Program, described in the response to Request C. was used to identify, evaluate, classify the safety

. significance, and resolve design bases discrepancies. These discrepancies included missing design documents. inadequate design documents, and documents containing conflicting information. These open items ranged from minor discrepancies. .nich had no impact on the safety of the plant, to issues which potentially required %ification of the NRC.

To ensure the design bases were maintained during the design basis reconstitution. OPPD adopted the following interim policy for assessment of configuration chdnges to FCS:

The Fort Calhoun USAR in conjunction with any pending changes to Fort Calhoun design drawings and Safety Evaluation Reports used for amendments to the technical specification shall be used for providing design bases l for act1vities such as modifications, procedure changes, and safety evaluations.

l In conjunction with the early phases of the design bases reconstitution. the OPPD engineering design, and configuration control processes were upgraded to ensure that subsequent changes were effectively controlled and the changes were verified to conform with the design bases.

OPPD has completed internal Safety System Functional Inspections (SSFIs) on 8 systems.

The intent has been to validate the effectiveness of the Design Basis Reconstitution Program by assessing the ability of the systems to meet their design bases. These SSFIs have confirmed the operational readiness of the systems and the effectiveness of l corrective actions implemented as part of the SEP. l l

4 LIC-97-011 Attachment 2 Page 54 i The Design Basis Reconstitution Project and closure of 080 Open Items successfully recovered and reconstituted the FCS design bases. This Project consolidated the design basis into the Design Basis Documents (DB0s). Independent confirmation that the design basis was successfully reconstituted is discussed in Section E.7.

$ Design Bases Verification E.2 Verification that the FCS physical configuration conformed to the reconstituted design 4

bases was accomplished by physical verification. functional verification and a safety evaluation check.

1. The design requirements were compared against the current plant configuration as shown on controlled documents (i.e., drawings specifications etc.).

2. Walkdowns were conducted to verify that the controlled documents conformed with

the physical plant.

3. Modifications implemented by 0 PPD after issuance of the FCS operating license were reviewed to include any new design requirements, and the associated 10 CFR 50.59 safety evaluations were reviewed to verify their adequacy. Concerns identified during this review were tracked and resolved by the DBD Open Item program.

4. The functional requirements identified in the DBDs were compared against existing i

OPPD procedures to insure surveillance and in-service testing was being adequately performed to verify system and component operability over time.

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4 Additional actions included physical walkdowns of plant components to verify the maintenance data base and an upgrade of the vendor manuals to incorporate updated vendor information. Procedures controlling both the maintenance data base and the vendor manuals were also upgraded to ensure both the data base and vendor manuals received

{

timely and appropriate updating.  !

Details of the Design Bases Verification are included in the response to Request C.

These verification actions provide adequate confidence that the physical plant conforms

'with the completed Design Basis Documents. The independent verification of this project, discussed in Section E.6, provided assurance that this project achieved its objective.

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'E.3 USAR Verification The USAR Verification Project described in the response to Request C. verified that l the information contained in the USAR conformed to the information gathered during the- ,

i- Design Basis Reconstitution Project into the DBDs. The USAR Verification Project also verified that a proper review and inclusion of information required by 10 CFR 50.71(e) had been completed. The Project also revised the USAR update and review procedures as ig necessary to maintain the USAR information current and consistent with the design bases.

E.41 ProceduresUpgradeProjects

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FCS procedures were rewritten / upgraded in the late 1980s through the early 1990s by i completion of the Procedures Upgrade Project- and the associated Design Bases / Safety Related Operating Procedure Compliance Project. Details of these projects are discussed '

p in the response to Request B. Based on the successful completion' of these projects, r J OPPD has reasonable assurance that the FCS design bases are reflected in the current operating, maintenance and testing procedures.

l E.5 Operating Experience Review (0ER) and Commitment Action Tracking System

The FCS OER Program and Commitment Action Tracking System are used to evaluate and 3

address NRC requirements and generic information distributed in the form of Generic letters. Bulletins. Information Notices. Administrative Letters, changes to regulations, y and applicable 10 CFR Part 21 Reports. The OER program was established to provide evaluations of industry operating experience, approve corrective actions and assure implementation of corrective actions. Each item is screened to determine potential applicability, significance and priority. The recommendations are tracked to completion '

in the Commitment Action Tracking System. Deficiencies considered to be adverse to quality are addressed through the Condition Report corrective action process.

An example of this process is t'le OPPD response to the requirements of Generic Letter 1

.88-20. Individual Plant Examination for Severe Accident Vulnerabilities. As a result j of these requirements. OPPD has completed various walkdowns, upgrades to physical structures and revisions to design basis documentation over the last several years.

The response to OER item may be classified as an Ongoing Comitment. Nuclear Operations Division Quality Procedure N00-0P-34. Ongoing Comf tment Program, provides administrative controls for processing and maintaining such commitments. An ongoing comitment is a docketed comitment to the NRC detailing a procedure or administrative corrective action. or a statement of intended compliance with a specific industry standard which is performed periodically or continuously and does not include an external reporting requirement References to ongoing comitments are identified in OPPD procedures to alert the users such that the licensing bases are maintained. I

LIC-97-011 l Attachment 2 l

Page 56 The implementation of the OER Program and Comitment Action Tracking System provide  !

adequate confioence that NRC requirements are evaluated and incorporated into the FCS design and licensing bases.

E.6 Availability 'of Design Bases Documentation i To assist personnel involved in design changes and safety evaluations. OPPD provides easy access to the USAR. Technical Specifications. Design Basis Documents and other licensing basis documents.

1. Controlled hard copies of the USAR. Technical Specifications, and Design tasis Documents (D80) are maintained in the plant and the administration building.

2. Full text search capability is available on the Fort Calhoun LAN for electrJnic copies of the USAR. Technical Specifications. NRC Safety Evaluations of Techaical Specification Amendments. OER documents. Design Basis Documents and proce'Jures.

Although these are not " controlled" copies, they are restricted such that they may only be updated by authorized personnel.

These actions provide reasonable assurance that personnel have ready access to the DBDs. USAR. Technical Specifications and other design / licensing bases documents.

E.7 Assessments of Design Bases Since 1987. OPPD has performed 9 (including one follow-up) audits based on NRC guidance for Safety System Functional Inspections (SSFIs). During these audits, vertical slice evaluations were performed on 8 systems: Auxiliary Feedwater. Instrument Air.120 VAC Vital Distribution. Component Cooling Water,125 VDC. Raw Water. Stator Cooling, and Diesel Generator. The purpose of these audits was to verify the operational readiness of the systems, including system modifications, testing, maintenance and operation, to perform required safety functions in conformance with the current licensing bases.

These audits also served to validate the effectiveness of the Design Basis Reconstitution Program.

The follow-up SSFI audit was performed to ensure that corrective actions were performed for deficiencies identified during the Instrument Air. Component Cooling Water, and Auxiliary Feedwater SSFI audits. This follow-up audit also included verifications that deficiencies identified by OPPD prior to the NRC Electrical Distribution System Functional Inspection (EDSFI) were corrected or on track for correction.

The subsequent SSFIs each generated approximately three concerns that dealt with items such as inconsistencies between USAR values and calculations, inadequate post-maintenance testing of equipment, and administrative deficiencies with Design Basis Documents, plant procedures, and drawings.

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These audits show a continuing trend of improved conformance of the plant to its design bases. Review of corrective action documents generated as a result of these inspections noted that corrective actions such as updates to procedures, System Training Manuals and drawings, changes to operation and testing procedures, and corrections to the Computerized History and Maintenance Program System (CHAMPS) equipment database have taken place to correct the identified deficiencies.

During the nonths of October and November 1994, OPPD personnel conducted the Service Water System Operations Performance Inspection (SWSOPI). This assessment determined that the Raw Water, Component Cooling Water and related supporting systems are in good condition. The assessment also concluded that the systems are operated within the design criteria for FCS.

The assessment activities described above provide OPPD with reasonable assurance that the configuration and performance of the FCS systems, structures and components are consistent with the design bases.

E.8 NRC Inspections of Design Bases In addition to assessments conducted by OPPO which evaluated effectiveness of the FCS Design Basis Reconstitution Project and configuration control process, the NRC has conducted several major inspections of OPP 0's programs. In April 1990, the NRC conducted an Electrical Distribution System Functional Inspection (EDSFI), as documented by inspection Report 91-01, dated May 20, 1991. The inspection team reviewed selected modifications, design calculations, and design basis documents. Additionally, the inspection team conducted interviews and reviewed training records, corrective action records, maintenance records and surveillance test records. The review focused on the off-site power supply grid and associated on-site, safety-related busses and loads.

During the EDSFI, the team noted a number of strengths and relatively few weaknesses.

The team noted that the design basis documents were generally comprehensive, user friendly, and provided a good information source to the engineering staff. Design information was readily retrievable and accurate.

An NRC/NRR Special Audit of Control Processes for Commitments and Current Licensing Basis was documented in a report dated October 19, 1993. The staff conducted audits at a cross section of reactor plants to assess the processes used by licensees for controlling commitments that affect the plants' current licensing basis, including maintaining and updating the final safety analysis report. At FCS. the audit team found that commitments which affected the USAR were captured by the USAR update process. The team's review of several plant modifications to identify changes to plant systems and verify incorporation of the changes in the USAR showed that all affected text descriptions and system drawings in the USAR were properly revised to reflect the associated modifications.

l LIC-97-011 Attachment 2 Page 58 An inspection of OPPD's Engineering and Technical Support (E&TS) capability was completed August 25. 1995, as documented by Inspection Report 95-11. dated October 19.

1995. This inspection was conducted to evaluate the ability of the Production Engineering Division to provide effective engineering and technical support to the plant. In addition to inspecting the activities of the engineering group, the inspectors also reviewed the processing of 10 CFR 50.59 safety evaluations. The team concluded that Design Engineering was accomplishing its goals and managing workloads. The team also noted that OPPD's response to an industry event involving pipe rupture because of a less than adequate erosion / corrosion program was exemplary. The team found the effectiveness of the Nuclear Safety Review Group to also be a strength. The inspectors identified that a plant modification was implemented using a process that specifically does not allow such changes to the plant design basis. The team considered this a failure to perform a 10 CFR 50.59 safety evaluation. Since this finding. OPPD has implemented extensive enhancements to the 10 CFR 50.59 process.

The inspection activities described above provide OPPD with reasonable assurance that the configuration and performance of the FCS systems, structures and components are consistent with the design bases.

E.9 Engineering Design and Configuration Control Engineering design control is conducted in accordance with 10 CFR 50 Appendix B.

Criterion III: ANSI N 45.2.11; 10 CFR 50.59; and 10 CFR 50.71 as described in detail in the responses to Requests A and C. Key provisions are listed below.

1. Design work is reviewed for conformance with the design bases or changes are made ,

to the bases. '

2. The Design Change Package (DCP) and supporting material (calculations, analyses. 4 specifications, drawings, etc.) are developed using approved procedures and are l maintained as controlled documents. t i

3. Design verification is conducted in accordance with established procedures.  ;

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4. System interaction analysis is used to assess the effects of a design change system requirements.

5. Plant and Engineering management approval processes are defined.

6. The impact of a design change on plant procedures and training is assessed.

7. Changes to the USAR. Technical Specifications, and procedures and mooifications are reviewed.

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8. Applicable procedure and design changes are screened to determine if changes to ,

the USAR or Technical. Specifications are required. i

9. The review and approval process for 10 CFR 50.59 evaluations is defined,

10. A process is in place for determining the information to incorporate into the l USAR. '

Thus, the FCS configuration control processes provide reasonable assurance that the plant's approved configuration is maintained.

E.10 Assessments of Engineering Design and Configuration Control Processes Internal Quality Assurance audits have evaluated the engineering design and configuration control processes during both normal and refueling operations. These were implemented as Safety System Functional Inspections and a Safety System Outage Modification Inspection.

Additionally. Quality Assurance personnel perform periodic audits of both the Engineering Configuration Management program and the Station Engineering program. To support the established audit program. Quality Assurance has established a surveillance program that periodically evaluates the engineering activities with potential to affect the Licensing Bases. 'These surveillances encompass activities associated with Station Engineering. Modifications Performed During Outages. Reload Analyses. Design Basis / Drawing Control, the 10 CFR 50.59 safety evaluation program, and fire protection.

Formal review of modification close-out is performed during the Quality Assurance audit of engineering configuration management. This review ensures that design bases documents that are required to be updated as a result of station modification have been updated and that modifications that are performed are done such that the licensing bases  ;

are maintained.

l Review of audit reports related to Engineering Configuration Management noted that the ,

Design Change Administration Audit, conducted in 1987 and 1988 reviewed approximately '

70 modification packages. Based upon these reviews deficiencies were identified that included: lack of design basis information, lack of engineering judgment / assumption documentation, inadequate work instructions, incomplete construction work packages. 4 insufficient 10 CFR 50.59 safety evaluations, and lack of administrative control for '

the training program for engineers. Subsequent audits noted that corrective actions 1

-had been effectively implemented and audit personnel who had participated in _ the previous audit noted improved modification package quality. The Design Change Administration Audit, performed in 1991, determined that previous audit findings had been corrected. Two corrective action reports were generated during this review. These 1 reports were administrative in nature and dealt with completion of documentation l reviews, i

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During the years from 1990 to 1993. 0A personnel further verified compliance with the l Appendix B Criterion III by performing the Design Engineering Audit. This audit was l performed on a yearly basis, covering the four disciplines of Engineering. These audits  ;

verified that, even though some isolated conditions adverse to quality were identified, i the engineering design process was effectively implemented.

The 1995 configuration processes audit identified weaknesses in the areas of documentation update, engineering services procurement, procedure quality and QA program compliance. These conditions adverse to quality were considered isolated and no programatic problems were identified.

A review of the QA surveillance reports notes that very few conditions adverse to quality were identified during the modification reviews. Over the past four refueling outages (1992 through 1996), approximately 65 modifications have been reviewed. Nine  !

(9) corrective action documents were issued to document concerns. None of the I identified weaknesses were considered to indicative of a systematic or programmatic l problem.

Review of these audits revealed that programs have been established and improved to  ;

ensure modifications performed at FCS that maintain the approved design and licensing i bases. Concerns identified during the early audits have been corrected by enhancing Engineering Instructions and providing necessary training to Engineering personnel.

The OPPD Nuclear Safety Review Group provides a monthly review of safety evaluations conducted in accordance with 10 CFR 50.59. These reviews have documented a steady improvement in the overall quality of these evaluations. The 1996 reviews show the number of " attention to detail" type errors have been reduced during the year.

. The OPPD internal assessments provide reasonable assurance that the desig1 change and configuration control processes are effectively implemented and maintain the physical configuration of the plant in conformance with the design bases.

E.11 NRC Inspection of Engineering Design and Configuration Control Processes Several NRC inspections have assessed the FCS engineering design and configuration control processes. A 1992 inspection (92-04) included the following statement:

The team concluded that the licensee had developed and implemented a program to control design changes and modifications in accordance with the TS. the USAR and regulatory requirements. The team considered the performance of safety evaluations to be exceptianal.

The same team also concluded that the temporary modification program effectively controlled temporary modification in accordance with the Technical Specifications. USAR and regulatory requirements.

LIC-97 011 Page 61 A 1993 inspection report (93-08) affirmed that FCS had a good program for performing safety evaluations in accordance with 10 CFR 50.59. The same inspection noted examples of incomplete evaluations, evaluations that lacked attention to detail, and one screening that was not adequate. A 1995 inspection report (95-11) also noted that the lack of attention to detail on the part of reviewers of 10 CFR 50.59 safety evaluations.

OPPD initiated actions to improve the overall quality of safety evaluations and attention to detail. The quality of the safety evaluations is evaluated and tracked by the Nuclear Safety Review Group as discussed in the previous section. The trend shows an improvement in the quality of evaluations since the corrective actions were implemented in early 1996.

A 1994 maintenance reliability initiative inspection repvrt (94-09) noted that the FCS procedures for modification control were comprehensive and provided an in-depth description of responsibilities and detailed instructions. The inspection team evaluated two modifications to determine effectiveness of OPPD's implementation of these procedures. The team concluded the procedures were effectively implemented.

A 1995 inspection report (95-11) identified several minor inaccuracies between the physical configuration, the DBDs and the USAR. The team determined OPPD had identified these problems and had good programs to correct the problems. The team concluded that the programs to correct the inaccuracies in the D80s and USAR appeared to be effective and good progress was being made to reduce the number of these inaccuracies.

E.12 Ongoing Verification Activities On a quarterly basis. System Engineers develop report cards providing commentaries on the health of their respective systems. These reports also identify any open modifications. Engineering Assistance Requests, and other configuration control issues.

In order to provide additional assurance of conformance with the design bases. OPPD is currently completing assessments of selected FCS systems using the Nuclear Energy Institute (NEl) Guidelines for Assessing Programs for Maintaining the Licensing Basis (NEI 96-05). Deficiencies found are entered into the Condition Report system for resolution. The assessment of the Chemical and Volume Control System (CVCS) is complete, and assessments of the Safety Injection and Instrument Air systems are ongoing. OPPD will complete assessments of all remaining FCS safety-related and safety-significant systems no later than February 1, 1999.

A comparison was made between information in USAR Section 9.2 (CVCS) and information in DBD #108 (CVCS). Approximately 90% of the USAR text was reviewed and verified consistent with text in the DBD. Overall, the results of the comparison were positive.

To a large degree, the USAR and DBD describe information related to the CVCS in a consistent manner, and the information corresponds to the plant physical configuration. j Three minor discrepancies were found during this comparison effort. j

LIC-97-011 Attachment 2 Page 62 J

Conclusions of the CVCS assessment included:

No licensing basis discrepancies that could result in a significant public health -

or safety concern were identified during the course of this self assessment.

One instance was found where the plant had not been operated in accordance with the Technical Specifications. Licensee Event Report 96-006 is being prepared to report that all charging pumps were rendered inoperable for approximately a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> period in violation of the Technical Specification requirements (CR 199601087).

Several items were identified that would improve the maintenance of the licensing

, basis:

1. Process improvements are needed for initiating and maintaining Technical Specification Interpretations.

2. Programatic guidance is needed to assess the need to incorporate current operating practices, i.e. operations memos. operator workarounds, old tag-outs, temporary modifications, old nonconformances, into the licensing basis documents.

3. Procedural guidance is needed for evaluating QA Progr6m changes for I reduction in comitment to the NRC prior to implementation. I

4. Process improvements are needed for more timely update of the USAR following configuration changes to the station. It was noted that a more efficient modification acceptance process is currently under i development. I l

The identified reportable condition and the areas for improvement were entered into the Corrective Action Program discussed in the response to Request D.

E.13 Condition Report System Trend The Condition Report system discussed in the response to Request D provides OPPD with the ability to effectively track plant configuration discrepancies. The Corrective Action Program is being effectively used to identify configuration control deficiencies and correct these deficiencies. The large majority of the configuration discrepancies are minor in nature, as defined in Section D.8.

E.14 Conclusion The OPPD processes, procedures and programs described in the responses to Requests A through E provide adequate confidence that the FCS physical configuration is maintained consistent with the design bases, and that design bases requirements are translated into operating, maintenance, and testing procedures. The results of assessments by OPPD and NRC confirm this conclusion.

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1 LIC 97 011 Attachment 3 Acronym Definitions and Referenced Procedures l

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LIC-97-011 I

Attachment 3 Page 1 Acronym Definitions Acronym Definition ALARA As Low As Reasonably Achievable ,

AOP Abnormal Operating Procedure CAG Corrective Action Group CAP Corrective Action Program CAR Corrective Action Report CAS Central Alarm Station CE Combustion Engineering CHAMPS Computerized History and Maintenance Program System CR Condition Report CRG Condition Review Group CVCS Chemical and Volume Control System CWO Construction Work Order DBD Design Basis Document DBR Design Basis Reconstitution DC-ECN Document Change Engineering Change Notice DCP Design Change Package E&TS Engineering and Technical Support EAR Engineering Assistance Request EDSFI Electrical Distribution System Functional Inspection E0P Emergency Operating Procedure ESP Engineering Support Personnel FC-ECN Facility Change Engineering Change Notice FCS Fort Calhoun Station FLC Facility License Change GEI General Engineering Instruction GL- Generic Letter HPES Human Per formance Enhancement System IEB Inspection and Enforcement Bulletins IET Incident Evaluation Team INA Independent Nuclear Assessment IPAP Integrated Perforriance Assessment Process )

IR Incident Report '

ISD instructional Systems Design JC0 Justification for Continued Operation LER Licensee Event Report MR Modification Request MWO Maintenance Work Order MWR Maintenance Work Request NEl Nuclear Energy Institute N00-0P Nuclear Operations Division Quality Procedure

LIC-97-011 Attachment 3

'Page 2 NSRG Nuclear Safety Review Group

, -NSSS Nuclear Steam Supply System OER Operating Experience Review ONC Operational Non-Conformance P&ID Piping & Instrumentation Diagram PED-GEI Production Engineering Division General Engineering Instruction PED-0P Production Engineering Division Quality Procedure PLDBD Plant Level Design Basis Document PRC Plant Review Committee PUP Procedures Upgrade Project

-QA Quality Assurance QAM Quality Assurance Manual QC Quality Control.

OP Quality Procedures OR Qualified Reviewer RCA Root'Cause Analysis

-SAC System Acceptance. Committee SALP Systematic Assessment of Licensee Performance SA0 . Safety Analysis for Operability SARC Safety Audit and Review Committee SDBD System Level Design Basis Document SEP Safety Enhancement Program SER Safety Evaluation Report SMART Station Modification Acceptance and Review Team S0 Standing Order SR Safety Related

. SRI-ECNi Substitute Replacement Item Engineering Change Notice SSC Structure. System or Component SSFI Safety System Functional Inspection SSOMI Safety System Outage Modification Inspection SWSOPI Service Water System Operational Performance Inspection TDB Technical Data Book TM Temporary Modification USAR Updated. Safety Analysis Report USQ Unrtviewed Safety Question

LIC-97-011 Attachment 3 Page 3 Referenced Procedures Eroceduce_110. Eraceduce_ Title N00-0P-3 10 CFR 50.59 Safety Evaluations N00-0P-7 Facility License Changes (FLCs)

N00-0P-12 Reporting of Defects and Noncompliance to the Nuclear Regulatory Commission N00-0P-16 Updated Safety Analysis Report (USAR)

N00-0P-19 Root Cause Analysis Program N00-0P-20 Human Performance Enhancement System Program N00-0P-22 Preparation and Approval of a Safety Analysis for Operation (SAO)

N00-0P-31 Operability and Reportability Determinations N00-0P-34 Ongoing Commitment Program ,

N00-0P-35 Licensee Event and Special Reports PED-GEI-3 Preparation of Design Change Packages PED-GEI-29 Facility Change Evaluation PED-gel-35 Preparation of EARS for Minor Configuration Changes / Replacements PED-GEI-60 Substitute Replacement Item Evaluations PED-0P-2 Configuration Change Control PED-0P-3 Calculation Preparation, Review and Approval PED-OP-5 Engineering Analysis Preparation Review and Approval PED-0P-19 Evaluation of Potentially Reportable Conditions PED-0P-29 Evaluating, Reconstituting, and Closing Design Basis Document Open Items OAM-14 Biennial Procedure Review Oversight Program SAP-35 Reporting of Safeguards Events 50-G-21 Modification Control S0-G-30 Procedure Changes and Generation S0-G-73 Fort Calhoun Station Writer's Guide S0-M-101 Maintenance Work Control 50-0-25 Temporary Modification Control 50-R-2 Condition Reporting and Corrective Action S0-R-3 Reportable Occurrences S0-R-11 Notification of Significant Events 50-R-12 Reporting of Physical Security Events

The Fort Calhoun intraNet WebPage Fort Calhoun Nuclear Station's recent effort to utilize the intraNet technology has provided may benefits in the storage and retrieval of documents for plant personnel. In January 1997, an electronic search engine was added to the web site to allow personnel to conduct searches to support 50.59 evaluations, modifications, and investigations from the following areas:

Design Basis Documents Administrative Manuals Plant Procedures Technical Specifications Fire Hazards Analysis USAR 1

Technical Databook / COLR NRC Documents INPO Documents 4

Offsite Dose Calc Manual Training Lesson Plans FCS Safety Manual The search engine provides search and retrieval in just seconds.

Personnel can screen through information quickly Searches can be limited to specific documents The search engine generates a HTML on the fly for the specific word / phrase search to speed up the search.

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1 Wednesday, February 19,1997 9:25 AM

i 4 FORT CALHOUN STATION January 1997 Monthly Operating Report OPERATIONS

SUMMARY

                                                                                  )

The Fort Calhoun Station (FCS) was at 30% of rated power on January 1,1997 following the steam leak repalm on the orifice downstream of MOV-CV that occurred on December 31,1996. At 2338 hours on January 1,1997, FCS began to increase power after clearing . a hold for Steam Generator chemistry. The Station reached 100% of rated power at 2053  ! hours on January 2,1997 and continued to operate at a nominal 100% power level for the remainder of the month. As a result of the Millstone Unit 2 Licensee Event Report (LER), which reported a problem with the failure to account for steam line pressure drops in the calculation of the Main . Steam Safety Valve (MSSV) setpoints, OPPD investigated if similar issues existed at FCS. i Although it was determined on January 22,1997 that MSSV chatter is not an issue at FCS, it was concluded that the number and location of inoperable MSSVs allowed by FCS Technical Specifications could have placed the plant outside of its design basis. OPPD is performing the necessary analyses to update the design basis of the plant and is providing additional guidance to the operating staff to ensure that the revised design basis is maintained. See LER 97-001 for additional details. l i i

]                    -o-  Fort Calhoun index Value                                                                                             i i

Industry bledien ladex Value i 100. I industry 4 Median l I 90 .. Projected u M M M M M _ ) E 3 70 . i s 4 SO .  ; Value for January 1987 31.s7 50 . ] 4 4 40 95/1 95/2 95/3 95/4 96/1 96/2 96/3 96/4 Jan Feb 97/1 I ' i DATES BY @ARTER (HOST RECENT W ARTER BY MONRl/WARTER)  ! l l PERFORMANCE INDEX TREND The performance index trend calculation is made up of eleven variables each weighted to arrive at an overall index value. The thermal performance, secondary system chemistry, ," and industrial safety accident rate values are calculated for a one-year period. Fuel reliability is calculated on a quarterly basis. The remaining values (unit capability factor, unplanned (unit) capability loss factor, unolanned automatic scrams per 7000 hours critical, j safety system performance, and collective radiation exposure) are calculated for a two-year i period. This method allows the index trend to be more responsive to changes in plant  !

performance, i l

INPO no longer uses the volume of low-level radioactive waste as a plant indicator. The value will still be tracked, but the value will no longer be used in calculating the Stations

!     Performance Index.

U l li

l l t a h%XNUMVALUE g December S6 O January17 8 i 16.00  ! i 14.00 - 8 8 a a t 1N- - 888 888 888 t 10~00 ~ 8 888 88 ' e 00 .

                                                                                                                          "                         """          ""         88 i                   e                  i               i        i                       i              i          i                i               i       y UCF                UCLF               HPSI               AFW          EACP       CRE                         uAS7      FRI         CPI                     TPI   ISAR                      ,

i i This graph shows the difference between the maximum number of points for each WANO indicator and the actual value achieved by Fort Calhoun for

the fourth quarter of 1996. t i

, UCF Unit Capability Factor TPI Thermal Performance Indicator l l UCLF Unplanned Capability Loss Factor CPI Secondary Chemistry Indicator t j HPSI High Pressure Safety injection ISAR Industrial Safety Accident Rate  ! AFW Auxiliary Feedwater l EACP Emergency AC Power j UAS7 Unplanned Auto Scrams / 7000 Hours  ; CRE Collective Radiation Exposure j FRI Fuel Reliability Indicator l i , Per INPO, the Performance Indicator for the Volume of Low Level Radioactive Waste buried will no longer be used in calculating the Stations Performance Index. All other parameters have been adjusted to reflect this change. j I lii [ f r

'l 2 FORT CALHOUN STATION PERFORMANCE INDICATORS REPORT January 1997-

SUMMARY

POSITIVE TREND REPORT POSITIVE TREND REPORT (cont.1 A performance indscator with data representing three Cun"- inine":,d Radiation Controlled Area consecutive months ofimprovmg poib.. nce or three (Page 54) consecubve months of performance that is superior to the stated goalis exhdubng a possbve trend per Nuclear Temoorary Mcdificetiens Opershons Division Quality Procedure 37 (NOD-QP. (Page 58) 37). The folomng performance mdicators exhdxted posstive trends for the reporbng month Safety System Failures ADVERSE TREND REPORT t l Hiah Pressure Safety Iniechon System A performana indicator with data representing three (Page 5) consecubve months of declining p b.. nc. or three consecutrve months of pedun. nce that is trending

Auxiliary Feedwater System toward dedining as determined by the Manager -

(Page 6) Station Engineering, conshtutes an adverse trend per j Nuclear Operabons Divisson Quality Procedure 37 { Cirersency AC Power System (NOD-QP-37). A supervisor whose performana j (Page 7) indicator exhituts an adverse trend by this definition l may specify in written form (to be published in this Semndary Chemistry report) why the trend is not adverse. < The foDomng pedonie.c. indicators exhibited adverse u--v Diesel Generator Unit Reliability trends for the reporting month. age 0) Maintenance Workload Backloo Diesel Generator Reliability (25 Demands) (Page 47) (Page 21) ] 4 Emeroency Diesel Generator Unreliability f (Page 22) i Missed Surveillance Tests Resultina in Licensee Event

  .Recorts (Page 28)

Unotanned Safety System Actuations (INPO Definition.1  ; (Page 34) Hazardous Waste Produced (Page 53) iv l l l

INDICATORS NEEDING INCREASED MANAGEMENT ATTENTION REPORT A performance indecator with data for the reporting period that is inadequate when cu,T@ ed to the OPPD goal is defined as "Needing increased Management Attention

• per Nuclear Operabons Division Quality Procedure 37 (NOD-QP-37).

BrJ8eliability Indicator (Page % Eautoment Farmd OutaQR1 (Page 38) Cents ner Kilowatt Hour (page 44) WANO INDICATORS (As wi@mi to previous month) Unit Capability Factor increasing Unplanned Capability Loss Factor No Change High Pressure Safety injeebon No Change Aux. Feedwater System No Change Emergency AC Power No Change Collecbve Radiabon Exposure No Change Unplanned Reactor Scrams No Change Fuel Reliability Decreasing Chemistry Indacator Decreasing Thermal Performance Decreasing industrial Safety Accident Rate No Change v

j FORT CALHOUN STATION i FUNCTIONAL AREA PERFORMANCE PANEL FOURTH QUARTER 1996

ADMINISTRATIVE CilEMISTRY COMPUTER EMERGENCY l SOFTWARE RESPONSE 1 l i

j ENGINEERING ENVIRONMENTAL EQUIPMENT FACILITY ! OPERATIONS 1 l 4 l 1 IlOUSEKEEPING INDUSTRIAL i SAFETY l a I k MAINTENANCE MATERIAL QUALITY RADIATION PROTECTION } i SECURITY m :- TRAINING i l LEGEND 1QTR 2QTR 3QTR RCm N m PERFORMANCF . Mf4 M ) 1996 1996 1996 f FOURTH QUARTER 1996 IWR j STATUS "~ 1 PERFORMANCE i 1 1 l

0 .. QUARTERLY TREND REPORT i i FOURTII QUARTER - 1996 a

                                                  ' TABLE OF CONTENTS" FUNCTIONAL AREA                                                                                                             PAGE NO.

3~ A D h I I N I ST R AT I V E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I C H E h 11ST R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

C0 51P UTE R SO FTW A R E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 Eh1 ERG ENC Y RESPONS E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

, EN G I N E E R I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4 E N V I R O N S I E NTA L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 EQ U IP h l E NT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 FACI LITY O PE RATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 FI RE PROTECTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 FUE L RELI A B I LITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 H O US E K E EPIN G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 I ND USTRI A L S A F ETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 6 SIA I NTE N ANC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 h l A T E R I A L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 Q U A L I TY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 RADI ATI ON P ROTECTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 S E C U R ITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 0 l S P E C I A L P R O C ES S ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 1 1 T RA I N I N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4 '

4

SUMMARY

The purpose of this report is to provide Omaha Public Power District (OPPD) Management and the Safety Audit and Review Committee (SARC) with a tool to identify any quality trends and to assess the quality of work in their areas of responsibility. The following table provides the number of docummts issued over eadi of the past four quarters. Numberissued 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Peramt Gunge 1996 1996 1996 1996 From Lwt Ouader NRC Violations /Non-Cited 4 Violations 2/0 2/2 5/2 3/1 CRs* 376 419 304 457 50.3 % Significant CRs* 14 7 3 13 333 % MNOls 11 4 8 14 75 % ) NSRG Recommendations 1 5 12 2 -83.3% This does not include Condition Reports (CRs) assigned a Condition Level of 6, because these events are not conditions adverse to quality (by definition). The number of CRs increased from the previous quarter; primarily due to increased work activities for the 19'X) Refueling Outage. The increases in Significant CRs and MNCRs were also the result of outage activities. The performance of the Nuclear Division's as defined by the Functional Area Managerindicated that seven (7) functional areas have declined in performance, one (1) functional areas has improved, while eleven (11) i functional areas have remairui the same. 1 i Declining performance was in the following areas: DECLINING PERFORMANCE STATUS l Department 3rd Quarter 1996 4th Quarter 1996 (Previous) (Current) Chemistry _ Cmi Engineenng Good  ! Envimnmmtal Good Fair l Fire Protection Fair i Housdeeping Good Fair I

;     Material                                             Fair Radiation Protection      Gooci                      Fair 1

I I

l l Improving pertomance was in the following areas i Ih1 PROVING PERFORh1ANCE STATUS Department 3rd Quarter 1996 4th Quarter 1996 (Previous) (Current) IndustrialSafety Fair Good Overall, one (1) areas was rated as " Excellent", eleven (11) areas were rated as " Good *, five (5) areas were rated " Fair", and two (2) area was rated as " Poor" Area rated as " Excellent" was: l SpecialProcesses l Areas rated as " Poor" were as follows: l . Fire Prutection FuelReliability Additional trending information is pruvided as follows: AppendixA " Human PerfomurreTrends" 1 Appendix B "Hardwan,CausalCode Distritmtion l AppendixC " Condition Report Event Type Distribution" l Appendix D " Overview of SpecialCodedIssues" l Appendix E " Trend Report Process" l l l l

                                                                                                                 )

l l 1 l

1 1 ADMINISTRATIVE W.J.Ponec l l l 4Q94 IQ95 2Q95 3Q95 4QTR SELF QV EXT 4Q95 1Q96 2Q96 3Q96 1996 CADS SIG PI l [ D CONCLUSION l l Administrative mlated issues have been grouped into the following three categories; mamgement related issues,  ; peam3 related issues andprucedums related issues. Them was an increase in management related issues from 6 last j qturier to ten (10) this qtarter, pesm3 related issues have increased from seven (7) last quarter to seventeen (17) this qturter; and procedure related issues have incmased fmm ten (10) to eleven (11) this quarter. The fourth quarter data i.rlicates that owrall perfomurre of this wincbwhas cbchned slightly. This dedination can be attributed to the mfueling outage arti tir significart irflux of non-nudear and temporary agency support personnel. 'The overall functional area's gwrforman for the fourth quarter is considered "Coxi". j SELF-ASSESSMENT RESULTS: No formal self-assessments were conducted during this quarter. Twenty-two (22) of the forty-four (44) CADS (50%) assigned to the Adnurustrative Window were self-identified; a notable improvement. Only four (4) of the forty-four (44) CADS were identified as a result of an audit, surveillance or formal observation. j l OU ALITY VERIFICATION RESULTS: Twenty-one (21) QA surveillances, three (3) QA l audits, and two (2) NSRG Assessments resulted in 4 CADS this quarter. CR 199601632 (Level 3) was issued as a result of Surveillance M-%5. CR 199601322 (Level 4) was issued as a result of Surveillance A-96-3. CR 19%01495 (Level 4) was issued as a result of Surveillance Z2-%1. CR 199601414 (Level 4) was issued as a result of QA Audit # 30 /13. EXTERNAL ASSESSMENT RESULTS: TheJUMA Auditidentified one(1)violationwith regard to procedural compliance. CR199601601 was issued to respond to this issue. CADS: A total of forty-four (44) CADS were assigned to the Admuustrative area during the fourth quarter. This is higher than the number of CADS assigned during the third quarter. The number of significant CADS nzmained at 0. SIGNIFIC ANT CAD SUMM ARY: 'There were no significant CADS assigned to the Admnustrative area during the fourth quarter. FCSTrend Report 4th Quarter of19% Page1

1 1 PER FORM ANCE INDICATORS: The published Document Review Perfonnance Indicator has shown a slight increase in the number of reviews more than six (6) months delinquent. i Documents currently included in this indicator are Special Procedures, the Site Security Plan, Maintenance Procedures, Preventive Maintenance Procedures, Operating Manual, PED ' Piocedures, NOD Procedures, Security Procedures and NOD Policies and Procedures. The 1996 Refueling Outage occurring dunng this period contributed to delayed responsiveness in completingbiennialreviews. Four (4) of the Nudear Adnunistrative Services Department internal performance indicators continued to trend positively; however, the indicator that reflects typographical errors showed a declining tmnd. This is attributed to the large volume of procedures that were issued during the outage. The percentage of typographical errors remained constant. j 1 l FCSTrend Report 4th Quarter of19% Page 2

l I i CIIEMISTRY D. E. Spires i 4Q94 mum 2Q95 l 3Q95 l 4QTR SELF l QV EXT 4Q95 l tQ96 l 2Q96 mmm 1996 CADS m m PI ( m CONCLUSION i i This quarter showed an increase in the number of conective action documents (CRs) generated, though no significant CRs were identified and 92% of the CRs were self identified. Four perfonnance indicators relative to chemistry degraded to poor this quarter. Plant shutdowns, leaking fuel pins, a difficult startup and abnormal operating conditions contn'buted to the dedine of these indicators. Several successes relative to the refueling outage were achieved. These were relative to off-site dose control, reduced duration of hold points, and degassing evolutions being performed in significantly less time than in the past. Overall, the Mndow designation this quarter is (considered "Cmi". ) SELF- ASSESSMENT RESULTS: A Chenustry Self-Assessment was perfomTed this quarter. Two (2) CRs were initiated as a result of this assessment. One (1) CR (Emergency Response), was for one of the RM-063 alternate sampling methods for high range gas, particulate and iodme not appeanng to work as written, i.e. equipment staged in an area not designated in the

                                                                                                                           ]

procedure. The other CR (Chemistry), was for the Reactor Coolant System (RCS) fluoride limit  ! being out of speofication. A number of strengths were identified along with several area:, for improvements _ OUALITY VERIFICATION RESULTS: Quality Assurance (QA) Surveillance Report C3-96-1, Instrumentation Calibrations and Function Checks, was performed this quarter. No CRs or CIDs were issued as a result of this surveillance.

                                                                                                                           ]

EXTERNAL ASSESSMENT RESULTS: A specialinspection was conducted by NRC Resident Inspectors on the Post Accident Sampling System (PASS). One (1) item was closed and five (5) potential Violations were identified, two (2) were relative to PASS operability and { maintenancepriority designation.  ! C A Ds: Thirteen (13) CADS (CRs) were issued in the Chemistry functional area during this { period. This is an increase from three CADS last quarter. Twelve (12) of the thirteen (13) CRs j (92%) initiated during this quarter were self-identified. Three (3) CRs dealt with soditun spikes in j the S / G blowdown and lor condensate systems after plant start-up from a refueling outage, and I were related to secondary system pump starts that pumped condenser blacklight water to the SlG.  ! Two (2) CRs wem related to high secondary dissolved oxygen levels, an ongoing problem that is l still being investigated and pursued.  ! FCSTrend Report 4th Quarter of 1996 l Page 3 l l

l i I i Two (2) CRs were due to a failed Al-110 hydrogen monitor, caused by excessive use of the instrument during plant shutdown and startup. Two (2) CRs documented two times that j - feedwater hydrazine was out of speofication due to pump problems, once with CF-13 and once I with CF-16B. One (1) CR was due to RCS hydrogen being low out of specification in preparation for plant shutdown due to low VCT overpressure. One (1) CR was due to CCW system having detectable cesium activity, in part due to a longer i count time to detect such. One (1) CR was due to RG fluoride being above the Action IEvel 1 , limit after plant startup from a refueling outage. The Technical Specification RCS fluoride limit l was never approached. One (1) CR, the only one not self-identified, was due to a Catergory 4 chemical being found stored in a Catergory 3 ama, the Turbine Building. SIGNIFICANT CAD

SUMMARY

No CADS (CRs) issued in the Chemistry functionalarea  !

during this period were considered significant. I PERFORM ANCE INDIC ATORS: The Performance Indicators (PIs) for the Chemistry functional area are shown below with their quarterly results and corresponding weights. Only two I (2) of the performance indicators were excellent this quarter he Chenustry Action Ixvels I Exceeded and the In-line Chemistry Instruments Out of Service continued their positive trends. All other performance indicators were poor this quarter. De Secondary Chemistry Index was poor this quarter due to high copper and dissolved oxygen levels in the secondary. RG Lithium

% of Hours Out of Limit was poor due to a difficult plant startup where power was raised and lowered several times diluting the lithium. The Collective Radiation Exposure was exceeded for all of Fort Calhoun due to fuel leaks. The Forced Outage Rate for the last 12 Months has been poor all year due to maintenance shutdowns. The PI actual weight total for the quarter was 0.615, or fair.

l I l l l FGTrend Report ' l 4th Quarter of 19% Page4 l

1 4 Performance Indicator Result Maximum Weight Actual Weight SecondaryChenustryIndex, 2.58 .3 0.07 QuarterlyAverage 4 7 RCS Lithium %of HoursOutof 5.11 .13 0.03 t Umit, Quarterly Average % CollectiveRadiation Exposure Greater thanOPPD .05 0.01 (Alnorreadings) Goal +10% Forced Outage Rate for last 12 5.38 .02 .005 Months Chenustry Action La'els 1.25 .2 .2 Exceeded /Mo.Overlast 12

Months' In-Line ChenustryInstruments 3.57 .3 .3

Out of Service, Quarterly 1 Average %

Total oH

• 1 0.615 This '.fuarter exhibited a variety accomplishments and a number of areas requinng attention.

Aommplishments induded a succescful completion of the self-assessment, using personnel from tw a (2) nudear utilities, and numerous outage activities. Successful outage activities induded ] nx luced hold point durations dunng the degassing evolutions, a fairly positive report on the steam l gemator health, an innovative use of the condensate cleanup trailer, and effective control of off- I site dose. Additionally, dunng this period, FCS chemistry personnel devised and wrote a  ! procedure for testing radioactive resin. This procedure has been published in the recent revision l to the EPRI Radioactive Waste Characterization Handbook. Gven the above described Performance Indicators, Self-assessment results and outage suc sses and innovative  ! contributions to the nudear industry, the chenustry program, overall, was deemed to be good. > l FCSTrend Report 4th Quarter of1996 Page5

                     ,m     -., ,                  ,        --,          . - + - ,         . , . , - .                , ;-

COMPUTER SOFTWARE T. J. McIvor 4Q94 IQ95 2Q95 3Q95 4QTR SELFlQV l EXT 4Q95 1Q% 2Q96 3Q96 19 % CADS - t - PI l f 3  ! CONCLUSION ' Perfonnance in the Computer Software functional ama continued at a Gagd level during the fourth quarter of 1996. l Positiw progmss has conbnued to be demcostrated with the ERF data server project and a number of wuk group applications. Several of the ITD efforts in support of the nudear orgaruzation have been completed; these indude  ; CIMS frIAZSHIP and the MERIMPP conversion. De EDITS conversion was scheduled for completion during the , fourth quarter; this has been delayed until the first quarter of 1997 due to extensive ITD support for CIMSI HAZSHIP l and MERIMPP. De nudear IT system study was completed during the quarter, and the results were pmsented to managenent m November 20. It has been decided to pursue reasonable enhanmments to the cunent system while conhnuing to monitor developments and plan for the eventual replacement of the system. De nudear organization is ruwprovicirg irpt to the corporate ITproject seyudu g process on a regular basis. %e areas of project planrung and prioritization and the overall computer system management program will continue to mquire incmased emphasis. He level of resourtes assigned to this area remains insufficient to meet the high priority user requests in a timely manner, j this situation is not expected to impmve in the short temt y SELF-ASSESSMENT RESULTS: The area mana ement team has determmed the the most significant factors which affect the Computer Software area are integrated priorities for business i unit applications, resoture issues and the overall computer system (hardware and software) management program Action plans are in either the developmental or active stages for all three (3) of these concerns, and progress is continuing to be made. The solutions, however, are longer term issues and require increased management attention. . 1 A total of five (5) Condition Reports (CRs) were assigned to the Computer Software area during the fourth quarter. All of these were self-identified by the work groups involved (Category 1). OUALITY VERIFICATION RESULTS: There were no intemal quality assessments completed during the quarter. One (1) Quality Assurance Surveillance was begun dunng the fourth quarter. This was requested by the Manager-Nuclear Process Computing Services as a result of CR199601282. The surveillance report was issued in January 1997 and will be addressed in the next quarterly report (Emergent QA Surveillance Report I-96-1). EXTERNAL ASSESSMENT RESULTS: There were no extemal assessments, such as NRC or INFO evaluations, of the Computer Software area during the quarter. l l l FCSTrend Report 4th Quarter of19% , Page 6 l

 . - ,  - ~= - - - .                -.    .-._      .    ---             . . - . - - - .                .-  . -

1 CADS: There were five (5) new CRs assigned to the Computer Software area during the third quarter One (1) was assigned to System Engineering and two (2) were assigned to Nuclear Process Computing Services for resolution.

The remaining two (2) were dosed to another CR These are summarized briefly below

4 j CR 199601281 This CR documents a disparity between the units of the QSPDS display of

wide range steam generator pressure and the calibration procedure for the

pressure instruments (psia vs. psig). It was assigned to System Engineering. A modification request will be submitted toinstall new i EPROMS in the QSPDS which have the correct engmeenng units (psia).

CR 199601282 This CR documents F ? i a software application for tracking outage worked hours was placed into production without having first been dassified by the Software Change Review Committee. The software was subsequently dassified. This CR was assigned to Nuclear Process Computing Services. Even though the problem originated in the user community, NPCS is in the best position to effect lasting corrective action. This CR was the basis for emergent QA Surveillance I-96-1, which will be chscussed in the next quarterly report. CR 199601429 An unexpected host failover of the ERF computer occurred on November 11,1996. An alarm printer in the TSC computer room was left off-line for several days. Alarm data was accumulated until an overflow condition was reached; this, in turn, caused the host failover. The changes for inadvertently talang this printer off-line have been lessened substantially by locking the TSC computer room. l Also, since this event occurred, extensive PMs and testing were conducted l on the entire ERE computer system as a result of the following two reports ' which were dosed to this one. Both host computers and all peripherals are now believed to be in top operating condition. l CR 199601536 The ERFCS was declared inoperable due to malfunctioning tenninals - closed to CR 199601429. CR 199601566 All Control Room ERF temunals locked up - dosed to CR 19%01429. l FCSTrend Report 4th Quarter of19% l Page 7  ! 1 1

1 l 1 l This functional area has one (1) CR that remains open from 1993 (199500426). This has an , action item relative to the Conunitment Tracking System that has been extended to December 31, 1997. Work also continues on a small number of CRs which remain open from earlier quarters of 1996. 1 SIGNIFlC ANT CAD SUMM ARY: There were no Significant CADS assigned to the Computer Software area dunng the second quarter. 1 1 PERFORMANCE INDICATORS: There are currently no formal performance indicators for  ! the Computer Software area. The key measure used by the management team is the degree to which planned and sdieduled activities am accomplished. Several major activities occurred I dunng the fourth quarter which cot.tinued to demonstrate the commitment of the computer groups to provide high quality information services to the nudear organization. These indude the followingitems: I

1. The ERF Plant Data Server was rewived, tested on the development system computer and installed on the production ERF computer in the Technical Support Center. Minor problems were being resolved at the end of the quarter. This computer system is now collecting data since the start up from the 1996 Refueling Outage. Several files have been constructed which can be called through the Fort Calhoun intranet web page for current plant information.

Additional such tables are under construction. Work is continuing on development of trending screens for use by System Engireers.

2. The most signifiant enhancement to the current CHAMPS maintenance management system that was identified by the Nudear IT Study is link provide word processmg capability to the maintenance planners for preparation of work instructions. Several options were exanuned for feasibility during November and December and a preferred option was selected. Work on full-scale development of a Work Instruction Management System began in January; the target production date is March 31,1997.

I 1 l 1 FCSTrend Report i 4th Quarter of 19% I Page 8 l l

EMERGENCY RESPONSE O. J. Clayton 4Q94 IQ95 2Q95 3Q95 4QTR SELF e EXT CNP 4Q95 IQ96 2Q96 3Q96 1996 i PI TRN CONCLUSION Omrall performance within the Emergency Response Functional Area continued to impmve during the fourth quarter of 1996, with several amas needing further enhanament and attentiort The gradual incmase in perfomaance is associated with incmased mamgement attention and support given to ERO related projects, evaluatior- and tratrung sessions. Impmvement was noted in the ama of self identification of program concems by EP Department personnel A thonxgh evaltution of dose assessment proficiency prior to plant start-up from the 1996 Refueling Outage indicated that improvements in the dose assessment training ar d pmficiency retention techniques implemented during the third quuter were strcessfti. A significart "fim1 ware" pmblem mmains to be msolved within the site's siren status feedback system Sewral ,rwgrarming failures wem identified and corrected within the site's automated ERO call out pmcess. The failum to iritiate the mquired offsite notifications within fifteen (15) minutes following the dedaration of a NOUE on 12 B1 B6 is a violation of procedure and mgulations, even though eqtupment failum contributed to the delay. Contiruxirmmgement attertion and support of conective and enhancement actions is required to bring this functional ama to an excellert perfomurre ratirg Omrall performance in the Emergency Resporise Functional Area is considemd { GOOD". j (SELF) SELF- ASSESSMENT RESULTS: One start-up action item (following the 1996 Refueling Outage) was conducted in the fourth quarter and approved by the Plant Review Committee. Validations of dose assessment capability in the Control Room were conducted near the end of the plant's refueling outage to determine if the level of proficiency was sufficient for reactor restart A significant percentage of qualified on-shift dose assessment personnel were observed perfomung one or more mini-drill scenario (s) usmg the equipment and supplies in the actual Control Room. Overall, dose assessment proficiency was demonstrated to be very high indicating a significant improvement in the training and requalification process over the past quarter. This functional area also continued to utilize an Emergerry Planrung Test program to verify and document routine requirements (similar to plant surveillances). Based on the results of the pre-start-up dose assessment evaluations, this area is considered " GOOD." (OV) OU A LITY VERIFIC ATION RESULTS: One (1) QA su2veillance was conducted dunng the fourth quarter of 1996 (final report was issued on January 3,1997) to conduct performance-based observations dunng a scheduled trairung drill on December 18,1996. The surveillance identified strengths, concems and suggestions for improvement. A strength was identified for the three way communications demonstrated amongst the operating crew in the simulator and in-plant Operators. Concems were identified associated with controller functions j which are drill related and would not affect an actual response. Five suggestions for impmvement i were provided for review. Based on the thoroughness of the surveillance conducted, including l the suggestions for improvement, the area of quality verification continues to be " Excellent." l FCSTrend Report . 4th Quarter of19%  ! Page 9 l t

(EXT) EXTERNAL ASSESSNIENT RESULTS: No "extemal" assessments were conducted in the area of Emergency Preparedness dunng the fourth quarter of 1996. Considering there was not any activity in this area, the performance is considered " Good" (or neutral, in this situation).

(C/N/P) CLASSIFICATION / NOTIFICATIONS / PARS: In the area of key responsibilities for actual emergency response, one emergency declaration was made during the fourth quarter.

On December 31,1996, the plant experienced a significant steam leak in the Turbine Building. An evacuation of the Turbine Building was conducted for personnel protection and safety. The Shift Supervisor declared a Notification of Unusual Event based on conditions that warrant increased awareness by plant staff and govemmental authorities. The dassification was 4 considered conservative and appmpriate. A failure of the Control Room Conference Operations Network phone set (the " COP Phone" is the primary means of conferencing all states and counties . together for emergency notifications) resulted in the Control Room Communicator proceeding to the backup method of notifications (individual commercial telephone contacts). The COP Phone was subsequently repatred and the notifications were completed seventeen (17) minutes following the declaration of the emergency classification. The Protective Action Recommendation given on l the NOUE notification was "NONE" v'hich is appropriate. The Alert Notification System (ANS), consisting of 101 sirens, performed at 9Z5% for the fourth quarter and 97.3% for the entire calendar year of1996. Due to the failure to complete offsite notifications of the NOUE within 15 minutes as required per 10 CFR 50, this an>a of performance declined to " FAIR." (CADS) CORRECTIVE ACTION DOCUh1ENTS: Three(3) Condition Reports (CRs)were issued to the Emergency Planrung Department in the fourth quarter of 1996 (one less than last quarter). All three were initiated by the Emergency Planning Department. All three (3) were not considered "significant" in nature, however, one (1) did identify the failure to complete offsite notifications within the required time frame as d2scussed previously. Considering that 100% of the CADS were self-identified and the minor significance of the CADS, tlis area of perfornunce is considered " Excellent." (SIG) SIGNIFIC ANT C AD SUNih1 ARY: There were no Significant CADS assigned to the Emergency Response area dunng this quarter, therefore, performance in this area continues to be

        " Excellent."

FCSTrend Report 4th Quarter of19% Page 10

(PIs) PERFORNI ANCE INDICATOR SUMNIARY: Overall 4th Quarter,19%: WHITE ANS (sirens) Performance: 4th QTR: 97.5% YTD: 97.3% WHITE Oct: 97.6% Nov: 97.5% Dec: 97.5% QTR Growl: 97.5 % Muumum ERO Staffing: All Minimum Staffmg Positions staffed (10%) GREEN Table B-1 Staffing: One Table B-1 Positions was not staffed (96.4%) WHITE Negative Feedback from ERO <0.5% (with training drill conducted) GREEN Negative Evaluation Results: Poor Controller Actions Dunng Training Drill Failure to staff one of two Field Team Technicians YELLOW Ratio of Self Identified Problems: 100.0% of all CADS GREEN Manager-EP Overall Assessment of Emergency Response Functional Area: WHITE (SUP) SUPPORT TO FUNCTIONAL AREA: The Vice President and Plant Manager fully supported the dose assessment evaluations conducted prior to plant start-up fmm the 1996 Refueling Outage. The Nuclear Services Division Manager and several department managers assisted with independent observations of the dose assessment evaluations. The Nuclear Process Computing Services Department continues to assist the department in the preparations and design of a new dose assessment model for the site. Increased attention to the "firmware" problem identified within the site's ANS electronic feedback system is needed to return the system to its intended monitoring function. Based on these activities and continued attention to EP issues dunng the moming Plan of the Day meetings, this area is considered " Excellent." (TRN) EMERGENCY PREPAREDNESS TRAINING: Fourth quarter EPtraming functions continued to improve with the review and rewrite of several EP related lesson phms. Dose assessment training was validated as an improvement through performance based evaluations. The fourth quarter training drill was given a high priority even though the site had recently completed a refueling outage. This area is considered " GOOD" and improving. l l FCSTrend Report , ' l 4th Quarter of19% Page11 i l l

ENGINEERING R. L. Phelps/S. K. Gambhir IQ95 2Q95 3Q95 4QTR SELF e' EXT 4Q95 1996 .I SIG PI CONCLUSION The Ergitterire furtiam! ama paformance has been acceptable. Initiatives taken to irnpmve Engineering timeliness have been successful in providing msponse to plant needs. Self mporting continues to be a domimnt factor in the rtnber of Cmition Reports (CRs). Ergirtering performance on outage pmjects, modtfications and system testing has bem good. Engineering continues to seek out optimal impmvements to plant equipment, however some complex pmjects strh as steam gprerator orifice plate mmoval, have not been entirely successful. Several NOVs and LERs have been written agurst engineering activities this quarter. Therefom, the overall Engtneenng functional ama is evaluated qs "Gmd"perfomunce. y SELF-ASSESSMENT RESULTS: During this quarter, the review of CRs continued using the PED self assessment guidelines. No adverse trends were noted. Initiatives to improve the timeliness of Engineering response to plant priorities have continued to prove effective. The use of the Fast ActionSupportTeams(FASTS),designprocess simplification,identificationof high priority issues, action plans, and daily attendance of PED at Emergent Work Meetings have continued to be effective. Actions to address Control Room Deficiencies, Operator Work Arounds and T-Mod reduction continue to be priorities, and success in meeting plant sdaedule requests for these areas has been acceptable. Utilization of the Condition Reporting System continues to be an effective tool for identifying and taking action on problems. Seventy-one issues applicable to the engmeenng area were noted during the fourth quarter. Review of the CRs revealed no adverse or degrading trends developing regarding current engineering processes Review.s of completed 50.59s, before issuance by designated readers, shows a continuing need to maintain attention-to-detail. Forty-six (46) of the seventy-one (71) CRs (65%) assigned to the Engineenng functional area were self-identified. 1 'Ihe 1996 Refueling Outage donunated engineenng support activities in the fourth quarter, and many activities demonstrated successful performance, such as the RCP 3B motor replacement, steam generator tubes inspection and repair, fuel reconstitution and reload redesign and actions to resolve the G.L. 96-06 issue. However, other issues such as USAR review, fuelloading by an  : inoperable WRNI detector, PASS operability and APD trip bypass technical speafication ) compliance concems have indicated that overall performance is not outstanding. In addition,  ; Engtneenng support has been involved in several LERs (some relating to the Issues above) and in I several violations. The move to WHITE is indicative of a caution flag at this time, rather tlun a negative trend. FCSTrend Report 4th Quarter of1996 Page 12

l l l l OUALITY VERIFICATION RESULTS: Two (2) QA surveillances and one (1) audit were I conducted in the Engineering functional area during the third quarter of 1996. The QA surveillances and audit resulted in no significant findings EXTERN AL ASSESSM ENT RESULTS: No fourth quarter inspection by an extemal agency specifically focused on Engineering. NRC Monthly Exits have reviewed Engineenng Support performance, and there have been some strengths and a few weaknesses reported. There may be a concern developing that Engineering support will receive irrreasingly greater scrutiny in light of I the industry / NRC concern exempli 6ed in the 50.MF letter on design basis. This is already evident - in many of our interfaces with the NRC, and the highest attention to thoroughness and quality will be thehallmarkof GREEN performan in this area. C ads: A total of seventy-one (71) CRs were assigned to the Engineenng area during the fourth quarter,1996. Comparisons with previous quarters showed that engineenng continues to be aggressive in identifying problems. In addition, action items resulting from CRs are being effectively managed. No adverse trends were identified during the fourth quarter. The CRs were reviewed for significance and indication of Engineering Program effectiveness. Special attention was given to CRs that requimd fonnal QP31s, and QP19s. Engineenng processes were found to be effective in identifymg, managing and resching identified problems. There is an increasing trend in CRs requinng QP-31 and QP-19 evaluations, and Engineenng must be attentive to the l priority and schedule for these. ' SIGNIFICANT CADS SUMM ARY: There were three (3) Condition Level l CRs in the area of Engineering dunng the fourth quarter of 1996. One Level 2 CR was issued in the area of Engineering. A review of the CRs did not result in a negative trend being identified, however the awamness of operability assessments is heightened. I

1. CR 199601351 The NRC has issued GL 96-06 which asks utilities to review the design of the containment air cooling system dunng a LOCA or MSLB event coincident with loss of offsite power (LOOP). The scenario of concern is a loop causes the CCW pumps to stop and the lack of CCW flow through I the coolers can cause steam flashtng in the cooling coils. (Level 1)

2. CR 199601519 During dTecking and reviewing of Revision 1 of EA FC 96-042, "HELB Steam Migration from Room 81 via Ductwork," the reviewed observed that a "CR" had not been issued for the conclusion that EEQ temperature l limits for Room 81 were exceeded for the " Dry Steam and Superheated steam cases. (Level 1) l FCSTrend Report 4th Quarter of19%

Page 13

1 1 - 3. CR 199601625 This CR is written to document a reportable condition which evolved from review of IE Bulletin 86-06 as prompted via CR 19%00737. Specifically, a plant corrective action document, Condition Report (CR) 199600737, l noted that certain ESF actuated valves changed position on ESF signal l reset when the change was not expected. (Irvel 1)  ; 1

4. CR 199601532 Per Technical Specification 1.3, the APD Trip is allowed to be bypassed I below 15% power. At less than 17% Delta Power, the APD Trip was 1 enabled on the RPS for all channels. Since one indication of power indicated greater than 15%, this CR was written to ensure that proper guidance would be placed into our procedures. (I.evel 2) ,

l PERFORM ANCE INDICATORS: The Performance Indicators (PIs)in the engineenng area 1 are good overall. The Indicator for Thermal Performance has continued to improve with year end performance above the goal. Fuel Reliability continues to indicate a negative trend which is expected to continue through Cycle 17 due to the remaming Westinghouse fuel assemblies that , are subject to vibration failures. The action plan for the fuel is being effectively managed. The Thermal performance indicator is expected to see additional improvement from planned use of an  ! EPRI performance monitonng computer code to be implemented by early 1997. Several areas . that Engineering is directly responsible for show positive trends. These indude Diesel Generator l Reliability, HPSI and AFW systems, Surveillance tests and Temporary Modifications. l Improvement continues to be needed for Equipment Forred Outage Rate. I FCSTrend Report 4th Quarter of19% Page 14

1 l ENVIRONMENTAL D. E. Spires l 4094 l IQ95 l 2Q95 3Q95 4QTR SELF EK8P - EXT 4Q95 memum 2Q96 3Q96 19 % CADS l SIG PI , I ( ) l CONCLUSION i The Environmental functional am decmased to yellow this quarter based on the relative number of Corrective Action Documents (CADS), one (1) Signification CAD, and radicactive effluent events. The number of reportable spills / NPDES violatiors daring the last twelve (12) months improved during this quarter. Quality verification results improved to green while CADS decreased to yellow. A high liquid effluent total txxiy dose, relative to previous years,is still a major concem in the Environmental Area. Overall, the performance of the Environmental functional ama for the fcurth quarter of 19% is " Fair". j SELF- ASSESSM ENT RESU LTS: Dunng this quarter a Chemistry self-assessment was conductect The results of the assessment showed two (2) areas ofimprovement and three strengths in this functional area. The improvements pertained to plant process monitors requmng frequentmaintenan and process monitor sampling procedures requiring some improvements. The strengths were in relation to the improvements of the chemical injection facility, computer trending of the rad effluent dose, and personnel " depth of capability" in the rad effluents area. It was also noted in the assessment that the Chemistry Department has never had a lost time accident. All of the CRs (100%) written dunng this period were self-identified. OUALITY VERIFIC ATION RESULTS: Two (2) QA Surveillance reports were issued during this quarter. B3-%2 Environmental Monitoring and B5-%1 Radwaste Refueling Activities were both completed with nc CRs or CIDs betng issued. The results of both reports were very positive. EXTERNAL ASSESSMENT RESULTS: Tl'em wem no extemalassessmentsperformed during the fourth quarter; however, personnel from other utilities participated in the afommentioned Chemistry self-assessment. CADS: Six (6) CADS were issued in the Erwironmental functional area during this period, one of which was significant. There were six CADS last quarter, none of which were significant. Two (2) of the CADS assigned to this functional area addressed spills or releases of hazardous matenals, which is identical to the number last quarter. All of the reported spills were minor in nature and were not reportable. The other four (4) CADS issued were in reference to rad effluents operations and documentation. These CADS concerred a VIAS during opening of the primary man ways, an unmonitomd release via Room 66, exceeding the noble gas admuustrative limit speofied in CH-AD-0021, and temporarily missing master purge paperwork for release %156. FCSTrend Report 4th Quarter of 19% Page 15

SIG NIFICANT CAD SUMM ARl: The Significant CAD stems from a VIAS that occurred during the 1996 Refueling Outage as the primary man ways were removed from the S / Gs. l PERFORM A NCE INDIC ATORS: The Performance Indicators (PIs) for the Environmental functional area are shown below with their quarterly results and corresponding weights. All of the PIs were excellent for the quarter except liquid effluents quarterly total body dose (poor), reportable spills / NPDES violations (fair), and other documented spills (fair). Reportable spills / NPDES violations Dunng the last 12 Months, a total of three (3) events, increased to fair. This is a positive trend. The liquid effluents quarterly total body dose of 5.65-1, decreased to poor as a result of plant transients in support of the 1996 refueling outage. This decrease was not unexpected due to fuel problems experienced during the last cycle. Other documented spills remamed unchanged at fair. Though most of the parameters showed a positive trend, the window color will be reduced to yellow, given the unmonitored release into Rm. 66, and the VIAS Performance Indicator Result Maximum Actual Weight Weight , 1 ReportableSpills/ NPDES 3 .25 .14 Violations Duringlast12 Months Other DocumentedSpills 8 .2 .14 Dunng Last12 Months > HazardousWaste Produced 15.9 .2 .2 per Month Over Last12 Months, Kilograms MixedWasteProduced per 0 .15 .15 Month Overlast12 Months, Kilograms Liquid EffluentsQuarterly 5.65 E-1 .2 .05 Total Body Dose, mrem Total 1 0.68 FCSTrend Report 4th Quarter of 19% Page16

l l EQUIPMENT M. R. Core l l 4Q94 IQ95 2Q95 3Q95 4QTR SEff QV EXT I 4095 IQ96 2Q96 3Q96 1996 CADS SIG CONCLUSION

    'Ihe number of CADS has increased som what from the third quarter, although many of the increases are due to        l I

equipment that cannot be serviced / Phra on-line and equipment problems were noted as a result of outage maintenance. There are no negative results from any assessments during the quarter. Category ( A1) equipment has been Increased by one, and all temaining corrective actions for the equipment remaming in category (A1) are in (place. Equipment periomunce dunng the third quarter has remained " Good". j l SELF-ASSESSN1ENT RESULTS: There were no Root Cause Analyses (RCAs) conducted that affected the equipment area dunng the forth quarter. OUALITY VERIFICATION RESULTS: No equipment concerns were identified as a result of normal Quality Control (QC) inspections, Quality Assurance (QA) surveillances and audits, Safety Audit and Review Committee (SARC) audits, and Nuclear Safety Review Group (NSRG) assessments and special reviews performed during this reporting period. EXTERNAL ASSESSh1ENT RESULTS: Numerous NRC inspections were conducted dunng the reporting period. There were no issues related to the Equipment Functional Area. i l CADS: A total of 129 CADS were assigned to the Equipment Functional Area during the fourth quarter, an increase of forty-two (42) from the third quarter 19% (NOTE: The number mismatch between third and fourth quarter is because the latest data from the Condition Report (CR) System is being used and has changed slightly since the third quarter mport was prepared). Only two (2) of the CRs were Significant. The large number of CRs this reporting period is largely the result of issues identified during the 1996 Refueling Outage on equipment that is only available for mamtenance dunng outages. SIGNIFIC ANT C AD SUh!51 ARY: There were two (Irvel 2) Significant CADS assigned to the Equipment Functional Area during the fourth quarter. One (1) on an anchor that had pulled l out of the ceiling in Room 13 (19%01183) and the other on RC-142 being out of tolerance for l the setpoint (199601272). PER FOR 51 ANC E INDIC ATORS: The Thermal Performance Indicator has continued to exceed both the FG and INPO goals. Only December data is available, due to the 1996 Refueling Outage, and the indicator is 99.6%. The year end average for 1996 is 99.69%. (The FCS goal is 99.6% and the INPO year 2000 goal is 99.5%.) FGTrend Report 4th Quarter of19% Page 17

M AINTEN A NC E RULE: During the fourth quarter, the air compressors and daarging pumps were added to category (A1) and the condenser was removed from category (A1). Additionally, all of the recommendations for removal of equipment from the (A1) category are in place. l FCSTrend Report 4th Quarter of 1996 Page 18

FACILITY OPFRATIONS R. W. Short 4Q94 IQ95 2Q95 3Q95 4QTR SELF QV EXT 4Q95 IQ96 2Q96 3Q% 1996 ' PI CONCLOSION The 1996 Refueling Outage (October 5 through November 27) comprised a significant }ution of this quarter. Operations performance in some areas (e.g., no loss of shutdown cooling) was very good and in other areas (e.g., safety system actuations and n> portable events) was poor. Causes of the poor performance induded an outage schedule that was not scheduled via system windows, high radioactivity caused ty failed fuel and inattention to detail. Self-assessments, quality verification and extemal c.ssessment results all indicate good to fair performance. However, based on the number of significant Condition Reports (CRs) (8) and associated reportable events, the Facility Operations area is rated as " Fair". j SELF-ASSESSMENT RESULTS: Routine management observations of the Control Room, Field Operators and the Outage Control Center (OCC) performance indicates good performance with mom for improvement in all phases. The 1996 Refueling Outage schedule not betng i completely arranged in a system window framework did challenge this functional area, espeaally the OCC. Seventy-six (76) of ninety-two (92) (83%) CRs assigned to facility Operations were self-identified by the Operations Department. l i OUALITY VERIFICATION RESULTS: Quality Assurance (QA) survetilances on the EOP / AOP Program and Contamment Integrity were conducted with no problems noted. The l Nuclear Safety Review Group (NSRG) completed its comprehensive report on the 1996 i Refueling Outage and this functional area's results were generally positive with the exception of I the Control Room and OCC coordination was noted as weak at times. EXTERNAL ASSESSMENT RESULTS: NRC inspections were conducted by the Resident inspectors and also by Region IV in the areas of Refueling Operations, Conduct of Control Room Activities, and Licensed Operator Requalification. A level IV violation was received for loading fuel next to an inoperable wide range detector and a non-cited violation was received for loading computer tagging databases enor that resulted in the wrong fuses being pulled for maintenance on a valve. The Refueling Operations inspection report was critical of several Operations activities in this area. The other inspection reports, especially Control Room activities and Licensed Operator Requalification, noted good Operator performance. FCSTrend Report 4th Quarter of 19% Page 19

CADS: A total of ninety-two (92) CRs were assigned to the i'metional area of Facility Operations, with eight (5-level 1s and 3-12 vel 2s) Significant CRs. This is an increase from forty-seven (47) CRs last quarter (with only one was Significant). Of the ninety-two (92) thtrty-nine (39) have been assigned to the Operations Department for rec,olution. Even considering the increased activity with the 1996 Refueling Outage fom10ctober 5 Srough November 27, the number of Significant CRs (all 8 resulted in LERs) was unacceptably high and this area reflects an i adverse trend. I SIG NIFICANT CAD SUMM A RY: Eight (8) Significant CRs were assigned to this area in the I fourth quarter. 'Ihis is an increase from one (1) last quarter. l PERFORM ANCE INDIC ATORS: Eleven (11) perfonnance indicators were established for use in assessing the Facility Operations area. Of these, three (3) exhibited a positive trend, one neutral trend and seven (7) exhibited a negative trend compamd to last quarter. This is considered fair performance based on the increased number of Operator manipulations required during the1996 RefuelingOutage. Forced Outage Rate (for last 12 months) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5. 3 8% Unplanned Automatic Rx Scrams (for last 36 months) . . . . . . . . . . . . . . . . . 0.00/17000 Hours Unplanned Safety System Actuations (NRC Def) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . O DailyThermalOutput (% days at 1495 Mwth) . . . . . . . . . . . . . . . . . . . . . . . . . . . 16%* l Personnel Related LERs . . . . . . . . . . ........ . . . ... . .. ...... ........ 2 PersonnelRelated NRC Violations . ... ... .. . .. ... ............ 1 Non-Cited Mas-Operation (Self-Checking) . . . . . . . .. ...... .... ........... ........ 4 MtsOperations (Procedure Adhemnce, Inadequate Communication, Inadequate Procedure) . . . . . . . . . . . ... . .. . .. . .. . . .. 2 EquipmereOut of Position . . . ..... .. . ................ . ........ . . ... 2 SignificantEvents . . .... .. . ..... .. ... ........... ..... .. ... .0 Tagging Events .. .... . .. .. . ...... . . . . .... . ... .2 1996 Refueling Outage from October 5,1996 through November 27,1996. FCSTrend Report l 4th Quarter of 19% Page 20

FIRE PROTECTION R. L. Jaworski -em IQ95 2Q95 3Q95 4QTR SELF QV 4Q95 l IQ96 2Q96 3Q96 1996 .I SIG PI CONCLUSION Thmnere na signficant CADS identified this quarter. Three (3) Condition Reports (CRs) were classified as Irvel 3. One (1) of the level 3 CRs regarded violation of Transient Combustible Limits as specified in SO-G91. This is an ongoing problem, two (2) Notice of Violations (NOVs) were issued against Standing Order G91 this year. A Lube Oil leak on RCP-3D has prompted an NRC mview of FCS License Basis on Appendec R, III.O. A Docketed Request for Further Information is forthconung. As noted in the System Enginwr's report FP system MWOs are continuing a two (2) year tretti of irrmasiry backlog. Based on the above, the Fire Protection functional area must be considemd in (an adverse trend this quarter. j SELF-ASSESSMENT RESULTS: Fire Protection (FP) System Report Cards indicate overall System Assessment for 1996 was good. However, there are four (4) areas of concern: 1) Appendix R, Section III.0 issues have surfaced this quarter due to a RCP fire at ANO in October, DEN-Mechanical is finahzing EA-97002 and assisting System Engineenng with QP-31 on the 'D' pump leak; 2) in response to CR 199601142, a HPES on transient combustible storage disempancies identified the need for labeling of containers of combustible liquids, the labehng has been performed and ' awareness' training provided via 3N pmsentations and memo; 3) in respnse to CR 199601591 regarding FP barrier penetrations the FPPET has taken responsibility for resolution and will meet later in January 1997 to develop corrective action plan; 4) FP system MWOs have continued to accumulate this quarter, and the adverse trend in MWO performan has continued for two (2) years. OUALITY VERIFICATION RESULTS: Quality Assurance $urveillance Report No. F1-96-3 was completed this quarter with a conclusion that the "FCS FP program is adequately implemented". EXTERNAL ASSESSMENT RESULTS: ANIInspection completed this period; no new recommendations; five (5) suggestions closed; four (4) new suggestions. Overall the ANI FP Inspection was positive, with exception of G91 Combustible Control Repeat Issues. NOTE: , Notice of Violation Level IV received for Standing Order G91 violation in the Intake Structure. ) l FCSTrend Report 4th Quarter of 19% Page 21

CADS: A total of thirty-three (33) Corrective Action Documents (CADS), all Condition Reports (CRs), were assigned to the FP functional area during the fourth quarter. An additional eight (8) , CRs related to the FP Program or FP equipment, but not assigned to the FP area, were generated I during the fourth quarter. None of the CRs was considered significant, but three (3) of the CRs assigned to the FP area were classified as Level 3. One (1) of the Level 3 CRs documented discovery of transient combustible aerosols that were not stored in accordance with Standing Order G91. Two (2) other CRs also dealt with violations of the requirements of Standing Order I G-91. l Violation of the requirements of Standing Order G91 has been an on-going problem, and has ) resulted in two (2) Notices of Violation in the past. Corrective actions are on-going to improve performance in this area. The other two (2) Level 3 CRs concerned the historical installation of a trap door in HVAC ductwork which would have potentially interfered with the operation of a fire damper, and a fire in Room 66 which involved smoldering rags. The remaining CRs were reviewed for significance and indications of general FP Pmgram effectiveness, speofically in the area of fim barrier configuration control, fire pmtection hardware, housekeeping and control of combustibles, and staff training program status. Ten (10) CRs documented instances where FP equipment was inoperable in excess of the time limits provided in the USAR and Standing Order G-103, mostly as a result of Refueling Outage (RFO) activities. Four (4) CRs addressed failures to obtain Fire Protection Impairment Permits (FPIPs) prior te removing FP equipment from service. Four (4) CRs addressed firewatch deficiencies associated with Contract Firewatch personnel dunng the RFO. The remaining CRs addressed minor equipment and documentation problems and personnel errors, with no particular trend observed. Here were no LERs or CADS which are defined as significant by the INFO Significant Event Criteria. The continuing instances of noncompliance with the requirements of Standing Order G91, as well as deficiencies  ; assocated with Contract Fimwatch personnel and initiation of FPIPs for inoperable FP equipment, is indicative of an adverse trend in the FP functional area for this quarter. SIGNIFICA NT C A D SUhf NI ARY: There were no significant CADS associated with the Fire Protection functional area this quarter. PERFORM ANCE INDIC ATORS: There are no Performance Indicators speafically associated with the Fire Protection Program and Fire Protection functional area. l l FCSTrend Report 4th Quarter of1996 Page 22 1 j

1 l 1 i FUEL RELIABILITY M. J. Guinn  ! 2Q95 3Q95 4QTR SELF QV EXT 4Q95 sq96 2Q96 3Q96 1996 CADS SIG fJNCLUSION Based an me ftel failure idatfied in December 1996 and the expectation that futum fuel failures will be identified, the Fuel Performance function area performance is to be considemd " Poor" for the fourth quarter of 1996. j SELF-ASSESSMENT RESULTS: The scope of the completed, planned and ongoing self-Assessments exhibit a STRENGTH in the Fuel Reliability Functional Area. A Fuel Inspection and Reconstitution effort was completed during the 1996 Refueling Outage by Nudear Engineenng and other plant personnel with Westinghouse assistance. Full core in-mast sipping and ultrasonic testing determined seventy-eigth (78) failed pins. Nine (9) assemblies retunung to the core had a total of fourteen (14) pins reconstituted. The nine (9) assemblies reconstituted induded tlute (3) thrice burned (Batch R), four (4) twice burned (Batch S) and two (2) once burned (Batch T). Primary failure mechanism in thirteen (13) out of the fourteen (14) mconstituted failed rods inspected was grid to rod fretting. Forty-four (44) Batch "U" assen blies with a revised fuel assembly grid design wem installed in the Cyde 17 core to minmuze this failure mechanism. Ongoing tests at the Westinghouse Fuel Assembly Compatibility Test System (FACIS) am being conducted to generate comparative impact characteristics of the original and revised grid designs. S. M. Stoller Corp. is an independent consultant overseeing this endurance testing scheduled for first quarter 1997. One (1) failed rod in a once bumed assembly, T045, had an incomplete end cap girth weld. Westinghouse conducted a detailed review of manufactunng and quality records for this failed md with no adverse findmgs. A new grooved end plug design is proposed to ehnunate incomplete I welds not detectable during the manufacturer's inspections. Pmsentations wem made to PRC, SARC and numerous Nudear Utility Managers conceming the i 1996 fuel inspection and reconstitution results. l A meeting with OPPD, Westinghouse and Stoller personnel is planned for February 11,1997 to develop the strategy and schedule for options to improve fuel performance with the goal of the ' l Fort Calhoun Station (FG) having zero fuel failures in Cycle 18. 1 FGTrend Report 1 4th Quarter of19% Page 23

i 4 OUALITY VERIFICATION RESULTS: There were no significant findings for the quarter. This is a STRENGTH in this performance category. Quality Assurance (QA) Audit Report No. 48, Nudear Fuel determined that activities associated with the performance of fuel mconstitution

have been effectively established. Surveillance Report on fabrication of FCS fuel at Westinghouse identi6ed no deficiencies. Particular attention was given to the process changes relating to the design dianges for the Batch "U" fuel mentioned above, There were no NSRG recommendations .

related to this project. EXTERNAL ASSESSNIENT RESULTS: The fuelinspection and reconstitution was observed by the NRC and no concems and violations were reported. C ads: The number of Condition Reports (CRs) and the recurnng fuel failures identified in the i Level 3 CRs indicate an ADVERSE trend. Six (6) CRs were assigned to the Fuel Reliability functional area during the fourth quarter; one (1) - level 3, two (2) - Level 4, and three (3) - . Level 5.

SIGNIFICANT CAD SUhlN1 ARY

No significant CRs were issued pertauvrg to fuel reliability in the fourth quarter.

PERFORN1 ANCE INDICATORS: The performance indicator, the primary contributor to this window, currently indicates a NEUTRAL trend. The Cyde 17 fuel performance is trending similar to Cyde 16. The monthly FUEL RELIABILITY INDICATOR (FRI) for December 19% was 4.5 X 10-' microcuries/ gram. This was the first monthly FRI value for Cyde 17. As of 4 December 31,19%, FG personnel assessed that there was one (1) failed pin in Cycle 17. This assessment was based on the previous cyde's fuel perfom1ance history; CADE, a feel failure J prediction code; and the radiochenustry samples showing an lodme 131 spike taken during the down power on December 31,1996. The fuel vendor, Westinghouse, has also confirmed this fuel failum pmdiction. l i i

FCSTrend Report l 4th Quarter of19%

, Page 24

. HOUSEKEEPING G. C. Bishop 4Q94 IQ95 muse 3Q95 4QTR SELE QV EXT 4Q9, IQ96 2Q96 3Q96 19 6 CADS SIG PI CONCLUSION  : Housekeeping performan is considered in need of impmvement for the fourth quarter of 1996 based upon Standing Order G-6 Observations, a level 4 NRC Violation, ard plant t m. Several observations arrl CRs were wntten on 1 luusekeeping problems during the outage. Incmased attention is needed dunng the high activity period of an outage y SELF-ASSESSNIENT RESULTS: Based upon plant tours, six (6) CRs were written to j document deficiencies in housekeeping. Repeat problems that need further attention are ladder storage and control of chemicals. Also, some housekeeping problems in contairunent will be addressed in the outage critique. A formal self-assessment of the housekeeping program was not completed this quarter. OUALITY VERIFICATION RESULTS: Quality Assurance continued perfomung tours within the plant dunng the outage. Housekeeping deficiencies were addressed through the outage organization. EXTERNAL ASSESSh1ENT RESULTS: A Level 4 NRCViolationwas writtenbecause Standing Order G-91 was violated when several gallons of paint was left muttended on the top floor of the intake structure at the end of the work day. CADS: Six (6) (Condition Reports) CADS were intemally identified compared to three (3) last quarter. These included one (1) on improperly stored ladders after the outage and three (3) on improper control of chemicals. Control of chemicals will be further emphasized especially dunng thehigh activity time of anoutage. SIGNIF1 CANT CAD SUNihlARY: NoSignificantCADswereidentified. PERFORhl ANCE INDIC ATORS: Housekeeping is not tracked in the monthly performance indicator book. The Housekeeping indicator color has been assigned " Yellow." There appears to be a degrade in performance because of the deficiencies noted and the Level 4 NRC Violation. However, the high activity of the outage contributed somewhat to the increase in deficiencies. A Ramp-up item will be added to review these CRs as part of pmparation for the next refueling outage. FCSTrend Report 4th Quarter of 19% Page 25

I I INDUSTRIAL SAFETY G. C. Bishop 4Q94 3Q95 4QTR SELF C o' 4Q95 IQ% 2Q96 3Q96 1996 CADS 1 PI i l CONCLUSION Industrial Safety's performance during the fourth quatter of 1996 was hi". For' Calhoun Station (FCS) employees not only achieved their goal of no lost Time Accidents during the 1996 Refueling Ourage, they worked  : safely enough to ;xevent a Recordable Injury as well Another significant achievement in regards to CADS is that I they did not increase even though the Outage presented incnnsed exposure to occupational hazards, and none written Qvere significant. y S EL F-ASSESSNIENT R ESULTS: Them was no formal "self-assessment" perfomied in the Industrial Safety discipline dunng the fourth quarter, however, the orogram is monitored by the IndustrialSafety Coordinator. OUALITY VERIFICATION RESULTS: There were no sigdficant results pmvided. EXTERNAL ASSESSNIENT RESU!.1S: No enernal assessments were conducted of the Industnal Safety Program during the fotn :h quarter of 1996. C ads: Condition Reports (CRs) did not increase dunng the fourth quarter vice the third quarter, and them were no siFaificant trends identified. SIGNIFICANT CAD SUNINIARY: None. PERFORNI ANCE INDIC ATORS: At the end of the third quarter the statistics are as follows: LostTime Frequency Rate 0.44 0.25 Recordable Fmquency Rate 2.03 0.10 1 Vehide Accidents 0 0.10 Corporate Violations 0 0.10 NRC Violations 0 0.25 i i 0.80 FCSTrend Report 4th Quarter of1996 Page 26

d MAINTENANCE H. J. Faulhaber 4Q94 IQ95 2Q95 3Q95 4QTR SELF QV EXT ' 19 % CADS SIG 4095 1Q96 2Q96 3Q96 PI CONCLUSION During the fourth quarter, pedormarre in the Maintemnce functional area continued to improve. Perfomunce

,           remains good in all but one (1) ama, the Pedormance Indicator index, which is rated fair. In addition, the Maintenance Self Assessment conducted this quarter indicates that the Maintenan Department is effective in

meeting the requuements of the INPO maintenance objectives. The on-line conective maintenance backlog did climb above the goal during the quarter. However, this was due to focusing on outage related conective maintenance during the Refueling Outage. The average backlog for the fourth quarter was still below the backlogs f

seen during the first and second quarter. The Performance Indicator Index continues to show a need for continued management attention in selected areas. As a msult of continued good perfomunce in most amas, the overall rating

qvill continue to be " GOOD" for the fourth quarter, j i

SELF-ASSESSMENT RESULTS: The Maintenance Department conducted a one (1) week self-assessment during the month of De mber. 'Ihe self-assessment team induded seven (7) 2 OPPD personnel and two (2) peer evaluators, one (1) from the River Bend Station and the other from the Brunzwick Nuclear Plant. The OPPD personnelinduded mpresentatives from Quality Assurance, Quality Control, Chemistry, Radiation Protection, Training, and Maintenance. The ! assessment was conducted using maintenance performance objectives and criteria from INFO publication 90-015, " Performance Objectives and Criteria for Operating and Near Term Operating License Plants." The assessment team concluded that, when examming the overall performan l objectives the Maintenance program is effectively meeting the requirements of each objective. However, when examining individual criteria for each objective, the Team did recommend improvements in several areas. In addition, the Team also identified several strengths Both recommendations for impmvement and stmngths were presented to Management in draft form on j December 6,1996 and the final report is bemg developed. Self identification of problems in the Mamtenance functional area continues to improve with approximately 62% (44 of 70) of the Condition Reports (CRs) being initiated by maintenance j pemannel Performance in this area continues to be rated as GOOD. i  ! FCSTrend Report 4th Quarter of 19% Page 27

1 OUALITY VERIFlCATION RESULTS: Quality Assurance perfomied one (1) audit, one (1) emergent surveillance, and two (2) scheduled surveillances during the quarter. Audit e 18 was i conducted in the area of " Corrective Maintenance Activities - Mechanical." The e udit did result in , one (1) CR and two (2 ) QDS. CR 19%01331, was not maintenance related and the CID 960667, recommended improvements when job briefings encompass several Msks. The emergent suweillance, M-96-5, was conducted in the area of M&TE usage. The surveillance  ! concluded that controls of M&TE "from issue to return" are inadequate. CR 199601632 was initiated to track corrective action for this programmatic deficiency. The first scheduled surveillance,01-%1, was conducted in the area of surveillance test and calibration procedure performance. He surveillance included obsewing the performance of thirteen surveillance test / calibrations. No deficiencies or discrepancies were identified. Finally, the second scheduled suweillance, FI-%3 was in the area of Fire Protection. This surveillance included observing attnb ites which included fire impairment initiatiort job briefings, system release for work, riggir: . .ammable storage and capacity verification, and safety. The . surveillance resulted n > .e (1) CID (CID 960780). However, the CID was not maintenance i related. During this quarter, Quality Control conducted fourteen (14) surveillances of maintenarre work l activities. QC noted three (3) concerns associated with those work activities and CRs were initiated where appropriate to document conditions adverse to quality. Specific issues related to the CRs are covered in the CAD section of this report. QC also noted many examples of good work practices, good attention to detail by craft personnel and quality work packages. As a result of minimal deficiencies identified in the area of Quality Verifications this area is rated as GOOD. EXTERNAL ASSESSMENT RESULTS: There were no extemal assessments perfomied dunng the fourth quarter. Corrective actions for the INPO findings identified dunng the second quarter are continuing to progress The NRC Resident Monthly inspections did not identify any violations dunng the fourth quarter that wem assigned to the Maintenance SALP category. Perfonnance dunng NRC inspections improved during the semnd quarter and the improvement was sustained through the third and fourth quarter. As a result of the continued improvement through the last three quarters, performance in this area is rated as GOOD. FCSTrend Report 4th Quarter of 19% Page 28

CADS: Seventy (70) Condition Reports were assigned to the maintenance functional area this quarter. The total CRs assigned to the maintenance functional area increased fmm 31 to 70 when ~ compared to the third quarter. However, the 1996 refueling outage occurred during this quarter resulting a dramatic increase in the number of activities and personnel on site supporting . maintenance. His greatly increases the number of individual tasks for which an error could occur. Causal factor distribution during the fourth quarter indicates that no one causal factor has a significant number of CRs. The seventy (70) CRs are. distributed across fifty one (51) different causal factors. The Procedural Compliance PI increased in October and decreased in November and December. Again, the increase was contributed to the increase in the number of personnel and number of tasks completed dunng the Refueling Outage. The Quarterly Maintenan Conference meetings will continue to emphasis the importance of procedural compliance. In addition, the Maintenance Training Advisory Committee is scheduled to consider several issues related to training that could improve procedural compliance. He numberof CRs assigned to the maintenance functional area during the fourth quarter are comparable to the numbers assigned in the first and second quarters considenng that the number of individual tasks for which an enur could occur greatly incmased during the fourth quarter due to the 1996 Refueling Outage. Performance in this area continues to be rated as GOOD. S1GNIFICANT CAD SUMM ARY: Here were no significant CADS assigned to the Mamtenance Functional area during this quarter. Performance for this quarter will be rated as GOOD. PERFORM ANCE INDICATORS: He calculated performance indicator index for this section indicates a rating of FAIR. The following table summarizes the performance indicators for the third quarter: l i l FCSTrend Report 4th Quarter of 1996 Page 29

MAINTENANG PEJOR, .NCEINDICATOR INDEX PEFFORMANCEINDICATOR h1EASUR %T htEASUR %T h1EASURE WT h!EASURE WT ALWAL INVEX E E VALUE VALUE HPSI Safdy Sptem Ibfinnarse (three nu oth value) .0011 .05 .004 .04 .005 .025 006 .015 .0001 .05 Auxiliary ftsslwater Sptem ibftwuunce(thne .0021 .05 .01 .04 .0125 .025 .0150 .015 .0017 .05 nuoth vatte) Emergency ACIbwerSptem Petfinmaixe(thne .0035 .05 .024 .04 .030 .025 .036 .015 .0036 .04 nunth valta) Overtive Oi>Line Gxtnil Rinnn Defitienies (ttmv <5 .05 <8 .04 <10 .025 <12 .015 8 .025 nundimvnyy) Gotn>l R u nn Def nitnies Completal By Targd >85% .05 >80% .04 >75% .025 >7t% .015 57.89 % 0 Date(thne nunth ralte) hlainten.nv Pemnn4 Relatal LERs None .10 1 .08 2 .04 >2 .015 Nme .10 htaintenance Pemunel Related NRC Viottims Niurlast 2 .10 0 .08 1 .04 >1 .015 Nnnein 3nior .10 quarters 4thquarter Genxtiw htaittenne Datikwg (thrw nwinth <350 .10 <400 .08 <450 .05 <500 .03 441.67 .05 awrage) Ownbe Prewntiw hiaintmne (time nuoth <1% .05 <2% .04 <3% .03 <5% .01 .97% .05 mwage) h taintenanw Rem nk (three nu oth average) <3% .05 <4.5% .04 <6% .025 <7.5% .01 1.1% .05 Ontne hiaintrinixv Owrtime (thrw nu oth awrage) <5% .05 <7% .04 <10% .025 <12% .01 26 % 0 Sdniule Perfinnanx (thnr n a nth mvrage) >85% .10 >80% .08 >70% .05 >65% .03 65 6 % .03 O nwtive htaintenance Gnnpletal By Target Date >80% .1 >75% .08 >70% .05 >65% .03 33 % 0 (thne nwwth merage) TOTAL 0.545 FCSTrend Report 4th Quarter of 1996 Page 30

MATERIAL S. J. Willrett 4Q94 IQ95 2Q95 3Q95 4QTR SELF QV EXT 4Q95 IQ96 e' 1996 CADS SIG PI CONCLUSION in the first quarter the performance in the material area was dassified as goal due to a decmase in the raw number of Condition Reports (CRs). Second quarter data indicates that pedormance has continued to improve in terms of number of CRs. The fourth quarter induded a refueling outage and a six (6) week Quality Assurance (QA) audit. Weaknesses in the materials work process, especially the Procumment Enginwring ama, were identified. Themfore, Qhe materials area is dassified as " fair" for the fourth quarter. y SELF-ASSESSMENT RESULTS: Self-Assessments in the fami of the following were conducted: Monthly WarehouseInspections Weekly PMO on Material Storage Area Walkdowns MaterialNon-Conformance Program Material Discrepancy Notice / Report Program MonthlyInventory PhysicalCounts MaterialReturn Process . l Shelf-Life Program, List of Expired items PerfomunceIndicators j l SOG6,(Housekeeping) Inspections l l I The self assessment programs worked to prevent further significant problems in the plant from occurring. l FCTrend Report 4th Quarter of19% i Page 31 l l l

 .-       .. =-             .-      . _ . _ - _ _ . _ -         .              ..     .          . .     . - _ . - .

i i 2 OUALITY VERIFICATION RESULTS: One (1) audit on material identification / control and 4 procurement control was conducted during the quarter. The audit concluded that there are a number of deficiencies in the material / procurement control program. Ten (10) CRs were wrHtm , as a result of the audit. ' EXTERNAL ASSESSNIENT RESULTS: Therewerenoexternalassessmentsof theMaterials area during the fourth quarter. 2 CADS: A total of nineteen (19) CRs were written during the first quarter associated with the

    ' Materials area. This is an increase of twelve (12) from the previous quarter. The CRs were generated from a variety of areas within the material functional area. Most of the problems are in the Procurement Engineering area. Those problems included issues of receipt inspection,                        !

material specification, CG Dedication enors and material discrepancies. These problems will all be addressed through the CR System and the PMT formed to improve work processes in  ; Procumment Engiraring. I 1 SIGNIF1 CANT C AD SUh15I ARY: No significant CADS were issued in tids area dunng the first quarter. l l P ERFOR A1 ANCE INDICATORS: The Spare Parts Inventory Value performance indicator showed a slight decrease from the previous quarter. This is primarily due to not ordering up to the maximum level and reductions reahzed through the investment mcovery program. The spare parts inventory remains at an acceptable level. The range of other Material Area performece indicators show no adverse trends. This shows that goals am being approached, met, or exceeded. There were no new performance indicators created during the fourth quarter. l l FCTrend Report 4th Quarter of19% Page 32

QUALITY C. J. Brunnert , em um IQ95 2Q95 3Q95 4QrR SELI l QV l EXT l 4095 lIQ96 2Q96 3Q95 1996 CADS em PI CONCLUSION in the thin:1 qturter, the prfomaance in the Quality area nrnamed good and impmwd significantly in most amas. Ewn 4 though the performance indicators and CADS declined during the fourth quarter, the dedine was small compamd to , incmase in activities during the refueling outage. While the JUMA audit had several im}mrtant findings and amas for inyrownut it also idertified sewral significant strengths. The continued lack of significant CRs and high level of self-idstfiedCRs rmnimd at last quuter lewis. Thus the performance of the quality functional area for the fourth quarter of 1996 has dedined some from last quarter but remains "gtxxi." j SELF-ASSESSMENT RESULTS: Noself-assessmentswereconductedintheQuality functional area during the fourth quarter of 1996. Seven (7) of the thtrty-eight (38), fourth quarter Condition Reports (CRs) assigned to the Quality Functional Area were identified by regulatory or audit groups while sixty-three percent (24 of 38) were self-identified. OUALITY VERIFICATION RESULTS: The 1996 JUMA audit was conducted during the fourth quarter of 1996. No other audits or surveillances were conducted during the quarter in the Quality functional area. 'Ihe JUMA audit assessed Quality Assurance, Quality Control, NSRG, Corrective Action and Receipt Inspection. Four (4) findings were identified in the areas of audit scheduling, corrective action response and QC inspection qualifications. Several strengths were also identified in the Quality functional area induding: performance of Root Cause analycis by NSRG; problem owr ership, senior management support and CRG meetings in the corrective action area; use of technical experts on QA audits; and QC participation in meetings to enhance support of plant activities. Finally the JUMA audit identified opportunities to improve in all areas of thequalityfuretionalarea. EXTERNAL ASSESSMENT RESULTS: No NRCinspections were conducted in the fourth quarter that specifically assessed the quality functional area. One resident monthly inspection did identify an unresolved item regarding the failure of the PRC to review eighteen (18) CRs for Operability per Standing Order R-2. This item was also identified in CR 199601542. FCTrend Report 4th Quarter of 1996 Page 33

l l 1 C A Ds: Thirty-eight (38) CRs were assigned to the Quality functional area during the fourth quarter which is over twice the fifteen (15) CRs assigned last quarter. Nearly thirty percent (30%) of these (11 of 38) continue to be from overdue CR responses or corrective actions. This I was also a finding on the JUMA audit. Another fifteen percent (15%) of the CRs were to track l completion of old incident reports rather than problems in the quality functional area per se. The l majority of the remaining CRs were related to QC inspections and plant outage activities. No I noteworthy trends regarding cause codes or group codes were identified. Although the increase in CADS indicates declining performance, it is commensurate with the increased activity of the outage. Much of this increase (especially in overdue CRs, QC inspection and holdpoints missed ) I is outage-related. l SIGNIFIC ANT CAD SUhlNI ARY: No significant CADS were assigned to the Quality I functional area during the fourth quarter of 1996 which is the sixth consecutive quarter without a significant CAD. ' PERFO Rhf ANCE INDICATORS: None of the five (5) Performance Indicators (PIs) for the Quality functional area wem classified as an adverse trend, a positive trend or needs increased management attention during the fourth quarter of 1996 (based on September-November 1996 PIs). The Significant Events P1 showed no change over the last three (3) PI reports. In the same petiod, the Preventable / Personnel Error LERs P1and LER Root Cause Breakdown PI showed large jumps in October while the Violation Trend PI degraded slightly early in the quarter but is now showing some improvement. Finally the Condition Reports By Level Pl showed a gradual increase in total number of open CRs and open Irvel 1, 2 & 4 CRs. Even though this represents a decline in performance, it is consistent with the incmased acFvity levels during the 1996 Refueling Outage and still is below the threshold of needing management attention. FC Trend Report  ; 4th Quarter of 19% Page 34 l l

l l RADIATION PROTECTION S. W. Gebers l 1 4Q94 IQ95 2Q95 3Q95 4QTR SEtJ V EXT l 4Q95 IQ96 2Q96 3Q)6 1996 CADS CONCLUSIONS I In the third quarter the Radiation Protection functional ama was dassified as good because of continued good perfomiance as indicated by assessments and audits. Results from self assessments, extemal surveillance, and NRC insp&tions in the fourth quarter show a stable tmnd in this functional ama. However, the poor perfomiance in i contmiling radiation exposure and contanunation suggests perfomaance in the fourth quarter was fair. Therefore, the performan in the Radiation Protection Functional area is " Fair". j SELF-ASSESSNIENT RESULTS: One Radiation Protection Self-Assessment per procedure RP-309 was perfonned in the fourth quarter.This assessment targeted mview of ALARA packages, primarily from the 1996 Refueling Outage. The assessment found that each package reviewed was complete and had the mquired information. However, improvement in consistency of format and documentation was noted; and a dear method for implementing lessons leamed l needs to beaddressed. OUALITY VERIFICATION RESULTS: Four (4) Quality Assurance Surveillance reviews were performed in the forth quarter. Surveillance (B5-96-1) focused on radwaste refueling activities, no condition reports or gds were issued. Surveillance (H5-96-1) reviewed the Respiratory Protection Pmgram. One Condition Report (CR-199601181) was issued. Fit testing of a newly trained fire watch was not performed according to procedure RP-509. Surveillance (H6-96-2) " Radiological Instruments," focused on radiological survey instrumentation preparations for the 1996 Refueling Outage. No condition reports or CIDs were issued. The last surveillance of 1996 (H14-96-1) was a comprehensive evaluation of radiation protection refueling activities. This surveillance included review of high radiation exposure jobs, contamination control, and dose control. No condition reports or gds were issued, but several recommendations for improvement were pmvided. EXTERNAL ASSESSSIENT RESULTS: One NRCinspection " Radiation Protection - Outage" was performed dunng the 1996 Refueling Outage. No violations or open items were issued during this inspection. Dunng NRC resident rounds, one violation of failure to follow procedure was issued in the forth quarter. A contaminated individual was not logged into the Personnel Contamuation tog FC-RP-207-2, as mquired by procedure RP-207. However, no l response was required as the NRC concluded that immediate corrective actions would prevent j reoccurrence. l FC Trend Report 4th Quarter of19% Page 35

i l C A Ds: Rere were seventeen (17) Condition Reports (CRs) initiated during the forth quarter, an upward trend from the third quarter. Three of the seventeen were identified by groups outside the Radiation Protection or Corporate Health Physics departments. 4

CR Identification Performance 4

l # ! 14-l 12-

                 ,,      [                                                                              External i                C:                                                                                       Audits OPPD Audits 2'                                                            Other 0                                                          Departrnents 1st        2nd              3rd   4th   Radiation Otr         Otr             Otr   Otr   Protection lERadiation Protection 30ther Departments DOPPD Audits OExternal Audits l                                 j Fifty-three percent (53%) of the CRs identified in the forth quarter were human performance related, as compared to seventymne dunng the third quarter. This mduction in human performance weaknesses is considered an impmving trend. One of the conditions reported was identified byQuality Assurance.

SIGNIFICANT CAD

SUMMARY

None. )

PERFORM ANCE INDICATORS: The Radiation Protection functioral area lus seven performance indicators used as a pmdictor of general perfomTance. 2

1. Clean Controlled Areas Personnel Contanunation's > 1,000 dpm /100 cm . There has been l a total cf eighty contamination events thmugh the end of December. ,

l FC Trend Report 4th Quarter of 1996 Page 36

l l l

2.  :

Maximum Individual Radiation Exposure < 1,500 mrem. The maximum indiddual exposure goal for this year is 1,500 mrem. Through the end of December, one individual accumulated 1,415 mrem, which is the highest individual exposure for the year. 'This exposure is less than the annual goal. l

3. Volume of low-level Solid Radioactive Waste < 1,050 ft 3. As of the end of December '

eight hundred eighty-four cubic feet of waste is in final fonn. This volume is 84.2% of the goal. This volume indudes six hundred fifty-seven cubic feet buried and two hundred twenty-seven cubic feet stored in final form. 'Ihis indicator shows the vohnue of radioactive waste buried from Fort Calhoun is less than the industry upper ten percent (10%).

4. Contaminated Radiologically Controlled Areas < 95% of the surface area. At the end of December 8.4% of the RCA was maintained contaminated. This value is less than the annualgoal.

5. Poor Radiological Work Practices < 15 for the year. At the end of December six Poor Radiological Work Practices have been documented. This number of poor practices is less than the annualgoal.

6. On-line collective radiation exposure less than 19 rem, and total exposure less than 138 reminduding outages.

The year-to-date collective exposure (TLD) through the end of December was 226 rem. This total exposure is greater than the annual goal. Despite several initiates put into place to reduce collective radiation exposure, the goal was exceeded. A plant team to review the maintenance schedule, frequency of Prevent Maintenance (PM) work, and plant operations was implemented to routine reduce radiation exposure. A PRC subcommittee implemented many source temi reduction tediniques in an effort to reduce radiation exposure to plant personnel and the environment. A Radiological Safety HIT addressed methods to reduce individual radiation exposure, increase worker efficiency, and ectucate the work force on radiological conditions from fuel defects. 1 I FC Trend Report 4th Quarter of 19% j Page 37 l

l Collective Radiation Exposure Self Reading Dosimeters Only (ALNOR) 120 - 250 100 - g 200 g g, 80 - 3 O 60 -

                                                                                   ~

40 - 20 -

• 0 0 50 h o e

*1 _ _ _- ,

_o a s a s a a a s a a s a 4

              ,4   4       s    i       k    ,i     s,     #   6     6    $    $

m s < m 4 o z o l M Monthly -*-Total l The current industry besM:artile is 145 person-rem per year. The yearly average for Fort Calhoun Station for the three years from October 1993 through September 19% is 116.78 , person-rem per year.

7. Contml the number of unplanned personnel contamination's resulting in skin exposure of 20 mrem or greater to less than thirty-nine. At the end of December there wer2116 unplanned personnel contamination's resulting in skin exposure of 20 mrad or greater.

This number of contamination events is 297% of the goal. The perfonnance index for the Radiation Protection Functional Area is 0.67 showing poor performance in this area. The indicator showed a 14 point decmase from third quarter perfonnance. Fuel performance significantly affected our exposum and contamination control efforts. Exceedmg these goals resulted in performance index of poor for 1996. The performance index is calculated by applying a set of weights to each performance indicator and comparing the sum of the weights to a performance criterion. A performance index of 0.9 or higher is considered excellent performance, an indicator of 0.9 to 0.8 is considered good,0.8 to 0.7 is considered fair and below 0.7 is considered poor. The performance index is calculated using the following table, by companng the present goal performance with the weights in the table. FC Trend Report 4th Quarter of 19% Page 38

____.m _ . _ _ ._. . . . . ._ . _ _ _ _ . ___ Crtterla IMeasure WeiQht Measure iWeiQht Measure i Wei@t i Measwe WeiQht Collective Dose < INPO 10% 0.40 < 10% Goal 0.36 < Goal 0.30 > Goal 0.20 Contaminated Area < 10% Goal 0.10 < 5% Goal 0.08 < Goal 0.05 > Goal 0.02 Volume LLW < INPO 10% 0.20 < 10% Goal 0.15 <(bal 0.10 > Goal 0.05 I Unplanned Contamination < 10% Goal 0.05 < 5% Goal 0.03 < Goal 0.02 > Goal 0.01 I CCA > 1.000 dom < 25 0.05 < 35 0.03 < 55 0.00 > 55 0.01 PRWP < 10% Goal 0.10 < 5%Cbal 0.08 < Goal 0.06 > Goal 0.03 Max. Individual Exposure 0 > 1.500 0.10 1 = 1.500 0.08 2 > 1.500 0.05 3 > 1.500 0.02 Total t l 1.00 0.811 0.58i 1 034 i l l l l I l l i l i l 1 1 l l 1 FC Trend Report 4th Quarter of19% Page 39

l SECURITY H. J. Sefick 4Q94 IQ95 2Q95 l 3Q95 4QTR 3 a QV EXT 4Q95 IQ96 mem 3Q96 1996 CADS PI CONCLUSION In the fourth quarter, the performance in the huity area was dassified as good performance generally meeting 4 management expatations. The huity Department's self-assessment and observation programs continue to enhance the overall effectiveness of the security program. Significant is the fact that of the 659 task perfomunce appraisals conducted during the past six month pericxt none requimd management follow-up action. Performance indicators continued an upward trend (51%) during this quarter. Management attenhon has been incmased in this ama. htrity survey results continue to suggest that plant personnel am satisfied with security performance. Based on the above assessment, the huity functional ama's perfomunce is rated as " Good". j SELF-ASSESSNIENT RESULTS: During the fourth quarter of 1996, thtrty-one (31)intemal security surveillances were performed. Of the sun eillances conducted, only one required minor  ! administr ative follow-up. Security personnel conducted 210 task performance appraisals, l continuing to enhance individual security officers' skills and job knowledge. Dunng the past six l month period the Security Department conducted a total of 659 task performance appraisals. None of the task performance appraisals required follow-up action. AnintemalSecurity Department customer service survey was again conducted during the fourth quarter of 1996. The i sun ey covered such areas as personnel safety, professionalism, motivation, and security support. l The general results of the stuvey revealed 92 percent of the respondents, up from 84 percent last l quarter, were satisfied with the support provided by the Security Department. OU ALITY VERIFIC ATION RESULTS: There were no Security Department Quality Assurance (QA) audits dunng the fourth quarter of 1996. EXTERNAL ASSESSN1ENT RESULTS: There were no extemalassessments conducted during the fourth quarter of1996. CADS: Six (6) CADS were issued to the security functional area during the fourth quarter of l 1996. Five (5) of the six (6) Condition Reports (CRs) (83%) were self-identified. Two (2) of i the CADS issued involved the incmased trend in the number of security personnel errors. This will continue to be the focus of attention by Security management personnel, not only during the first quarter,but throughout 1997. SIGNIFIC ANT C AD SUNINT ARY: No significant CADS were issued in the security functional area dunng this quarter. PERFORNI ANCE INDIC ATORS: Dunng the fourth quarter of 1996, there were eighty-thme FC Trend Report 4th Quarter of19% Page 40

i l l (83) loggable security incidents. This resulted in a 51 percent increase from the third quarter of 1996. During the 1996 Outage, an increase in non-system failures, primarily in security force errors (13), and system failures caused this negative tmnd. Although none of the perfomTance indicators were classified as an adverse trend, the results above have required increased management attention. Meetings were held during the fourth quarter with all security force personnel. The focus of these meetings were to identify ways that security personnel could pmvent recurrence of the errors and to identify new areas for improving effectiveness and I efficiency. Security Performance Indicators are rated as fair requiring further management attention. l i i l l l l 1 FCTrend Report 4th Quarter of 19% Page 41 l l

l l SPECIAL PROCESSES R. L. Wylie o CONCLUSION The Special Process functional ama perfomiance continued with strong, positive performance for the fourth quarter. This assessment by the Special Pmcesses PEP team was based on a continued decmase in the number of CAD's issued to this functional area relative to the significant incmase in work activity due to the Refueling Outage this quarter, which is attributed to effective conective actions in the welding area that wew implemented prior to the start of the RFO. Also the extremely low weld repair rate continued on a positive trend for this wporting period. Therefore, the overall assessment of perfomiance for the Speaal Processes functional ama continues to te { Excellent" for the fourth quarter of 1996. y SELF-ASSESSMENT RESULTS: The Speaal Processes PEP team concluded that there were no adverse trends requiring special attention in the fourth quarter. Several process enhancements were identified in the welding ama which will be implemented in 1997. 'Ihese indude changing the Welder Roster format to be more usec friendly, placing accountability on the welder for ensunng weld processes are correct for the applicationhe / she is working on, and revising weld definitions for additional controls on speofic weld applications. OUALITY VERIFICATION RESULTS: There were noQuality Assurance (QA) audits I surveillances speafic to Speaal Processes during the fourth quarter. Three (3) weekly surveillances and a scheduled QA audit were reviewed for any concems related to Special Prcresses and none were noted. There were no NSRG evaluations conducted during the fourth quarter of 19%. EXTERNAL ASSESSMENT RESULTS: There were no external assessments of theSpecial Processes functional area during the fourth quarter of 1996. CADS: During the fourth quarter, four (4) Condition Reports (CRs) were assigned to the Special Processes functional area. One of the CRs was assigned to this functional area which was i assessed as a maintenance plannmg issue with no impact on the Welding Program. This is a l reduction of two (2) CRs from the third quarter. The assessment of the three remaining CRs all related to the welding / NDE area, confinned that two (2) CRs were related to procedure non-compliances due to lack of attention to detail with both being repeat issues from earlier quarters in rod control and welder qualifications. The cause of the third CR, related to incorrect information compounding a welding / NDE issue, is still being assessed under action items for final l I deternunationofthecause. FC Trend Report 4th Quarter of1996 , Page 42 1

The " CADS" performance area will remain " good" for the fourth quarter based on the decrease of two (2) CRs from the previous quarter and the repeat issues from earlier quarters to verify that corrective actions are effective. SIG NI F1C A NT C AD SU M M A RY: No significant CADS were issued in this functional area in the fourth quarter of1996. PERFORM ANCE INDIC ATORS: The Weld Repair Indicator continues with an excellent repair rate of less than one percent in the fourth quarter. This is especially good performance considering the significant increase in controlled welds performed due to the 1996 Refueling Outage. 9 l

                                                                                                                                          )

FCTrend Report 4th Quarter of19% Page 43 j

TRAINING R. G. Conner 4@4 IQ95 2Q95 4QTR SElf QV EXT

 ;1 " -

_. 4Q95 IQ96 _ ie. 19 % i 3 PI CONCLUSION Continued efforts are still needed to complete conective actions outlined in the master action plart Self assessments am effectively being used to ensure the training progmms meet the ACAD regturements. Training is rated " Good" with an improving tamd during the third quarter j SELF-ASSESSMENT RESULTS: herewasoneself-assessmentperformedinOctoberof 1996 in the non-accmdited traming programs admmistrated under the direction of the Maintenance Training Supervisor, (Quality Assuran (QA) / Quality Control (QC), Procurement and Warehouse, Planner and Schedulers, and Safety Training programs). Weaknesses found in these areas are as follows:

1. Lack of QA / QC management oversight of the QA I QC training pogram.

2. Lack of management direction for the OPPD and FCS safety training programs  !

3. Training pmgram ownership and TPMP maintenan hasbeennunimal.ne program and associated TPMP have not been updated for approximately four (4) years.

The following strengths were noted:

1. He content of the QA IQC training programs was developed with the intent of beconung INPO accredited. The programs can be administrated like accrc<lited pitgrams.

2. He QA / QC lab is fully equipped with NDE inspection equipment, test pieces and mock-ups, used to certify OPPD as well as contract QC inspectors.

3. The QA / QC Instructor mamtains Level 111 certificates in mechanical, electrical, visual weld, and VT 1-3, as well as an AWS Certified Welding Inspector (CWI) and Certified Welding Educator (CWE).

4. Good resources to support safety trauung at FCS.

i

5. A full time qualified safety trauung instructor is available at FG l

i FC Trend Report 4th Quarter of1996 Page 44 i 1

6. The FCS safety instructor as well as the Industrial Safety Coordinator is OSHA certified as an instructor.

This category is assigned a performance level of fair with an improving trend. OUALITY VERIFlCATION RESULTS: NonewsignificantitemswereidentiDedasa result of intemal audits or inspections performed during the fourth quarter. This category is assigned a performance level of good with an improving trend. EXTERN AL ASSESShf ENT R ESULTS: An NRC inspection was conducted on November 11-22,1996. The purpose of this inspection was primarily to evaluate effectiveness of the licensed operator requalification program. The inspectors conduded that the operations training < staff developed, administemd. and evaluated the requalification examinations professionally and , accurately. This category is assigned a performance level of good with an impnwing trend. CADS: Two (2) Condition Reports (CRs) were identified by the cognizant work group. This is a decrease of one from the previous quarter (13 for the second quarter and 3 for the third quarter). Of particular interest is CR199601603, which identifies a problem with the retention of training records. This is a significant issue requiring Information Technology (IT) support to correct. This category is assigned a performance level of excellent with a neutral trend. SIGNIFICANT CAD SUMM ARY: No significant CADS were issued in this ama dunng the quarter. No significant CADS have been issued for 1996. This category is assigned a pedormance level of excellent with a neutral trend. P ERFOR M ANC E INDIC ATORS: The Training Performance Indicators are published in the monthly Training Report. Line management obsen ations improved during the third quarter and have been effective. This category is assigned a performance level of good with a neutral trend. FC Trend Report 4th Quarter of 19% Page 45

' l i i l i i i i I e 4 i 1 f 4 APPENDIX A 1 i. !, HUMAN PERFORMANCE TRENDS i i.: d

)
.a 1

4 i i I T 4 8 d

4 APPENDIX A - HUMAN PERFORMANCE TRENDS Human Performance Trend Categories Human Performance trending and monitoring is provided through the collection and analysis of Condition i Report Cause Codes assigned to each Condition Report (CR). These Cause Codes are assigned by the 3 CAG based upon the cause determination provided by CR Owners. The CR Cause Codes provide a means of characterizing the type or nature of causal factors which have influenced or directly contribute to the creation of the situation being reported under the CR system. The categories of human performance causal factors described below are designed to provide a means of

,       assessing human performance using the cause code information provided by the Condition Report i        System. These categories are based, in part, upon those described within the INPO HPES Coordinator Manual.

4 COMMUNICATIONS Causal Codes: CMW, CMXA-C, MSWB , This category is defined to include all causal factors which indicate that a communication error contributed to the event being reported. Errors would include both misunderstood verbal communications as well as errors in the timing or quality / completeness of the communication which is at fault. PROCEDURES Causal Codes: PRYA-D, PRYF-G,1,J, PRWA, DSXF, DSXG This category is defined to include all causal factors which indicate that an approved procedure or other document which carries the intent of procedural adherence is in error. The types of errors covered would include: unclear guidance, wrong step sequences, omitted steps, typographical errors, and simply, not having a procedure. The inadequacy of the content of a procedure due to poor technical review is not covered under this category since the emphasis here is on how the document, right or wrong, contributed to an event as a human performance factor. PROCEDURE USAGE / ADHERENCE Causal codes: PRWB-D, PRX, PRYE This category includes those factors which resulted in a failure by personnel to follow approved procedural guidance. These failures are not necessarily due to individuals failure to use a given procedure, but could also include faibres due to confusing formats, incorrect procedural usage level, or a failure of the procedure in r,Ilowing for the ability to self-check. Errors associated with the technical content or intent of a procedure would not be included here, but rather under the category of " Assessment Methods", WORK PRACTICE Causal codes: PEW, PEX, MSWG This category includes those factors which typically contribute to a lack of attention to detail or an inconsistency in work practice standards. Carelessness, over-reliance on favorite indications, inadequate preparation on the part of the worker, or slips in the application of a skill which is not covered by a procedural requirement are some of the items covered under this category. Failures to self-check / verify are also an input to this category. A-1

APPENDIX A - HUMAN PERFORMANCE TRENDS WORK ORGANIZATIONIPLANNING Causal codes: ISW, CMXD This category includes factors related to the preparation, planning, and scheduling of work. l SUPERVISORY METHODS Causal codes: ISX, MSWE, MSWF This category reflects situations where either "no supervision

• or
• inadequate supervision during the performance of a work assignment contributed to errors. In addition, situations where inadequate enforcement or accountability of management policies, standards, or administrative controls are included .

under this category. l MANAGERIAL METHODS Causal codes: MSWA, MSWB-D, MSWH This category includes those factors which are associated with the Standards, Policies, and Administrative Controls (SPAC) provided by management. The impact of SPACs on human error is typically due to either the lack of individuals having this guidance or the guidance not being clear or up-to-date. 1 CHANGE MANAGEMENT Causal codes: MSY, RSXB-C, RSWC This category includes causal factors which are associated with corrective action identification, development, or implementation as well as the failure to adequately apply intemal and extemal operating j experience. ' ASSESSMENT METHODS Causal codes: MSX, MSZA, DSWC, DSWD, DSXA-E, l HFYA l This category includes factors which indicate a failure has occurred in the areas of technical reviews and l assessments. These reviews could be design reviews, inadequate testing or procedural reviews. l TRAINING / QUALIFICATION Causal codes: TRW, TRX, PEX This category covers those aspects of an event which are found to have resulted from errors made in the i training or qualification process for individuals. ' INTERFACE DESIGN Causal codes: HFW, HFY, HFZ This category includes causal factors associated with the situations where the interface between personnel and equipment / systems has resulted in personnel errors. Examples would include poor human factoring in the design of labels, layouts, or processes which by design are overty complex, or require complex manipulations. Interfaces of personnel with processes or systems which do not allow for the A-2

APPENDIX A - HUMAN PERFORMANCE TRENDS ability to detect or recover from a fault condition are included in this category. , ENVIRONMENTAL EFFECTS Causal codes: HFX This category includes work environment factors such as noise level, lighting, and temperature effects which contribute to personnel error. t i 1 l I A-3 l

APPENDIX A - HUMAN PERFORMANCE TRENDS QUARTERLY CAUSE CODE DATA BY CATEGORY (TABLE 1)

    ~NE5                                            95 1 95-2 95 3 95-4 TOTAL 96 1 96-2 96 3 l96-4 TOTAL COMMLNCATimS                                        5     3   5      27       40        26      11     9    13          59 PROCEDWES                                          14   19   15      34       82        42    46      53    56         195 PROCEDWE WAGE /ADFERATCE                           39   50   47      21      157        21     27     18    18          84 WORK PRACTICE                                      28   23   31     125      207 "141        123      79    94(        43Q VORK ORGANZATION/Pl#NtG                             7   10    2      15       34        18    26      18    30          92 SlttMVISORY METHODS                                 9    3    1       7       20        10     15     11    16          52 MAfMGERIAL METFDDS                                 11   12   10      63       96        40    37      29    34         140 Ct%NGE MAfMGEMENT                                   3    8    6       6       23        19     14     15    13          61 ASSESSMENT METFDDS                                 29   17   15      39      100        49    41      43    33         166 TRAINPG/QlMUFICATION                                4    2    6       9       21          4    11      7     7          29 INIt:to ACE DESIGN                                  8    1    3       6       18          6      7    11    19          43 ENAROtNENTAL ttet: CTS                              2    5    0       1        8          1      1     0     4          7 QU4 tit:r<LY TOTALS                              159   153  141    353       806       377  359     293    337        1364 300 - -                                                         300 i
                        ~                                                                                                      l 250                                                             250 200
              &         _            m     @-

200 150 -- m 150 100 - f f '( 100 -- 7-

                        ~      #

Q d --- 7 W

50 -l i

g"g#' 50 --"

                                                                            $M       s 0-         -           -                  -

0 i i 4 95-4 96-1 96-2 96-3 96-4 95-4 96-1 96-2 96-3 96-4 0 co mcanoa D w=esumemoe D mocanvan 0 cames ==esummT O accansnaas O useu=rmmemos O wounnexe 0 mmm ous, manos a oa ano.wu= g sumvsoevern.co. m =maaoa E suvuommensemet DESCRIPTION: Data presented in the above table represents the total number of CRs or irs reported for l each of the last 6 quarters which were assigned cause codes that fell within one of the prescribed human ' performance categories listed. The condition report process respectively establishes 30 and 45 day response due dates for all level 3 and 4 condition reports. As a result, approximately 80 CRs from the 4th quarter require CR Owner response 4 before they can be coded for inclusion in the above chart (as of 1/15/97). A-4

APPENDIX A - HUMAN PERFORMANCE TRENDS 1 l l l QUARTERLY CAUSAL OCCURRENCE RATES l (TABLE 2) l l U TECORIE5 95-1 95 2 95-3 95-4 AVRG. 96-1 96-2 l 96-3 96-4 AVRG. COMMLNCATIOt6 0.03 0.02 0.04. 0.08 0.04 0.07 0.03 0.03 0.04 0.04 i PROCEDLRES 0.09 0.12 0.11 0.10 0.10 0.11 0.13 0.18 0.17 0.15 l PROCEDLRE WAGE /ADFERAfCE 0.25 0.33 0.33 0.06 0.24 0.06 0.08 0.06 0.05 0.06 4 VORK PRACTICE 0.18 0.15 0.2; 0.35 0.23 0.37 0.34 0.27 0.28 0.32 VORK ORGANZATIOffPLRNN3 0.04 0.07 0.01 0.04 0.04 0.05 0.07 0.06 0.09 0.07

 '5[FER91SORY METHDDS                                0.06 0.02 0.01   0.02     0.03  0.03 0.04    0.04   0.05    0.04 MAtAGERIAL METFODS                                 0.07 0.08 0.07   0.18     0.10  0.11 0.10    0.10   0.10    0.10 CHVJGE MANAGEMENT                                  0.02 0.05 0.04   0.02     0.03  0.05 0.04    0.05   0.04    0.04 ASSESSMENT METFODS                                  0.18 0.11 0.11   0.11     0.13  0.13 0.11    0.15   0.10    0.12 TRAINNG/%AUFICATION                                 0.03 0.01 0.04   0.03    0.03   0.01 0.03    0.02   0.02    0.02 INitHPAl I!2 SIGN                                  0.05 0.01 0.02   0.02    0.02   0.02 0.02    0.04   0.06    0.03 l

EPMROtNENTAL ttthCTS 0.01 0.03 0.00 0.00 0.01 0.00 0.00 0.00 0.01 0.00 l 0.7 - 0.8 0.8 - - - - - - 0.7 0.5 - - - - 0.6 0.4 --- hg- f> 0.4 u-- ..

                                                                                                                      )
                                                                     ,3__         _                       _        _
  "                                        -       ~
                                                                     "~

iG a ees

                                             ~

0- i sa i m mi i i 0 i i i i rv-95-4 96-1 96-2 96-3 96-4 95 4 96-1 96-2 96-3 96-4

                    =      ,

c 8 ==--' 8 = = ==. i 8 =.=="_. - 8 ==: u i DESCRIPTION: Due to the changes in reporting sensitivities and methods, as well as mxj s in the amount of work activity between one quarter and the next, a means of normalizing the inforNtion provided on the previous table is required. Data presented in the above table is based upon cividing the , number of cause codes for each category in a given quarter by the total number of cause codes assigned I to all categories for that quarter. This information represents the percentage of all human performance errors which are attributed to a particular category of influence. AdditionaHy, since the number of cause codes assigned is a function of the number of CRs and irs which are created, and the number of CRs and irs created are likely to be proportional to the amount of work activity being performed in the plant, these percentages represent a rate of occurrence. A-5

l APPENDIX A - HUMAN PERFORMANCE TRENDS QUARTERLY HUMAN PERFORMANCE TREND i (TABLE 3) CATEGORIES 95 1 95-2 95-3 95-4 Average 96 1 96-2 96 3 96-4 Average COMML.NCATIOf6 -0.24 0.21 -0.90 -3.45 -1.10 -1.14 0.72 0.96 0.54 0.27 PROCEDWES 0.04 -0.21 -0.62 0.27 -0.13 -0.49 -1.60 -5.29 1.00 2.10 PROCEDL.RE LEA 3E/ADFERATCE -3.43 -2.50 -1.27 2.28 1.23 1.45 0.75 0.51 1.12 0.96

                                                                                        ~

MORK PRACTICE 1.10 1.48 1.87 -5.96 -1 32 -1.64 -0.63 0.76 1.23 -0.07 MORK ORGANZATIOffPLAtNtG 0.96 04; 1.67 -0.20 0.60 -0.30 -1.41 -0.72 2.44 1.22 St.atHVISORY METFODS -0.16 1.73 1.76 0.39 0.93 -0.03 -2.90 -0.95 -1.60 1.37 MAN 0GERIAL METFODS 0.92 0.57 0.61 -6.96 -1.22 0.13 0.11 0.34 0.55 0.22 CFMNGE MANAGEMENT 1.11 -0.30 0.35 1.19 0.59 -1.01 0.10 -0.98 0.05 0.46 ASSESSMENT METFDDS 0.04 0.90 0.86 0.69 0.62 -0.07 0.03 -3.05 1.65 -0.36 , TRAINtG/QL.MLIFICATION -0.57 2.29 -3.03 0.15 -0.29 1.32 -0.53 0.26 0.22 0.32 l INI t.to. ACE DESIGN -5.42 0.63 -0.12 0.11 -1.20 0.42 -0.70 7.89 3.34 2.88 ENARONVIENTAL t-ttt-GTS {

                                                      -0.58 -4.70     0.73 0.55        1.00    0.63 0.44        1.50     7r10 -1.13  '
 %ElGhit:u AVERAGE                                     7.20 -1.35 -0.58 -5.40        -3.63    -1.73 -0.82       0.13    0.23  -0.56 7 5 3 1 o1 2 3
                                      , , , , ,,,,,,                             Q u arte r Total     T re n d      T re n d Communication -       3
                                            ' '          I f,E                     13        59       0 54          0 96 Procedures -     i         l  l 3      E l          l 56       195      .l.00           5 29 Procedure Usage -       4

• i l I A E i is 84 8 12 0 $1 l Work Practice - l 94 437 8 23 0.76 Work Organization - I I *
• i
• i 30 92 2 44 0.72

    ,f Supervisory Methods -             ,               f M,                       16        52        1 60         0 95 fm Managerial Methods -          l I        *El         '           34       140       0 55          0 34             l Change Management -                                                        13       61        0 05          0.9s f                         fl Assessment Methods -         1  ,

i1 6 t M; 33 166 1 65 3 05 Training / Qualification - ' I 7 29 0 22 0 26 l fl interface Dealgn - ] g [ g M i 19 43 3.34 7 89 ' Environmental Effecta - M . 4 6 7.10 1.50 Weighted Average - l ll lEf 337 1364 0 23 0.13 DESCRIPTION: The data presented in the above table represents how the number of CRs and irs for a given category (using the occurrence rates from the preceding table) compare to those of the previous icur quarters. This is accomplished for the current quarter by subtracting the occurrence rate assigned to ' a category, from the average occurrence rate assigned under the same category for the previous four quarters, and then dividing by the standard deviation for the previous four quarter average. The value obtained represents how the current quarter compares to the average of the previous four quarters in units of standard deviation. The sign (+/-) of the values reported in a given quarter indicate whether the quarterly data is showing a decline (-) or improvement (+) from the previous four quarters. Values <-1 or >1 would indicate tr.at the rate of occurrence for a given category is outside the 68 percentile range for the data being analyzed. A-6

APPENDIX A - HUMAN PERFORMANCE TRENDS 1 MONTHLY HUMAN PERFORMANCE INDICATOR  ! (TABLE 4) Human Performance ladicator Nov 95 Dec 95 Jan 96 f eb 96 Mar 96 Apr 96 May 96 Jun 96 Jul 96 Av896 Sep96 Oct 96 Nov 96 FHP Related CR1 bli 127 123 124 133 100 121 148 98 106 110 169 149 Hours Worked 48932 59991 49675 49776 58479 62535 49347 52885 55807 47139 58908 64844 65839 FCRs/100 Hours Worked 0.24 0.21 0.25 0.25 0.23 0.16 0.25 0.28 0.18 0.22 0.19 0.26 0.23 4 Month Moving Average 0.00 0.00 0.00 0.24 0.23 0.22 0.22 0.23 0.22 0.23 0.22 0.21 0.23 4 Month 5tanadard Deviation 0.00 0.00 0.00 0.02 0.02 0.04 0.04 0.04 0.05 0.04 0.04 0.03 0.03 l upper Error Level (+ 1s) 0.00 0.00 0.00 0.25 0.25 0.26 0.26 0.27 0.26 0.27 0.26 0.25 0.25 Lower Error Level ( Is) 0.00 0.00 0.00 0.22 0.22 0.18 0.18 0.18 0.17 0.19 0.18 0.18 0.20 0.300 0250. . -

                                ?'~___~~                                   ,..    -

j 0200 g ..-.. .

                                              '         * ^

g / '. . scRs/100 % Worked l M _._.. 4 uare % Aerm l 1 0.i50 . ..upon error Lew (+ t.) 8 .. Lower Error Leel(-1s) ii oc

   \t 0.100 -

0.050 0.000 , , s s s 3 3 3 s s s 8 a s s

            $      $      4    8      $      1          I mone
                                                                $        4      I       A     8      $

DESCRIPTION: Recent interest in providing a means of trending overall human performance improvements or declines has resulted in the development of the Monthly Human Performance Indicator shown above. This indicator is based on plotting the number of human performance related CRs divided by a measure of the amount of work being performed during a month. Currently, work hours are combined from Maintenance, Operations, Chemistry, Radiation Protection, and Security as these groups work hours tend to fluctuate dependent upon the overall amount of work being performed throughout the plant. The amount of work performed is assumed to reflect a measure of the amount of risk or potential for performance errors. A four month moving average and 1o standard deviation is plotted to assist in understanding how a given month compares to previous monthly trends. A-7

APPENDIX A - HUMAN PERFORMANCE TRENDS l

SUMMARY

ANALYSIS Data used to compile the trends described in this appendix are based primarily on the causal codes assigned to each condition report since the fourth quarter of 1995. Since the Condition Report process establishes a 20-45 day response time, complete data for the fourth quarter will not be available until February. General Analysis i WORK PRACTICE, PROCEDURES, ASSESSMENT METHODS, and MANAGERIAL METHODS remain  ! as the top four significant contributors to human performance errors with ORGANIZATION AND WORK i PLANNING showing a marked increase frem previous quarters (reference Table 1). In addition, the 337 ( cause codes assigned to all human peformance categories represents 73% of all cause codes assigned for the 4th quarter. Significant events during the fourth quarter which are related to these categories include: CR# 199601279 Loss of containment closure during fuel movement activities occurred twice l CR# 199601346 during a one week period of time. These failures to met Technical Specification requirements were considered to have resulted, in part, from inadequate . procedural guidance for controlling isolation points, and failure to follow work practice requirements during the performance of maintenance. CR# 199601272 A less than adequate review performed on the design of the insulation provided for testing of the pressurizer code safeties resulted in a test failure and the inability to prove that the lift set point was with in Technical Specification  ! requirements. CR# 199601476 A failure to adequately assess the effect of jogging reactor coolant pumps on RCS temperature resulted in a cooldown event in excess of Technical Specification requirements. CR# 199601649 Identified work practme failures assocsated with torquing and gasket replacement  : skills that resulted in a steam leak leading to a plant shutdown for repairs. l CR# 199601508 The replacement of a non-CQE gate valve with a globe valve was not discovered until after the job was completed. The failure to verify that the correct valve was being installed by several individuals was considered to be the primary cause of this event. Human Performance Category Analysis Communication: The number of CRs which directly, or indirectly resulted from errors in communication increased slightly from the third quarter, however, this increase is considered to fall within a normal range of statistical fluctuation and is not considered to represent a significant change. Procedures: An slight improvement in the number of CRs attributed to proceduralinadequacy is indicated by a review of Table 2 and 3. This category still remains as the second leading contributor to human performance error. A review of specific cause code data related to procedural inadequacies indicates that the significant contributors to this category are " situations A-8

APPENDIX A - HUMAN PERFORMANCE TRENDS l l which are not addressed by procedures", or" procedures which have unclear guidance". This trend remains negative for the fourth straight quarter. Procedure Usaae/ Adherence: An improvement in trend during the fourth quarter is indicated by Table 3. Considering the probable increase in the number of procedures utilized during the 1996 refueling outage, and the increase in "outside" support personnel, this may indicate that when procedural guidance is adequate, personnel are able to appropriately adhere to pr',cedures. l , it should be noted that specific cases of failure to follow procedures due to reasons other than those identified under this category are generally coded as being caused by a

• lack of attention to detail". These types of errors are coded under the work practice category.

Work Practice: This category remains as the largest single contributor to human performance errors at Fort Calhoun Station since the introduction of the condition report system. Tables 2 and 3 would indicate that an improving trend has occurred during the 3rd and 4th quarters of 1996. This trend may be primarily due to the dropping of the 1995 third quarter data from the four quarter comparison provided in Table 3. The actual number of work practice errors increased over the third quarter (Table 1). Two significant contnbutors to this category during the fourth quarter are ' Skill-based slips

• and " Failure to self-check / verify. i Work Oraanization and Plannina: The number of work organization and planning errors increased during the fourth quarter to the highest level (30 CRs) recorded thus far. The increase in work load associated with planning and scheduling the 1996 refueling outage was a likely factor in this change. This appears to be supported by the increase observed during the second quarter of 1996 when the plant entered an extended maintenance outage in response to the reactor coolant pump ARD failure. The primary contributors to this category are " inadequate Scheduling", end

    " Inadequate Job Briefings".

Suoervisorv Methods: The negative trend indicated by Table 3 has continued since the beginning of 1996. A review of cause code data indicates that " inadequate Supervision", " Inadequate Enforcenunt", and " inadequate Accountability" are the significant contributors to this category (15 of 16 CRs) during the fourth quarter. Manaaerial Methods: While this category remains as one of the top four contritsutors to human performance error (reference Table 1), no significant change in trend has been identified for this category. l I Chance Manaaement Based on a review of Tables 1,2 and 3, the number of conditions resulting from errors associated with change management appear to be stable. Assessment Methods: A review of Table 3 indicates that an improving trend may exist in this j category. A review of causal code data indicates that the change is due to a reduction in the number of CRs attributed a " Lack of Depth in Evaluation"(6 of 33 in the fourth quarter as compared to 21 of 43 in the previous quarter). Trainina and Qualification: No significant change in trend has been identified for this human performance category. Interface Desion: A review of Table 1 and 2 indicates that an increase in the number of interface design related conditions has occurred. The primary contributors to this change are events j resulting from man-machine interfaces such as " Inadequate Labels", " Monitoring Alertness", and from complex system interactions which result in

• Monitoring too many items simultaneously".

l l l A-9 l i l

l t APPENDlX A - HUMAN PERFORMANCE TRENDS 1 Environmental Effects: This category experienced a large increase (as indicated by Table 3) during this quarter. The total r. umber of environmentally induced human performance errors I during 1996 is currently six. flour of these CRs occurred during the fourth quarter. Two of these l condition were due to high ri.diation area environments which may become a greater factor in human error during refueling outages. Due to the low number of events, this trend may not be considered significant at th'.s tirne. 1 Monthly Human Performance Indicator This indicator was recently developed as a means of providing an overall assessment of human performance trends on a monthly basis. The monthly trend indicates that the months of June and October,1996 saw an increase in the number of human performance related errors beyond a level of assumed, normal fluctuation. Review of cause code information for these time periods indicates that an 4 increase in the number of conditions having resulted from errors in the WORK PRACTICE and PROCEDURE categories were the primary contributors. 4 The negative trend in WORK PRACTICE during these months appears to be concentrated in the, area of

 " Skill Based Slips" and " Failure to Perform Self-CheckingNerification" The negative trend in the area of PROCEDURES is primarily due to an increase in the number of identified " Drawing Errors' and " Situations not adequately covered by proosdures*

The fact that the 1996 refueling outage and the unplanned outage related to the reactor coolant pump ARD failure coincide with these time periods may indicate ! hat outage related activities resulted in these increases, i 1 l l I i 1 A-10 l

k l , 1 l l 4 i , I l i i i ) i 1 ! l . \ . \ l l APPENDlX B J l HARDWARE CAUSAL CODE DISTRIBUTION - l i i 4 i J l l l l l l l l

APPENDIX B - HARDWARE CAUSAL CODE DISTRIBUTION Th3 purposs of this Appendix is to allow managtment to ass:ss overall" Hardware" cause code trends  ! l Although the functional area analyses in this report do include condition report and cause code information I when appropriate, some cause code trends may be hidden because the trend is divided among several ' functional areas. The following table provides specific data on the number of Condition Reports (or incident Reports) which have been assigned one of the " Hardware" cause code types as of January 15,1997.

 ~ CATEGORIES                                95 1 95 2 95 3 95-4           TOTAL]961 96-2 96 3 l96-4 TOTAL                        l DESIGN SPECIFICATIONS                          33    24     16        28      101   21          11        9   12             53 DESIGN REVIEW                                    6     3      2         7       18    13         14       21   19             67 COMPUTER SOFTWARE                                4     0      0         5        9     4          4        4      3           15 INADEQUATE PREVENTIVE MAINTEfMNCE                2      1     3         9       15      3         4        3      3           13 UNRELIABLE EQUIPMENT                             4     5      2         6       17   18          12        8   14             52 FABRICATION ERROR                                5     4      4         6       19      5         7        7     6            25 UNFORSEEN FAILURE                              42    40      32         6     120    11          12       11   14             48 ASSUMED RISK                                     6     8      4         4       22     2          4        6     8            20 UNFORSEEN CALIBRATION FAILURE                   12    12     14         4       42     5          1        6   11             23 UNFORSEEN MATERIAUEQUIPMENT FAILURE            20     12     18        50     100    44          71      65    67            247 l TOTALS                                        134   109      95     125       463   126       140                            563 140l 157 Note: As of January 15,1997 approximately 80 condition reports have not been cause coded.
                                          "ile o" Char:

160

                                                                                                                    ^

1* y m; , 120 -7 ( 5 DESIGNSPECIFICATIONS '

       =                                                100 -    .

3 DESIGN REVIEW 8 [ COMPUTERSOFTWARE 80 -

    @ INADEQUATEPREVENilVEMAINTENANCE                                                      -

UNREllABLE EQUIPMENT 60 - - - FABRICATION ERROR ' ' ' 40 - UNFORSEEN FAILURE [ ASSUMEDRISK 20 - j UNFORSEEN CAUBRATION FAILURE l

     -                                                     0       i           i           i             i            i UNFORSEENMATERIAUEQUIPMENTFAILURE 95-4        96-1       96-2           96-3        96 4 i

B-1

d APPENDIX B - HARDWARE CAUSAL CODE DISTRIBUTION 1

SUMMARY

ANALYSIS With approximately 80 condition reports remaining to be cause coded (as of 1/15/97) a review of the Hardware Cause Code Distribution Table indicates the following trends:

             " Unforseen material / Equipment Failures" and " Design Review" are the two largest causal factors categories in the area of Hardware.

Hardware Category Analysis Unforseen Material /Eauioment Failures: A review of causal code data for this category indicates that " Material / Equipment Wom" is the largest contributor to this category (67 of 157 CRs). Of

these 67 CRs 26 were attnbuted to worn material or equipment and 18 due to damaged material or equipment.

Desian Review: A review of causal code data for this category identified 8 of the 19 CRs for this quarter were to design reviews being inadequate. Additionally,8 CRs were due to wrong drawing information. I i I O l 1 l l I l l B-2

1 1 l a f 4 t J i 4 APPENDlX C i

CONDITION REPORT EVENT TYPE DISTRIBUTION 1

1 r i 1 e 4

APPENDIX C'- CONDITION REPORT EVENT T(PE DISTRIBUTION l l l The following tables provide a breakdown of the number of Condition Reports (CRs) which have been assigned a particular " Event Type

• code for each of the 19 "Funcuonal Areas" identified under the Condition Report System as of January 15,1997 This information does not include all CRs which have been assigned to a particular functional area, however, those which are identified are considered to  ;

represent the significant portion of the reports assigned. It is expected that this information can be utilized to assist in identifying those areas where increase management focus may be warranted. Functional Area: Administrative Functional Area: Environmental Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Administrative 14 11 3$ Chemical 2 2 9 Performance Release / Controls (ADP) (Clut) Administrative 11 6 25 Radioactive 3 3 10 Mix-up (ADX) Releases (REL) Documentation 7 4 14 Equipment (EQU) I 1 5 (DOC)

      ..                                                     Chemistry (CIO          I         O           2 Administrative      11         2            14 Event (ADE)

Equipment 1 0 2 Actuations (EQA) Functional Area: Chemistry Functional Area: Engineering Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Chemistry (ClO 11 2 34 Engineering 23 17 69 (ENG) 3 0 13 9hn=t Plant 16 15 68 Modifications Chemical 1 1 7 (MOD) (C Configuration 14 11 48 Control (CON) i Drawings 7 12 36 (EQO) (DWG) Instrumentation Equipment 11 4 35 2 0 4 (INS) (EQU) e C-1

APPENDIX C - CONDITION REPORT EVENT TYPE DISTRIBUTION Functional Area: Fire Protection Functional Area: Housekeeping Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Fire Protection 24 12 60 floucekeeping 5 2 12 (FP) (floU) Administrative 2 2 7 Administrative 4 2 8 Peformance Performance (ADP) (ADP) Instrumentation 2 7 Material Contml 1 1 0 3 (INS) (MAT) Equipment (' Qti) 2 0 5 Fire Protection [ 2 3 (FP) Functional Area: Radiation Protection Functional Area: Computer Software Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned - Currest Previous Year to Currest Previous Year to Quarter Quarter Date Quarter Quarter Date Radiation 13 4 42 Computer (CPr) 6 6 25 Protection (HP) Radiological 3 4 12 U) Work Practices (RW) Documentation 0 0 2 Equipment 2 0 7 (EQU) Radiation 0 0 2 Administrative 2 2 4 Event (ADE) Functional Area: Fuel Reliability Number of Condition Reports Event Type Assigned Current Prwious Year to Quarter- Quarter Date Chemistry (CH) 1 0 4 operational 3 0 3 Nonconformance (NCR Equipment (EQU) 2 0 2 C-2

1 i APPENDIX C - CONDITION REPORT EVENT TYPE DISTRIBUTION Functional Area: Maintenance Functional Area: Material l l Number of Condition Reports Number of Condition Reports l Event Type Assigned Event Type Assigned Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Maintenance 18 11 75 Material control 16 6 43 Work Document (MAT) (MWO) Administrative 7 0 9 Equipment 17 13 74 Performance Maintenance (ADP) 1 (MNT) i Equipment 2 2 g l l Administrative 10 40 Maintenance 11 Performance (MNT) (ADP) Plant 0 1 5 Equipment (EQO) 7 4 29 g Modifications Equipment (EQU) I 1 4 Functional Area: Facility Operations Functional Area: Quality - l l Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned I Current Previous Year to Current Previous Year to  ! Quarter Quarter Date Quarter Quarter Date Equipment (EQU) 16 41 Administrative 1 23 11 $4 {

Performance 1 surveillance Tests 18 10 39 (ADP)

(ST) Quality 14 4 28 Equipment 15 9 35 Assurance & Operability Control (QA) (EQO) Documentation 3 1 10 Administrative (DOC) l 5 13 30 ' 3 Performance 4 (ADP) Equipment 7 4 23 Actuations (E0A) c-3

APPENDIX C - CONDITION REPORT EVENT TYPE DISTRIBUTION Functional Area: Emergency Response Functional Area: Security Number of Condition Reports Number of Condition Reports , Event Type Assigned Event Type Assigned i Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Emergency 5 7 16 Security (SEC) 5 5 27 Response (ER) Quality 0 1 2 Regulatory 0 1 2 Assurance & (REG) Control (QA) i l Functional Area: Training Functional Area: Industrial Safety Number of Condition Reports Number of Condition Reports Event Type Assigned Event Type Assigned Current Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Training (TRG) 2 Ig Industrial Safety 1 10 7 40 (SAF) , Administrative 0 2 9 Perfonnance Administrative 4 5 14 (ADP) Performance (ADP) Equipment i 1 6 (EQU) l Radiation 1 1 4 Protection (HP) l l h l Functional Area: Special Processes Functional Area: Equipment l Number of Condition Reports Number of Condition Reports l Event Type Assigned Event Type Assigned " Currest Previous Year to Current Previous Year to Quarter Quarter Date Quarter Quarter Date Welding (WLD) 4 4 13 Equipment (EQU) 108 64 322 l Maintenance 2 0 5 Equipment 40 29 115 l Work Document Operability ' (MWO) (EQO) Administrative 2 2 $ Operational 12 26 102 Performance Nonconformance (ADP) (NCR) Instrumentation 10 5 35 (INS) Equipment 7 6 25 Actuations (EQA) C-4

4 4 1 i. J q l 4

   ^

1 1 I l APPENDIX D l 4 I OVERVIEW OF SPECIAL CODED ISSUES 1 4 e 1 I 1 l ? j l l i f i

l APPENDIX D - OVERVIEW OF SPECIAL CODED ISSUES GENERAL i The purpose of this appendix is to provide management the ability to assess the distribution of condition l reports or cause codes relative to issues which have been determined by management to be of special interest. These issues, or areas of interest, are typically identified through the assignment of the , appropriate 'special codes" to each condition report by the CAG group. In addition to providing  ! information concerning the distribution of special codes, an assessment of the distribution of causal codes I assigned to condition reports which were identified as having a technical soecification violation has also been provided. TECHNICAL SPECIFICATION VIOLATIONS The following distribution of causal codes are related to condition reports which have been identified as resulting in technical specification violations. The large majority of these violations are a result of a failure to fo"ow procedures as described in T.S. 5.8.1, While these types of violations are individually, nonreportable, they are individually assessed by the Condition Review Team (as described in S.O. G-5, and R-2)in order to determine if an Un-Reviewed Safety Question (USQ) exists. The process of reviewing T.S. 5.8.1 violations was initiated in the third quarter of 1996, and involved performing a retrospective review of approximately 650 condition reports generated since September 1995. This review process is not yet complete. 4 TABLE-1 DISTRIBUTION OF CAUSE CODES FOR T.S. VIOLATIONS CODE DESCRIPTION # CRs PEW LACK OF ATTENTION TO DETAll 30 PEWB OVERSIGHT-LACK OF DIRECTION 29 MSXB LACK OF DEPTH IN EVALUATION / REVIEW 18 PRYD SITUATION NOT COVERED 18 MSWC SPAC NOT USED 15 PEWG SKILL BASED SLIP 14 PEWA CARELESSNESS 14 Note: The above data represents the seven largest cause code categories as of 1/15/97. l 1 l i, .1 D-1

APPENDIX D - OVERVIEW OF SPECIAL CODED ISSUES 1 l SPECIAL CODES Special codes are assigned to condition reports by the CAG group as a means of tracking and identify;ng unique concerns related to a wide variety of issues. This appendix is intended to provide an overview of the distribution of selected special codes which have been assigned to condition reports during the fourth quarter of 1996. SPECIAL CODE CATEGORIES Identified by NRC Resident inspector: The following conditions were identified during the fourth quarter of 1996 as having been brought to attention by the resident NRC inspector. I CR#199601336 Concems regarded the loading of fuel near an inoperable excore detector during core reload. A root cause analysis was performed in support of the response to this condition report. Identified by the Self-assessment for Licensina Basis. Conformance Prolect: As of January 15,1997 the self-assessment of the Fort Calhoun Stations conformance to its licensing basis had initiated 1 condition report during the fourth quarter. A review of these CRs identified the following l distribution; I CR# 199601618 Identified a concem associated with a failure to update the USAR on heavy loads following changes which resulted from reducing the boric acid concentration in the boric acid storage tanks.  ; l Good Catches: As of January 15,1996 eleven (11) condition reports were identified as good , catches during the third quarter. ' CR# 199601279 Identification by a licensed operator, during an informal discussion of work activities, that containment closure requirements may have been violated l CR# 199601309 Identification by an operator that a wooden ladder was smoldering in the i turbine building sump. l CR# 199601346 Identification by a control room operator that containment closure requirements had been violated. CR# 199601363 Identification of a design flaw in the control circuit for switch HC-101 CR# 199601368 Identification of degraded pressurizer heater cables during an ISI weld inspection. CR# 199601395 Identification of a personnel hazard regarding work being performed in j the intake cells / tunnels in conjunction with a 480 volt buss outage l l CR# 199601412 Discovery that a potentially contaminated valve had not been properly l stored. l CR#199601454 Identification of panels on Al-50 that pose an electrical hazard to personnel and instrumentation and which are no longer required. CR#199601472 Identification that test connections for LLRT testing on penetration M-80 were not tapped through during original construction of plant. D-2 l l

 . .. .. .   .  -.. .    .     .-      . .  -      -           -     ----            -~-- --       --- - ~- -

. APPENDIX D - OVERVIEW OF SPECIAL CODED ISSUES CR#199601514 S*curity gurrd id!ntifi;d MS-282 was Inking in room 81. CR#199601526 Discovery that the spent regenerant tank crane could cause damage to fire protection piping when traveling west upon its track. M D-3

                                   -                                                         w   n            -
                                                                                                                -y

=r A - + APPENDIX E TREND REPORT PROCESS I

APPENDIX E Trend Renort Process i

The Trend Report Process starts with the assignment of group, functional area, and cause codes to OPPD's Corrective Action Documents. The groups code represents the management level whose personnel caused the finding (e.g., Supervisor - Quality Control, Supervisor - Chemistry, Manager - Training, etc.). The functional area code represents what type of problem occurred ,

(e.g., Fire Protection, Maintenance, Engineering, etc.). The cause codes are representative of the , factors which resulted in the problem occurring. The cause codes are based upon the  ! classification contained in Nuclear Operations Division Procedure NOD-QP-19, " Incident Investigation / Root Cause Analysis Guidelines". Under the new Condition Report (CR) System, each Corrective Action Document (CAD) can be assigned as many group codes, functional area codes, and cause codes as necessary to trend the ' condition. This allows the CAD to identify if more thar om 'rea or cause was involved in the incident or finding. As a result of this process, a CAD cm ifect the trending of multiple groups, cause, and functional areas. CRs have preliminary functional area and group codes assigned soon after they are issued. As a result of the preliminary codes, infonnation in the report may change slightly as the CADS are closed and their final codes are determined. Finally, CRs are not cause-coded until the response  ; is received from the owner which is between twenty and forty-five days. This is reflected in the -l

high number of unassigned cause codes as mentioned previously.

The next step in the Trend Report Process is the analysis of the trend data by the lead manager for each functional area. Since this report serves as a tool to identify any quality trends and to assess the overall quality of work in each functional area, the lead manager must consider more than CAD trends when determining the performance of an area. As a result, each lead manager

                                                                                                                          )

includes the following information in the analysis of their functional area. SELF ASSESSMENT RESULTS: Results of any self assessments that were conducted in the functional area and the number of problems which were self-identified. l OUALITY VERIFICATION RESULTS: Key results of any Nuclear Safety Review Group (NSRG) assessments, Quality Control (QC) inspections, and Quality Assurance (QA) audits or j surveillances. EXTERNAL ASSESSMENT RESULTS: Key results of any NRC, INPO, State of Nebraska, or other external inspections including any violations or other findings resulting from external assessments. E-1

  - -          -        - . _ -             -.       .       .       .  . - -       -      .     - - ~ - -

a j CADS: Any trends or recurring problems associated with CADS. SIGNIFICANT CAD

SUMMARY

A summary of significant CADS which occurred during the j quarter.

PERFORMANCE INDICATORS: The status of any Performance Indicatorsp) for the functional area. This includes those indicators in the Monthly PI Report distributed by Station Engineering and any other key indicators that may be maintained separately. 4 , Based on this data and any other pertinent information, the lead manager for each functional area j classified the area's performance as Excellent (green), Good (white), Fair (yellow), or Poor (red). j To assist in classifying an area, senior management has provided that the following general

j. performance criteria:

j Excellertt: Sustained excellent performance. Performance recognized as among the best in the industry. Performance which consistently exceeds expectations. Performance which reflects a strong commitment to maintaining excellence. . Good: Performance which consistently meets expectations. Performance which reflects a strong commitment to weakness identification and correction. Emir: Performance which occasionally does not meet expe.ctations. Performance which indicates the need for further management attention. Ener: Performance which does not meet expectations and for which corrective action is being taken promptly to alleviate a significant safety, program, or regulatory concern. Performance characterized by significant recurring problems. Performance which, is assessed against NRC SALP criteria, would be a Category 3. Since this report uses functional areas which emphasize teamwork, it should be noted that the lead manager for a functional area may not be directly responsible for the problems that are included in that area. For example, the Manager - QA&QC is the lead manager for the Quality functional area. This area includes problems with QC holdpoints which may be caused by a i worker failing to notify QC or by a QC Inspector failing to sign the holdpoint after completing the inspection. Even though both incidents would be Quality problems, QA&QC personnel would not be directly responsible in the first instance. This same analogy can be applied to the other functional areas. To assess the problems associated with a department, the group codes,  ; which document who caused the problem, would have to be reviewed. j E-2 1

FORT CALHOUN STATION TREND REPORT DISTRIBUTION LIST Ouality Assurance Prouram Mananement W C. Jones 8W/EP 1 W. G Gates 8W/EP 1 (5 Extra Copies) R. L. Andrews FC-2-4 T. L. Patterson FC 2-4 S. K. Gambhir FC-2-4 V. Kroon SE/EP 3 - W. Steele SE/EP 1 j N. L. Marfice 9E/EP 1 G. J. Krause 43rd l R. L. Sorenson 2E/EP 1 D. D. Kloock 8E/EP 1 J. W. Chase FC-1-1 C. J. Brunnett FC-2 2 D. R. Trausch FC-I-l O. J. Clayton FC 2-4 (2 Extra Copies) J. W. Tills FC-2-4 (1 Extra Copy) H. J. Sefick FC-2-1 S.1. Willrett FC-I-5 J. L. Skiles FC-1-1 - R. W. Short FC-I-l R. L. Jaworski FC-2-4 T. J. McIvor FC-2-2 1#.J.Ponec FC-2-4 R. L. Wylie FC-1-8 R. G. Conner FC-3 1 - M. A. Tesar FC-1-1 i Other SARC Members D. P. Galle J. H. MacKinnon R. D. Martin l J. W. Shannon B. Lisowyj FC-2-3 Other PRC Members S. W. Gebers FC 1-1 H. J. Faulhaber FC-1-1 M. R. Core FC-I 9 C. P. Stafford FC-1 1 D. E. Spires FC-1-1 G. C. Bishop FC-1 1 Others L. T. Kusek FC-1 12 C. K. Huang FC-2-4 M. G. Burggraf FC-1-1 J. B. Herman FC-1-1 D. S. Booth FC-1 2 J. K. Kellams FC-1-1 J. K. Gasper FC-2-4 L. A. Howell FC-I-I QA Records FC-2-2 R. G. Haug FC-2 1 E. R. Lounsberry FC-2-2 R. L. Plott FC-1 1 F. C. Scofield FC-2 4 J. E. Zelfel FC-2-2 J. G. Keppler M. L. Ellis FC-1 2 Distribution til Extra M_ Total Copies 76

FORT CALHOUN STATION

                                                                                                                                                  " ORGANIZATION CHART" VICE PRESIDENT DIVISION MANAGER                                                                                                                                                                   AGER-DIVISION MANAGER                                                     MANAGER-
           - ENGINEERING &                                                                                               MANAGER-FORT                                           STRATEGIC                  MANAGER-CORRECDVE     AGER MA a

OP NS MOW STADON NG # UD q y ASSESSMENTS fq ACTION GROUP - Manager-Station - Manager- Esmergency - Manager-Security - Noclear Planning Engineering Nuclear Planning Services Department

                                                                                                                                                                                  "'8*   "*'*"

- Manager-Design - Assistant Manager- - Corporate Health -

                                                                                                                                                                                        ""P     E Engineenna Nuclear                                                                                      Fort Calhoun Station     Physicist

{" - hianager Construction Manager- Operations - A ity

                                                                                                                                                                           ~

8 8 A n iw Mces Control - Manager-Nucleas - Manager-Radiation - Manager-Nuclear Procurement Services Protection Safety Review Group Nuclear Safety - Manager-Nuclear - Manager-Planning & - has hop Ucensing 'cheduling Coordinator - Vice President

                                                                                                                      - Manager- Maintenance Manager- Chemistry

         --       -.          - - _ - .       . - . . . . - _ .  ~   -    - - - . -     . - -

1 GROUP EXPECTATIONS 1

1. To relentlessly pursue an INPO 1 and SALP 1 by focusing on excellence l in everything we do. First and foremost an uncompromising dedication to reactor safety.

2. Personally maintain a high energy level.

e Mentally -integrate big picture e Physically - rest, recreation, conditioning e Philosophically " freedom for" not " freedom from" e Winning Attitude - The only reason we are not as good as other people is because we choose not to be.

3. Practice teamwork and respect for individuals in a positive environment created by ourselves. Tough times and adversity will be regarded as challenges and learning opportunities. Treat every action we as leaders take as having more impact than all our words, procedures, and public speeches.

l

4. Use this guide for your business life.

e coaching /in the plant / observation / performing personal self- I assessment e equipment / daily operations / meetings e pinnning/ industry participation / corporate participation

5. We currently see ourselves as:

e technically very good a great in a crisis Keep these attributes and add: e practice moving forward with a plan aware of current industry issues / trends e become known as a strong finisher e great teamwork - seek to understand first before making a decision

6. Treat people right - like you would want to be treated. Show respect; listen; show appreciation for people's effort / time; show support.

l l I Workshop Series Performance Coaching: Developing Your Team Response to request from NLDSC l l Workshop description-  ! This workshop will be a forum for Nuclear Senior Managers to communicate their i performance expectations and make assignments. i l The workshop format will utilize group discussion, application exercises, and a text book 1 to equip managers, supervisors, and leads with the performance coaching skills needed to help transform a work group into a high-performance team. Learners will master the same proven management techniques perfected by successful coaches to lead, inspire, motivate, mentor, counsel, and create winning performance. Learning objectives: Upon completion of the workshop, the learner will be able to: l Apply the five-step FCS coaching model ' Employ the multiple roles of the performance coach

                 - Coaching (the high-performing " player").
                 - Mentoring (the willing " player").                                                        I
                 - Counseling (the problem " player").

Apply strategies for creating and maintaining a high performance team, including use of personal effectiveness tools. Understand OPPD's Performance Management process and available tools:

                 -PPR
                 - Performance Improvement Plan (PIP)
                 - Development Action Plan (DAP) l
                 - Standardized use of Performance Measures
                 - Methodology for conducting a value-added analysis of work processes /                     f activities l
                 - Methodology for setting Performance Expectations at the at the work group level Link coaching skills development with FCS Team Building (Unity Council) activities.

Understand Nuclear Senior Management's expectations, and the process they will use to monitor implementation of expectations: Use PM tools beginning 1997 Complete PPRs focusing on feedback dialogue by 3/31/97 Complete analysis of work to identify non-value-added work; take initiative j to do whatever needs to happen to modify / streamline work as appropriate  ; by 12/31/97; Seek sponsorship support as needed. Complete definition of work group level Performance Expectations with employee input (customizing from generic set provided by HPL team) by 12/31/97. N:\ DEPT 611\PRGMs\PRFCOACH.Ho1 January 23,1997

Performance Co:ching Workshop Page 2 Learners: All FCS managers, supervisors, coordinators, crew leaders and other group leads. This is approximately 225 people. Proposed Time line: Workshop series begins week of January 27,1997. Allleamers will complete . Parts 1,2, and 3 by April 30,1997. Logistics: Each class of learners will attend three 4-hour sessions (with intersession

                " homework" assignments).

Number of classes is approximately 15. Maximum class size is 15 leamers. The book The Manager's Role As Coach will be provided to each leamer and will serve as a " text book" for Parts 2 and 3 of the workshop. l l l l l I l l l i l l l l l l N:OEPT611PRGMs\PRFCoACH. Hot January 23,1997 l

1

Performance Coaching

Developing Your Team i

Part 1 - Management Direction (four hours) J Management Expectations

• Updates re: nuclear-related

                    - Competencies
                    - Position Description / Competency Profile i
                    - Performance Planning and Review Process and form Revisions
                    - Performance Measures
                    - Performance improvement Plan
                    - Development Action Plan
                    -Targeted Development Planning
                    - On-going Follow-up Personal Effectiveness Tools
                   - Meeting Management
                   - Time Management
                   - Stress Management Conducting a value-added analysis of work processes / activities Part 2 - Coaching Behaviors (four hours)                                                                       i Specific performance coaching behaviors
                   - Coaching (the high-perforrning " player")
                   - Mentoring (the willing " player")
                   - Counseling (the problem " player")                                                                 '

Part 3 - Team Support Behaviors (four hours) High performance team support behaviors Defining work group performance expectations Integration of Perfonnance Coaching workshop with other FCS initiatives

                   - FCS Coaching Model
                   - Unity Councils Note: Nuclear Senior Management support activities include teaming the workshop content; identifying management expectations; identifying on-going follow-up/ monitoring mechanisms; and actively participating in the delivery of Part 1 workshop content (i.e., Management Expectations) with each group of leamers.

N:\ DEPT 611\PRGMS\PRFCoACH.Ho1 January 23,1997

, OBJECTIVES At the end of this course the leamer will be able to:

 +        Apply the 5-step FCS coaching model
 +         Employ the multiple roles of the performance coach:

coaching 4 mentoring counseling

 +       Apply strategies for creating and maintaining a high performance team, including use of personal effectivaness tools
 +        Understand OPPD's Performance Management (PM) process and available tools:

PPR Performance improvement Plan (PIP) Development Action Plan (DAP) standardized use of Performance Measures methodology for conducting a Value-Added Analysis of work processes / activities ' methodology for setting Performance Expectations at the work group level l

 +        Link coaching skills development with FCS team-building activities (Unity Council)        !
 +        Understand Nuclear Senior Management's expectations, and the process they will use to monitor implementation of expectations I

use PM tools beginning in 1997 1 complete PPR's focusing on feedback dialogue by 3/31/97 complete analysis of work to identify non-value-added work; take initiative to do whatever needs to happen to modify / streamline work as appropriate by 12/31/97; seek sponsorship as needed complete definition of work group level Performance Expectations with employee input (customizing from generic set provided by HPL team) by 12/31/97 N:\ DEPT 611PRGMSPRFCoACH. Hot January 23,1997

Performance Relationship Map Business Need: Operate Fort Calhoun Station safely at optimal performance while reducing costs. Operational Results On-the-Job Performance

1. Should (desired) Causal *

                                                               %                             2. Should (desired)

(Business and operational goais) Linkage (On-the-job performance requi ments established for employees (of performance required) to ensure that goals are met)

      . INPO 1 rating,1998                                           Nuclear Oraanization Overall SALP 1 rating.1997                                        .

Mgmt>supv. with Ees define expectations at work group level

     . 3 Year Capacity factor       ,f   %g                      .

MgmtJsupv. previoe on going accurate, honest feedback; pay 85.9%.1997. 87%. 2000 g,7 attention ask questions so Ees feel accountable

     -     Cost competitive: 1.8
                                                                     . Mgmtisupv. use tcots available to monitor and take action re:

cents /KWh. 2000 performance (e.g., PPR. PIP, DAP)

           $66.7 Msttion budget.1999                                 .

• Other Mgmtisupv. teacn and coach Ees, model desired behaviors Mgmtisupv. ask for performance feedback from Ees; Ees provide feedback to mgmtisupv.

Everyone does work as efficiently, effectively as possible (e.g., meet ngs are well-run) Mgmtisupv. and work group rnembers study, simplify, optimize work processes Ees monitor themselves, their work and their leaming; participate in self improvement Gap Gap 4.Is Causal 9 3. Is (Current performance is yielding ("-

• Linkage (Indicates the current / actual performance of current performance results) (performance to results) employees when compared to the should)

     . INPO2 SALP 2 Nuclear Oroanization Overall Capacity factor =88.1%                                       Expectations not clearly stated and understood by work groups                                    j st att levets 2.79 cents /KWh                                          .                                                                                                    l
    .      1996 budget = $76.7 million                                  Mgmt/supv. tend to avoid difficult feedback dialogues needed to enhance accountability
          - O and M. $64.8 million                                  .

Capital, 311.9 milion Mgmtisupv. do rot effectively monitor perform. accountability

    . Other                                                         via tools available to them People are well-trained technically, but may not take responsibility for continuous coaching & learning on the job
                                                                    . Mgmtisupy tend not to seek perform. feedback from Ees Mgmt /supv. do not have/take time to analyze, simplify work processes People not in habit of self-monrtoring

--> 5. Environmental Factors impacting Performance 4--- External Causes internal Causes (Causes outside the contr .e (Causes wrthin the control of management which contnbute to the management whr.h contr%;, ; gap in operational and perfctmance results) the gap in operational and Lack of skill: e.g., performance results)

                                                                    . Performance management . Work p ocess analysis
                                                                    . Coaching / feedback                       . Confrontation / confhet resolution
    . Regulatory requirements
    . Insurance requirements Other factors: e.g.,

• Deregulation Too many unproductive meetings leaving insufficient time to

    . Competition (Generation,                                      attend to entical managing / supervising responsibilities
                                                                    .   "F'irefighting*/ urgencies etc.)
    . Marketplace saturation Ineffective planning / scheduling information systems (product; open market Inadequate consequence management practices
                                                                    . Inadequate performance reinforcement mechanisms services)                                                 . Overty-complex work processes Current systems /s'.ructures do not optimany support performance requirements (e.g., org. structure, policies, rewards. PPR, business processes, job design, job tools, supervision. training, etc.)

NSDEPT611\PRGMS\ COURSES \PRFCOACH\PRFCOACH.OHT 1/23/97

Current Imorovement initiatives Designed to Achieve Desired Operational Results (See Performence Relationshin Mao - box 1 *Should") Initiatives Objectives

• 100-Day Team Recommendations
• Focus FCS activities after reorganization
• Integrated Development Plan > Increase organizational and individual capability
• PPR Revisions *
• Improved understanding of performance expectations, on-going performance feedback
• Maintenance Work Flow Project
• Streamline maintenance process; reduce cycle time
• Washington Energy Intemational Group
• Position FCS for deregulation and competition Recommendations
• Radiation Protection PMT
• Reduce personnel radiation exposure; increase efficiency; reduce cost
• Unity Councils
• Unity Councils
• Project Success
• Improve leadership, teamwork, communication, and human performance
• 1.8 cents /KWh, SALP 1, INPO 1 by 1999 1 Nuclear Techni::al Training Five-Year improve efficiency of training program; greaterline ,

Plan; Training Performance Measures management accountability for training results

• Competencies
• Define / communicate performance expectations 1

Technleal Area-Specific PEPS

• Problem solving; technical enhancements initiatives - Proposed Objectives - Proposed
• Performance Coaching Workshop *
• Manager / supervisor skill development re: PPR revisions, coaching / feedback skills
• Leadership Development Plan Revisions
• l
• Leadership Development Plan Revisions

                                                 * *ldentify/ develop key FCS leaders                     !

Note:

• indicates major efforts that emerged from IDP and/or 100-Day Teams N^ DEPT 611\PRGMS\PRFCOACH. Hot 4- 1/23/97

Performance Relationship Map Business Need: Operate Fort Calhoun Station safely at optimal performance while reducing costs.

2. Should (desired)

(On-the-job performance requirements established for employees to ensure that goals are met) Nuclear Oraanization Overall i'

• Mgmtdsupv. with Ees define expectations at work group level
• MgmtJsupv. provide on-going accurate, honest feedback; pay attention, ask questions so Ees feel accountable {
• MgmtJsupv. use tools available to monitor and take action re: performance (e.g.,

PPR, PIP, DAP)

• Mgmtlsupv. teach and coach Ees, model desired behaviors
• MgmtJsupv. ask for performance feedback from Ees; Ees provide feedback to mgmtlsupv.
• Everyone works as efficiently, effectively as possible (e.g., meetings are well-run)
• Mgmtlsupv. and work group members study, simplify, optimize work processes
• Ees monitor themselves, their work and their learning; participate in self-improvement 1

l NM)EPT611PRGMSPRFCOACH.HO1 1I23/97

PERFORMANCE COACHING WORKSHOP SERIES SUPERVISORY DEVELOPMENT  : 1

 + Training Program Master Plan (TPMP) for FCS Supervisory Development                           )
    - Complete 1997                                      l
    - action HRD and Training                            <
    - TPMP require following courses
       + EEO/ Affirmative Action Application
       + CBOP (initial)
       + CBOP (requal) l 1   l l                                                         l

l SUPERVISORY DEVELOPMENT

                   + Introduction to Supervisory Skills
                   + Making a Difference (new conununication skills course)
                   + Union Bargaining Agreement Interpretation (new)             I
                   + BehavioralInterviewing Skills i

l l PERFORMANCE COACHING 1

                -FEEDBACK                                                         l
                -INTERVENTION 2

PERFORMANCE COACHING + Performance shoulds and related support / solutions needed to reach business and performance goals

  - Better tools to manage employee performance
  - Learn to be better coaches
  - Review work processes for non-value added work
  - Identify work group-level performance expectations PERFORMANCE COACHING

+ Learning Objectives

  - Apply the five-step FCS coaching model
      + plan
      + teach
      + monitor
      + intervene
      + learn i

3 t

1 1 1 l I PERFORMANCE COACHING

    + Employ the multiple roles of the performance coach
       - Coaching those who exceed expectations
       - Mentoring those who meet expectations
       - Counseling those who do not meet expectations 4

1 PERFORMANCE COACHING s

    + Understand OPPD's Performance Management process and available tools
       - PPR
      - Performance Improvement Plan
      - Development Action Plan

. - Use of Performance Measures

      - Conducting value added analysis
      - Setting Performance Measures l

4

PERFORMANCE COACHING + Understand Nuclear Senior Management's expectations and the process

 - Use ofPM tools
 - Complete PPRs focusing on feedback
 - Complete analysis ofnon-value added tasks
 - Complete dermition ofPerformance Expectations PERFORMANCE COACHING

+ Training

  - Three days
  - Four hours each day

+ Started January 27,1997 5

i 4 1 i } i 1 1 i i

OPERATIONAL &

PERFORMANCE l STANDARDS 4 i l

l l I l

1 i t i Operational Results j 3-Year Capacity Factor i 87 % 87 - l 86-85-84 - 83.5 % 83 - A Q,i g!;f Et. 82 - J ., L ^.5 81

                                                         ' 'kSk   ~

Current Desired l l l l l 1

1 4 i I I i Cost Competitive j 3 2.8 a 1 p. . .g .if, ' 2.5 - p$!c' .<+ ,. n - vy:.. ,+. l . 2 W pg .l.91 .gy 1.8 , w. i'e;4, u@.. l 1.5 - d ..M !  %.M l 1 MU 3; A l $ l 0.5 - N('[k

                                                          .q,      9 0-                                      i                         i cents /kwH               Current                 Desired i

f i ! Operational Results - SALP l 1 l I 1.25 - i 1.5 - -. 1 . i h g% :3} fr j 1.75 - EM* ?@.

                                                          ;f-40 -V i                                                          b.t ' ::/

l

2-  %?.! % i

) Current Desired i 1 i e i .,i 2

1 I i 1 Operational Results - INPO 1-2- ;w, 3lpgh 4 d,.' 3__ , , Current Desired Desired On-The-Job Performance Standards i

 + Model desired behaviors
 + Set, communicate and follow up on           l expectations
 + Provide and encourage on-going accurate and honest feedback 3

i Desired On-The-Job Performance

Standards

+ Mentor, coach and counsel l + Work is routinely performed as efficiently and as effectively as possible 4

1 4 i r i Desired On-The-Job Performance Standards

      + Work processes are periodically reviewed then
         - simplified
         - optimized
      + Self monitoring and selfimprovement are practiced by each member of the Port Calhoun team 4

l Causes Which Contribute to the Gap in Operational and Performance Results

 + External
    - Competition                             1
    - Regulatory requirements and practices l

l Causes Which Contribute to the Gap in Operational and  ; Performance Results  ! l

 + Internal
    - Unproductivemeetings J
    - Unnecessary firefighting
    - Ineffective planning and scheduling
    - Overly complex work processes I

5 i

4 E 4

Causes Which Contribute to the Gap in Operational and Perfonnance Results

   + Internal (cont.)

i

      - Lack ofcertain skills (primarily non-technical)

e.g.

          + perfonnance management
          + work process analysis

, + coaching and feedback

          + constnictive confrontation and conflict resolution i

6

4 i i i i 4 i y

%. THEREIS AN?INGREAStDE Ae

I i,a NUMBER'OFPNANTS!bNhTHEM e ..> -, : ;; . ..mm me:w ,, m i, #. e

a. WATCHE,.S,,

                                                    .n      .T,.n  isi We.

v r.ec%m, j

• Configuration Control

{ -More plants are going to enforcement l on configuration control. l Lack of Questioning Attitude j - Complacency l Lost Safety Focus 5 l 1 j p .. j ,.._..-_..... ,

                                       ,. {i l                               . iip 40:s t w >           -
                                                                        - ;, c
                                         +                /

i L . ._. _ . 1 A _ __ . _ _ _ . _ _ . _ . . . . _.-_.__1__.. ! Pre-decisionalEnforcement , Conference focused on i configuration control. l

• Recent violations focused on 4

configuration control. i i i I i

l l l I i ' I i l

• Be aware of what configuration controlis.

I

• Get out in the plant.

I !

• Look for differences and report them.

CIVIL PENA LTY A SSESSMENT l ! - v - i .  :=- - 3* mm1 , :r

                                                               +
                                                                                      +

r. b Timely resolution of material deficiencies.

                            - Questioning Attitude Look for a way on continued improvement.

Stay focused on safety.

k l d f

r. . !  !:.} l , e .

                                                                                                  ~ ~-W I$ ' :     _, *

• i

                                                          ' p.UNITY! en - ;

COUNCIL 1 CHARTER!;

                                                                       .      g:             x             .                       . .     ,

} j The purpose of the Unity Councilis to provide a process for 1 identifying and implementing ideas for change in the way

business is conducted at every level of the Nuclear Organization

] The primary goal of this process is to enhance our performance l and teamwork among the nuclear work groups. This will d position FCS as a viable contender in an unregulated, competitive

environment without sacrificing the safe operation of the plant.

j In recognition of the fact that those closest to the work prosess j are most fcmiliar with the barriers to optimize efficiency, j members of the Unity Council will be open to and will be seeking l input from all members of the Organization. Therefore, solutions ) to issues addressed by the Unity Council will ilm to enhance trust, { integrity and respect. ) 1 1 l c' .,

1. Al.1.lSSUES WL1.EE FEARD.1 1 KEEP TIE DEOSION MMONG PROCESS AT TIE NTIATING COUNOL '

LEVEL 1 1 COWA.NCATE NORMATION ON ISSUE RESOLLmONS NO STATUS l ACCURATB.Y TO AI.L EWFLO(EES1 l 4 RESPECT CONFIDENHAUTY OF ODERS INVOLVED IN SLRFAONG #O RESOLVING ISSLES

                                                                          'c'e'v eta $
                                                                       "";;0'.'t"                    "'====               **1Z':"'*"   !

u , l ,

                                                                    ""u:'"3",'"' l

 - . . . . . . - . _ . - . . . . - . _ . . - - -                               _ . - . - . . - . . . . _ . . ~ . _ . . - . . _ ~ . -                                                 - _ _ - - . _ . _ _ - . - . - . _ _ . . - - _ _,

I a i i

y . . . ~.. . _ _ _ . _ . . . . . . . . . . . . . . . . . . . _ . . . . . . . . _ . . . ,

1 I

                                                 } '{  ~hI h                                               '
                                                                                                                                       !7

h' b.h ..

(5- Y l l -_.._._..A_. .#...._____.._.___.2_._____._ 1 1 ! . WHY IS IT AN ISSUE 71 1 . HOW WILL THE ISSUE IMPROVE LEADERSH'P, COMMUNICATIONS, OR ! PERFORMANCE 71 l . IS THE ISSUE COST EFFECTIVE 7 WILL THE ISSUE RESOLUTION j REQUIRE ADDITIONAL STAFFING, EQUIPMENT OR LARGE USE OF l EXISTING RESOURCES?1 1 } . DOES THE ISSUE REQUIRE COORDINATION COUNCIL RESOLUTION?1 i . Could the issue have impact on existing corporate policies.1 ) . Could the issue have impact on union bargaining agreements.1 s . Could the issue have impact on Nuclear Programs, procedures or policies?1 g3% . .p -o 7 .i . .

                                                                                                        , . ;csaw,,
                                                                                                                                              ..s -
                                                                                                                                                                      - p _ 'fm
                                                                                                                                                                                           ,  w-
                                            $. 4.. A. . Wm
                                            -                                          h r@. %_M.7                    .                        3.. W1.+             S,nM       +i.

n @cy.,.,%;,

                                            ?lPERF.QRMANGE!QOAGHING,g gog

p. m +. w4- .

                                                                                                                       ;;;Og g gg
                                                                                                                             ~                                        me m; g f 3j 1

1

1 j j { pra-r~ . y ) 1 } 3 e ' i 1

• Model desired behaviors l
• Set, communicate and follow up on j expectations i Provide and encourage on-going accurate and honest feedback 4

i i

                                                               -.                              ;     1      o l                                                             ..
                                                                                                          ..._..____..__.A_m

• Mentor, coach and counrei i
• Work is routinely performed as j efficiently and as effectively as possible d

4 1 i 4 a

im mg 8 i!

• Work processes are periodically reviewed then

     -simplified
     -optimized Self monitoring and self

/ improvement are practiced by each i member of the Fort Calhoun team i jJ.,i'i '.y i ; 1G' (T! .?( : n p{r f;>; i e Learning Objectives '

   - Apply the five-step FCS coaching model plan teach a monitor intervene learn 1

_ _ _ _ _ - _ -}}