ML17331B174

Submits Response to Violations Noted in Insp Repts 50-315/93-12 & 50-316/93-12.Corrective Actions:Licensee Failed to Perform Evaluation to Determine Changes Made to Feedwater Pump Speed Control Sys by Temporary Mod

ML17331B174
Person / Time
Site:	Cook
Issue date:	01/04/1994
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO.
To:	Martin J NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:1184H, NUDOCS 9401110333
Download: ML17331B174 (88)

Text

Inaiana Miciugan Power Company P.O. Box 16631 Columbus, OH 43216 Z

INWAIIIA NICHIGAN PO Vlf'EP:NRC:1184H 10 CFR 2.201 Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 NRC INSPECTION REPORTS NO. 50-315/93012 (DRS)

AND 50-316/93012 (DRS)

REPLY TO NOTICE OF VIOLATION U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Attn: Mr. J. B. Martin January 4, 1994

Dear Mr. Martin:

This letter is in response to a letter from G. E. Grant dated November 24, 1993, which forwarded a Notice of Violation associated with a System Based Instrumentation and Control Inspection conducted by Zelig Falevits and others of your office during August 17 through September 28, 1993. The violations are associated with errors in a setpoint calculation for refueling water storage tank level instrumentation and for the lack of an unreviewed safety question determination for Temporary Modification 2-93-015, which installed an I to I converter in the Unit 2 feedwater pump control circuitry.

Our reply to the Notice of Violation is contained in Attachment 1.

We were also requested to respond to open and unresolved items from the inspection report. Our response is contained in Attachment 2.

In addition, we were requested to address non-cited deficiencies associated with Temporary Modification 2-93-015. Our response to this request is contained in Attachment 3.

This letter is submitted pursuant to 10 CFR 50.54(f) and, as such, an oath statement is attached.

Sincerely, EF. f" E. E. Fitzpatrick Vice President dr Attachments

Mr. J. B. Martin AEP:NRC:1184H cc: A. A. Blind G. Charnoff T. E. Murley - NRC NFEM Section Chief NRC Resident Inspector J. R. Padgett

ATTACHMENT 1 TO AEP:NRC:1184H REPLY TO NOTICE OF VIOLATION to AEP:NRC:1184H Page 1 NRC Violation I (Severity Level IV)

"10 CFR 50, Appendix B, Criterion III, states, in part, that measures shall be established to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions and that the design control measures shall provide for verifying the adequacy of the design.

Contrary to the above, on September 1, 1993, the team noted that:

RWST instrumentation loop setpoint calculation No. 1-2-I9-03, dated August 25, 1993:

(1) Erroneously derived the setpoint uncertainty value based on the use of Model N-E13 RWST transmitters. However, the installed RWST level transmitters were Model E13DM-HSAHl.

(2) Provided no justification for using transmitter elevation 599'3" to derive the setpoint uncertainty value in the setpoint calculation.

(3) Did not consider the error effects of the velocity head of the ECCS pump flow (Safety Injection and Residual Heat Removal pumps) during design basis accidents.

b. Flow diagram OP-1-5144-13, "Containment Spray System Unit ¹1,"

incorrectly identified RWST level transmitter ILS-950 as having a minimum level alarm at 638'll". However, the design basis setpoint value was 637I 2tl ff Response to Violation I dmission or De ia of t e Alle ed Vio atio Indiana Michigan Power admits to the violation as cited in the NRC Notice of Violation.

2. easons o t e V o at o The examples of the violation will be addressed individually.

l.a.(1) The cause of the violation is attributed to lack of attention to detail in cross-checking the specific details in the documentation. The instrument data sheet used was from a slightly different model (N-E13 versus the correct model E13 of the same vendor). Both types of instruments are used in the plant and have nearly identical performance characteristics. The correct model data sheet was compared to the incorrect model data sheet and minor differences in some of the uncertainty terms were found which required the to AEP NRC 1184H Page 2 calculation to be revised. However, no significant differences were found that changed the end results of the calculation adversely. Therefore this error did not have an adverse effect on safety.

l.a. (2) As noted in Revisions 2 and 4 of the calculation, the elevations used in the calculation were based on field walkdown measurements, which were taken in"1979. During the inspection, the elevations were remeasured but did not exactly match the previous field walkdown measurements.

Because of the time that has elapsed, it was not possible to It is positively identify the reason for the discrepancy.

noted, however, that the differences between the as-built measurements taken during the inspection and the measurements used in the calculation differed only slightly. The differences in elevation between the as-built heights and the heights used in the calculation varied between 1 inch and 2.5 inches, compared to an instrument span of 363 inches. The worst case error in the non-conservative direction (as-built lower than calculation) was 2.5 inches, which equates to an error of approximately 0.7%. In other words, the RWST indication and alarm will occur 0.7% lower than actual level.

However, since there is considerable margin built into the alarm setpoints (approximately 21% for the low alarm and approximately 7% for the .low-low alarm), this measurement error had no adverse effect on safety.

l.a. (3) The alarms and trips associated with the RWST level instrumentation include both high and low level type functions. There are two high level functions. The High Level Alarm is used to ensure the operators do not inadvertently overflow the RWST when filling, and a Minimum Level alarm is used to alert the operators that Tech Spec required minimum RWST volume requirements are being encroached. There are also two low level functions. The Low Level alarm is used to alert. the operator that RWST inventories are approaching a low level at which the operator should begin to transfer to ECCS recirculation mode, and a Low-Low Level alarm and RHR pump trip is used to alert the operator that RWST inventory is depleted and to protect the RHR pump from damage due to low NPSH.

Because the RWST level transmitter is tapped off the ECCS suction line at the bottom of RWST tank, velocity head effects can be induced when the ECCS pumps are running. The high level functions are not affected by this because these functions are only used when the ECCS pumps are not running.

The low level functions are affected because the ECCS pumps are running when they are required to function. However, to AEP:NRC:1184H Page 3 computation of the velocity head effect shows it will only affect the low level functions in a conservative manner. The velocity head effect induces a negative bias that results in a level indication that is lower than actual level. The effect on the low level alarms and RHR pump trip functions is that they will occur sooner and therefore does not jeopardize the RHR pump or plant safety.

Velocity head effect was not addressed due to unawareness of its influence on this application. Criteria for when this effect is to be considered were not included in the procedure used for the preparation of this calculation. It should be noted that most tank level instruments are tapped on the side of the tank and are therefore not affected by the velocity head effects of the tank suction line.

l.b. Setpoint information for the Cook Nuclear Plant is controlled through the Plant Setpoint Document, rather than through the flow prints. Because of this, there was no systematic process to have setpoints placed on flow prints, or to update the flow prints if the setpoint changed. The setpoints displayed on the drawings are for reference and are used for understanding the drawing only.

3. Cor ective Actions Taken and Results Achieved The examples of the violation will be addressed individually.

l.a. (1) The correct model data sheet has been compared with the incorrect model instrument data sheet and no significant differences were found. The calculation will be revised to incorporate the correct instrument data sheet information.

l.a. (2) The calculation has been revised to reflect the correct as-built elevations for the RWST level transmitters.

l.a. (3) The calculation has been revised to address the velocity head effects. Review of other tank level applications found the CST tank level to have a similar velocity head effect which was not addressed. The CST calculation will also be corrected.

l.b. Drawing OP-1-5144 and OP-2-5144 were revised to reflect the correct setpoints.

to AEP:NRC:1184H Page 4

4. ctions Taken to Avoid Further Violations As a general comment, it is noted that numerous instrument setpoint calculations are being updated as part of the Reactor Protection and Control Systems Upgrade Project which will be implemented during the 1994 refueling outages. The specific examples of the violation are addressed individually, below.

l.a. (1) Training sessions were held for Corporate Z&C engineers which emphasize that self-checking and engineering reviews and

'erification are expected to be such that errors in instrument model numbers and similar documentation data is discovered and corrected prior to document issue.

l.a. (2) A sampling of safety-related instruments will be checked to verify correct as-built elevations are incorporated into setpoint calculations. The sampling will be completed by May 31, 1994. Further preventive measures will be established depending on the results of the sampling.

l.a. (3) The engineering guide governing setpoint calculations was revised on November 15, 1993, to require that process measurement effects, such as velocity head effects, are considered, as necessary, as part of the calculation preparation.

l.b. Since the intent of setpoint information on flow prints is only to help in the understanding of the drawing, a note will be added to all affected flow prints which states:

"Cautionl The setpoints indicated are provided only to assist in understanding the drawing. Refer to appropriate setpoint control document for actual device setpoints." The notes will be added by July 6, 1994.

5. Date When Full Com liance w 1 be Achieved The examples of the violation will be addressed individually.

l.a. (1) Full compliance will be achieved by January 10, 1994, when the calculation is revised to reflect the correct model performance characteristics.

l.a.(2) Full compliance was achieved on November 5, 1993, when the calculation was revised to incorporate the as-built transmitter elevations.

to AEP:NRC:1184H Page 5 l.a. (3) Full compliance was achieved on November 5, 1993, when the calculation was revised to include consideration of velocity head effects. The calculation for the CST will be revised by January 10, 1994, to correct the similar deficiency identified during the review of the RWST calculation.

l.b. Full compliance was achieved on September 15, 1993, when the affected Unit 1 and 2 OP drawings were revised.

to AEP:NRC:1184H Page 6 NRC Violation ZZ (Severity Level ZV)

"10 CFR 50.59 states licensees may make changes to the facility as described in the safety analysis report without prior Commission approval unless the change involves an unreviewed safety question. A written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question is required.

Section 10.5.1.1 of the UFSAR states that the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators. In addition, Section 14.1.9 analyzed a loss of normal feedwater from pump failures which could result in a reduction of the secondary system to remove heat generated in the reactor core.

Contrary to the above, on April 7, 1993, the licensee failed to perform an evaluation to determine that changes made to the feedwater pump speed control system by temporary modification 2-93-015 did not, involve an unreviewed safety question."

'I Response to Violation ZZ Admission o Denial of the Alle ed Violation Indiana Michigan Power denies the violation as cited in the NRC Notice of Violation.

2. Reasons for Denial of the Violation At the Cook Nuclear Plant, temporary modifications undergo a screening to determine if an unreviewed safety question determination is required to be performed pursuant to 10 CFR 50.59. The process we use is based on the guidance of NSAC 125 (June 1989), entitled "Guidelines for 10 CFR 50.59 Safety Evaluations." This document was prepared jointly by the Nuclear Management and Resources Council (NUMARC) and the Nuclear Safety Analysis Center of the Electric Power Research Institute (EPRI).

The inspection report (page 14) states:

"By failing to recognize that the speed control system was described in the UFSAR, the licensee concluded that 10 CFR 50.59 was not applicable, therefore, no safety evaluation was performed.

The licensee's failure to perform a safety evaluation is considered to be a violation of 10 CFR 50.59."

to AEP:NRC:1184H Page 7 We disagree with the statement that the speed control system is described in the UFSAR. The UFSAR (Section 10.5.1.1) specifically states that "the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators." There is no description of the circuitry provided in this statement.

NSAC 125 recognizes that changes made to the facility may implicitly impact the UFSAR. For these cases, NSAC 125 states that:

"If the SSC (structure, system, or component) is part of a larger SSC described in the SAR and if the change affects the design, function, or method of performing, the function of the larger SSC AS DESCRIBED IN THE SAR (emphasis added) then a safety evaluation is required."

As discussed in the UFSAR, the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators. The temporary modification only added a device to isolate noise in the speed control circuitry. Neither the function of the main feedwater pumps as described in the UFSAR (to provide feedwater flow to the steam generators) nor the method of performing the function as described in the UFSAR (with variable speed pumps) were impacted by the temporary modification. Additionally, since there was no description of the feedwater control system in the UFSAR, neither the design of the control system nor the design of the feedwater system as a whole as described in the UFSAR were impacted by the change. Based on these considerations, we conclude that no unreviewed safety question determination was required pursuant to 10 CFR 50.59.

As described in NSAC 125, "The purpose of 10 CFR 50.59 is to preserve the original licensing basis in the information submitted to the NRC as part of the application for an operating license and in the final safety evaluation report (SER) issued by the NRC staff. The NRC relies on this information to conclude that an operating license can be issued without undue risk to the health and safety of the public.

This regulation allows the licensee to make changes without prior NRC approval while maintaining the licensing basis. It defines conditions that must be met in determining if prior regulatory review is needed."

The level of detail included in the UFSAR regarding the feedwater system is relatively small. This is commensurate with the fact that the system is non-safety related. The NRC staff review of the system during the original licensing of the plant would be expected to be of a different level of detail than for those systems which are safety related. Since the staff review did not rely on a detailed description of the speed to AEP:NRC:1184H Page 8 control system in order to conclude that the feedwater system was adequate from a safety perspective, it is not reasonable to conclude that It 10 CFR 50.59 would be applicable to changes to the circuitry. is noted that Section 14.1.9 of the UFSAR, entitled "Feedwater System Malfunctions," analyzes a complete loss of feedwater (due to no specific reason) and concludes that:

"a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves."

It is also noted (as acknowledged in the inspection report, page 14) that failure of the I to I converter would have the same effect as the fail'ure of the hand/auto station already in the circuit. In other words, the change did not introduce a new failure mode into the system,

ATTACHMENT 2 TO AEP:NRC:1184H RESPONSE TO OPEN AND UNRESOLVED ITEMS to AEP:NRC:1184H Page 1 The cover letter for Inspection Report 50-315/316 93-012 (DRS) requested we provide a written response to open and unresolved items in the inspection report. There is only one item in this category, Unresolved Item 315/316 93012-04(DRS). This unresolved item involves a calculation performed to verify that an instrument sensing line associated with the Unit 2 auxiliary feedwater system was adequately supported. The inspection report states that:

"The licensee was in the process of performing a calculation to determine whether the sensing line installation was adequate. Pending review of the calculation, this item is considered unresolved."

The subject calculation was completed following the inspection, and, at the inspector's request the information was mailed in an overnight package to NRC Region III on October 20, 1993. The information, documented in a memo from S.

J. Jarrett/H. P. Damasco to M. S. Ackerman dated October 19, 1993, follows.

AMERICAN ELHI IC

~bVZ2 Oate October 19, 1993 Subject Cook Nuclear Plant NRC SBICI Inspection Assistance NEDS Review of As-Found Conditions From S. J. Jarrett/H. P. Damasco To M. S. Ackerman As per your request, the Nuclear Design - Structural 8r, Analytical Section (NEDS) has reviewed. the as-found condition detected by an NRC inspector during a NRC SBICI Inspection. The 3/8"P, I/2"P, and 3/4"Q downstream low pressure instrument lines associated with 2-FFS-257, and branch lines off the feed water header line 2-FW-48 were believed to have piping/tubing syans that exceeded Alternate Analysis Criteria.

Based on the above findings, Nuclear Engineering Site Design Section (NESD) performed a detailed walkdown of the piping system (see Attachment A). In order to satisfy the seismic overlap criteria, NESD included additional piping and suyports beyond the area addressed by the NRC inspector. In so doing, they found that two support components in the overlap region of the continuation line were missing .

Nuclear Design - Mechanical Section (NEDM), performed a pxeliminary assessment of the piping system as found by the NRC inspector and concluded in their E-Mail, dated September 17, 1993 to Stan Farlow (see Attachment B), that the minor deviations in some support span lengths have an insignificant effect on the operability and/or the design basis of the instrument line. I NEDS performed, the as-found and the as-designed DBE seismic evaluations on the piping system, in Calculation No.

DC-D-02-MSC-36, by using walkdown information (see Attachment A),

and Ebasco/P-Delta (E/PD) STRUDL integrated computer program.

The piping stresses and the valve accelerations in these analyses were found to be well within the design basis criteria allowable limits.

Attachments C through F highlight the very small stress interaction ratios, and vertical and horizontal valve acceleration ratios resulting from the E/PD STRUDL analyses. These analyses were performed to confirm the NEDM preliminary engineering review.

NRC SBICI Inspection Assistance October 19, 1993 Page 2 Although the piping system in the as-found condition meets the design basis criteria limits, the two missing components are being installed on the piping system to reflect similarity of pipe supports on identical piping systems in the vicinity, and further to conform with good engineering practice.

It is NEDS understanding that Job Order No. C19393 has been initiated by Plant Maintenance to install the two missing components.

If you have any further questions please contact these writers at extension 3157.

. Jarret . P. Damasco Approved by

+0 N. Ruccia, Manager Nuclear Design - Structural k Analytical Section

/sjj xc: R. C. Armstrong/S. P. Hodge - w/o attachment A. K. Dey - w/o attachment S. K. Farlow - w/o attachment R. C. Carruth - w/o attachment T. H. Cummings - w/o attachment

Attachment A Copy'f Pf.elcLvralk from NZSD

Sheet 1 of 1 p

Attaakxment.

~pry cC M~ihuL ~ ~m Wax9mv (cuba~ 9/3.V/BS) N~n ~mrpa. D~

Sheet 1 of 1 To STAN K. FARLOWQNEDQAEPSC ROBERT C. CARRUTHQNEDQAEPSC Cc: Thomas H CummingsQSDOQCOOK,STEVE P- HODGEQNEDQAEPSC CC'm:

AMIYA K- DEYQNDGQAEPSC b ject: NRC Inquiry Relative To 2-FFS-257 Date: Friday, September 17, 1993 13:44:19 EDT Attach:

Certify:

Forwarded by:

Instrument Line 2-FFS-257 I understand, that an NRC inspector has noted the piping line associated with the Instrument g2-FFS-257 to be under supported as compared to MDS-601 design standard.

Our findigs are given below:

The Standard MDS-601 was issued in Feb.1988 and, I believe this instrument line was installed much earlier. This system and other instrument piping/tubing systems were installed per EDS Corp. developed guideline titled"Fz.eld Fabri-cated Support Guide For Small Diameter Pipes and Copper Tubing".This guide-line was included as a section in a larger guideline known as the Alternate Analysis Criteria( AAC ).

The AAC provides cookbook type directions in supporting piping/tubing systems and is very conservative.Our experience, with pzping/tubing systems analysed and supported per AAC,indicates that:

• these systems generally are well designed and meet applicable code allow-ables easily when computer analysis is performed to verify system adequacy where limitations of the AAC have been marginally exceeded.

The instrument line being reviewed is 1/2 in. sch.80 pipe upto a TEE onnection and then reduces to 3/8 in.sch. sections.Our walkdown review of this systems indicates minor deviations pipezn some span lenghths which we believe have no significance on the operability and/or the design basis of this insturment line.

If you need additional information, please contact me at 3860.

Thanks, Amiya Q

Attachment C Metho clolo Meth.os usec9. to analyze the piping system in Calculation Mo 30 C-30-OQ-ZvXR C-36

Sheet 1 of 1 A" PSC PAGE 6. 1 DONALD C. COOK NUCLEAR PLANT, UNIT 2

6. 0 METHODOLOGY The piping system has been rigorously analyzed using walkdown information and Ebasco/P-Delta (E/PD) STRUDL integrated computer program. Additional hand calculations, where necessary, have been performed to develop problem input data and/or to justify acceptance of the piping system.

The information relevant to the final output is given below:

COMPUTER PROGRAM & VERSION:

E/PD STRUDL Version 0193 VAX VMS 5.5 GOB No. RUN DATE & TIME COMMENTS 2448 10-18-93 08:21 AP run - see comm. below 2449 10-18-93 08:23 DSN run - see comm. below ADDITIONAL COMMENTS:

Job no. 2448 is the run simulating the as-found condition with Design Basis Earthquake'esponse spectra.

Job no. 2449 is the as-design axn that includes the two missing pipe clamps, and using the Design Basis Earthquake response spectra.

NUCLEAR DESIGN GROUP CALCULATION No. DC-D-02-MSC-36 i~jlljg /0 (Q cy Bv DATE DATE

A<t:aakxmenC D Problem Description

httacnment u Sheet 1 of 1 AEPSC PAGE 7.3.

DONALD C. COOK NUCLEAR PLANT, UNIT 2 7.0 PROBLEM DESCRZPTZON GENERAL DESCRIPTION:

Analysis on the as-found 3/8", 1/2", 3/4" downstream low pressure instrument header and branch lines off feedwater line 2-FW-48 to address overspan concerns raised by an NRC inspector.

NUMBER OF SUPPORTS:

8 (including the two missing unistrut clamps)

PRIMARY BUILDING:

Turbine - Area 2T4 ELEVATION OF MAIN,PIPING:

593'-0" MAJOR EQUIPMENT:

None NUCLEAR DESIGN GROUP CALCULATION No. DC-D-02-MSC-36 (o/i/(/yy l0 (8 g5 DATE CHK. DATE

Attachment E T his ratios attachment valve show s s tres s acceleration as fpunt. anal sis ancL ratios from whichortezclucLes the tw o are the that missing sup p comp onents w e11 rnri.thin the cLesign basis criteria allow able U.mits Sheet 1 S tress Xnteraction Ratio Hox izontal Sheets 2%3 AccelerationValve Ratios S heets 4 8c5 Vertical Valve Acceleration Ratios

• • P-DELTA / HPDS *CALCULATIOH No. DC D 02 HSC-36 *JOB HO 00002448*18.0CT-1993 08:21t56~PG 00066 ~"

//-PIPIHG /-TBLHAH-/--CODE--/--------/--NODE--/EQUATION/-----APPLIED FORCES-----/--------STRESS DATA-.---- --//--LOAD--/ RESULT //

'ASS HEHBER SECHAH VERSION FX/HX FY/HY FZ/HZ PRES/HONP TOT/ALM RATIO 31.1-4 0.0 0.0 0.0 .

0.0 0.0 0.0 826.22 6509.64 7335.86 37500.00' '.20

'UHHARY OF THE PIPING HEHBER CHECKS BY EQUATIOH EOUATION HAXIHUH IR HEHBER LOAD 31.1-1 0.4748 98 13 31.1-2NU 0.5887 44 1000 ~V lC/2,/Tv+OH +P y'OS 31.1-2E 0.5290 . 44 2000 31.1-3 0.0979 30 14 31.1-4 0.0979 98

• • • ~****"~FOLLOWING IS A SUHHARY OF THE CODE CHECKS PERFORHEO ABOVE *~**~*~~*~

ALL 54 PIPE HEHBERS, THAT ARE CHECKED, PASSED CODE CHECKS.

$ CHECK PIPE BREAK LOC ALL SCHECK PIPE CRACK LOC ALL SCHECK PIPE BRANCH DISPL ALL PRIHT PIPE ADD STRESS IHT FACT

~* P-DELTA / HPDS 'CALCULATIOH Ho. DC-D-02-HSC-36 "JOB HO 00002448*18.OC'I 1993 08:21:56~PG 00068

• 4 lI i 1 AA AA t ii li i i 1

• A 4 &*A A 1 1 A 1 A AA A*l*A*A*A% II ****** il I i

• STRUDL PIPIHG VALVE HEHBER CHECK I TRACE 28 RESULTS

• iiAAAAAIIN111AA1AA*AIA*N1AAAl**AlliAAII*A1*NAAA*%****A JOB ID - FFS257 JOB TITLE - CALCULA'IIOH Ho. DC.D.02.HSC-36 ACTIVE UHITS - LENGTH ME IGHT AHGLE TEHPERATURE TINE HASS IHCH KIPF RAO FAH SEC LBH

//-VALVE-./-TBLHAH-/-----.TITLE------/CRITICAL/-HEHBER-/-"- - ---APPLIED DATA - - --- -/----ALLOWABLE DATA .---/--RATIO /-RESULT-//

HODE SECHAH LOAD AX/HX AY/IIY AZ/HZ ACC 'OH . 'QUATIOH 308 GATE1 123 308 L 0.00 27.39 0.00 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 ONE 123 308H 0.00 0.00 772.80 1000000.06 0.04 PASS 0.00 0.00 ~ ONE 456 308L 0.01 54.24 0.01 772.80 1000000.06 . 0.07. PASS 0.00 0.00 0.00 ONE 456 308H 0.01 54.24 0.01 772 F 80 1000000.06 0,07 PASS 0.00 0.00 0.00 OHE 508 GATE3 123 508L 0.00 27.39 0.03 772.80 1000000.06 0.04 PASS 0.00 0.00 ONE 0.00'7.39 0.03 1000000.06 0.04 pAss

'72.80 123 508H 0.00 0.00 0.00 0.00 OHE 456 508L 0.01 54.24 0.05 772.80 1000000.06 0 07 ~ PASS 0.00 0.01 0.00 ONE 456 508H 0.01 54.24 0.05 772.80 1000000.06 . 0.07 PASS 0.00 0.01 0.00 OHE 608 GATE2 123 608L 0.00 49. 27 0.00 772.80 1000000.06 0.06 PASS 0.00 0.00 0.00 ONE

• • P.DELTA / HPDS .*CALCULATIOH Ho. DC.D 02-HSC-36 *JOB HO 00002448*18.0CT-1993 08:21:56'PG 00069 ~"

//-VALVE--/-'IBLHAH/------TITLE------/CRI'TICAL/-HEHBER-/----------APPLIEDDATA----------/----ALLOUABLEDATA -- / RATIO /-RESULT-//

NODE SECHAH LOAD AX/HX AY/HY AZ/HZ ACC HOH EOUA'I ION 123 ~

608H 0.00 49.27 0.00 772.80 1000000.06 0.06 PASS 0.00 0.00 0.00 OHE 456 608L 0.01 97.46 0.01 772.80 1000000.06 Q.13 " PASS 0.00 0.00 0. 00 OHE 456 608H 0.01 97.46 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 OHE 438 Y075T38 123 438L 0.00 27.32 0.00 772.80 1000000.06 ~ 0.04 PASS ~

0.01 0.02 0.02 ONE 123 438H 0.00 27.32 0.00 772.80 1000000.06 . 0.04 PASS 0.01 0.02 0.02 ONE 456 438L 0.01 54.06 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 OHE 456 438H 0.01 54.06 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 ONE

~ **" *

• FOLLSIIHG IS A SUHHARY OF THE CODE CHECKS PERFORHED ABOVE ******

• ALL 4 PIPIHG VALVE JOIHTS , THAT ARE CHECKED, PASSEO CODE CHECKS.

S SCHECK VALVE HORIZONT. ACCEL. FOR 3G S

PIPE VALVE CHECK PARA AXPFF 1.0000 LOADS ALL AYPFF 0.0001 LOADS AI.L AZPFF 1.0000 LOADS ALL AAF 1.5000 Jol ALL TRACE 2 EHD LOAD LIST 1000 2000 CHECK PIPE VALVE IHT EOU JOI-308 508 608 438 .

HOCH TWO CA ID rt g7 N 0 8

tA '8 0

'" P-DELTA / HPDS ~ *CALCULATIOH Ho. DC-0-02-HSC-36 *JOB HO 00002448118-OCT-1993 08:21:561PG 00070 *A AAAAAAIAttAAAAAAAAIAAAAAAAAAAAAAIIAAAAAAAAAAAAAAAAAAAAA STRUDL PIPIHG VALVE HEHBER CHECK ,TRACE 2, RESULTS *

• AA***illllll*AAAA*11*11*1***1*1*AAAAAIA*11**A**i*Alii JOB IO - FFS257 JOB TITLE CALCULATIOH Ho. DC-D-02-HSC-36 ACTIVE UNITS - LENGTH HEIGHT ANGLE TEHPERATURE TINE HASS IHCH KIPF RAD FAH SEC LBH

//-VALVE--/.TBLHAH-/.---"TITLE-----./CRITICAL/-HEHBER-/----------APPLIED DATA--------.-/--"ALLOMABLEDATA-.-./-.:RATIO./-RESUL'I-//

NODE SECNAH LOAD AX/HX AY/HT AZ/H2 ACC HOH 'EQUATION I

308 GATE1 1000 308L 41.19 0.00 41.'IO 1159.20 1000000.06 0.05 PASS 0.00 0.01 0.00 ONE 1000 30NI 41.19 0.00 41.10 1159.20 1000000.06 0.05 'ASS 0.00 0.01 0 ~ 00 OHE 2000 30BL 81.58 0.01 81.39 1159.20 1000000.06 0. 10 PASS 0.00 0.01 0.00 OHE 2000 308H 81.58 0.01 81.39 1159.20 1000000.06 0.10 PASS 0.00 0.01 0.00 ONE 508 GATE3 1000 508L 40.48 0.00 259.97 1159.20 1000000.06 0.23 . PASS 0.00 0.00 0.00 ONE 1000 50BH 40.48 0.00 259.97 1159.20 1000000.06 0.23 PASS 0.00 0.00 0.00 OHE 2OOO 5OBL 80.16 0.01 524.42 1159.20 1000000.06 0.46 PASS 0.00 0.01 0.00 ONE 2000 508H 80.16 0.01 524.42 1159.20 1000000.06 0,46 PASS 0.00 0.01 0.00 OHE 608 GATE2 1000 60BL 40.89 0.00 46. 15 1159.20 1000000.06 0.05 PASS 0.00 0.00 0.00 OHE

• • P-DELTA / HPDS "CALCULATIOH No. DC-D-02 HSC.36 *JOB NO 00002448~18.0CT.1993 b8:21:56'PG

~ ~ 00071

//.VALVE--/-TBLNAH-/-----TITLE-- ---/CRITICAL/-HEHBER-/--- ------APPLIED DATA----- ----/--- ALLONABLE DATA----/--RATIO-/ RESULT-//

AX/HX AY/HY AZ/HZ ACC HOH . . EQUATION NODE SECNAH LOAD 1000 608H 40.89 0;00 46.15 1159.20 1000000.06 0.05 PASS 0.00 0.00 0.00 ONE 2000 608L 80.98 0.01 90.76 1159.20 1000000.06 0:10 PASS 0.00 0.00 0.00 ONE 2000 60BH 80.98 0.01 90.76 1159.20 1000000.06 0. 10 PASS 0.00 0.00 0.00 ONE 438 Y075T38 1000 438L 39.75 0.00 45.44 1159.20 1000000.06 ~ 0.05 PASS 0.04 0. 12 0.07 ONE 1000 438H 39.75 0.00 iÃ.44 1159.20 1000000.06 0.05 PASS 0.04 0. 12 0.07 ONE 2000 438L 78.19 0.01 89.34 1159.20 1000000.06: 0.10 PASS 0.05 0.14 0.09 ONE 2000 438H 78.19 0.01 89.34 1159.20 1000000.06 0.10 PASS 0,05 0.14 0.09 ONE

• • • A****A* A SUNHART Of TNE CODE CHECKS PER fORHED ABOVE as***a***~

fOLLOUING IS ALL 4 PIPIHG VALVE JOIHTS, THAT ARE CHECKED, PASSED CODE CHECKS.

PIPING LOAn TYPES ALEVEL 11 701-BLEVEL 1000 TO 1001 CLEVEL 2000 TO 2001 DEAD 11 .

TNERHAL SECTION fR HS 2 0.0 1.0 OUTPUT DECIHAL 5 OUIPUT BT HEHBERS LOAD LIST 11 1 701 1000 TO 1001 2000 TO 2001 S

UNIT INCH POUND RAD LIST DISPLACEHENTS '6PIPEJOI' Q

h 8 'h h 0 Ln g 0

Attachment P This attachment ratios ancL valve showers str ess acceleration ratios from the as- desi

>which incluc9.es the tw o missing algal s is support thecomponents vrithin design that, are smell basis criteria allow.a&le limits Sheet S tres s Xnteraction Ratio Sheets 2%3 Horizontal Valve Acceleration Ratios Sheets 4%5 Vertical Valve Acceleration Ratios

• ~ P-DELTA / HPDS *CALCULATIOH Ho. DC-0-02-NSC-36

//-PIPIHG-/-TBLHAH-/--CODE--/---- ---/--NODE--/EGUATIOH/-- --APPLIED FORCES-----/--------STRESS DATA-- - --//--LOAD /-RESULT-//

HEHBER SECNAII VERSIOH FX/HX FY/HY FZ/HZ PRES/HOHP TOT/AL1I RATIO 31.1-4 0.0 0.0 0.0 826.22 7368.35 'ASS 0.0 0.0 0.0 6542.13 37500.00 0 '0

~ SUHHARY OF THE PIPING HEHBER CHECKS BY EQUATIOH EQUAT IOH NAXINUH IR HENBER LOAD 31.1.1 0.4766 98 13 lure/2aaT(ON ghee'N 31.1-2HU 0.4753 98 1000 31.1-2E 0.4770 44 2000 31.1-3 0.0973 30 14 31.1-4 0.0973 98

• ~~*~~**~* FOLLOHIHG IS A SUHHARY OF THE CODE CHECKS PERFORHED ABOVE +~*""**~*~

ALL 54 PIPE HEHBERS, THAT ARE CHECKED, PASSED CODE CHECKS.

SCHECK PIPE BREAK LOC ALL SCHECK PIPE CRACK LOC ALL SCHECK PIPE BRANCH DISPL ALL PRINT PIPE ADD STRESS IHT FACT C/l +,

Al Et lD rt o vg ID 0

W ft

"" P-DELTA / MPDS *CALCULATIOH Ho. DC-D-02-MSC-36 *JOB HO 00002449118-OCT-1993 08:23:19*PG 00097 AIAA*l*l*AIA111*111111*IAIAIAAI*11111IAI*1*AAAAA**AAII

" STRUDL PIPIHG VALVE MEMBER CHECK ,TRACE 2, RESULTS

• IIAIAAAAAAAIAAAAAIIIAAIAAAAAAAAAAAAAAAAAAIAAIIAAAAAAAA JOB ID - FFS257 JOB I I TLE - CALCULATIOH Ho. DC-D-02-MSC-36 1

ACTIVE UNITS - LEHGTH IIEIGNT AHGLE TEMPERATURE TIME MASS IHCH KIPF RAD FAH SEC LBM

//-VALVE--/-TBLHAM-/------TITLE-----/CRITICAL/-MEMBER./-------- -APPLIED DATA-- -------/----ALLOWABLEDATA----/--RATIO./ RESULT //

HOOE SECHAM LOAD AX/MX AY/MY AZ/MZ ACC MOM EQUATION 308 GATE i 123 308L 0.00 27.39 0.00 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 OHE 123 308M 0.00 27.39 0.00 772.80 1000000.06 0.04 . PASS 0.00 0.00 0.00 OHE 456 308L 0.01 54 '4 0.01 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 ONE 456 308M 0.01 54.24 0.01 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 OHE 508 GATE3 123 508 L 0.00 27.39 0.03 772.80 1000000.06 0. 04 PASS 0.00 0.00 0.00 ONE 123 508M 0.00 27.39 0.03 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 ONE 456 508L 0.01 54.24 0.06 772.80 1000000.06 0.07 PASS 0.00 0.01 0.00 OHE 456 . 508M 0.01 54.24 0.06 772.80 1000000.06 0.07 PASS 0.00 0.01 0.00 OHE 608 GATE2 123 608 L 0.00 51.23 0.00 772.80 1000000.06 0.07 ~ PASS 0.00 0.00 0.00 OHE

• • P-DELTA / HPDS *CALCULATIOH Ho. DC-D-02 HSC.36 *JOB HO 00002449~18-OCT-1993 08:23:19'PG,00098

//-VALVE--/-TBLHAH-/------TITLE-- -/CRITICAL/-HEHBER-/----------APPLIEDDATA------ --/;-- ALLOMABLE DATA----/--RATIO/-RESULT-//

NODE SECHAH LOAD AX/HX AT/HY AZ/HZ ACC HOII ~ EQUATION 123 608H 0.00 51.23 0.00 772.80 1000000.06 0.07 PASS 0.00 0;00 0.00 OHE 456 608L 0.01 101 ~ 30 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 OHE 456 '08H 0.01 101.30 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 ONE

\

438 Y075T38 123 'QBL 0.00 27.64 0.00 772.80 1000000.06 0.04 PASS 0.01 0.02 0.02 OHE 123 438H 0.00 27.64 0.00 772.80 1000000.06 0.04 PASS 0.01 0.02 0.02 ONE 456 43BL 0.01 54.70 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 OHE 456 438H 0.01 54.70 0.0'I 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 OHE

• • • *~** FOLLOMIHG IS A SUHHARY OF THE CODE CHECKS PERFORHED ABOVE *" * ****

ALL 4 PIP IHG VALVE JOIHTS , THAT ARE CHECKED, PASSED CODE CHECKS.

S SCHECK VALVE HORIZOHT ~ ACCEL. FOR 30 S

PIPE VALVE CHECK PARA AXPFF 1.0000 LOADS ALL ATPFF 0.0001 LOADS ALL AZPFF 1.0000 LOADS ALL AAF 1.5000 JOI ALL TRACE 2 EHD LOAD LIST 1000 2000 CHECK PIPE VALVE IHT EQU JOI-308 508 &08 438 .

HOCH TMO

• 1 P.DELTA / HPDS *CALCULA'IIOH Ho. DC-0-02 HSC.36 *JOB NO 00002449118-OCT-1993 08:23: 19*PG 00099

• ill*11*1**111111111*1*11*1111111*1*1111111111*1111*1*

• STRUDL PIPING VALVE MEHBER CHECK ,TRACE 2, RESULTS

• 111111111*la*1*111111**1**lllll11111**111111111*1*111*

JOB ID - FFS257 JOB Tl'ILE - CALCULATIOH Ho. DC-D-02 HSC-36 AC'IIVE UHITS . LEHGTH 'EIGHT ANGLE TEHPERATURE TIME HASS INCH KIPF RAD FAH SEC LBH

//-VALVE--/-TBLNAH/------TITLE------/CRITICAL/HEMBER-/------- --APPLIED DATA--------- /----ALLOMABLE DATA ---/--RATIO-/-RESULT-//

NODE SECHAH LOAD AX/HX AY/HT AZ/HZ ACC HOH ,  : EQUAllOH 308 GATE1 1000 308L 41.19 0.00 41.10 1159.20 1000000.06 0.05 PASS 0.00 0.01 0.00 OHE 1000 ~ 308M 41.19 0.00 41.10 1159.20 1000000.06 0+05 PASS 0.00 0.01 0.00 OHE 2000 ,308L 81.56 0.01 81.39 1159.20 1000000.06 0.10 PASS 0.00 0.01 0.00 ONE 2000 308H 81.56 0;01 81.39 I 159.20 1000000.06 0.'10 PASS 0.00 0.01 0.00 ONE 508 GATE3 1000 508L 40.63 0.00 308.47 1159.20 1000000.06 0.27 PASS 0.00 0.00 0.00 ONE 1000 508H 40.63 0.00 308.47 1159.20 1000000.06 0.27 PASS 0.00 0.00 0.00 ONE 2000 508L 80.47 0.01 620.58 1159.20 1000000.06 0.54 PASS 0.00 0.01 0.00 OHE 2000 508H 80.47 0.01 620.58 1159.20 1000000.06 0.54 PASS 0.00 0.01 0.00 OHE 608 GATE2 1000 608L 40.76 0.01 48.66 1159.20 1000000.06 0.05 PASS 0.00 0.00 0.00 ONE P't+

CA Io rt 0 g

• • P-DELTA / HPDS *CALCULATION No. DC-D.02-HSC-36 *JOB NO 00002449*18-0CT.1993 08:23:19*PG 00100

// VALVE--/-TBI.HAH-/------TITLE------/CRITICAL/-HEHBER-/" - -----APPLIED DATA----------/-- -ALLOMABLE DATA.-.-/--RATIO-/.RESULT'-//

HODE SECNAH LOAD AX/HX AY/HY AZ/M? ACC NOH EOUATIOH 1000 608H 40.76 0.01 48.66 1159.20 1000000.06 PASS 0.00 0 F 00 0.00 ONE 2000 60SL 80.72 0.01 95.58 1159.20 1000000.06 0.11 PASS 0.00 0 F 00 0.00 ONE 2000 60BH 80.72 0.01 95.58 1159.20 1000000.06 0.11 PASS 0.00 0.00 0 F 00 ONE 438 Y075T38 1000 438L 41.17 0.00 45.88 1159.20 1000000.06 0.05 PASS 0.03 0.11 0.07 ONE 1000 438H . 41.17 0.00 45.88 1159.20 1000000.06 0.05 PASS 0.03 0.11 0.07 ONE 1000000.06'.05 2000 438L 80.98 0.01 90.20 1159.20 1000000.06 0.10 PASS 0.05 0.13 0.09 ONE 2000 438H 80.98 0.01 90.20 1159.20 0.10 PASS 0.05 0.13 0.09 ONE

• "** ~ FOLLNING IS A SUHHARY OF THE CODE CHECKS PERFORHEO ABOVE

• • ~ *"

ALL 4 PIPIHG VALVE JOIHTS , THAT ARE CHECKED, PASSED CODE CHECKS.

'IPIHG LOAD TYPES ALEVEL 11 701-BLEVEL 1000 TO 1001 CLEVEL 2000 TO 2001 DEAD '11-THERHAI. 'I SECTIOH FR NS 2 0.0 1.0 OUIPUT OECIHAL 5 OUIPUT BY HEHBERS LOAD LIST 11 1 701 1000 TO 1001 2000 TO 2001 S I UNIT INCH POUND RAD LIST DISPLACEHEN'IS I6PIPEJOII U) ',

I (D

fD I ft (

Vi I I

0 ~

ATTACHMENT 3 TO AEP'NRC'1184H INFORMATION REGARDING TEMPORARY MODIFICATION 2-93-015 to AEP:NRC:1184H Page 1 As requested in the cover letter for inspection report 50-315/316 93012 (DRS),

this attachment provides the results of our investigation into the issues associated with Temporary Modification (TM) 2-93-015. This TM installed current-to-current (I/I) converter modules in the control circuitry for the Unit 2 East and West Main Feedwater Pumps. The purpose of the temporary modification was to isolate the control circuitry from the presence of signal grounds in the field wiring that were interfering with proper operation of the D/P slave controller. The modification was processed as a temporary modification because the source of the grounding problem could not be resolved with the unit on line.

In addition, the control room portion of the feedwater control system circuitry has been scheduled for replacement in 1994 as part of the reactor protection system upgrade project.

Two design errors occurred that were found during the post-installation checkout process. Specifically, the design as released by Plant Engineering, had the I/I wired in backwards (i.e., input and output reversed), and a necessary 100 ohm input resistor omitted. These errors were discovered during post-installation circuit checks, prior to post modification testing (PMT) and prior to placing the affected circuitry back into operation. Since the errors were identified and corrected prior to PMT, no condition report was required. Although not required, a condition report was subsequently generated to document our investigation.

The TM did not involve safety related equipment, and therefore the quality assurance requirements of 10 CFR 50 Appendix B and the requirements of ANSI-N45.2.11 do not specifically apply. However, since the temp mod process used at the plant is common to both safety-related and non-safety related changes, our review of this event was conducted to consider its implications for safety-related temp mods.

The following conclusions were drawn from our investigation of this event:

The design error occurred as the result of attempting to copy the design details of an existing I/I circuit from another approved plant drawing to the drawing markup which was used to serve as the installation drawing for the temp mod..While possible to correctly identify the input and output leads on this drawing, the input/output lead designations (W,X & D,F) were inadvertently transposed during their transfer to the mark-up. The 100 ohm input resistor was overlooked and was not transferred to the mark-up.

2. A ~endor document detailing the design and installation requirements for the I/I was available, but not as an 'pproved plant document. The engineer chose not to use it, because he felt confident that he was adequately familiar with the hardware and that reference to the vendor information would be unnecessary. Use of the document would likely have allowed the engineer to detect the transposition error, as well as the omission of the input resistor.

to AEP:NRC:1184H Page 2

3. The engineer assigned to develop and implement the temp mod request was adequately qualified to perform this task.

4. The supervisor who reviewed the marked-up drawings was also adequately qualified to perform this function. However, the supervisor became involved with the development of the drawing markups. As a result, his review was not independent, which may have compromised his ability to detect the design errors.

5. The temp mod procedure does not clearly convey design review requirements, The Engineering Supervisor initialed the drawings in the temp mod package indicating his concurrence. However, this review is not a step required by the temp mod procedure.

6. The Plant Engineering review required by the temp mod procedure calls for "...a review of the request to ensure that it is needed, correct, practical, and that it accomplishes the intended purpose."

It does not require a technical verification of the design change documents (i.e. drawing mark-up in this case) that are used to implement the request.

This event has identified a weakness in the temp mod procedure design review requirements. In order to prevent recurrence of an incident of this nature, the TM procedure, PMP.5040.MOD.001, will be revised to strengthen the requirements for verifying the technical accuracy of the design. The revision will clearly identify expectations and responsibilities for the review. These revisions will be completed by April 30, 1994.

The unapproved vendor technical document has been forwarded to the Vendor Information Control System section and is currently in the approval cycle.

There are currently a total of 33 temporary modifications installed at the Cook Nuclear Plant. After a preliminary review, two modifications were identified whose failure could cause a unit trip. Plant Engineering will conduct an independent design verification of these two temp mods. This review will be completed by January 31, 1994.

In the interim, the Plant Engineering Superintendent is taking precautions by closely scrutinizing design complexity of all temp mods that are submitted for Plant Engineering review, and will require independent technical verification if deemed necessary.

ACCELERATED DISTRIBUTION DEMONSTPATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

CESSION NBR:9401110333 DOC.DATE: 94/01/04 NOTARIZED- YES DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co.

RECIP.NAME RECIPIENT AFFILIATION MARTIN,J.B. Document Control Branch (Document Control Desk)

SUBJECT:

Submits response to violations noted in Insp Repts 50-315/93-12 & 50-316/93-12.Corrective actions:licensee failed to perform evaluation to determine changes made to feedwater pump speed control sys by temporary mod.

DISTRIBUTION CODE: IEOID COPIES RECEIVED:LTR i ENCL / SIZE:

TITLE: General (50 Dkt)-Insp Rept/Notice of Vi&o ation Response NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 PD 1 1 HICKMAN,J 1 1 INTERNAL: AEOD/DEIB 1 1 AEOD/DSP/ROAB 1 1 AEOD/DS P/TPAB 1 1 AEOD/TTC 1 1 DEDRO 1 1' NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 NRR/DRIL/RPEB 1 1 NRR/DRSS/PEPB 1 1 NRR/PMAS/ILPB1 1 1 NRR/PMAS/ILPB2 1 1 NUDOCS-ABSTRACT 1 1 OE DIR 1 1 OGC/HDS2 1 1 1 1 RES/HFB 1 1 6 FILE 01 1 1 EXTERNAL EGGG/BRYCE E J ~ H~ 1 1 NRC PDR 1 1 NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 22

~k Cl

Inniana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 Z

lNDfANA NICHIGAM POMP'EP:NRC 1184H 10 CFR 2.201 Donald C. Cook Nuclear Plant Units 1 and 2 Docket: Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 NRC INSPECTION REPORTS NO. 50-315/93012 (DRS)

AND 50-316/93012 (DRS)

REPLY TO NOTICE OP VIOLATION I

U. S, Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Attn: Mr. J. B. Martin January 4, 1994

Dear Mr. Martin:

This letter is in response to a letter from G. E. Grant dated November 24, 1993, which forwarded a Notice of Violation associated with a System Based Instrumentation and Control Inspection conducted by Zelig Falevits and others of your office during August 17 through September 28, 1993. The violations are associated with errors in a setpoint calculation for refueling water storage tank level instrumentation and for the lack of an unreviewed safety question determination for Temporary Modification 2-93-015, which installed an I to I converter in the Unit 2 feedwater pump control circuitry.

Our reply to the Notice of Violation is contained in Attachment 1.

We were also requested to respond to open and unresolved items from the inspection report. Our response is contained in Attachment 2.

In addition, we were requested to address non-cited deficiencies associated with Temporary Modification 2-93-015. Our response to this request is contained in Attachment 3.

This letter is submitted pursuant to 10 CFR 50.54(f) and, as such, an oath statement is attached.

Sincerely, E. E.

p Fitzpatrick Vice President dr Attachments /((()

9401110333 940104 PDR ADOCK 05000315 8 PDR

C Mr. J. B. Martin 2 AEP: NRC: 1184H cc: A. A. Blind G. Charnoff T. E. Murley - NRC NFEM Section Chief NRC Resident Inspector J. R. Padgett

y7 STATE OF OHIO)

COUNTY OF FRANKLIN)

E. E. Fitzpatrick, being duly sworn, deposes and says that he is the Vice President of licensee Indiana Michigan Power Company, that he has read the forgoing response to the NOTICE OF VIOLATION FOR NRC INSPECTION REPORTS NO. 50-315/93012 (DRS) AND 50-316/93012 (DRS) and knows the contents thereof; arid that said contents are true to the best of his knowledge and belief.

Subscribed and sworn to before me this ~g day of 19 ~p NOTAR PUBLIC AITA D. H!LL NOTARY PUCLIC. STATE Of OH;.

ATTACHMENT 1 TO AEP'NRC:1184H REPLY TO NOTICE OF VIOLATION to AEP:NRC:1184H Page 1 NRC Violation I (Severity Level IV)

"10 CFR 50, Appendix B, Criterion III, states, in part, that measures shall be established to assure that the design basis is correctly translated into specifications, drawings, procedures, and instructions and that the design control measures shall provide for verifying the adequacy of the design.

Contrary to the above, on September 1, 1993, the team noted that:

a. RWST instrumentat'ion loop setpoint calculation No. 1-2-I9-03, dated August 25, 1993; (1) Erroneously derived the setpoint uncertainty value based on the use of Model N-E13 RWST transmitters. However, the installed RWST level transmitters were Model E13DM-HSAH1.

(2) Provided no justification for using transmitter elevation 599'3" to derive the setpoint uncertainty value in the setpoint calculation.

(3) Did not consider the error effects of the velocity head of the ECCS pump flow (Safety Injection and Residual Heat Removal pumps) during design basis accidents.

b. Flow diagram OP-1-5144-13, "Containment Spray System Unit 01,"

incorrectly identified RWST level transmitter ILS-950 as having a minimum level alarm at 638'll". However, the design basis setpoint value was 637I 2l1 tl Response to Violation I dmission or Denial of the lie ed Violatio Indiana Michigan Power admits to the violation as cited in the NRC Notice of Violation.

2. easons for the Violatio The examples of the violation will be addressed individually.

l.a.(1) The cause of the violation is attributed to lack of attention to detail in cross-checking the specific details in the documentation. The instrument data sheet used was from a slightly different model (N-E13 versus the correct model E13 of the same vendor). Both types of instruments are used in the plant and have nearly identical performance characteristics. The correct model data sheet was compared to the incorrect model data sheet and minor differences of the uncertainty terms were found which required the in'ome to AEP:NRC'1184H Page 2 calculation to be revised. However, no significant differences were found that changed the end results of the calculation adversely. Therefore this error did not have an adverse effect on safety.

l.a. (2) As noted in Revisions 2 and 4 of the calculation, the elevations used in the calculation were based on field walkdown measurements, which were taken in 1979. During the inspection, the elevations were remeasured but did not exactly'atch the previous field walkdown measurements.

Because of the time that has elapsed, it was not possible to positively identify the reason for the discrepancy. It is noted, however, that the differences between the as-built measurements taken during the inspection and the measurements used in the calculation differed only slightly. The differences in elevation between the as-built heights and the heights used in the calculation varied between 1 inch and 2.5 inches, compared to an instrument span of 363 inches. The worst case error in the non-conservative direction (as-built lower than calculation) was 2.5 inches, which equates to an error of approximately 0.7%. In other words, the RWST indication and alarm will occur 0.7$ lower than actual level.

However, since there is considerable margin built into the alarm setpoints (approximately 21$ for the low alarm and approximately 7% for the,low-low alarm), this measurement error had no adverse effect, on safety.

l.a. (3) The alarms and trips associated with the RWST level instrumentation include both high and low level type functions. There are two high level functions. The High Level Alarm is used to ensure the operators do not inadvertently overflow the RWST when filling, and a Minimum Level alarm is used to alert the operators that Tech Spec required minimum RWST volume requirements are being encroached. There are also two low level functions. The Low Level alarm is. used to alert the operator that RWST inventories are approaching a low level at which the operator should begin to transfer to ECCS recirculation mode, and a Low-Low Level alarm and RHR pump trip is used to alert the operator that RWST inventory is depleted and to protect the RHR pump from damage due to low NPSH.

Because the RWST level transmitter is tapped off the ECCS suction line at the bottom of RWST tank, velocity head effects can be induced when the ECCS pumps are running. The high level functions are not affected by this because these functions are only used when the ECCS pumps are not running.

The low level functions are affected because the ECCS pumps are running when they are required to function. However, to AEP:NRC:1184H Page 3 computation of the velocity head effect shows it will only affect the low level functions in a conservative manner. The velocity head effect induces a negative bias that results in a level indication that is lower than actual level. The effect on the low level alarms and RHR pump trip functions is that they will occur sooner and therefore does not jeopardize the RHR pump or plant safety.

Velocity head effect was not addressed due to unawareness of its influence on this application. Criteria for when this effect is to be considered were not included in the procedure used for the preparation of this calculation. It should be noted that most tank level instruments are tapped on the side of the tank and are therefore not affected by the velocity head effects of the tank suction line.

l.b. Setpoint information for the Cook Nuclear Plant is controlled through the Plant Setpoint Document, rather than through the flow prints. Because of this, there was no systematic process to have setpoints placed on flow prints, or to update the flow prints if the setpoint changed. The setpoints displayed on the drawings are for reference and are used for understanding the drawing only.

3. Corrective Actions Taken and Results Achieved The examples of the violation will be addressed individually.

l.a.(1) The correct model data sheet has been compared with the incorrect model instrument data sheet and no significant differences were found. The calculation will be revised to incorporate the correct instrument data sheet information.

l.a.(2) The calculation has been revised to reflect the correct as-built elevations for the RWST level transmitters.

l.a.(3) The calculation has been revised to address the velocity head effects. Review of other tank level applications found the CST tank level to have a similar velocity head effect which was not addressed. The CST calculation will also be corrected.

l.b. Drawing OP-1-5144 and OP-2-5144 were revised to reflect the correct setpoints.

to AEP:NRC:1184H Page 4

4. ctions Taken to Avoid Further Vio at ons As a general comment, it is noted that numerous instrument setpoint calculations are being updated as part of the Reactor Protection and Control Systems Upgrade Project which will be implemented during the 1994 refueling outages. The specific examples of the violation are addressed individually, below.

l.a.(1) Training sessions were held for Corporate ZSC engineers which emphasize that self-checking and engineering reviews and verification are expected to be such that errors in instrument model numbers and similar documentation data is discovered and corrected prior to document issue.

l.a.(2) A sampling of safety-related instruments will be checked to verify correct as-built elevations are incorporated into setpoint calculations. The sampling will be completed by May 31, 1994. Further preventive measures will be established depending on the results of the sampling.

l.a.(3) The engineering guide governing setpoint calculations was revised on November 15, 1993, to require that process measurement effects, such as velocity head effects, are considered, as necessary, as part of the calculation preparation.

l.b. Since the intent of setpoint information on flow prints is only to help in the understanding of the drawing, a note will be added to all affected flow prints which states:

"Caution} The setpoints indicated are provided only to assist in understanding the drawing. Refer to appropriate setpoint control document for actual device setpoints." The notes will be added by July 6, 1994.

5. Date When Full Com liance will be Ach eyed The examples of the violation will be addressed individually.

l.a.(l) Full compliance will be achieved by January 10, 1994, when the calculation is revised to reflect the correct model performance characteristics.

l.a.(2) Full compliance was achieved on November 5, 1993, when the calculation was revised to incorporate the as-built transmitter elevations.

to AEP:NRC:1184H Page 5 1.a.(3) Full compliance was achieved on November 5, 1993, when the calculation was revised to include consideration of velocity head effects. The calculation for the CST will be revised by January 10, 1994, to correct the similar deficiency identified during the review of the RWST calculation.

l.b. Full compliance was achieved on September 15, 1993, when the affected Unit 1 and 2 OP drawings were revised.

to AEP:NRC:1184H Page 6 NRC Violation II (Severity Level IV)

"10 CFR 50.59 states licensees may make changes to the facility as described in the safety analysis report without prior Commission approval unless the change involves an unreviewed safety question. A written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question is required.

Section 10.5.1.1 of the UFSAR states that the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators. In addition, Section 14.1.9 analyzed a loss of normal feedwater from pump failures which could result in a reduction of the secondary system to remove heat generated in the reactor core.

Contrary to the above, on April 7, 1993, the licensee failed to perform an evaluation to determine that changes made to the feedwater pump speed control system by temporary modification 2-93-015 did not involve an unreviewed safety question."

Response to Violation II dmission or Denial of the Alle ed Violation Indiana Michigan Power denies the violation as cited in the NRC Notice of Violation.

2. easons for Denial of the Violation At the Cook Nuclear Plant, temporary modifications undergo a screening to determine if an unreviewed safety question determination is required to be performed pursuant to 10 CFR 50.59. The process we use is based on the guidance of NSAC 125 (June 1989), entitled "Guidelines for 10 CFR 50.59 Safety Evaluations." This document was prepared ]ointly by the Nuclear Management and Resources Council (NUMARC) and the Nuclear Safety Analysis Center of the Electric Power Research Institute (EPRI).

The inspection report (page 14) states:

"By failing to recognize that the speed control system was described in the UFSAR, the licensee concluded that 10 CFR 50.59 was not applicable, therefore, no safety evaluation was performed.

The licensee's failure to perform a safety evaluation is considered to be a violation of 10 CFR 50.59."

to AEP:NRC:1184H Page 7 We disagree with the statement that the speed control system is described in the UFSAR. The UFSAR (Section 10.5.1.1) specifically states that "the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators." There is no description of the circuitry provided in this statement.

NSAC 125 recognizes that changes made to the facility may implicitly impact the UFSAR. For these cases, NSAC 125 states that:

"If the SSC {structure, system, or component) is part of a larger SSC described in the SAR and ifthethe function change affects the design, of the larger function, or method of performing SSC AS DESCRIBED IN THE SAR (emphasis added) then a safety evaluation is required."

As discussed in the UFSAR, the variable speed turbine driven main feedwater pumps are designed to provide the required feedwater flow to the steam generators. The temporary modification only added a device to isolate noise in the speed control circuitry. Neither the function of the main feedwater pumps as described in the UFSAR (to provide feedwater flow to the steam generators) nor the method of performing the function as described in the UFSAR (with variable speed pumps) were impacted by the temporary modification. Additionally, since there was no description of the feedwater control system in the UFSAR, neither the design of the control system nor the design of the feedwater system as a whole as described in the UFSAR were impacted by the change. Based on these considerations, we conclude that no unreviewed safety question determination was required pursuant to 10 CFR 50.59.

As described in NSAC 125, "The purpose of 10 CFR 50.59 is to preserve the original licensing basis in the information submitted to the NRC as part of the application for an operating license and in the final safety evaluation report (SER) issued by the NRC staff. The NRC relies on this information to conclude that an operating license can be issued without undue risk to the health and safety of the public.

This regulation allows the licensee to make changes without prior NRC approval while maintaining the licensing basis. It defines conditions that must be met in determining if prior regulatory review is needed."

The level of detail included in the UFSAR regarding the feedwater system is relatively small. This is commensurate with the fact that the system is non-safety related. The NRC staff review of the system during the original licensing of the plant would be expected to be of a different level of detail than for those systems which are safety related. Since the staff review did not rely on a detailed description of the speed to AEP:NRC:1184H Page 8 control system in order to conclude that the feedwater system was adequate from a safety perspective, it is not reasonable to conclude that to the circuitry. It is 10 CFR 50.59 would be applicable to changes noted that Section 14.1.9 of the UFSAR, entitled "Feedwater System Malfunctions," analyzes a complete loss of feedwater (due to no specific reason) and concludes that:

"a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves."

It is also noted (as acknowledged in the inspection report, page 14) that failure of the I to I converter would have the same effect as the failure of the hand/auto station already in the circuit. In other words, the change did not introduce a new failure mode into the system.

ATTACHMENT 2 TO AEP:NRC:1184H RESPONSE TO OPEN AND UNRESOLVED ITEMS

to AEP;NRC:1184H Page 1 The cover letter for Inspection Report 50-315/316 93-012 (DRS) requested we provide a written response to open and unresolved items in the inspection report. There is only one item in this category, Unresolved Item 315/316 93012-04(DRS). This unresolved item involves a calculation performed to verify that an instrument sensing line associated with the Unit 2 auxiliary feedwater system was adequately supported. The inspection report states that:

"The licensee was in the process of performing a calculation to determine whether the sensing line installation was adequate. Pending review of the calculation, this item is considered unresolved."

The subject calculation was completed following the inspection, and, at the inspector's request the information,was mailed in an overnight package to NRC Region III on October 20, 1993. The information, documented in a memo from S.

J. Jarrett/H. P. Damasco to M. S. Ackerman dated October 19, 1993, follows.

Z AMERICAN ELECTREC

.Qe PCMR-Date October 19, 1993 Subject Cook Nuclear Plant NRC SBICI Inspection Assistance NEDS Review of As-Found Conditions From S. J. Jarrett/H. P. Damasco To M. S. Ackerman As per your request, the Nuclear Design - Structural 5 Analytical

.Section (NEDS) has reviewed the as-found condition detected by an NRC inspector during a NRC SBICI Inspection. The 3/8"qb, 1/2"$ ,

and 3/4"Q downstream low pressure instrument lines associated with 2-FFS-257, and branch lines off the feed water header line 2-FW-48 were believed to have piping/tubing spans that exceeded Alternate Analysis Criteria.

Based on the above findings, Nuclear Engineering Site Design Section (NESD) performed a detailed walkdown of the piping system (see Attachment A). In order to satisfy the seismic overlap criteria, NESD included additional piping and supports beyond the area addressed by the NRC inspectox. In so doing, they found that two support components in the overlap region of the continuation line were missing o Nuclear Design - Mechanical Section (NEDM), performed a preliminary assessment of the piping system as found by the NRC inspector and concluded in their E-Mail, dated September 17, 1993 to Stan Farlow (see Attachment B), that the minor deviations in some support span lengths have an insignificant effect on the operability and/or the design basis of the instrument line.

NEDS performed the as-found and the as-designed DBE, seismic evaluations on the piping system, in Calculation No.

DC-D-02-MSC-36, by using walkdown information (see Attachment A),

and Ebasco/P-Delta (E/PD) STRUDL integrated computer program.

The piping stresses and the valve accelerations in these analyses were found to be well within the design basis criteria allowable limits.

Attachments C through F highlight the very small stress interaction ratios, and vertical and horizontal valve acceleration ratios resulting from the E/PD STRUDL analyses. These analyses were performed to confirm the NEDM preliminary engineering review.

hta-System

NRC SBICI Inspection Assistance October 19, 1993 Page 2 Although the piping system in the as-found condition meets the design basis criteria limits, the two missing components are being installed on the piping system to reflect similarity of pipe supports on identical piping systems in the vicinity, and further to conform with good engineering practice.

It is NEDS understanding that Job Order No. C19393 has'een initiated by Plant Maintenance to install the two missing components.

If you have any further questions please, contact these writers at extension 3157.

W y'C.

. Jarret P. Damasco

/

Approved by Pd N. Ruccia, Manager Nuclear Design - Structural h Analytical Section

/ski Kc: Ro C. Armstrong/S. P. Hodge - w/o attachment A. K. Dey>> w/o attachment S. K. Farlow - w/o attachment R. C. Carruth - w/o attachment T. H. Cummings - w/o attachment

. Attachment A Copy of Field.vralk from NESD

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Attachment B Sheet 1 of 1 To STAN K. FARLOHQNEDQAEPSC ROBERT C. CARRUTHQNEDQAEPSC Thomas H CummingsQSDOQCOOK,STEVE P. HODGEQNEDQAEPSC m AMIYA K. DEYQNDGQAEPSC

Subject:

NRC Inquiry Relative To 2-FFS-257 Date: Friday, September 17, 1993 13:44:19 EDT Attach:

Certify:

Forwarded by:

Instrument Line 2-FFS-257 Itheunderstand that an NRC inspector has noted the piping line associated Instrument g2-FFS-257 to be under supported as compared to MDS-601 with design standard.

Our findigs are given below:

The Standard MDS-601 was issued in Feb.1988 and I believe this instrument line was installed much earlier. This system and other instrument piping/tubing systems were installed per EDS Corp. developed guideline titled"Field Fabri-cated Support Guide For Small Diameter Pipes and Copper Tubing".This guide-line was included as a section in a larger guideline known as the Alternate Analysis Criteria( AAC ).

The AAC provides cookbook type directions in supporting piping/tubing systems and is very conservative.Our experience, with pxping/tubing systems analysed and supported per AAC,indicates that:

• these systems generally are well designed and meet applicable code allow-ables easily when computer analysis is performed to verify system adequacy where limitations of the AAC have been marginally exceeded.

instrument line being reviewed is 1/2 in. sch.80 pipe upto a TEE nection and then reduces to 3/8 in.sch. pipe sections.Our walkdown review this systems indicates minor deviations xn some span lenghths which we believe have no significance on the operability and/or the design basis of this insturment line.

If you need additional information, please contact me at 3860.

Thanks, C Amiya Q

Attachment C Metho ciolo Method. us eel to analyze the p9yf ng system xn Calculation No DC-D-02-MSC-36

Attachment C Sheet 1 of 1 AEPSC PAGE 6.1 DONALD C. COOK NUCLEAR PLANT, UNIT 2 6~0 METHODOLOGY The piping system has been rigorously analyzed using walkdown information and Ebasco/P-Delta (E/PD) STRUDL integrated computer program. Additional hand calculations, where necessary, have been performed to develop problem input data and/or to justify acceptance of the piping system.

The information relevant to the final output is given below:

COMPUTER PROGRAM & VERSION:

E/PD STRUDL Version 0193 VAX VMS 5.5 JOB No. RUN DATE 6 TIME COMMENTS 2448 10-18-93 08:21 AF run - see comm. below 2449 10-18-93 08: 23 DSN run - see comm. below ADDITIONAL COMMENTS:

Job no. 2448 is the run simulating the as-found condition with Design Basis Earthquake response spectra.

Job no. 2449 is the as-design run that includes the two missing pipe clamps, and using the Design Basis Earthquake response spectra.

NUCLEAR DESIGN GROUP CALCULATION No. DC-D-02-MSC-36 i~j/8'ATE BY DATE

ACtaakxment D Px obl.em Description

Attachment D Sheet 1 of 1 AEPSC PAGE 7 ~ 1 DONALD C. COOK NUCLEAR PLANT, UNIT 2 7.0 PROBLEM DESCRIPTION GENERAL DESCRIPTION:

Analysis on the as-found 3/8", 1/2", 3/4" downstream low pressure instrument header and branch lines off feedwater line 2-FW-4'8 to address overspan concerns raised by an NRC inspector.

NUMBER OF SUPPORTS:

8 (including the two missing unistrut clamps)

PRXMARY BUILDING:

Turbine - Area 2T4 ELEVATION OF MAIN.PIPING:

593'-0" MAJOR EQUIPMENT:

None NUCLEAR DESIGN GROUP CALCULATION No. DC-D-02"MSC-36 gn lg/yy BY DATE CHK. DATE

Attach. ment E TM.s ratios at tach.

and.

mentvalve sh.ops stx ess acceleration ratios fx om th.e as- fpunt anal sis wh.ich. emclucles th.e two missing supp ort the rvithin comp onents basis d.esign that axcriteria e vrell allows ah le limits S h.eet Stress Interaction Ratio Sh.eets 28c3 H ori zontal Valve Acceleration Ratios Sheets 455 Vertical Valve Acceleration Ratios

"* P.DELTA / MPDS "CALCULATIOH No. DC.D-02-MSC.36 *JOB NO 00002448*18'OCT.1993 08:21:56*PG 00066 ~

I

//-PIPING-/-TBLNAM-/- CODE--/

MEMBER SECHAM VERSIOH

"""/- NODE--/EQUATIOH/---- -APPLIED FORCES-----/--------STRESS DATA-.-

FX/MX FY/MY FZ/MZ PRES/HONP TOT/ALU

-- .-//--LOAD--/-RESUL'I-//

RATIO 31.1-4 0.0 0.0 0.0 0.0 0.0 0.0 826.22 6509.64 37500.00'ASS 7335.86 0.20 PIPIHG

'UMMARY OF THE MEMBER CHECKS BY EQUATION EQUATION MAXIMUM IR MEMBER LOAD 31.1-1 0.4748 98 13 31.1-2NU 31.1-2E 0.5887 0.5290 .

44 44 1000 2000 pl TCR/IQVIOk ~yS05 31.1-3 0.0979 30 14 31.1.4 0.0979 98

• • ~**"*~*~FOLLOHIHG IS A SUMHARY OF THE CODE CHECKS PERFORMED ABOVE *******~*~

ALL 54 PIPE MEMBERS, THAT ARE CHECKED, PASSED CODE CHECKS.

SCHECK PIPE BREAK LOC ALL SCHECK PIPE CRACK LOC ALI.

SCHECK PIPE BRANCH DISPL ALL PRINT PIPE ADD STRESS INT FACT rid 0 rt lD, Q rt O M EI Q

0 W rt

• • P DELTA / MPDS *CALCULATIOH Ho. DC D.02-MSC-36 *JOB HO 00002448*18-OCT-1993 08:21:56iPG 00068 *

~

ii*iili*iiiiiiiii11***il**iii*liiliiili*l*i*iiii*iiii*

i STRUDL PIPING VALVE MEHBER CHECK ,TRACE 2, RESULTS

• i*iii****i******i*****i**iil***i****i*lii******i*****

JOB ID - FFS257 JOB TITLE - CALCULATIOH Ho. DC-D-02-MSC-36 ACTIVE UNITS - LEHGTH HEIGHT ANGLE TEMPERATURE TIME MASS INCH KIPF RAD FAH SEC LBM

//-VALVE--/-TBLHAM-/------TITLE.-----/CRITICAL/-MEMBER-/----------APPLIED DATA-- ----- -/- --ALLNABLE DATA-.---/--RATIO-/-RESULT-//

NODE SECHAM LOAD AX/MX AY/MY AZ/MZ ACC 'OM 'QUATION 308 GATE1 123 308L 0.00 27.39 0.00 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 OHE 123 308M 0.00 27.39 0.00 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 OHE 456 308L 0.01 54.24 0.01 772.80 1000000.06 ~

0.07. PASS 0.00 0.00 0.00 OHE 456 308M 0.01 54.24 0.01 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 I ONE 508L 0.00 27.39 0.03 772.80 1000000.06 0.04 PASS 508 GATE3 123 0.00 0.00'.00 'ONE 123 508M 0.00 27.39 0.03 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 OHE 456 508L 0.01 54.24 0.05 772.80 1000000.06 0.07'. PASS 0.00 0.01 0.00 OHE 456 508M 0.01 54.24 0.05 772.80 1000000.06 . 0.07 PASS 0.00 0.01 0.00 OHE 608 GATE2 123 608L 0.00 49.27 0.00 772.80 1000000.06 . 0.06 PASS 0.00 0.00 0.00 ONE N +g CA rt lD (D,Io.

rt 0 lo IB 0 5 H) rt Ln +

'h

• • P.DELTA / HPDS "CALCULATIOH Ho. DC-D 02 HSC.36 *JOB HO 00002448~18.0CT.,1993 08:21:56*PG 00069 *

//-VALVE- /-TBLNAII /------TITLE- ----/CRITICAL/-HEHBER-/----------APPLIEO DATA--- -----/----ALLOWABLEDATA-- / -RATIO-/-RESULT-//

HODE SECNAH LOAD AX/HX AY/HY AZ/HZ ACC HOII EQUATIOH 123 ~

608H 0.00 49.27 0.00 772.80 1000000.06 0.0& PASS 0.00 0.00 0.00 ONE 456 ,

608L 0.01 97.46 0.01 772.80 1000000.06 Q.13 " PASS 0.00 0.00 0.00 ONE 456 608H 0.01 97.46 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 OHE 438 Y075 138 123 438L 0.00 27.32 0.00 772.80 1000000.06 . 0.04 PASS.

0.01 0.02 0.02 OHE 123 438H 0.00 27.32 0.00 772.80 1000000.06 0.04 PASS 0.01 0.02 0.02 ONE 456 4381. 0.01 54 F 06 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 ONE 456 438H 0.01 54.06 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 ONE

• • • • • • • • • • FOLLOWIHG IS A SUHHARY OF THE CODE CHECKS PERFORHED ABOVE **********

ALL 4 PIPING VALVE JOINTS , THAT ARE CHECKED, PASSED CODE CHECKS.

SCHECK VALVE HORIZOHT. ACCEL. FOR 30 S

PIPE VALVE CHECK PARA AXPFF 1.0000 LOADS ALL AYPFF 0.0001 LOADS ALL AZPFF 1.0000 LOADS ALL AAF '1.5000 JOI ALL TRACE 2 EHD LOAD LIST 1000 2000 CHECK PIPE VALVE INT EQU JOI-308 508 608 438-HOCH TWO M

rt O lA 8

~" P DELTA / HPDS *CALCULATION No. DC.D-02-HSC-36 *JOB NO 00002448*18-OCT-1993 08:21:56~PG 00070 ~"

• • • • • • • l NI*IA1**l***i*i***0k*****A**%1**A****4k**%%*%I*

8 STRUDL PIPING VALVE MEHBER CHECK, TRACE 2, RESULtS

• 1ll*AI**0k***I*1**lii11111*III'%**%A*ii***1111******A*

JOB ID - FFS257 JOB TITLE - CALCULATION Ho. DC-D-02-HSC-36 ACTIVE UNITS - LENGTH HEIGHT ANGLE TEHPERATURE TINE HASS INCH KIPF RAD FAH SEC LBH

//-VALVE--/-TBLNAH-/.-----TITLE---.--/CRITICAL/.HEHBER-/----------APPLIED DATA-------- -/-- -ALLSIABLE DATA--"/-...RATIO./-RESULT-//

' .EQUATION NODE SECNAH LOAD AX/HX AY/HY AZ/HZ ACC HOH 308 GATE1 1000 308L 41.19 0.00 41.10 1159.20'000000.06 0.05 PASS 0.00 '

0.00 0.01 ONE 1000 308H 41.19 0 F 00 41.10 1159.20 1000000.06 '.05 PASS 0.00 0.01 0.00 ONE 2000 308L 81.58 0.01 81.39 1000000.06 0.10 '159.20 PASS 0.00 0.01 0.00 ONE 2000 308H 81.58 0.01 81.39 1159.20 1000000.06 0.10 PASS 0.00 0.01 0.00 ONE 508 GATE3 1000 508L 40.48 0.00 259.97 . 1159.20 1000000.06 0.23 . PASS 0.00 0.00 0.00 ONE 1000 508H 40.48 0.00 259.97 1159.20 1000000.06 0.23 PASS 0.00 0.00 0.00 ONE 2000 508L 80.16 0.01 524 '2 1159. 20 1000000. 06 0.46 PASS 0.00 0.01 0.00 ONE 2000 508H 80.16 0.01 524.42 1159.20 1000000.06 0.46 PASS 0.00 0.01 0.00 ONE 608 GATE2 1000 608L 40.89 0.00 46.15 1159. 20 1000000.06 0.05 PASS 0.00 0.00 0.00 ONE N+

V1 ff Io fo rt 0 C El.

Vl ~

• ~ P DELTA / HPDS "CALCULATION No. DC-D 02 MSC 36 *JOB HO 00002448*18-0CT.1993 08:21:56*PG 00071 *"

//-VALVE"/-TBLNAH-/----"TITI.E"----/CRITICAL/ HEMBER-/--- ------APPLIED --- - --/-- --/--RATIO-/ RESULT-//

NODE SECNAM LOAD AX/HX DATA-AY/MY AZ/HZ 'CC

-ALLSIABLE DATA-HOM .

' EOUATION 1000 608M 40.89 0.00 46. 'I5 1159.20 1000000.06 0.05 PASS 0.00 0.00 0.00 ONE 2000 608L 80.98 O.OI 90.76 1159.20 1000000 06 F 0:10 PASS 0.00 0.00 0.00 ONE 2000 608M 80.98 0.01 90.76 1159.20 1000000.06 0.10 PASS 0.00 0.00 0.00 ONE 438 Y075T38 1000 438L 39.75 0.00 45.44 1159.20 1000000.06 ~

0.05 PASS 0.04 0.12 0.07 ONE 1'000 438H 39.75 0.00 45.44 1159.20 'I000000.06 0.05 PASS 0.04 0.12 0.07 ONE 2000 438L 78.19 0.0'I 89.34 1159.20 1000000.06." 0.10 PASS 0.05 0. 'I4 0.09 ONE 2000 438H 78.19 0.01 89.34 1159.20 1000000.06 0.10 PASS 0.05 0.14 0.09 ONE

• • "******* FOLLOUING IS A

SUMMARY

OF THE CODE CHECKS PERFORMED ABOVE "*~~******

ALL 4 PIPING VALVE JOINTS ~ THAT ARE CHECKED ~ PASSED CODE CHECKS ~

PIPING LOAD TYPES ALEVEL 11 701 .

BLEVEL 1000 TO 1001 CLEVEL 2000 TO 2001 DEAD 11 THERMAL 1 SECTION FR NS 2 0.0 1.0 OUTPUT DECIMAL 5 OUTPUT BY MEHBERS LOAD LIST 11 1 701 1000 TO 1001 2000 TO 2001 S

UNIT INCH POUND RAD LIST DISPLACEHENTS EPIPEJOI

Attachment P This attachment ratios ancl valve showers stx ess acceleration ratios from the as-desi n anal sis w hich includes the trio missing suppox.t components that are

@within the d.esign basis w ell cx iteria allowable limits-Sheet 1 Stress Ratio Interaction Sheets 2 f43 Horizontal AccelerationValve Ratios Sheets 48z;5 V ertical ValveRatios Acceleration

• • P-DELTA / MPDS *CALCULATIOH Ho. DC-D.02-MSC-36 *JOB HO 00002449~18-OCT 1993 08:23:19*PG 00095 ~

//-PIPING-/-TBLNAM-/--CODE--/ - ---/--NODE -/EQUATIOH/-- ---APPLIED FORCES- ---/- - STRESS DATA //--LOAD- /-RESULT-//

MEMBER SECNAM VERSION FX/MX FY/MY FZ/MZ PRES/NONP TOT/ALM RATIO 31.1-4 0.0 0.0 0.0 826.22 7368.35 PASS 0.0 0.0 0.0 6542.13 37500.00 0 '0

~

SUMMARY

OF THE PIPING MEMBER CHECKS BY EQUATION EQUAT I OH MAXIMUM IR MEMBER LOAD 31.1-1 0.4766 98 13 IPJ icuue,T/og g8fsN 31.1-2NU 0.4753 98 1000 31.1-2E 0,4770 44 2000 31.1-3 0.0973 30 14 31.1-4 0.0973 98

• "***~**** FOLLOHIHG IS A

SUMMARY

OF THE CODE CHECKS PERFORMED ABOVE ALL 54 PIPE MEMBERS, THAT ARE CHECKED, PASSED CODE CHECKS.

SCHECK PIPE BREAK LOC ALL SCHECK PIPE CRACK LOC AI.L SCHECK PIPE BRANCH DISPL ALL PRINT PIPE ADD STRESS INT FACT

• • P DELTA / MPDS *CALCULATION Ho. OC-D 02.MSC-36 *JOB HO 00002449*18-OCT-1993 08:23:19 PG 00097 *" ~ i-i**liii*iii**iiiiiiii*iiilii*ii*iiiliilii*1*iiii***i*1

" STRUDL PIPING VALVE MEMBER CHECK ,TRACE 2, RESULTS

• iiiii*i**********i***1***i%A****i**i*I***i*i*A**i*ii*

JOB ID - FFS257 JOB TITLE - CALCULATION Ho. DC D-02 MSC-36 1

ACTIVE UNITS - LEHGTH HEIGHT ANGLE TEMPERATURE TIME MASS IHCH KIPF RAD FAH SEC LBM

//-VALVE--/-TBLHAM-/"---TITLE------/CRITICAL/-MEMBER-/----------APPLIEDDATA--------"/-- ALLOIABLE DATA-"-/- RATIO / RESULT //

NODE SECNAM LOAD AX/MX AY/MY AZ/MZ ACC MOM EQUATION 308 GATE1 123 308L 0.00 27.39 0.00 772.80 1000000.06 0.04 PAss 0.00 0.00 0.00 ONE 123 308M 0.00 27.39 0.00 772.80 1000000.06 0.04 . PASS 0.00 0.00 0.00 OHE 456 308L 0.01 54.24 0.01 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 ONE 456 308M 0.01 54.24 0.01 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 ONE 508 GATE3 123 508L 0.00 27.39 0.03 ?72.80. 1000000.06 0.04 PASS 0.00 0.00 0.00 ONE 123 508M 0.00 27.39 0.03 772.80 1000000.06 0.04 PASS 0.00 0.00 0.00 ONE 456 508L 0.01 54 '4 0.06 772.80 1000000.06 0.07 PASS 0.00 0.01 0.00 ONE 456 . 508M 0.01 54.24 0.06 772.80 1000000.06 0,07. PASS 0.00 0.01 0.00 OHE

~ 608 GATE2 123 608 L 0.00 51.23 0.00 772.80 1000000.06 0.07 ~ PASS 0.00 0.00 0.00 OHE CA h7 5 Ln ~

• • P.DELTA *"

/ MPDS *CALCULATIOH Ho. DC.D-02 MSC-36 *JOB HO 00002449~18-OCT-1993 08:23:19~PG 00098

//-VALVE--/-TBLHAM-/-----TITLE- ----/CRITICAL/-MEMBER /----------APPLIED DATA----------/- - ALLOWABLE DATA----/--RATIO-/ RESULT-//

LOAD AX/MX AY/MT AZ/MZ ACC MOM ~ EQUATIOH NODE SECNAM 123 608M 0.00 51.23 0.00 772.80 1000000.06 0.07 PASS 0.00 0.00 0.00 OHE 456 608L 0.01 101.30 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 OHE 456 '08M 0.01 101.30 0.01 772.80 1000000.06 0.13 PASS 0.00 0.00 0.00 OHE 438 Y075T38 123 '438L 0.00 27.64 0.00 772.80 1000000.06 0.04 PASS 0.01 0.02 0.02 ONE 123 438M 0.00 27.64 0.00 772.80 1000000.06 0.04 PASS 0.01 0.02 0.02 OHE 456 438L 0.01 54.70 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 OHE 456 438M 0.01 54.70 0.01 772.80 1000000.06 0.07 PASS 0.02 0.04 0.04 OHE

~********* FOLLOHIHG IS A

SUMMARY

OF THE CODE CHECKS PERFORMED ABOVE

• ~********

ALL 4 PIPING VALVE JOI HTS t THAT ARE CHECKED, PASSED CODE CHECKS ~

S SCHECK VALVE HORIZOHT. ACCEL ~ FOR 3G S

PIPE VALVE CHECK PARA AXPFF 1.0000 LOADS ALL ATPFF 0.0001 LOADS ALL AZPFF 1.0000 LOADS ALL AAF 1.5000 JOI ALL TRACE 2 EHD LOAD LIST 1000 2000 CHECK PIPE VALVE INT EQU JOI-308 508 608 438-NOCH TllO Vl Q

rt O

~5

• • P-DELTA / HPDS *CALCULATION Ho. DC.D-02-HSC 36 *JOB NO 00002449*18 OCT 1993 08:23!19iPG QQQ99

• '*liilil**i*ii*liiiiiiiiilllliiiiiiii*i***liiiii*i*i*1

• 2, *

• • i*iiiiiiili*ili*liiiiiiii*i*iiiiiiiiii***ili*iiiiiii STRUDL PIPING VALVE MEHBER CHECK ,TRACE RESULTS JOB ID - FFS257 JOB TITLE - CALCULATION Ho. DC-D-02-HSC-36 ACTIVE UNITS - LENGTH HEIGHT ANGLE TEHPERATURE T IHE HASS--

INCH KIPF RAD FAN SEC LBH

//-VALVE--/-TBLNAH-/--"-TITLE- ----/CRITICAL/ HEHBER-/----------APPLIED DA'TA------- --/--- ALLOMABLE DATA----/--RATIO/-RESULT-//

LOAD AX/HX AT/HY AZ/HZ ACC HOH EQUATION NODE SECNAH ,

308 GATE1 1000 308L 41.19 0.00 41.10, 1159.2Q 1000000.06 0.05 PASS 0.00 0.01 0.00 ONE 1000 - 308H 41.19 0.00 41.10 1159.20 1000000.06 0.05 'ASS 0.00 0.01 0.00 ONE 2000 ,308L 81.56 0.01 81.39 1159.20 1000000.06 0.10 PASS 0.00 0.01 0.00 ONE 2000 308H 81 ~ 56 0;01 81.39 1159.20 1000000.06 0:10 PASS 0.00 0.01 O.OD ONE 508 GATE3 1000 508L 40.63 0.00 308.47 1159.20 1000000.06 0.27. PASS 0.00 0.00 0.00 ONE 1000 508H 40.63 0.00 308.47 1159.20 1000000.06 0.27 PASS 0.00 0.00 0.00 OHE 2000 508L 80.47 0.01 620.58 1159.20 1000000.06 0.54 PASS 0.00 0.01 0.00 OHE 2000 508H 80.47 0.01 620.58 1159.20 1000000.06 0.54 PASS 0.00 0.01 0.00 ONE 608 GATE2 1000 608L 40.76 0.01 48.66 1159.20 1000000.06 0.05 PASS 0.00 0.00 0.00 ONE CO lD rt lo rt O EI 0 5 A rt

• • P DELTA / MPDS *CALCULATIOH Ho. DC.D.02.MSC.36 "JOB HO 00002449*18-OCT-1993 08:23:19*PG 00100 *

//-VALVE--/ TBLHAM-/--""TITLE--- -/CRITICAL/-MEMBER-/-------- -APPLIED DATA----- ----/-" ALLSIABLE DATA""/--RATIO-/-RESULT-//

NODE SECHAM LOAD AX/MX AY/MY AZ/MZ ACC MOM EQUATION 1000 608M 40.76 0.01 48.66 1159.20 1000000.06 PASS 0.00 0.00 0.00 ONE 2000 608L 80.72 0.81 95.58 1159.20 1000000 06 F 0 11 PASS 0.00 0.00 0.00 ONE 2000 608M 80.72 0.01 95.58 1159.20 1000000.06 0.11 PASS 0.00 0.00 0.00 OHE 438 Y075T38 1000 438L 41.17 0.00 45.88 1159.20 1000000.06 0.05 PASS 0.03 0.11 0.07 OHE 1000 438M " 41.17 0.00 45.88 1159.20 1000000.06 0.05 PASS 0.03 0.11 0.07 OHE 1000000.06'.05 2000 438L 80.98 0.01 90.20 1159.20 1000000.06 0.10 PASS 0.05 0.13 0.09 ONE 2000 438M 80.98 0.01 90.20 1159.20 0.10 PASS 0.05 0.13 0.09 OHE

• ~*~~****~FOLLNIHG IS A

SUMMARY

OF THE CODE CHECKS PERFORMED ABOVE

• • • • • • ~~**

ALL 4 PIPING VALVE JOIHTS , THAT ARE CHECKED, PASSED CODE CHECKS.

PIPIHG LOAD TYPES ALEVEL 11 701-BLEVEL 1000 TO 1001 CLEVEL 2000 TO 2001 DEAD 11-THERMAL

'8PIPEJOI'O 1

SECTION FR NS 2 0.0 1.0 OUTPUT DECIMAL 5 OUTPUT BY MEMBERS LOAD LIST 11 1 701 1000 TO 1001 2000 TO 2001 S I UHIT IHCH POUND RAD LIST DISPLACEMENTS P't (D

rt O Ln g 0

W rt

ATTACHMENT 3 TO AEP'NRC:1184H INFORMATION REGARDING TEMPORARY MODIFICATION 2-93-015 to AEP:NRC:1184H Page 1 As requested in the cover letter for inspection report 50-315/316 93012 (DRS),

this attachment provides the results of our investigation into the issues associated with Temporary Modification (TM) 2-93-015. This TM installed current-to-current (I/I) converter modules in the control circuitry for the Unit 2 East and West Main Feedwater Pumps. The purpose of the temporary modification was to isolate the control circuitry from the presence of signal grounds in the field wiring that were interfering with proper operation of the D/P slave controller. The modific'ation was processed as a temporary modification becaus e he source of the grounding problem could not be resolved with the unit on control li ne.

circuitry Inn aadd ition, the control ioom portion of the feedwater system has been scheduled for replacement in 1994 as part of the reactor protection system upgrade project.

Two design errors occurred that were found during the post-installation checkout process. Specifically, the design as released by Plant Engineering had the I/I wired in backwards (i.e., input and output reversed), and a necessary 100 ohm input resistor omitted. These errors were discovered during post-installation circuit checks, prior to post modification testing (PMT) and prior to placing the affected circuitry back into operation. Since the errors were identified and corrected prior to PMT, no condition report was required. Although not required, a condition report was subsequently generated to document our investigation.

The TM did not involve safety related equipment, and therefore the qua uality ass urance requirements of 10 CFR 50 Appendix B and the requirements of ANSI-N45 2 .111 do not specifically apply. However, since the temp mod process used at the plant is common to both safety-related and non-safety related chan es, our review of this event was conducted to consider its implications for safety-related temp mods.

The following conclusions were drawn from our investigation of this event:

The design error occurred as the result of attempting to copy the design details of an existing I/I circuit from another approved plant drawing to the drawing markup which was used to serve as the installation drawing for the temp mod. While possible to correctly identify the input and output leads on this drawing, the input/output lead designations (W,X & D,F) were inadvertently transposed during their transfer to the mark-up. The 100 ohm input resistor was overlooked and was not transferred to the mark-up.

2. A vendor document detailing the design and installation requirements for the I/I was available, but not as an approved

'plant document. The engineer chose not to use it, because he felt confident that he was adequately familiar with the hardware and that reference to the vendor information would be unnecessary. Use of the document would likely have allowed the engineer to detect the transposition error, as well as the omission of the input resistor.

to AEP:NRC:1184H Page 2

3. The engineer assigned to develop and implement the temp mod request was adequately qualified to perform this task.

4. The supervisor who reviewed the marked-up drawings was also adequately qualified to perform this function. However, the supervisor became involved with the development of the drawing markups. As a result, his review was not independent, which may have compromised his ability to detect the design errors.

5. The temp mod procedure does not clearly convey design review requirements. The Engineering Supervisor initialed the drawings in the temp mod package indicating his concurrence. However, this review is not a step required by the temp mod procedure.

6. The Plant Engineering review required by the temp mod procedure calls for "...a review of the request to ensure that it is needed, correct, practical, and that it accomplishes the intended purpose."

It does not require a technical verification of the design change documents (i.e. drawing mark-up in this case) that are used to implement the request.

This event has identified a weakness in the temp mod procedure design review requirements. In order to prevent recurrence of an incident of this nature, the TM procedure, PMP.5040.MOD.001, will be revised to strengthen the requirements for verifying the technical accuracy of the design. The revision will clearly identify expectations and responsibilities for the review. These revisions will be completed by April 30, 1994.

The unapproved vendor technical document has been forwarded to the Vendor Information Control System section and is currently in the approval cycle.

There are currently a total of 33 temporary modifications installed at the Cook Nuclear Plant. After a preliminary review, two modifications were identified whose failure could cause a unit trip. Plant Engineering will conduct an independent design verification of these two temp mods. This review will be completed by January 31, 1994.

In the interim, the Plant Engineering Superintendent is taking precautions by closely scrutinizing design complexity'f all temp mods that are submitted for Plant Engineering review, and will require independent technical verification if deemed necessary.