ML20203G245

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-440/97-08.Actions Will Be Examined During Future Insp

ML20203G245
Person / Time
Site:	Perry
Issue date:	12/11/1997
From:	Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Myers L CENTERIOR ENERGY
References
50-440-97-08, 50-440-97-8, NUDOCS 9712180086
Download: ML20203G245 (2)

See also: IR 05000440/1997008

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Du h 11, 1997

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Mr. Lew W. Myers 4

Vice President- Nuclear l

Centerior Service Company l

P.O. Box 97. A200 l

Perry, OH 44081 l

SUBJECT: NOTICE OF VIOLATION (NRC INSPECTION REPORT NO. 50-440/97008) .

Dear Mr. Myers: 1

This will acknowledge receipt of your letter dated November 24,1997, in response to

our letter dated September 23,1997, transmitting a Notice of Violation associated with the . I

above mentioned inspection report at the Perry Nuclear Power Plant We have reviewed your

corrective actions and have no further questions at this time. These corrective actions will be

examined during future inspections.

.

Sincerely, ,

Original Signed By J. Jacobson

John A. Grobe, Director

Division of Reactor Safety

Docket No. 50-440

Enclosure: Ltr dtd 11/24/97, L. W. Myers,

Centerior Energy, to USNRC w/ encl

See Attached Distribution

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OFFICE Rlll/DRS- [d Rill /DRS lc- Rill /DRP c. . Rill /DR$ff

NAME Miller:sd' M Lougheed %f- Kolsk J t Q Grobe ( W )

DATE 12/sr/97 12/v /97 12/i/97

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L. W. Myers 2

cc w/ encl: H. L. Hegrat, Manager, Regulatory Affairs

T. S. Rausch, Director. Quality and

Personnel Development

R. W. Schrauder, Director, Perry

Nuclear Engineering Department

W. R. Kanda, General Manager,

Nuclear Power Plant Department

N. L. Bonner, Director, Nuclear

Maintenance Department

H. W. Bergendahl, Director, Nuclear

Services Department

State Liaison Officer, State of Ohio

Robert E. Owen, Ohio Department of Health

C. A. Glazer, State of Ohio

Public Utilities Commission

Distribution

Docket File w/enci Rlli PRR w/enct Rlli Enf. Coordinator w/enci

PUBLIC IE-01 w/enct SRI, Perry w/ encl TSS w/ encl

LPM, NRR w/enci J. L. Caldwell, Rlli w/enci DOCDESK w/enci

DRP w/ encl A. B. Beach, Rif t w!enci CAA1 w/enci

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• CENTERIOR

O ENERGY

Power Generation Group

Pony Ndear Power Piat Mad Addsess: 216 280-6915 Lew W. Wrert

to Cerder Nwj Po. Don 97 FAX:21fr?804029 We Presdort

P,n y. Cruo 440t 1 Puery. Oi 44081

November 24,1997

PY CEl/NRR-2228L

United States Nuclear Regulatory Commission

Document Cor. trol Desk

Washington, DC 20555

Perry Nuclear Power Plant

Docket No. 50 440/97008

Reply to Notice of Violation

I adies and Gentlemen: ,

Enclosed is the Perry Nuclear Power Plant staff s reply to the Notice of Violation and

Notice of Deviation contained in NRC Inspection Report 50-440/97008, which was

transmitted by letter dated September 23,1997. The inspection report documents the

results of the NRC's follow-up on issues identified during the Design inspection as

described in inspection Report 50-440/97201. The required 30 day response due date

was extended an additional 30 days by the NRC during a telephone discussion that was

held between Mr. Mark A. Ring, Branch Chief- DRS, Region 111 and Mr.1lenry L

llegrat from the Perry plant on October 10,1997.

Each issue identified during the NRC Design Inspection was individually investigated as

part of the Perry Corrective Action Program, in addition, the Design Engineering staff

perfonned a Collective Significance Assessment which assesses the global implications

of the individual issues under Potential issues Form (PIF) 97-1024. A primary corrective

action for PIF 97 1024 was an initiative to develop and implement a Design Engineering

improvement Plan to address key areas needing enhancement, such as design control,

corrective action, calculation quality, and Updated Safety Analysis Report accuracy and

consistency The improvement plan initiative was discussed during the NRC Exit

meeting on August 27,1997, and is cun'ently under development at this time. A full

time project manager has been assigned to develop and monitor the engineering plan for

implementation.

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Novemter 24.1997

Page 2 of 2

If you have questions or require additional information, please contact

Mr. IIenry L. liegrat, hianager - Regulatory Affairs at (216) 280 5606

Very truly yours,

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DTG:dtg

Enclosures

cc: NRC Region 111 Administrator

NRC Resident inspector

NRC Project Manager

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Novemter 24, 1997

1, Lew W. Myers, being duly sworn state that (1) I arn Vice President - Nuclear, of the

Cente ior Service Company, (2) I am duly authorized to execute and file this

certification on behalf of The Cleveland Electric illuminating Company and Toledo

Edison Company, and as the duly authorized agent for Duquesne Light Company, Ohio

Edison Company, and Pennsylvania Power Company, and (3) the statements set forth

herein are true and correct to the best of my knowledge, information and be:lef,

O % cu3

Lew W. Mye@

Swom to and subscribed before me, the

J{ day of b /94 7 *

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h idk J. BOS!ACK!

Notary Public, State of Ohio

Conniasion[xpires

(Recorded in LakeApril 23,)2001

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REPLY TO NOTICE OF VIOLATION

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Violation 97008-01 ,

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RestatnnenLaf the Violation .

10 CFR Par 150, Appendix B, Criterion til states in part that the design control measures

shall provide for verifying or checking the adequacy of the design.

The Peny Operations Quality Assurance Program, USAR 17.2 commits to compliance ,

with Regulatory Guides and Standards as listed in USAR Table 1.8 2.  !

USAR Table 1.8 2 commits to following ANSI N45.2.ll 1974 for Quality Assurance

Requirements For The Design of Nuclear Power Plants.

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Nuclear Engineering Instruction NEl 0341 Revision 5 " Calculations" applies to all

calculations to establish design bases or to change design documents. Paragraph 6.2,

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Calculation Revisions states " Design Engineers are to monitor calculations to detennine

if a revision is required e.g. receipt of new/ revised design input, confirmation of

assumption etc," Paragraph 6.3, Review and Approval, states, " Verification / review and

approval of calculation should precede use of the results for design, but must be ,

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completed prior to the component, system, or structure being declared operable."

ANSI N45.2.11 states that aesign analyses shall be perfonned in a planned, controlled

and correct manner. In addition, the design activities shall be prescribed and

accomplished in accordance with procedures which provides adequate checking or

verifying the results of the activity.

(b) Contrary to the above, as of March 27,1997, the Perry Plant design control

measures did not ensure that calculations or analyses were verified and controlled

adequately for the following examples:

(1) Calculation PIl t2,"P11 Level Setpoints in CST for E22 and

ESI Instruments" dated March 12,1985, determined the CST low-

level swapover setpoint limit required to ensure that the llPCS

system has adequate net positive suction head and that no vortex

occurs before suction valve swapover to the suppression pool. The

calculation was inadequate, in that, it used a non-conservative

(improper) HPCS flow rate and did not consider the impact of

valve timing when establishing the CST low level swapover

setpoint.

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(2) Engineering input for safety evaluation 96128 dated October 10,

1996, associated with USAR change request (CR) 96150 for

Section 9.2.2 revision, which changed the emergency closed I

cooling (ECC) system surge tank sizing basis from a 7 day supply

without necessary makeup to a 30 minute supply, used a non-

conservative flooding rate of 60 ppm for surge tank overflow on

the basis of the minimum calculated emergency service water

(ESW) makeup flow to one ECC surge tank. Since the operating

procedure directs the operator to initiate ESW makeup to both i

tanks, the flooding rate should have consider the flooding of both

tanks (i.e.,120 gpm). Additionally, the flooding potential should

have been calculated using assumptions that ESW flows would be

the maximum (117 gpm to each tank) rather than minimum (60

ppm to each tank).

(3) Calculation p42 31,"ECC A Ileat Exchanger Test Results- 1995,"

Revision 0, dated September 15,1995, contained an assumption

that the test instrumentation was within calibration limits. Post-

test calibration was specified to confinn this assumption. As of

March 27,1997, the outstanding assumptions, open for 18 months,

which could afTect calculation acceptability, had not been

confmned or closed.

(4) Calculation E22-2," Overpressure Protection Analysis," Revision

0, dated February 23,1983, which performed an overpressure

protection analysis on a portion of the lipCS system was

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inadequate. The analysis did not identify operating conditions

, under which pressure relief devices were required to ftmetion. The

I maximum pressure considered for the suction piping was 31.25

psig whereas the suction side relief was set at 100 psig. Maximum

discharge pressure considered was 1130 psig whereas the discharge

side thermal relief valve was set at 1560 psig. Funher, the analysis

for the suction piping did not evaluate other pressurization

potentials such as post accident alignment from the suppression

pool with consideration of containment overpressure, or conditions

of back leakage from the reactor pressure vessel (RpV).

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Tne perry plant management accepts the violation as described.

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Reasonfor the Violation

The cause of these discrepancies is inattention to detail during the development phase of

!he calculations. The errors were associated with the supporting assumptions and were

not the main focus of the engineering product. Additionally, reasonable values were used

instead of conservative values without documenting an adequate basis to support these

assumptions or deviations from worst case scenario analysis.

Conective Steps Taken and Results Achieved

The assu' ,tions used in Calculations Pil 12, P42 31, and E22 2 have been corrected by

revisions of these calculations on September 8,1997, March 31,1997, and September 10,

1997, respectively. Safety Evaluation 96128 was superseded by a USAR change and

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supporting 10 CFR 50.59 Safety Evaluation on June 14,1997. These actions taken were

discussed in a letter dated August ' t,1997 (PY-CEl/NRR 21941.), which provides the

response to the NRC Inspection Report 50-440/97 201, dated June 10,1997, Unresolved

items (URis) 97 201 01,22,25, and 18, respectively.

These cirors were discussed with the Design Engineering staff during the September

1997 section meeting. The Manager Design Engineering re-emphasized management's

expectations that the engineering staff maintain an appropriate focus on being self-

critical, ensure that conservative assumptions and values are utilized in calculations, and

verify that any deviations from the conservative assumptions are adequately justined and

documented.

Contcliyg.. Steps That Will 13e Taken To Avoid Further Violations

An Enginc: ring Improvement Plan is currently being developed. This plan will

incorporate lessons learned from the collective sigaiticance review. The plan will be

finalized by the end of 1997, 1

A computerized calculation database, including tracking of the open assumptions has

been implemented and is being utilized to maintain design calculations with the exception

of the structural / piping stress calculations. The latter calculations are currently being

entered into the database. Open assumptions will become more accessible and easily

identified when a review is conducted for future calculation revisions.

Date When Full Comnliance Will lie Achieved

Full compliance was achieved on September 10,1997, with the completion of the last

corrective action taken as described above.

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Violation 97008-02

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RestatemenLnOhe_Yiolation

10 CFR Part 50, Appendix B, Criterion XVI requires conditions adverse to quality such

as malfunctions and deviations be promptly identified and corrected.

The Perry Operations Quality Assurance Program. USAR Chapter 17.2, requires in

Section 17.2.16 that significant conditions adverse to quality be identified and have

action taken to prevent recurrence.

Contrary to the above, as of March 27,1997, the following examples of untimely or

ineffective corrective action were identified:

1. The testable rupture discs for emergency diesel generators had failed to _

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properly operate on more than 12 occasions over the past 12 years. The root

cause evaluation for the most recent failure (February 19,1997) documented

in PlF 97- 0325 attributed one of the causes to untimely corrective action.

2. Test results for the llPCS keep full pump (from TXI 229, dated March 19,

1996) showed that the pump was not capable of delivering the flows and  :

pressure specified in USAR Section 6.3. This degraded condition had existed

since July 24,1993, when the surveillance test was conducted. As of March

27,1997, this issue had not been resolved.

3. The licensee's safety system functional inspection of the llPCS system in

1992, recognized that the basis for not installing the llPCS pump motor

overfrequacy protection re!ay during construction, as documented in FDDR

KLI 3890, dated May 28,1985, was not well founded. Consequently, the

licensee performed Calculation E22-19," Justification for Elimination of

IIPCS Overfrequency Relay," Revision 1, dated July 23,1992, to evaluate the

efrect of not installing the relay. The licensee's corrective action decision in

1992 was improper for the following reasons:

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The calculation referenced Section 111, NB3654 1. of the ASME Il&PV

Code in order tojustify exceeding the system design pressure in the

event that the Division 111 emergency diesel generator frequency goes

above 60 llz. The licensee did not identify the specific edition or

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addenda of the Code; however, Design Specification (DSP) E22 1 '

4549 00,llevision 3, dated April 18,1986, specifies the 1974 ASME

B&PV Code with addenda up to and including the winter 1975 issue, ,

Section Ill, Division 1. This code edition and addenda did not provide

adequate basis to enable the licensee to justify the allowances for  ;

exceeding the design pressure in accordance with NB 3654 1,

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The licensee's calculation methodology provided a relief path to limit

pressure using the minimum flow valve and its actuation circuitry as

overpressure protection devices. Ilowever, the licensee was unable to

demonstrate compliance with the requirements for this valve and its

actuation circuitry, as specified in ASME Code Section 111, Article NC-

7000," Protection Against Overpressure."

4. On October 8,1996, emergency closed cooling boundary valves were

included into the licensee's In service Test program as part of corrective

actions. This corrective action was ineffective in that test procedure PTI-

P42 P0008, "P42 (ECC) System Leak Rate Test Procedure," revision 1, was

inadequate and did not demonstrate that equipment could perform

satisfactorily under accident conditions. Specifically, as of March 27,1997,

the differential test pressure specified for system boundary valve seat leak

testing was only approximately % of the pressure that the valves would be

subjected to under accident conditions. No extrapolation of test data to

compensate for this difference in test conditions was included in the

procedure.

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' Die Perry plant management accepts the violation as described.

hason for the Violation

The cause for not fully resolving these issues within a more reasonable time frame was

identified as a work management issue. The Design Engineering section, while under a

large work' load combined with a personnel reorganization effort, did not effectively

maintain an issues prioritization schedule during this period. As a result, these issues did

- not receive the appropriate level of management attention.

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CDDEClinjieps Taken and Results Achieved .

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The diesel generator rupture disc setpoint was changed, and the testing frequency

increased on March 19,1997, and May 28,1997, respectively. These changes have

prevented additional excessive setpoint lift problems. The final resolution is a design

modification as discussed in a letter dated August 11,1997 (PY CEl/NRR 2194L), which

. provided the response to NRC Inspection Report 50 440/97 201, Unresolved item (URI)

97 201 05.

The keep full pump calculations revisions were completed on September 4,1997. The

calculations now include minimum required performance criteria. This information

supports a USAR revision scheduled to be completed by December 12,1997, as

discussed in letter PY-CEl/NRR 2194L, which provided the response to NRC Inspection

Report 50 440/97 201. URI 97 20102.

Calculation E2219," Justification for Ehmination ofliigh Pressure Core Spray

Overfrequency Relay," was revised on October 16,1997. The calculation addresses the

concerns cited and is presently under GE Design Interface Evaluation. This evaluation is

scheduled to be completed by December 5,1997, e discussed in letter PY-CEl/NRR-

2194L, which provided the response to NRC Inspection Report 50-440/97 201, URI 97-

201 04.

The results of recent Emergency Closed Cooling system valve and system testing were

documented in a letter dated June 13,1997 (PY-CEl/NRR 2180L), Tids testing was

completed on June 17,1997, and established that the allowable system leakage criteria

has been met, restoring the plant to the USAR design basis for allowable leakage.

C,bnective Stens That Will lle Taken To Avoid Further Violations

The engineering issues have since been prioritized and effectively communicated to

management. The Engineering issues List is now a section management process that

maintains engineering issues with a high degree of visibility and establishes appropriate

priorities to ensure timely and effective resolution. This process was implemented as an

engineering management work prioritization tool during March 1997,

Date When Full Compliance Wilule Achieved

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Full compliance was achieved on October 16,1997, with the completion of the last

conective action taken as described above.

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Violation 97008-03

Restatement of the Violation

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10 CFR Part 50, Appendix II, Criterion 111 states in part that measures shall be established

to assure that the design bas!s for structures. systems and components are correctly

translated into specifications, drawings, procedures, and instructions.

Section 3.5.1.4 " Missiles Generated by Natural Phenomena," of the Updated Safety

Analysis Report (USAR) states in part that: those systems or components listed in Table

3.2 1 that are required to ensure the integrity of the reactor coolant pressure boundary or

maintain safe shutdown conditions are provided with tornado missile protection by

location within seismic Category I structures, unique missile barriers or by the shielding

of an adjacent seismic Category I stmeture.

Contrary to the above requirements, from initial licensing until June 15,1997, the suction

piping for high pressure core spray (llPCS) and reactor core isolation cooling (RCIC)

from the condensate storage tank (CST), and the emergency service water discharge

piping were listed in Table 3.21 of the USAR; however, they were not protected from

tornado generated missiles as described in the USAR. These components respectively,

are required to ensuie the integrity of the reactor coolant pressure boundary and to

maintain safe shutdown conditions.

lleply

The Perry plant management accepts the violation as described.

Reason for the violation

Prior to receiving the plant's operating license, an apparent communication error occurred

in 1984 between the Architect Engineer (AE) and CEI engineering personnel following

the plant's construction and subsequent NRC review of the Final Safety Evaluation

Report (FSAR). The AE perfomied calculations in 1984 which utilized a probability

methodoi% ,,, to provide protection to the plant from tomado missiles, These calculations

provided protection from tomado missiles that was different than the physical protection

description in the FSAR as reviewed by the NRC and documented in the NRC's Safety

Evaluation Report dated May 1982. This condition went unidentified until a subsequent

review was performed by the design engineering staff following the NRC Design

inspection as documented in NRC Inspection Report No. 50-440/97 201.

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Consstive Steps Taken and ResulttAddered

A change was made to Off Normal Instruction (ONI) ZZZ 1," Tornado ni liigh Winds." ,

to include shifting IIPCS/RCIC suction to the suppression pool in the e,ent of a cited

tornado and temporary shielding was installed in areas of potential vt,lnerability. A plant

walkdown was conducted to identify other areas of potential concern and a preliminary

probability comparison was performed to provide reasonable assurance that, in

conjunction with the other compensatory actions, a cumulative probability calculation

would provide a basis to support the current design. These actions, along with the short

and long term corrective actions, were discussed in letters dated June 13,1997 (PY-

CEl/NRR 2180L), and August 11,1997 (PY-CEl/NRR 2194L), which provided the

response to NRC Inspection Report 50 440/97 201, URI 97 201-08.

Conective Steps That Will Be Taken To Avold Further Violations

A license amendment was submitted in a letter dated August 14,1997 (PY CEl/NRR-

2200L), and supplemented by letters dated September 26,1997 (PY-CEl/NRR 2221L),

ano dated October 1,1997 (PY CEl/NRR 2224L). Amendment No. 90 to the Perry  !

operating license was issued by the NRC on November 4,1997, which approved the

probability methodology for protection from tornado missiles.

Date When Full Comnliance Will Be Achieved

Full compliance was achieved on November 4,1997, with the issuance of Amendment

No. 90 to the Facility Operating License No. NPF 58 for the Perry Nuclear Power PI nt,

Unit 1.

Violation 97008 04

Restatement of thc Ylolatiom

10 CFR 50.59, " Changes, tests and experiments," pemiits the licensee, in part, to

make changes to the facility as described in the safety analysis report without

prior Commission approval provided the change does not involve an unreviewed

safe:y question (USQ). A proposed change, test, or experiment shall be deemed

to involve a USQ if, in part, a possibility for an accident or malfunction of a

ditTerent type than any evaluated previously in the safety analysis report may be

created. The licensee shall maintair. records of changes in the facility and these

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records must include a written safety evaluation which provides the bases for the

determination that the change does not involve a USQ.

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USAR Sections 6.3.2.2.1 and 6.2A.2.2.2 describe the normal alignment of the

llPCS system and the position of containment isolation valves for the suppression l

pool cleanup system (SPCU). Specifically, Section 6.3.2.2.1 indicates that llPCS

suction is normally abgned to the CST and that it will automatically switchover to l

the suppression pool. Section 6.2.4.2.2.2 for the SPCU system containment

isolation valves states that the compliance for containment isolation requirements

for the return line is satisfied,in part, on the basis that the line is normally closed.

Contrary to the above requirements, from the String of 1993 until March 31, '

1997, the SPCU system was in operation almost continuously resulting in a

11PCS alignment different than described in the USAR and resulting in the SPCU

system containment isolation valves to be normally open versus closed, a=

specified in the USAR. This continuous operation of the suppression pool system

was not supported by a written safety evaluation as required by 10 CFR 50.59.

Reply

The Perry plant management accepts the violation as described.

Reason for the Violation

Since the 1992 Emergency Core Cooling systems strainer problem was encountered,

maintaining suppression pool cleanliness received heightened attention. A decision was

made to increase operation of the SPCU system. As a result, the SPCU system was

operated more than what was originally intended without making procedure changes in

accordance with 10 CFR 50.59 to preclude operation differently than described in the

USAR. 'Ihis change in operation of the SPCU system should have been addressed as a

change to tw USAR to provide clarification of the system's operation.

Corrective Stens Taken and Resuhs Achieved

Administrative controls were put into piwe on March 31,1997, to limit the operation of

the SPCU system. Additionally, on September 4,1997, the matter was again discussed

with operations to further clarified the operational limitations that were provided on

March 31,1997.

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Correcliyejilms That Will Be Taken Tdvoid Further Violations t

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A USAR change which clearli describer. SPCU operation limitations. if any, is scheduled

to be completed by I ecembe; 12,1997, as discussed in a letter dated August i1,1997  ;

(PY-CEl/NRR 2194L), which provided the response to NRC Inspection Report 50-

.-40/97 201, URI 97 20109.

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D;de When Full Compliance Will Be Achievsd

Full compliance was achieved on March 31,1997, with implementation of administrative

controls which clarifies operation of the SPCU system limiting its use maintaining

compliance with the USAR.

Violation 97008-07

Ecstatementsf_the Violation

10 CFR 50.71(c) requires the licensee to update the Final Safety Analysis Report

(FSAR) originally submitted as part of the application for the operating license to

as ure that the information included in the "*AR contains the latest material

developed. The updated FSAR shall be reviscd to include the effects of,in part,

all safety evaluations perfonned by the licensee in suppor: of conclusions that

changes did not involve a unreviewed safety question._ Updates must be filed

annually or 6 months after each refueling outage. The updates must reflect all

changes up to a maximum of 6 months prior to the date of filing.

Contrary to the above, as of March 27,1997, the licensee failed to update the

FSAR (currently referred to as the Updated Safety Analysis Report (USAR)) to

reflect plant conditions, which existed more than 6 months prior to the previous

USAR update, as evidenced by the following examples ofinaccurate or non

updated USAR information:

1. USAR Table 8.3-1 listed fuel oil transfer pumps 1R45C001C and 2C as 0-

second loads. IR45C001C and 2C were 40 minute automatic cyclic loads for

both loss of offsite power (LOOP) and loss of coolant accident (LOCA).

2. USAR Table 8.3 1 identified a 9 kW load for IE22C004B, which did not

agree with the 8 kW load in Calculation PSTG 0014.

3. US AR Table 8.3 1 listed the inrush currents for HPCS fuel oil transfer pumps

IR45 C001C and 2C as 109A, whereas Calculation PRMV-0017 listed the

inrush current as 130A.

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4. USAR lable 8.31 listed the inmsh currents fo ' A 3 diesel generator room

fans OM43 C001C and 2C as 362A, whereas C . tion PRMV-0017 listed

the inrush cunent as 376A.

5. USAR Table 8.31 listed the Full Load Amperage (FLA) ofl1PCS diesel

generator starting air compressor iE22 C004B as 13A, whereas Calculation

PRMV-0017 listed the FLA as ll A.

6. USAR Table 8.31 listed the llPCS ESW pump IP45 C002 is as 75 hp,88.5

FLA. and 557A inrush, whereas Calculation PRMV-0017 listed the same

load as 75 hp,85.4 FLA, and 543A in rush.

7. USAR Table 8.3 1 listed the rating ofilPCS diesel generator space heater

iE22. D011 as 2 kW, with a load current of 3 amp. Calculation PRMV-0017

listed the same space heater as 1.6 kW, with the load current of 2.01 amp.

Drawing D 206 029/BB," Electrical One Line Diagram, Class IE,480 V Bus

EFID," listed the same space heater as 2.4 kW.

8. USAR Table 3.9 30 listed active valves not associated with the nuclear steam

supply system (NSSS). Tids table had not been updated to reflect several

ECC system modifications. Valves P42 F315A,B,C should have been

deleted from the table, since they were converted from automatic to manual

valves by DCP 92 0060. Valves P42 F550 and P42 F551 should have been

added to the table, since they were converted from manual to automatic

valves by DCP 90 0012.

9, USAR Tables 9.2 18 (ECC Pumps) and 9.2-19 (ECC lleat Exchangeru listed

two different values for ECC system operating flow rate (1860 versus 1820

gpm). Since all pump flow was delivered to the heat exchanger, the flow

values should have been the same.

hPh

The Peny plant management accepts the violation as described.

hason for the Violation

The majority of the discrepancies identified occurred due to a past practice which was

limited to the Electrical Power Element of the Design Engineering Section. This practice

allowed discrepancies between difTerent series of calculations, and between calculations

and the USAR to exist if they were judged to be technically insignificant. Design

Engineering discontinued this practice several years ago, but prior discrepancies between

calculations and the USAR remained unchanged.

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I'mlosme l

Pap 12 of 15

CunceliYtSkns Taken and Results2chieved

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Changes to USAR tables 3.9 30,9.218, and 9.219 were completed on November 20,

1997, which address items 8 and 9 above to accurately reflect the current plant

conditions.  !

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A change to the USAR will be made to address items I through 7 described above as a

coordinated effort upon completion of the corrective actions associated with Violation

97008-09. The USAR changes are scheduled to be completed by June 10,1998. These

actions were discussed in a letter dated August 11,1997 (PY-CEl/NRR 2194L), which

provided the response to NRC Inspection Report 50 440/97 201, URis 97 201 15,

and 24.

Correctivt1gpsThat Will He Taken To Avoid Furiter Violations

,

The Design Engineering staff has recognized this old work practice as unacceptable. As a

result of the Collective Significance review performed under PlF 97-1024, engineering

expectations with respect to USAR compliance are being developed into a consolidated

plan to maintain higher standards within the Design Engineering section. The plan will

be developed by the end of 1997 ar.d will become pan of the organization's 1998

enhancements to improve overall performance in engineering products.

Date When Full Compliance Will Be Achieved

Full compliance will be achieved when the corrections to items 1 through 7 are

incorporated into the USAR which remains on schedule as discussed in CEI letter (PY-

CEl/NRR 2194L), and will be completed by June 10,1998.

Violation 97008 09

11cslaicngnt of the Violation

10 CFR Part 50, Appendix B, Criterion 111 states in part that the design control

measures shall provide for verifying or checking the adequacy of the design.

The Perry Operations Quality Assurance Program, USAR 17.2 commits to

compliance witii Regulatory Guides and Standards as listed in USAR Table 1.8 2.

USAR Table 1.8 2 commits to following ANSI N45.2.11 - 1974 for Quality

Assurance Requirements For The Design of Nuclear Power Plants.

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Enclosme 1 l

!' age 1.1 of 15

Nuclear Engineering Instruction NEl 0341 Revision 5 " Calculations" applies to

all calculations to establish design bases or to change design documents.

Paragraph 6.2, Calculation Revisions states " Design Engineers are to monitor

calculations to detennine if a revision is required e.g. receipt of new/ revised

design input, enfirmation of assumption etc." Paragraph 6.3, Review and

Approval, states, " Verification / review and approval of calculation should precede

use of the results for design, but must be comple'ed prior to the component, ,

system, or structure being declared operable." i

ANSI N45.2.11 states that design analyses sh::ll be performed in a planned,

controlled and conect manner, in addition, the design activities shall be

prescribed and accomplished in accordance with procedures which provides

adequate checking or verifying the results of the activity.

(a) Contrary to the above requirements, as of March 27,1997, the

licensee had modined various systems as reflected in design

drawings and did not update / revise 15 calculations as described in

the following examples:

(1) Electrical drawing D 206-029, " Electrical One- Line

Diagram, Class IE,480 V Bus EFID," Revision BB,

identified the installation of a 10 hp electric motor for

compressor IE22-C004A. Calculation PRMV-0017 "EHF-

1 E Transformer Breaker E111305," Revision 0, did not list

the compressor motor. USAR Table 8.31 also did not

correctly identify the motor loads. Calculation PRMV-

0017 was last updated on March 11,1985 (12 years ago),

and did not reflect the current plant loads and settings.

(2) Calculation PSTG-0003 "480 V Safety Related Motor

Starting Voltage Drop," Revision 2, dated June 29,1995

(page 6), contained an open assumption that required

confirmation. Calculation PSTG 0001,"PNPP Auxiliary

System Voltage Study," Revision 2, approved on August

24,1995, provided the information to resolve the open

assumption. As of March 27,1997, calculation PSTG-

0003 was not updated to close the open assumption, even

though the infonnation to close the assumption was

available on August 24,1995.

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Enclosure i

Page I ) of 15

(3) Calculation PRDC-0006, " Load Evaluation and Battery

- Sizing of Division 111 Class IE DC System," Revision 0,

dated April 8,1991, did not adhess Division 111 high

pressure core spray (liPCS) pump iE22C001 breaker l

E111304 spring charging motor load at t=0 second, the load

profile for 01 min for continuous (L2) load, and the DC

contml circuit loads (L2 loads) of the breakers.

(4) Calculation PRDC-0004, " Class IE DC Control Circuit

Coordination," Revision 2. dated May 30,1995, did not

address switch #12 added to drawing D206-051, " Electrical

Main One-Line Diagram, Class IE DC System" Revision

RR, dated May 15,1992, in accordance with DCP 90-0012.

(5) Calculation PRLV-0004,"480 V Breaker Coordination,"

Revision 2, dated April 30,1996, was reviewed against

associated electrical D- 206 series drawings for 480-V

motor control centers (MCCs). Various discrepancies and

typographical errors were found between thc calculations

and the drawings as noted below:

Item # Calculation PRLV-0004 DIaw. ing D-206 series

1B21-F065A 6.6 lip 6.4 lip

P42-F551 MISSING 0.13 lip

P45 D004A 7 IIP 1 IIP

P42-F550 MISSING 0.13 lip

M25-C001B 100 IIP 60 lip

1G33-F001 3.0 HP 3.9 HP

Reply

The Perry plant management accepts the violation as described.

Reason for the Violation

The Electrical Power E(ement of the Design Engineering Section maintained a past work

practice that did not require incorporating minor changes that did not sigrificantly impact

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- the conclusions of the calcula. tion. This practice allowed minor discrepancies to exist

until the calculation was revised for a more significant change, at which time the minor

- discrepancies would be included in the revision.

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Enclosure l-

l' age 15 of 15 j

Corrective Steps Taken and Results Achieved

The electrical engineering staff performed a comprehensive review of safety-related i

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electrical calculations with a current revision older than five years. This review was

completed on October 31,1997, with no safety significant concerns identified. 1

As a coordinated follow-on corrective action, the calculation and drawing discrepancies

identified above are curTently being revised to reflect the current plant conditions as

discussed in a letter August i 1,1997 (PY CEl/NRR-2194L), which provided the

response to NRC Inspection Report 50-440/97-201, URI 97401-24. These calculations

are being corrected as soon u possible to support the necessary changes to the USAR

tables as described in the response to Violation 97008-07 and are scheduled to be

completed by May 4,1998.

Items 3 and 4 described ibove have been corrected by a Design Change Control (DCC)

for Calculations PRDC-004 and PRLV-0004 on November 18,1997, and March 4,1997,

respectively.

Corrective Stens That Will Be Taken To Avoid Further Violations

The Design Engineering staff has recognized this previous practice as unacceptable.

Management expectatious have been established to eliminate this practice. As a result of

the collective significance review performed under PIF 97-1024, Design Engineering

Section expectations are being developed into a consolidated plan to maintain higher

standards 'vithin the section, including calculation quality .ssues. The plan will be

finalized by the end of 1997.

Date When full Comnliance Will Be Achieved

Full compliance will be achiesed no later than May 4,1998.

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Enclosure 2

Pageiof4

REPLY TO A NOTICE OF DEVIATION

DEVI ATIONS 97008-05 and 97008-06

Restatement of the Deviation

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During an NRC inspection conducted on July ' ' ugust 27,1997, deviations of

your Generic Letter (GL) 89-13 commitments s i u. i in accordance with the

" General Statement of Policy and procedure f; 1 ^ Cafon ment Actions," NUREG-

1600, the deviations are listed below:

A. PNPP respose to GL 8913," Service Wau. WW L 3 Affecting Safety-Related

equipment," PY-CFl/NRR il21, dated January 190, stated that " The ESW

air-to-water heat exchanger (HPCS room cooler) will be inspected and cleaned at

enh refuel outage, fin and tube side, as an alternate to perfonnance testing."

This commitment to clean and inspect the heat exchangers was identified in the

Perry Regulatory Infonnation Management System as commitment L01181.

Contrary to the above, as of March 27,1997, - cre was documentation that the

HPCS room cooler was inspected, however,;.. re was no documentation to

substantir.te that the room cooler was cleaned as committed in PY-CE!/NRR-1121

and confimied in PY-CEI/NRR-1734L. The licensee stated that to their

knowledge, inspections were performed, but no cleaning was done or the HPCS

Room Cooler.

B. PY-CEl/NRR-1734L, dated April 8, I '4, implementation of GL 89-13. " Service

Water Problems Affecting Safety -Rek.ed equipment," states "With available

improvements in methodology cited above, the IIPCS Room Cooler will now be

tested, or alternate monitoring methods will be determined in accordance with

Electric Power Research Institute (EPRI) NP-7552," and "PNPP will maintain the

present testing frequencies of once per cycle until such time as our testing

demonstrates that a reduced frequency is warranted."

The HPCS Room Cooler was tested in June 1995. Calculation M39-6, HPCS

Room Cooler Perfonnance Test Results 1995," deemed the test results

! inconclusive. From June 1995 through August 27,1997, no other operability test '

was conducted. The licensee had not been successful iin establishing a

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performance test program for the HPCS room cooler and had reverted to the

inspection program, but had extended the frequency beyomi each cycle without a

history of testing to demonstrate that a reduced frequency was warranted.

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Enclosure 2

Page 2 of 4

Erph'

The Perry management accepts the deviations as described, it was the interpretation of

PNPP plant staff that cleaning as discussed in a letter dated January 26,1990 (PY-

CEl/NRR 11211 ), was to be performed when required based on inspection results. A

. letter dated April 8,1994 (PY CEl/NRR 1743L), was issued to substitute performance

testing for inspection of the High Pressure Core Spray (liPCS) room cooler. This change

was based on the belief that a test had been developed that would provide accurate and

reliable results. After the letter was issued and the performance test was perfonned, it

was determined that the test methodology produced inclusive results.

. Reason for the Deviations

A. The issue was evaluated in PlF 97-0463 which acknowledges that, although not

clearly documented, a deviation from a commitment made in letter PY-CEl/NRR-

1121 occurred. However, the wording used in the commitment was in error; it

was intended that cleaning of the HPCS room cooler would be performed "as

necessary" and plant procedures were modified to reflect this intention. The

commitment as written implies that the unit was to be cleaned even if the

irispection had found tha cooler to be in a clean condition. This was not the intent

of the commitment. Electric Power Research Institute Report (EPRI), NP-7552,

" Heat Exchanger Performance Monitoring Guidelines," contains a cautionary

statement that excessive cleaning of a heat exchanger may cause wear and tear on

the component. Subsequently, when the HPCS room cooler was inspected, it was

detennined that cleaning was not necesrary. A review of the work orders

associated with inspection of the HPCS room cooler documented that the cooler

was ci.:an. In addition, further analysis was performed to confirm that cleaning of

the cooler remains unnecessary.

D. . A perfonnance test was conducted on the HPCS room cooler during the fifth

cycle in July 1996. The test data was evaluated in August 1996 and determined

that the test was inconclusive. PIF 96-2671 was initiated to address this issue.

The corrective actions from this PIF were to review and revise the test protocol

and associated calculations as necessary. The PIF also concluded that no

significant fouling was expected as the current cooler coil had been installed in

1994. However, it was not recognized that the commitment ofletter PY-

CEl/NRR-1734L may not have been satisfied.

~ PlF 96-3642 was initiated in December 1996 and questioned whether compliance

with the commitment in letter PY-CEI/NRR-1743L, since a conclusive

performance test had not been performed.

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Enclosure 2

Page 3 of 4 -

PlF 96 3642 states : hat although the commitment specifically states that three

tests are needed before a trend can be developed, it is reasonable to apply the

- same logic to require three inspection', for trending purposes. Three inspections

had been performed and found the cooler to be acceptable. Additionally, the

cooler coil had been repbced with a new coil in 1994. Since PNPP was meeting

the requirements of EPRI NP-7552, which allows the performance of visual

inspections and cleaning, and water treatment as alternate methods, it was

concluded that the frequency of inspection could be extended to five years

consistent with Generic Letter 89-13. These methods had been completed prior to

conducting the perfonnance test. PlF 96-3642 also concluded that the cooler was

. clean.

PlF 96-3642 states that PNPP will test the cooler in accordance with the

commitment once per cycle for at least three cycles, assuming a performance test

can be developed. After three cycles the test frequency may be revised but will

not exceed five years. The assessment of P!F 96-3642 concluded that the

commitment had not been violated, was accurately stated, and did not require

revision.

- The issue was also evaluated in PIF 97-0463 which acknowledges that although

not clearly documented, d:viations from commitments occurred. Perry personnel

understood the " testing" aspect of the commitment to permit an " alternate

monitoring method" as determined in accordance with EPRI NP-7552, which

allowed credit for the inspections being performed. Accordingly, it was

determined by the engineering staff, through an evaluation of past 11PCS room

cooler inspections (and cleaning where necessary), that justification to extend the

frequency existed.

The specific intent of the commitment in PY-CEI/NRR-1743L was not clearly

documented. While prudentjudgment was applied to extend the frequency of

liPCS room cooler inspection and conclude that the intent of the commitment was

met, it was not recognized that the actions taken had changed the commitment.

Corrective Stens Taken and Resultr Achieved

Evaluations conducted in the cited PIFs concluded the HPCS room cooler was clean.

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Enclosme 2

Page 4 of 4

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Corrective Steps that Will be Taken to Avoid Further Deviations

Plant management identined that expectations for " literal compliance" required

strengthening; these expectations were reinforced as part of Engineering Suppon

Personnel (ESP) Continuing Training conducted during the second calendar quarter of

1997. The circumstances surrounding this event were discussed as an example increase

the sensitivity toward and reinforce the meaning ofliteral compliance.

Plant Administrative Procedure (PAP)-610," Regulatory Commitment Tracking

Program," was revised to incorporate the NEI guideline " Managing NRC Commitments,"

revision 2,12-19-95. This provides the guidance regarding modification, revision, or

deletion of commitments, and notification to the Nuclear Regulatory Commission.

Date When Full Comnliance Was Achieved

Full compliance was achieved on August 11,1997, with the submittal of a letter dated

August i 1,1997 (PY-CEI/NRR-2194L), that clarified the commitments in question and

further established revised commitments to inspect and clean as necessary and use

alternate monitoring methods as described in EPRI NP-7552 while developing

performance based testing.

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Enclosure 3

PageIofI

' COMMITMENTS-

The following table identifies those actions which are considered to be regulatory

commitments. Any other actions discussed in this document represent intended or

planned actions, are described for the NRC's information, and are not regulatory

commitments. Please notify the Manager-Regulatory Affairs at the Perry Nuclear Power

Plant of any questions regarding this document or any associated regulatory

commitments.

.

Commitments

By May 4,1998, calculation and drawing discrepancies identified in Violation 97008-09

will be revised to reflect the current plant conditions. The calculations are being

corrected first to support a change to the USAR tables as described in the response to

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Violation 97008-07.

By June 10,1998, the USAR will be revised to address the discrepancies described in

Violation 97008 07, items I through 7 as a coordinated effort following completion of

the committed corrective action assceiated with Violation 97008-09.

.