ML20212F911

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-331/97-06

ML20212F911
Person / Time
Site:	Duane Arnold
Issue date:	11/03/1997
From:	Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Franz J IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
References
50-331-97-06, 50-331-97-6, NUDOCS 9711050154
Download: ML20212F911 (1)

See also: IR 05000331/1997006

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_ Novembsr 3.1997

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1 Mr. John F. Franz, Jr.

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< LVice President, Nuclear-- '

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IES Utilities, incorporated -

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200 First Street SE-

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TCedar Rapids, IA 52406-0351

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? SUBJECT:1

NOTICE OF VIOLATION (NRC INSPECTION REPORT 50-331/97006(DRS)

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i Dear Mr. Franz:

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Ths .will acknowledge receipt of your letter dated October 2,1997 in response to our

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letter dated September 2,1997, transmitting a Notice of Violation associated with corrective

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actions, testing, procedure adherence, and design control at the Duane Arnold Energy Center.

LWe have reviewed your corrective actions and have no further questions at this time. These

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corrective actions will be examined during future inspections.-

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Sincerely,

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s/JMJacobson/for

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John A. Grobe, Director

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Division of Reactor Safety

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Docket No. 50-331

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i Enclosure:

~'J. l F. Franz letter to USNRC

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dated October 2,1997

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-ice w/ encl:

~ L. Root, President and Chief '

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G. Van Middlesworth, Plant Manager .

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K. E. Peveler, Manager, Regulatory Performanoa

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Chairperson, Iowa Utilities Board

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DOCUMENT NAME: G:DRS\\Dua97006.TY

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UTILIT!ES

John f. FranL Jr.

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October 2.1997

NG-97-1728

U. S. Nuclear Regulatory Commission

Attn: Document Control Desk

Mail Station Pl-37

Washington, D.C. 20555-0001

,

Subject:

Duane Arnold Energy Center

Docket No: 50 331

Op. License No: DPR-49

Reply to a Notice of Violation Transmitted with Inspection Report 97006

File:

' A-105, A-102

Dear Sir:

This letter is provided in response to a Notice of Violation transmitted with NRC

inspection Report 97006.

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This letter contains the following new commitment:

Complete a review of operability determination requirements contained in Administrative

Control Procedures and take actions as necessary by November 15,1997.

If you have any questions regarding this matter, please contact my office.

Sincerely,

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/

ohn F. Franz -

Vice President, Nuclear

Attachment:

' Reply to a Notice of Vmlation Transmitted with inspection Report 97006

cc:

R. Murrell

L. Root

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G. Kelly (NRC-NRR)

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A. B. Beach (Region 111)

NRC Resident Office

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IES Utilities Inc.

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Reply to a Notice of Violation

Transmitted with Irispection lleport 97006

VIOLATION ONE

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- C'riterion XVI of.10 CFR Part 50; Appendix B, states, in part, that meas

established to assure that conditions adverse to quality are promptly identified an

and that measures Sall assure that the cause of the condition'is de

i actions taken to precjude repetition.

Contrary to the above:

a

a)- As' of April-25,1997, corrective actions taken in response to a presious violation

involving procedure deficiencies and cited in Inspection Report 96005, dated

27,1996, had not. been adequate to preclude repetition.

Seven occurrences were

identified'where the licensee had failed to identify and correct pmcedure deficiencie

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- (50-331/97006-01a).

b) As of April 7,;1997,-adequate prompt corrective actions had not been taken for a fla

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~ identified in the body of the Residual Heat, Removal (RHR) pump D discharge

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valve (Vl9-001) in March of 1995. The flawed valve was retumed to

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1995, without' adequate nondestructive examinations to verify the subsurface extent

T flaw. . The flaw was subsequently removed by grinding it out on April 10, 1997

(50-331/97006-Olb).

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0This is a Severity Level IV violation (Supplement I).

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RESPONSE TO VIOLATION ONE

d(EASON FOR THE VIOLATION:

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VIOLATION la

NRC. Inspection Report 96005 cited a violation for failure to comply with Append

> B of 10_CFR. Specifically, a violation was issued regarding four identified procedu

- deficiencies. lAs a result, the procedure deficiencies were corcected, and the Q

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W Assurance Department reviewed the identified deficiencies and determined, that while

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the procedure review and approval process was adequate, heightened awareness for

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procedure owners was needed. Therefore, a memo was issued to all procedure

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owners to refresh the need for awareness while reviewing propased changes to

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Page 2 of 13

procedures. Specifically, the memo stressed the administrative requirements of the

procedure owner to review procedure revisions for the following:

. technical accuracy

compliance with requirements

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completeness

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adequacy of the Safety Evaluation Applicability Review

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impact on other procedures

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impact on the temporary revisions in effect for applicable impacted procedures

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NRC. inspection activities conducted in April 1997, identified seven procedure

inaccuracies as follows:

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O Operating Instruction (01) 149, " Residual Heat Removal System," Section 11.0,

Step 5.b, identified the torus level recorder on Panel IC03 as LR-4325; the correct

designation was UR-4325. The torus level recorder on Panel IC29 was identified

as LR-4385; the correct designation should have been LR-4385B. 01149 was

subsequently revised to reflect the proper recorder designations on August 26,

'1997,

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Q 01454, " Emergency Service Water System," Step 3.14, stated, " Verify that

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system operability has been established by successful performance of Surveillance

Test Procedure (STP) 48C001 within required surveillance frequency." The

correct statement should have been " Verify that system operability has been

established by successful performance of STP 48E001-Q within required

surveillance frequency." 01454 was revised to designate the proper STP on May

22,1997.

O Annunciator Response Procedure (ARP), "A" Recira MG Drive Motor Trip or

Overload, Step 3.1, stated, "If A MG SET MOTOR AMPS are greater than 565

amps and does not trip, immediately reduce the speed of the A Recirc MG until

this alarm clears. Comply with Tech Specs 3.6.F for Recirc Pump Speed

Mismatch Limitations"; the correct statement should have been, "If A MG SET

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MOTOR AMPS are greater that 565 amps, trip the A Recire MG."

The

procedure also contained two steps marked as 3.3. The ARP was revised to state

?.he correct actions and to correct the step numbering on April 16,1997.

O STP 46E001-SLO, Step 7.1.11.c, stated, "lF the answer to 7.1.11.a is NO, notify

the OSS immediately and follo'v the instmetions given in General Instmetion 4.3.

N/A Step 7.1.11.b." The correct sta ment should have been "lF the answer to

7.1.11.a is No, notify the OSS immediate5 and follow the instmetions given in

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General Instruction 4.4. N/A Step 7.1.11.b." The same error as above existed in

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Page 3 of 13

Step 7.2.11.c of the proceouse. The STP was subsequently revised to renect the

correct statement on July 1,1997.

O STP 47A016-Q, Step 7.1.12, failed to contain the pressure requirements for the

test. The test pressure was assumed to be the same pressure stated in Step 7.1.9

which was 43 pounds per square inch gauge (psig) (+2,-0 psig). However,

because of the time necessary to perform Steps 7.1.10 and 7.1.11, the pressure had

the potential to decay below 43 psig and thus make the test invalid ' Die STP was

subsequently revised to contain the appropriate acceptance criterion on July 1,

1997.

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O STP 47A016-Q, Step 7.1.15, stated, " Confirm " Pressure Drop" obtained

7.1.13 is less than (<) 10 psi over the ten minute period." The correct statement

should have been "Confinn " Pressure Drop" obtained in Step 7.1.14 is less than

(<) 10 psi over the ten minute period." The above error had existed since the

a

procedure was changed from STP 47A016, Revision 8 to STP 47A016-Q in 1995.

The procedure had been performed onc] a quarter for a total of six times since the

revision and the error had not been identified. The STP was revised to refl

correct statement on July 1,1997,

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O STP 45A002-Q, Step 7.4.55, failed to have American Society of Mechanical

Enginces (ASME) indicated after the closed stoke time acceptance criteria. The

STP was revised on July 16,1997, to contain the ASME designation.

The corrective actions implemented as a result of the violation contained in IR 96005

were focused on ensuring that proper mechanisms and cultures were in place to

identify and correct pacedural weaknesses. The corrective actions taken were not

broad enough to ensure that the appropriate level of detail was being applied during

the procedure usage, review, and approval process to identify and correct all potential

technicalinaccuracies.

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VIOLATION 1h

,On March 23,1995, a 3/4 inch long linear surface indication (flaw) was discovered

during an ASME Section XI Code required magnetic particle examination of valve to

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pipe weld, RH1-CF013, which was documented in Action Request (AR) 95-0464.

The flaw was in the base metal of the check valve body of the RHR pump D

discharge check valve, Vl9-000),0.06 inches outside the Code mandated inspection

boundary and was oriented circumferentially along the measured length. The valve

was accepted for continued service based on an engineering e aluation using a

fracture mechanics analysis which assumed an initial flaw depth of % thetalve body

wall thickness as an input to this calculation. An informational Ultrasonic Test (UT)

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Attachment I to

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Page 4 of 13

.was performed at the time but was riot documented because the flaw could not be

seen ultrasonically.

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No further actions were taken based on Duane Arnold Energy Center's (DAEC

' determination that the flaw was located outside the ASME Section XI boundary

clearly related to fabrication (surface lap), not service induced, and evaluated with

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fracture mechanics. As a result of discussion with your staff, it was determined that

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further actions would conservatively be taken. De UT inspection was performed and

documented. Results were the same as the ori;inal, un documented UT, the flaw

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could not be seern The tiaw was ground out with minimal material removal.

2. . CORRECTIVE ACTIONS TAKEN AND THE RESULTS ACHIEVED _

VIOLATION la .

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He DAEC continues to take actions to ensure that high quality, accurate, and human

factored procedures are maintained. Several actions have been taken to ensure that

procedures are properly maintained. These include:

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Administrative Control Procedures (ACPs) were revised to melude a flow

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. procedure usage.

Specifially, the flow chart -indicates when to stop using a

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- procedure when technical inaccuracies are identified and those mechanisms to utilize

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' for correcting identified technical inaccuracies

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Operations- Department completed a review of Operating Instructions and

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Annunciator Response Procedures to identify potential technhal inaccuracies and

procedure enhancements. As a result, approximately 200 Procedure Work Requests

were generated.

Operations Department conducted several ' peer visits' to other licensees to assess-

procedure usage, review, and revision processes. As a result, an intemal team has

'been assembled to review the recommendations and observations from the ' peer

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L visits' and implement appropriate actions as necessary.

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t Additionally, management continues to emphasize the need for technically accurate

-. plant procedures and the need to review all procedures prior to use (if possible) to

' determine 'if any procedure deficiencies exist.

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VIOLATION lb ~

- The flaw in question was completely rentoved by grinding. The removal process

conflumed the nature of the flaw as originally characterized.-

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Additional examinations have been completed as a conservative measure to assure

integrity of similar components. No indications were identified.

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' CORRECTIVE STEPS TIIAT WILL BE_7

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VLQLATIONS

AKEN TO AVOID FURTIIER

VIOLATION la

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- All~ actions to avoid funher violations have been completed.

VIOLATION Ib

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All actions to avoid further violations have been completed.

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DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED

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VIOLATION la

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Full compliance was achieved on April 24,1997, when all NRC identified procedure

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. deficiencies were entered into the DAEC corrective action program,

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- VIOLATION 1b

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" ' Full compliance was achieved on April 10, 1997, when the identified flaw wa

- removed from the valve oy grinding.

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_VIOLATIQb' ~nVQ

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criterion XNf 10 CFR Part 50, Appendix B, requires, in part, that all testing required t

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-l demonstrate that structures, systems, and components will perform satisfactorily in

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(are identified and performed in accordance with written test procedures which inco

requirements and acceptance limits contained in applicable design documents.

' C' ontrary to the above:

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' As;ofApril 24,1997, design valve closure time acceptance limits as described in the U

Final Safety Analysis Report (UFSAR) for valves MO-2003 and MO-1905 had not been

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-incorporated:into surveillance test procedure (STP) 45A002-Q, " Low Pressure Coolant

injection Operability." Further, on April 24,1997, the licensee identified that for valves MO-

2117 and MO-2137, design valve closure time acceptance limits as described in the UFSA

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had not been; incorporated into stuveillance tesi STP 45A001-Q, " Core Sppy System

Quarterly Operability"(50-331/97006-03).

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This is'a Severity Level IV violation (Supplement 1).

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RESPONSE TO VIOLATION TWO

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REASON FOR THE VIOLATION

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On April 24, 1997, the NRC SOPI team identified that STP 45A002-Q, " Low

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Pressure Coolant injection Operability," Revision 19, listed the maximum acc

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closing time for MO 2003 as less than or equal to 19.5 seconds and less tha

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to 18.4 seconds for MO 1905. These times exceeded the design closure time of 18

seconds for the RifR system shutdown cooling discharge isolation valves described

UFSAR Section 7.3.1.1.1.7. A review of previously perfonned tests using the no

conservative acceptance limits showed that no closure times in excess of the UFSAR

value had been recorded for these valves. As a result of these discrepancies, an AR

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. was initiated. He reason for the acceptance limits centained in the STPs being n

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conservative with respect to those contained in the UFSAR was failure to adequa

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review and maintain the UFSAR while implementing the Inservice Testing Pr

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requirements into applicable STPs.

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CORRECTIVE ACTIONS TAKEN AND THE RESULTS ACHIEVED

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fA review of other valve stroke times located in the UFSAR identif

Spray system injection isolation valves, MO 2117 and MO-2137, had recorded

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! closure times in excess of the UFSAR value. .Specifically, on February 10,1997, the

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valve closure times for MO 2117 and MO-2137 of 8.13 seconds and 8.16

respectively had been recorded. The valves had acceptance criterion of 10 seconds

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. maximum in STP 45A001-Q, " Core Spray System Qua terly Operability," Revision

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L23.- These valve closure times and maximum allowable acceptance criteria were

contrary to UFSAR Section 7.3.1.1.!.7, which specified a design closure time of 8

seconds for these valves. 'An immediate operability determination was completed and

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determined that the valves were operable.

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Additionally, the following valves have been identified as having ASME Code

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Section XI acceptance critena difkrent from those established in the UFSAR; Core

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- Spray System valves MO-2115, MO-2135; and RHR System valves MO-2004 and

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. MO4904. However, the actual recorded stroke time values for these valves are

conservative with respect to the UFSAR values.

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lSTPs 45A001-Q (for Core Spray), and 45A002-Q (for RHR) were revised to reflect

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CORRECTIV

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STEPS TilAT WILL llE TAKEN TO AVOID FURTHER

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VIOLATIONS

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All corrective actions to prevent further violations have been completed.

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4.

DATE WHEN FULL COMPLIANCE WILL HE ACHIEVED

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Full compliance was achieved on July 17.1997, with the completion of the proced

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- VIOLATION THREE

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, ; Criterion V of 10 CFR Part 50, Appendix B, states, in part, that activities affe

4 shall. be prescribed by documented instructions, procedures, or drawings, o

appropriate to the circumstances and shall be accomplished in accordance with these

L instructions, procedures, or drawings.

Engineering Department procedure 1203.21, " Engineering Calculations," Revision 3, Sec

3.2 stated, " Engineering calculations shall be verified independently.by another individua

technically. qualified in the same subject. and who did not participate-in the' origina

calculation," and, " Engineering calculation numbers for all IES Utilities Inc. and all s

generated calculations shall appear as: CAL-XYY-ZZZ, . . . ."

Engineering Department procedure 1203.31, " Design Verification Procedure," Revisio

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.Section 3.4 required, " Design review of design document (s).or modifications will be

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y sufficient to verify the appropriateness of the design input; including assumptions, de

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- basis, and applicable regulations; codes and standards; and that the design is adequate

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- intended application. . . ."

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.' Administrative Control Procedure (ACP) 114.5, Revision' 9, paragraph 3.1.(2).(a) req

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review and documentation of operability evaluations on the Action Request (AR) form.

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Contrary to the above:

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a)' L As' of April 7,1997, a fracture mechanics calculation performed on March 29,199

accept a flaw in the body of check valve Vl9-001, lacked a documented independent

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review of the calculation ~and a calculation control number (50-331/97006-04

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b) ' As of April'24,1997, ~a calculation in engineering maintenance action A26702G, date

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September 19,1996, which demonstrated the acceptability of a rew relief valve on the B

residual heat removal heat exchanger did not' meet procedure 1203.31 requirements in

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that. the incorrect design code was used for stress level acceptance criterion, an

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% correct /non-conservative input ' alue was used, design assumptio

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the applicable code edition was not documented in this calculation (50-3

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Ec) On~ July.1,1997, following1the pressure locking 'of valve Vl9-148, an o

Jassessment in accordance with XCP ll4.5. was not documented in

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request,' AR 97-0094 (50-331/97006 04c).

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.nis is a Severity LevelIV violation (Supplement I).

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_ ; BESPONSE TO VIOLATION THREE

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REASON FOR THE VIOLATION

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VIOLATIdN 3'a:

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Ori March 23,1995, a 3/4 inch long surface breaking linear indication was d

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L in the base metal of the check valve body of the RHR pump D discharge chec

=(see response to Violation Ib abbve). This valve was accepted for continued serv

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i based on; a: fractwe mechanics ' analysis performed on March 29

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- documented in AR 95-0464.

1995, and

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on March 29,1995, to accept the flaw in check valve Vl9-0001, lacked a documented

independent review of the calculation and a calculation control number.

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Engineering Department procedure 1203.21,." Engineering Calculations," Rev

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Sectibn3.2Aroouired that engineering calculations be verified independe

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tanother individual technically. qualified in the same subject and who did not

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1The procedure also required engineering

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icalculatiorisumbers for all IE3 Utilities Inc. and all supplier generated calculations,

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dThe cause was determined to be failure to follow the calculation procedure.

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EVIOLATION34

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Engineering Maintenance Action (ENA) A267_02G revision 2, replaced the existin

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RHR Heat Exchanger _(HX) shell side relief valve,~ PSVl953 and gagged and

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abandoned in place PSVl952. J As a result of an engineering review associated with

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the preparation of the EMA,'a concern arore as to a discussion contained in revision I

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"The Engineering Evaluation in Rev i discusses the bellows seal on the 4 by 6

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steam service valve'and takes credit for the ability of this. seal to isolate the

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exhaust line to the torus from the top wmks of this valve. An additional featu

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the 6ellows seal is16t it also isolates the effect of back-pressure in t

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line from the set pressure of the valve. The torus has a potential to be p

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iduridd and after an accidsnt shich would have the affect of raising the

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a valve without"a bellows s.:al by an amount equal to the~ back-precure. The

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J installed 4 by 6 relief valve will be gagged and neither tne 3/4 by i valve w

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installed nor the Crosby 900 series to be installed have a bellows.? D

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Lis acceptable th defeat this feature "

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iln response to this concern an engineering evaluation' stated the following:

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' "The stresses caused by this short-lived over pressurization on the vessel a

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~ than ASME Code allowable "

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Ein support of this statement an engineering calculation was documented in

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verification summary, report section of this EMA; This calculation evaluated the

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, ? potential stress affects on the B RHR HX 'shell thtt would be created if the HX were

to be over pressurized and a 25 psig backi ressure from the torus was applied to

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[HX shell relief valve (LOCA^ could create this back-pressure). This calculation was

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l performed to: help LtheJverification engineer! understand"the relatively minor

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contribution an additional 25 psig of back-pressure would have on the vessel stresses

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iinduced bylthe HX design ' ressure limitc The calculation,was an estimate of the

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Jeffect.using nominal stress formula assumptions and typical modem code criteria.-

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lhe engineer did not intend this calculation to be a design review, rather a informal

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check of the physical significance of this theoretical additional pressure load. The

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[ basis for th'e relief valves not requiring back pressure protection was act

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'the fact that the relief valve setpoint was below the shutoff head of the RHR p

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. and therefore no credible sequence of events exist that would challenge the i

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• of the RHR HXl

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jAireviewlof th[calcstationjby members of your staff identified the fallowin

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r incorrect assumptions made in the calculation:

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calculatior;iCalculated stresses initheLHX shell were compared against

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allowableL stresses from? the ASME Code Sectica III,: which' were less

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conservative that the original design Code, ~which wasSection VIII, Division 1

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of the 1%8 Edition, Winter' Addenda.

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Anachment I to

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NG-971728

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Page 10cf D

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The full wall'RHR shell thickness of 7/8 inch was used in th

calculation vice reducing the wall thickness to an appn priate original

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value by subtracting the co Tosion allowance from the wall thickness, which

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introduced a non conservative error.

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O' Design essumptions, . applicnble Code F<tition and Addenda were not

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documented in this calculation.

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Upon further review, it was determined ~ that Engineering Department proc

1203.31, " Design Verification Procedure," Revision 6, Section 3.4, requ

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review of design document (s) or modifications will be sufficient to verify th

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appropriateness of the design input; including assumptions, design basis, and

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applicable regulations; codes and standards; and that the design is adequate f

intended application. . . ." Although the calculation performed was not intende

be a design review, it appears that the procedure could imply that this calcul

could require proper design verification in accordance with the procedure.

.

VIOLATION 3e

. On January 15,1997, during a plant shutdown to repair a leaking valve, the ma

gate valve Vl9-0148, RHR shutdown cooling suction header manual isolation v

was isolated for maintenance. Upon compledon of the maintenance, Vl9-0148 c

-

not be opened. It was postulated that the valve had likely pressure locked and

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' instructions were given to vent the bonnet by opening the valve's body vent

This venting process allowed the valve to be smoothly opened by the opera

'

= Engineering performed an evaluation of the pressure effects on the valve and

l determined that no damage was done as a result of the pressure locking incide

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As a result of. the engineering evaluation conclusions, no formal operabil

evaluation was completed. However, ACP 114.5 requires operability evaluation

be completed when safety related components are degraded such that perfo

(operability is called into question.

should have been performed in this instance and documented in the

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2.

CORRECTIVE ACTION 3 TAKEN AND THE RESULTS ACHIEVED

VIOLATION 3a.

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AR 971033971033was initiated. The calculation was properly verified, with no identified

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Attachment I to

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NG.971728

Page 11 of 13

[

VIOLATION 3b

Several actions have been recently taken to enhance the calculation proce

A detailed team review was perfonned of procedure

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Calculations," and several improvements were made. A new revision

_

7/3/97. Improvements included better description of responsibilities of sup

'

verifiers, and preparers, clearer definitions of when calculation designation

,

design c'ontrol and action requests. A bri'.:are required, and exp

,'

with preparing or verifying :alculation;,.Img was held for those personnel involved

The briefing highlighted the procedure

changes and restated the basic expectations for calculations The briefin

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reviewed the concerns which were noted for calculations by the NRC SOP 1

Engineering Managemc .. has noted a significant increase in sensitivity to

calculation process since these corrective actions were taken.

VIOLATION 3e

Two industry experts (Dominion Engineering and Anchor Darling) perform

reviews of the pressure locking incident and determined that the valve in

'3

not affected by the pressure locking. An additional inspection was perfor

24,1997.

The inspection, conducted by the valve manufacturer and DAEC person

reached conclusions supporting the original and subsequent evaluations that t

.

,

did noisustain any yielding as a result of the pressure locking, therefore furt

justifying the operability of the valve. However, as a result of this issue and o

similar issues,is has been determined that the Action Request System

not clear on when operability determinations need to be documented. Theref

'AR has been initiated to clarify operabliity determination requirements and take

.

actions as necesary.

3.

CORRECTIVE STi:PS TIMT WILL BE TAKEN TO AVOID FURT

^

VIOLATIONS

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VIOLATION 3a

All actions to avoid Srther mlations have been completed.

VIOLATION 3b

All actions to avoid further violations have been completed.

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NG-971728

Page 12 of13

. VIOLATION 3e

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. As stated above,'an AR has been initiated to review operability determinati

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requirements and- take actions as. necessary. This review will be co

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November 15,1997;

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DATE WHEN FULL COMPLIANCE WILL BF ACHIEVFR

- VIOL' ATION 3a

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' Full compliance was achieved on April 16,1997, with the verification of the

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calculation performed for AR 950464950464

VIOLATION 3b ~

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Full compliance was achieved on July 7,1997, with the issuance of the revised

-1203.21..

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VIOLATION 3e

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- Full compliance was achicved on May 24,1997, when the inspection of Vl9-014

l detennined that no yielding had occurred.

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VIOLATION FOUR

'

Criterion III of 10 CFR Part 50,- Appendik.B, states, in part, that measures stal

' verifying or checking the adequacy of the design, and that changes shall be

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i control mea:ures commensurate with those applied to the original design.

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i Contrary to the above:

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A'dequate design control measures were not used in calculation CAL M97

.

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February 3; 1997. A non-verified input assumption was used for the bonnet temp

' CAL M97-002,.which evaluated the pressure locked valve Vi9-0148. This calculation

- a' straight average between the drywell temperatu e and the reactor coolant system w

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supporting thermal analysis to confirm or bound the valve bonnet temperature he

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(50-331/97006 10).

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This is a Severity Level IV violation (Supplement I).

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At+achment I to

NG 97-1728

Page 13 of 13

RESPONSE TO VIOI:ATION FOUR

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1.-

REASON FOR THE VIOLATION

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'On January 15,1997, during a plant shutdavn to repair a leaking valve, the manual

-

gate valve Vl9-0148, RilR shutdown cooling suction header manual isolation v

was isolated for maintenance. Upon completion of the maintenance, Vl9-0148 co

not be opened.

It was postulated that the valve had likely pressare locked and

instructions were given to vent the bonnet by opening the valve's body vent va

His venting process allowed the valve to be smoothly opened by the operator. A

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result, the valve operability was not a concern and the plant was started up. As

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conservative action, engineering performed an evaluation of the pressure effects on

the valve and determined that no damage was done as a result of the p

incident.

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Further review of the engineering calculation used to evaluate the pressure effects

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discovered a non-conservative temperature assumption, which may have affected

evaluation. Therefore, the original operability determination (which was not

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documented on the AR) was brought into question.

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- CORRECTIVE ACTIONS TAKEN AND THE RESULTS ACHIEVED

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Two industry experts (Dominion Engineering and Anchor Darling) performed

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reviews of the pressure locking incident and determined that the valve integrity

not affected by the pressure locking. An additional inspection was performed

24,1997.

The inspection, conducted by the valve manufacturer and DAEC personne

,

reached conclusions supporting the original and subsequent evaluations that the va

' did not sustain any yielding as a result of the pressure locking.

.

..

C' ORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID

3.

3'IOLATIONS

All actions to avoid further violations have been completed

4,

DATE WHEN FULL COMPLIANCE WILL BE ACIHEVED

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Full compliance was achieved on May 24,1997 with the satisfactory completion

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the valve inspection.

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