ML20078C557

Forwards Response to Violations & Proposed Imposition of Civil Penalty in Amount of $300,000 Noted in Insp Rept 50-298/94-14.Corrective Actions:Walkdowns of Primary Containment Penetrations Performed

ML20078C557
Person / Time
Site:	Cooper
Issue date:	01/18/1995
From:	Horn G NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ENFORCEMENT (OE), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS950028, NUDOCS 9501260384
Download: ML20078C557 (19)

Text

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5 GENERAL OFFICE

.t P O BOX 499. CoLUMOUS. NEBRASKA 68602-0499 Nebraska Public Power District

T /xi @ d n ' "

GUY R. IIORN Yke-President, Nuclear (4:32) 563-5518 NLS950028 January 18,1995 Director, Ofuce of Enforcement U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Gentlemen:

Subject:

Reply t, a Notice of Violation and Proposed Imposition Civil Penalties; NRC Inspection Report Nos. 50-298/94-14,50-298/94-16, and 50-298/94-19; Cooper Nuclear Station, NRC Docket 50-298, DPR-46

Reference:

Letter from Mr. J. L. Milhoan (USNRC) to Mr. G. R. Horn (NPPD), dated December 12,1994, Notice of Violation and Proposed Imposition of Civil Penalties - $300,000 (NRC Inspection Reports 50-298/94-14, 50-298/94-16, and 50-298/94-19).

This letter, including Attachments 1 and 2, constitute Nebraska Public Power District's (the District) reply to the referenced Notice of Violation (NOV) and Proposed Impo"ition of Civil Penalties in accordance with 10 CFR 2.201. Attachment 2 is a certified check in the amount of $300,000, for payment of the civil penalties. Per conversation with Mr. G. F. Sanborn, the submittal date of this response was extended to January 18,1995.

The referenced inspection reports document the results of three NRC inspections conducted from May 23 through August 12,1994, to specifically review the circumstances regarding: (1) the identification, in May and June,1994, by the Cooper Nuclear Station (CNS) staff of numerous Primary Containment penetrations that were installed in configurations that were contrary to NRC requirements and had never been local leak rate tested, (2) the extent of the conditions causing both Emergency Diesel Generators (EDGs) to be declared inoperable, which had resulted in the declaration of a Notification of Unusual Event and plant shutdown on May 25,1994, and (3) the identification by the CNS staff of numerous hardware deficiencies that resulted in the failure of the Control Room envelope pressurization test on April 11,1994.

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l As discussed in Attachment 1 to this letter, each of the " problems" and violations had common causes relating to NPG management and culture. As you are aware, NPG management changes have been implemented to address these issues. The breadth of personnel changes reflect the District's commitment to incorporating broad industry /-

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experience and achieving higher performance standards. More specifically, management i l oversight of the Condition Reporting Process has been increased to ensure adequate rigor and i

i urgency are applied in the evaluation of non-conforming plant conditions as they are khkhk bo98

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pirector, Office of Enforcement U.

S. Nuclear Regulatory Commission January 18, 1995 Page 2 of 3 identified. Also, focused attention is being placed in the review of Operational Experience, a' program that is considered key in early identification and timely resolution of potential issues.

Management communications have been improved through: (a) training for NPG Managers

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and Supervisors on important management topics including teamwork and communications, i

(b) daily management meetings to communicate priorities and standards, and to ensure inter-departmental coordination, and (c) implementation of a revised corrective action program that provides focus on organizational, programmatic and human performance concerns.

l In sununary, the District has taken, and will continue to take aggressive actions responsive to the mancgement issues that have negatively impacted the NPG safety culture. Steps l

l taken that have resulted in the progress made to date have been significant, and the District believes that the CNS " culture" issue is no longer a factor with regard to the recurrence of these violations. The majority ofissues and corrective actions discusced herein have been i

addressed in the July 28 and August 8,1994 responses to NRC Confirmatory Action Letters (CAL) dated May 27, June 16, July 1, and August 2,1994; during the September 16,1994 Enforcement Conference; and in the November 7,1994 letter to the NRC. As such, the District has not addressed in this response, any perceived differences of opinion since the cited violations should not have occurred in any event, and since they required significant attention before CNS could be considered ready for restart.

l Should you have any questions concerning this matter, please contact my office.

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G.

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ViceWsident - Nuclear l

Attachments cc:

Regional Administrator USNRC Region IV NRC Resident Inspector l

Cooper Nuclear Station l

l NPG Distribution i

pirectior, Office of Enforcement U.

S.

Nuclear Regulatory Commission January 18, 1995 Page 3 of 3 STATE OF NEBRASKA)

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PLATTE

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G. R. Horn, being first duly sworn, deposes and says that he is an authorized representative of the Nebraska Public Power District, a public corporation and political subdivision of the State of Nebraska; that he is duly authorized to submit this response on behalf of Nebraska Public Power District; and that the statements contained herein are true to the best of his knowledge and belief.

6,m tt, ___

G. R. Horn

/ E r8 day of 26 W,1995.

Subscribed in my preser. e and sworn to before me this l

6 d6d NOTARY PUBLIC d

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Page 1 of 16 l

i REPLY TO DECEMBER 12, 1994 NOTICE OF VIOLATION AND i

PROPOSED IMPOSITION OF CIVIL PENALTIES - EA NOS.94-164, 94-165,94-166 COOPER N'3 CLEAR STATION l

I NRC DOCKET NO-50-298, LICENSE DPR-46 l

During NRC inspections conducted from May 23 through August 12, 1*)94, violations 1

of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, Appendix i

C, the Nuclear Regulatory Commission proposed civil penalties pursuant to Section 234 of the Atomic Energy Act of 1954, as amended, 42 U.S.C 2282, and 10 CFR

)

2.201.

j For each of the violations and " problems," inadequate rnanagement performance significantly contributed to the existence or perpetuation of the deficiencies.

Therefore, management related corrective actions addressed in the cover letter to this response and in other District correspondence to the NRC (i.e., District letters to the NRC dated July 28, 1994 (NLS940001); August 8, 1994 (NLS9400026);

and November 7, 1994 (NLS940111)) should be considered part of the District's corrective actions. For brevity, this corrective action is not restated in each violation response.

The particular violations and the District's replies are set forth below:

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PROBLEM AREA A-Primrv Containment Intecrity Violations I.

Violation A.1 4

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Violation A.1 contained in Reference 1 cites the following:

" Technical Specification 3. 7. A.2.a

• Containment Integrityr* states, in r

parto that

• primary containment integrity shall be maintained at all times when the reacCor is critical or when the reactor water tespperature is i

above 212*F and fuel is in the reactor vessel... "

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• Technical Specification Burveillance Requires;ent 4.7.A.2.f.lo " Leak Jtate Testing, " states, in part, that *... local leak rate tests (LLRTs) shall be i

performed on the primary containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shutdown for refueling, or other convenient: intervals, but in no case at intervals no greater than two years....

The total acceptable leakage rate for all valves and penetrations other than the NBZVs (main steam isolation valves) is 0.60 La.

" Technical Specification 1.Y, "Burveillance Frequency," states, in parto that

• performance of a Burveillance Requirement within the specified time interval shall constitute contpliance with operability requirements for an LCO [ limiting condition for operation] unless otherwise required by the specifica tion. "

• Contrary to the above frcun January 18, 1974, until Nay 27, 1994, primary r

containment integrity was not maintained at all times when the reactor was critical or when the reactor water tentperature was above 212*F and fuel was in the reactor vessel in that the Burveillance Requirement for the local leak rate testing of 82 congponents had never been ingpiemented at an interval not to exceed two years. As the result of testing conducted on June 23, 1994, Isolation Valve 7?-65CV (one of the 82 coagponents) failed the LLRT, resulting in a tota" leakage v&1ve that significantly excseded the 0.60 La limit.

The 0.60 La limit corresponds to a leakage rate of

5. 3 7 scab (189. 60 scfh). The LLRT failure of Valve IA-65CV resulted in a total leakage rate that exceeded 17.66 scab (623.57 scfh).*

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Attacliment 1 -

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i Aam4s'sion or Den 4al t3 violation' I

The District admits the. violation.

I Reasons-for Violation

.The immediate cause.for not performing local leak rate testing (LLRTs) on 82 components is that they had not been - previously identified for inclusion in the CNS 10 CFR 50 Appendix J Program.

As discussed in LER 94-011 Revision 1, the root cause for this failure was lack of management-commitment to program implementation, in that the organizational focus for problem identification and resolution was primarily ' compliance-based.

This resulted in insufficient attention being paid to evolving regulatory:

issues in this

area, for which a compliance-based standard was-inappropriate,

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When the CNS Operating License was issued (1974), the specific testable!

Primary Contsinment penetrations and Primary Containment Isolation Valves l

(PCIVs) were listed in the CNS Technical Specifications. Eventually, this list was shifted to the' Updated Safety Analysis Report (USAR) throughLa.

License Amendment. The original licensing basis was that performance of-l l

these specific tests constituted a method for Appendix-J compliance that l

was acceptable to the Atomic Energy Commission (AEC).

The mindset was that since the design of many of the CNS systems pre-dated the issuance of l

Appendix J

for public

comment, adherence to the

. Technical, Specification /USAR list (along with other specific commitments that might l

be made) was all that was required, despite'later regulatory positions j

that contradicted this approach..Although these deficiencies:were self-l identified as a result of broad corrective action' in response to Inspection Report 93-17, the District duly acknowledges that this condition should have been recognized and corrected long ago.

corrective Steos Taken and-the Recults Achieved i

In addition to the generic NPG culture improvements that ' address.the root cause as previously noted, other more programmatic corrective actions have been taken. As stated in LER 94-011, Revision 1, walkdowns of the Primary l

l containment penetrations have been performed. This activity contributed r

l to the District's confidence that the scope of the Appendix J non-compliance has been comprehensively identified.

(See also, Confirmatory Action Letter Response dated July 28, 1994.)

The following courses of action were then followed:

(1)

As-found testing was performed for penetrations that had not previously been Type A, B,

or C tested and for which this testing l

was determined to be immediately practicable. Those that were not

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tested were either modified and then tested, or were designated as

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candidates for Appendix J exemption. The total os-found leak rate-l for the testable penetrations, except X-22 which curained drywell l

pneumatic supply check valve IA-CV-65CV, was 26 SCFH.

With regard I

to penetration X-22, leakage through IA-CV-65CV was simificant, preventing pressurization of the penetration using normal leak rate testing apparatus.

Accordingly, worst-case leakage was examined during safety consequence assessments that were performed.

Penetration modifications and component maintenance / replacement were performed.

Subsequent testing has verified that the total Primary containment as-found leak rate is less than the Technical Specification limit.

t (2)

Design changes were implemented, which included addition of test connections, installation of welded caps on spare penetrations, i

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Attach' ment 1 to NLS950028 Page 3 of 16 1

complete redesign of some containment isolation barriers, and l

installation of caps on vents, drain lines and test connections.

These design changes have been completed.

(3)

For penetrations and components that have been deemed impractical to test in accordance with the ' requirements 'of Appendix J,

NRC i

exemptions have been obtained.

'l Corrective Stens That Will Be Taken to Avoid Further Violations Appendix J compliance accountability has been improved by redefining

" program. owner" responsibilities to require the primary duty tc, be 1

ensuring the integrity of the overall program, rather than merely' functioning as a testing facilitator.

To assist in this objective, the i

current licensing basis for Appendix J testing will be formally captured in an Appendix J testing basis document.

This will provide a readily available, controlled source of comprehensive information to the Appendix J program owner which will facilitate the correct disposition of future Appendix J issues as they occur.

Date When Full Comoliance Will Be Achieved i

CNS is now in full compliance with the testing requirements necessary to demonstrate Primary Containment Integrity.

l II.

Violation A.2 Violation A.2 contained in Reference 1 cites the following:

• 10 CPR Part 50, Appendix B, Criterion XIa " Test Control," states in o

part, that *(a) test program shall be established to' assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and perfo2med in accordance with written test procedures which incorporate the requiramants and acceptance limits contained in applicable design documents. "

"CNB Quality Assurance Program for Operation Policy Directive Revision r

10, Section 2.11, written to implement the requirements of 10 CPR Part 50, Appendix B,

Criterion XI requires that each type of test program

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p;erformed will be defined by written procedures and instructions, and it l

requires that acceptance tests will be developed for structureso systems, and components to demonstrate their capability to perform satisfactorily following repairs or modification.

" Contrary to the above, the licensee did not assure that all testing was idenC1fied and periormed in accordance vith written teat procedurea which incorporated the requirements and acceptable limits. Specifically as of

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Nay 14, 1994, 68 components passing through 54 primary containment penetrations, each required to be local leak rate tested in accordance with che requirements of Technical Specification Burveillance Requirement

4. 7. A. 2. f.1,

" Leak Rate Testing,

• had not been identified in a procedure as requiring local leak rate testing,- as required by CNg Quality Assurance Progran for Operation Policy Directive, Revision 10, Section 2.11.

These l

components had never been local leak rate tested.

" Contrary to the above, the licensee did not assure that all-testing was j

identified and performed in accordance with written test procedures which iacorporated the requirements and acceptable limita. As of Jtano 21, 1994a instrument pressure switches PC-PB-12Ar Br C, and Ds PC-PR-101A, m C, sad a

Di PC-PR-119A, B,

C, and Di PC-PR-161 and PC-PT-512A and B, each required to be local leak rate tested in accordance with the requirements of I

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to NLS950028 Page 4 of 16 Technical Specification Burveillance Requirement 4.7.A.2.f.1,

" Leak Rate Testing," had not been identified in a procedure as requiring local leak rate testing after being isolated from the containment integrated leak rate test, as required by CNS Quality Assurance Program for Operations Policy Directive, Revision 10, Bection 2.11.

These switches had never been local leak rate tested."

Admission or Denial to Violation The District admits the violation.

Reasons for Violation The causes for not establishing written LLRT procedures for the components listed in the violation are identical to those discussed in response to violation A.l.

A lack of rigor in the Appendix J program (causert by a compliance based management philosophy) resulted in not identifying all of the components for which Type C LLRTs were practicable, and in isolating components which should have been left unisolated for inclusiin in the Type A ILRT boundary. Had all components within the Appendix J scope been previously identified and properly evaluated, they likely would have been included within this testing control program.

Please refer to the discussion under Violation A.1 for a more complete review of this issue.

Corrective Steos Taken and the Results Achieved The corrective actions discussed for Violation A.1 describe the efforts that have been made to comprehensively identify the additional components that require Appendix J testing.

These efforts have been completed, and changes to the CNS Appendix J testing procedures have been made which reflect current licensing basis testing requirements.

Corrective Steos That Will Be Taken to Avoid Further Violationa As discussed in the District's response to violation A.1, a testing basis document is being prepared that will provide a clear connection between the Appendix J requirements, the CNS licensing basis with respect to Appendix J compliance, and the testing procedures that implement the CNS licensing basis.

1 Date When Full Comoliance Will Be Achieved CNS is now in full compliance with the requirement to have written procedures that encompass the full scope of the 10CFR50 Appendix J testing program.

III.

Violation A.3 i

violation A.3 contained in Reference 1 cites the following:

• 10 CFR Part 50, Appendix B, Criterion III, " Design Control," states in part, that "[aleasures shall be estabilshed to assure that... the design basis...[is]

correctly translated into specifications,

drawings, procedures, and instructions. These measures shall include provisions to l

assure that appropriate qus11ty standards are specified and included in l

design documents and that deviations from such standards are controlled. *

• Draft General Design Criterion 53, a measure written to contply with the requirements of 10 CFR 50, Appendix B, Criterion III, as committed to in Appendix F of the tipdated Safety Analysis Report (liBAR), states that

• (p]enetrations that require closure for the containment function shall be i

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to NLS950028 Page 5 of 16 1

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protected by redundant valving and associated apparatus.

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• Draft General Design Criterion la as comunitted to in Appendix F of the tTBAR, states that *Et] hose systems and components of reactor facilities which are easential eo the prevention af accidenta which could eifect the a

public health and safety or mitigation of their consequences aball be l

identified and then designed, fabricated, and erected to quality standards i

that reflect the importance of the safety function to be performed. "

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• Qeneral Electric Design Specification No. 22A1153, " Codes and Industrial

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Btandard, " Revision la states, in Note 3 of the Appendix, that * (pfipings l

which is an integral part of the primary containment for isolation' i

purposes, shall have at least the same quality and levels of assurance as i

the primary containment."

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• Contrary to Criterion ZZZ, the licensee did not assure that the above E

l design bases were correctly translated into specifications and i

instructions and did not assure that deviations from quality standards l

were controlled. Specifically:

s a.

As of May 14, 1994, numerous primary containment penetrations had no redundant valving.

These penetrations included, but were not l

Limited to, Penetrations X-21, X-22, X-25, X-293, X-30E/F, X-33E/F, I

X-209A/B/C/D, and X-218.

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b.

As of February 22, 1994, 10 penetrations consisting of manually i

operated vents, drains, and test connections and requiring closure j

for the containment function, had a singin manual isolation valve

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for containment isolation as opposed to che required redundant valving and associated apparatus.

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c.

During an NRC inspection conducted June 13, 1994, through August 12, 1

1994, it was determined that approximately 300 containment O

penetrations had not been designed, fabricated, or installed to the l

same standards as the primary containment because these congponents l

had not been correctly classified as essential. As a result, these penetrations had not been designed, fabricated, or erected to l

quality standards that reflect the ingportance of the safety function to be performed (e.g., some welds were not nondestructively tested, i

some penetrations were not local leak rate tested, and the penntrations were not treated as safety-related by the licensee *s quality assurance program)."

Admission or Denial to Violation The District admits the violation.

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Reasons for Violation As noted in this Violation, there were two areas of non-conforming Primary

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Containment penetration design:

(1) the lack of Primary Containment 1

penetration barrier redundancy for all process lines passing through the 4

i Primary Containment, and (2) the improper classification and maintenance 4

of many penetrations as non-essential. These areas are discussed in more detail below.

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j (1)

Barrier Redundancy-The cited Primary Containment penetration barrier redundancy discrepancies involve manual valve configurations j

on penetrations designated for process lines. As noted by the NRC in Enclosure 2 of the NOV, these configurations were part of the i

original plant design. Specifically, the configurations were built l

to be in conformance with the 1967 Draf t AEC General Design Criteria 1

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(GDC), which did not explicitly require redundancy, for process lines-i isolated by manual valves.

In addition, the response to Final l

Safety Analysis Report (FSAR) ' Question. 5.5 clearly identified ' the

. j District's position with respect to GDC 55 and 56, and'Eafety Guide l

11.

In response to Question 5.5, the District stated that a single manual valve would be employed for instrument' lines and lines to-control systems or devices inside the Primary Containment, including pneumatic lines for valves,-dampers, etc.

(Regulatory Guide 1.11 still permits the use of single manual valves for some applications l

r and the GDC 'still provides forJ the acceptability of containment

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isolation provisions "on some'other defined basis".)

The CNS SER makes it clear that the AEC's' technical review, for '

initial licensing was performed based on 10CFR50 Appendix A j

criteria, within which, criterion 55 provides more prescriptive redundancy requirements.. It is unclear to what e degree the < AEC i

reviewed the redundancy issue, but the conclusion was' reached in Section 3.1 of the SER that the plant conformed to the intent of the Appendix A criteria.

The failure of the NPG to. resolve the ambiguity of the licensing basis was a key factor in the perpetuation of non-redundant Primary-Containment penetration-l barrier configurations, both in the original design, :and after.

various design, changes.

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In two of the cited examples, it is apparent that there was a. lack, of rigor in maintaining configuration-controls, Therefore, apart from the licensing basis ambiguities, inappropriate configuration management contributed to the existence of non-redundant-or unqualified barriers.

l (2)

Eenetration classification-The penetration misclassifications l

resulted from an original design error, in that, the requirements of General Electric Specification 22All53 for piping passing through the Primary Containment were not correctly. translated into the f

l piping specifications. The passive function of helping to maintain i

Primary Containment integrity was not recognized as - a safety I

function for otherwise non-essential piping.. As a result, piping segments were inappropriately procured, fabricated and maintained to 1

l the requirements of USAS B31.1, rather than to a level of quality

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commensurate to that applied for the Primary Containment (as in USAS l

B31.7).

The impacts of this discrepancy were that in certain

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instances:

(a) material traceability was not maintained as applicable for the piping and components; (b) appropriate non-destructive evaluations (NDE) were not performed on all applicable welds and piping; and (c) applicable non-destructive testing (NDT) was not performed on pressure retaining welds.

Although both of these non-conforming areas were self-identified as a result of design basis reconstitution ef forts and broad corrective acti'ons taken in response to Inspection Report 93-17, the District recognizes.that these conditions should have been identified and corrected much earlier..

1 Similar to the previous two violations, the failure to more promptly I

adentify and correct these deficiencies was due in part to a compliance-based focus, such that, undue reliance was placed in the continuing adequacy of the original plant design, based on AEC approval during initial licensing.

Corrective Steos Taken and the Results Achieved l

The District addressed actions regarding barric-redundancy and I

penetration classifications in a letter dated Augus:

8, 1994.

The following is a summary of corrective actions noted in that letter, as well y

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as additional responsive activities.

(1)

Barrier Redundanev-Walkdowns'of Primary Containment penetrations were conducted which verified the as-built barrier configurations.

As a result of identified discrepancies associated with inadequate Primary Containment penetration barrier redundancy, design changes have been developed and completed which have brought them into conformance.

Also, programmatic enhancements have been made to control vent, drain, and test line barrier configurations as discussed in the District's response to Inspection Report 94-03, t

l (2)

Penetration classification-A document review was performed for all Primary Containment piping and instrument penetrations to determine the scope and extent of the misclassifications.

Welds in penetration-attached process lines, for which original construction NDE was insufficient, were identified..Those that were found to be in non-compliance or indeterminate were subjected to additional NDE.

Five welds were found to have rejectable NDE indications and were repaired or replaced as deemed appropriate.

The piping and instrument line segments up to and including the PCIVs have been determined via a Design Change to be of equivalent quality to the Primary Containment, and a reconciliation has been performed between USAS B31.1 and B31.7 for these segments.

These penetrations and components are now treated as Essential IIN, and will be maintained j

under the District's ASME Section XI Program.

Corrective Steos That Will Be Taken to Avoid Further Violations I

No further directly related corrective steps are planned.

However, as stated during the Enforcement Conference, the two ongoing actions i

described below help to prevent recurrence of similar types of violations.

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(1)

To help identify other licensing basis issues stemming from the original design, the design basis reconstitution effort is being expedited.

(2)

The ASME Section XI boundaries are being reviewed to identify and l

resolve other potential pressure boundary classification errors. A section XI classification boundary basis document is being prepared that identifies the Section XI boundary and defines its bases.

Date When Full Comoliance Will Be Achieved l

CNS is now in full compliance with the requirements for Primary Containment penetration safety classification and PCIV redundancy, as the District understands that the one remaining single-valve PCIV process line is acceptr.ble to the NRC.

PROBLEM AREA B-Ocerability of the 480 Volt and 4160 Volt Buses I.

Violation B_1 l

Violation B.1 contained in Reference 1 cites the following:

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" Technical Specification 3.9.A.1.c

• Auxiliary Electrical Equipment,"

a requires, in part, that the reactor shall not be made critical from a cold Bhutdown Condition unless the 4160 volt critical buses 1F and 1G and the l

480 voit critical buses 17 and 1G are energised, and the us1dervoltage and loss of voltage relays, as well as their auxiliary relays, are operable.

" Technical Specification Surveillance Jtequirement 4.9.A.1.a

• Daergency s

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Page 8 of 16 Buses undervoltage relays, " states that "once every 18 months, loss of voltage on emergency buses is simulated to demonstrate the load shedding j

from emergency buses and the automatic start of diesel generators. " USAR i.

Section 2.2.7.2.1.a

• Rtandby A-C Power (Diesel Generators)

Test s

Capabilityo* defines the function of the protective scheme as providing for the clearing the buses of all motor loads estcepting supply to the 480 voit critical unit substation.

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• Technical Specification 1.Yr "Burveillance Frequencyo* statec in part, a

that

• performance of a Burveillance Jtequirement within the specified time interval shall constitute congpliance with operability requirements for an l

LCO [ limiting condition for operation) unless otherwise required by the specification.

" Contrary to the above from January 18, 1974, until May 25, 1994, the r

reactor had been made critical without 4160 volt critical buses 1F and 10, and 480 volt critical buses 1F and 10 being operable in that the undervoltage relays associated with several.of the electrical loads supplied by these buses had never been tested to demonstrate their operability or upon testing, failed to perform their intended function of shedding their respective electrical loads from these buses. "

Admission or Denial to Violation The District admits the violation.

Reasons for Violation The root cause of this violation was the CNS failure to view existing programs and methods with a self-critical end questioning attitude. With respect to fulfilling the surveillance requirements that demonstrate operability, this resulted in surveillance procedures that did not fully test the load shedding function.

This function encompasses the circuit path from the sensing of undervoltage on the 4160 volt and 480 volt busses i

l through the opening of the respective circuit breakers on undervoltage.

The requirement to test this function stemmed from a 1979 License Amendment that specifically incorporated this Surveillance Requirement l

into the CNS Technical Specifications.

The reconciliation performed in 1979 between the load sequence testing procedures and these new requirements was unsatisfactory in that all the required loads were not verified to shed on undervoltage (or loss of voltage), and that the logic i

system functional testing (LSFT) associated with actuation of the undervoltage relays was not comprehensive.

Numerous opportunities occurred for earlier identification. Most notably during the design basis reconstitution effort, through NRC information to the industry on Westinghouse DB-50 circuit breakers and deficient safety-related LSFTs, and through an Electrical Distribution System Functional j

Inspection (EDSFI) perfc,rmed at CNS.

However, given the NPG organizational focus that was previously in place, these opportunities were not utilized as vehicles for broader programmatic inquiry.

Corrective Stens Taken and the Results Achieved Several steps were taken to address this violation:

l (1)

LSFTs for 4160 volt buses IF and 1G, and 480 volt buses IF and 1G l

were satisfactorily performed.

(2)

The applicable surveillance procedures were revised to verify that i

the load shedding function occurs as required.

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(3)

An electrical calculation was performed which demonstrated that even if load shedding of all of the non-essential 480 volt' loads failed to occur, the diesel generators would have performed their intended safety function.

To correct the long-term reliability issues associated with the load shedding capability of Westinghouse DB-50 circuit breakers, a Design Change was implemented which resulted in replacement of the undervoltage trip devices with shunt trip devices within those safety-related circuit-breakers that are credited with shedding a non-essential load.

Industry experience has shown this to be an effective configuration.

j corrective Stens That Will Be Taken to Avoid Further Violations 1

This violation has prompted a broader inquiry into the adequacy of the CNS surveillance Testing Program. The intent of this - Cort is to verify that 11 of the surveillance testing requirements have been correctly I

cranslated into the surveillance procedures.' This effort is more fully discusred in the corrective action for Violation B.2.

Date When Full Comoliance Will Be Achieved i

CNS is now in full compliance with the requirements of Technical Specifications 3.9 A.l.c and 4.9.A.l.a.

i II Violation B.2 Violation B.2 contained in Reference 1 cites the tol)Jwing:

• 10 CFR Part 50, Appendix Br Criterion KZe

• Test Ctatrolo* staten in s

part, that *[a] test program shall be established to assure that all i

testing required to demonstrate that structures systems, and connponents o

will perfom satisfactorily 'in service is identified and perfomed in accordance with written test procedures which incorporate the requir===nts l

and acceptable limits contained in applicable design documents."

" Contrary to the above the licensee did not assure that all testing was r

identified and perfomed in accordance with written test procedures which incorporated the requirements and acceptable limits. Specifically a.

During an NRC inspection conducted Nay 23, 1994, through August 12, 1994 Procedure 6.3.4.3o

• Requential Esoading of Rmergency Diesel Generators " Revision 31, which is performed to satisfy Technical o

8pecification Burveillance Requirement 4.9.A.1.a "Esons of Voltage o

Relays,

• was determined to be inadequate because it did not assure that the emergency diesel generators and critical buses would perfom satisfactorily in service in that the procedure did not l

contain requirements to verify that the 400-volt supply breakers for safety-related and nonsafety-related loads would shed from their l

electrical buses within a specifled time, nor did the procedure l

identify that the control rod drive puanp motors and station air coalpressors were required to be shed from the electrical bus.

b.

During an inspection conducted Nay 23, 1994 through August 12, 1994r the NRC identified that Procedure 6.3.20.1o "RRR Bervice Sta ter Booster Pualp Flow Test and Valve Operability Testa

• Revision 27, did not provide for the testing of the load shedding feature of the supply breakers associated with the 4160 voit residual heat removal service water booster pualps. "

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-Admission or Denial to Violation i

i The District admits the violation.

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f Reasons for Violation The root-cause for not establishing written ' test procedures that-i adequately. reflect the Technical Specification requirements-is-i attributable to the same NPG cultural issues discussed in Violations A.1 and B.l.

Furthermore, as discussed in' Violation B.1, this generic cause manifested itself through:

I (1)

Not ensuring the incorporation of all load shedding verifications

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(including associated LSFTs) into the surveillance procedures when they first became recognized as surveillance requirements.

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(2)

Failure to recognize and correct the procedural deficiency in a more timely manner, particularly with respect to the' opportunities that occurred that might have prompted such recognition. These included the design basis reconstitution effort, industry operational l

experience with respect to Westinghouse DB-50 circuit breakers and j

inadequate LSFTs, and a previous EDSFI.

j Corrective Stens Taken and the Results Achieved The surveillance procedures cited in this violation have been revised to l

reflect appropriate load shed testing. Additionally, as discussed in the l

District's August 8,

1994, response to an NRC Request for Additional' Information, the investigation into the deficiencies of Surveillance Procedure

6. 3. 2 0. l' prompted a review of the Logic System Functional i

Testing performed for se eral key safety systems.

This review has identified significant testing omissions that are being addressed as documented in LER 94-009.

i Corrective Stens That Will Be Taken to Avoid Further Violations

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I CNS is currently verifying that all surveillance requirements contained in the CNS Technical Specifications have been adequately translated into surveillance procedures.

In summary, each Technical Specification surveillance line item is being compared with its analogous implementing procedure to determine exactly how the requirement is met, and whether the procedure is satisfactory. This judgment is being made with reference to various source documents such as elementary diagrams, flow diagrams, the USAR, and Design Criteria Documents, as applicable.

Upon completion of this project, CNS will have system packages that fully document compliance with the Technical Specification surveillance requirements.

Date When Full comoliance Will Be Achieved Verification that the key safety system surveillance requirements are adequately described by written procedures will be completed prior to startup. As discussed in the CNS Phase 1 Plan, these key systems include i

the Automatic Depressurization System, Core Spray System, High Pressure Coolant Injection System, Low Pressure Coolant Injection System, Reactor Protection System, Standby Gas Treatment System, Control Room HVAC System, and Reactor Building HVAC System. Verification that appropriate written procedures encompass all surveillance requirements will be achieved by July 31, 1995.

PROBLEM AREA C-Onerability of the Control Room Emeroency Filter System I

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Violation c.1 Violation c.1 contained in Reference 1 cites the following:

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• S'echnical Specification 3.12. A.1r " Control Room Rlmergency Filter Systen

• o states, in parte that *...the Control Room Rlsergency Filter system...shall be operable at all times when containment integrity is required.

"The Order Confirming Licensee Commitments on Post-SWZ Related Issues o dated Otsly 10, 1981, confirms NPPD's comaltmnt to connplete JRTREG-0737,

" Clarification of Sw! Action Plan Requirements

• Item ZZZ.D.3.4s " Control o

Room Eabitability." Item ZZZ.D.3.4 involves the review of facility design requirements against the Standard Review Plan.

The N9PD response to' Qeneric Letter 80-90, dated December 30, 1980, submitted the control room habitability evalueCion, which aCatedo ia part, *the CNB control room ventilation system is designed to maintain the control room at about 1/4 in. E,0 (0.031 kPa) positive pressure by supplying air at a high enough pressure that even when system losses and the booster exhaust fan t

pressures are accounted for, the control room pressure is still l

positive...'

'A Befety Evaluation Report for the Cooper Station from the Accident Kvaluation Branch on JR1 REG-0 73 7.

Item No, IZZ.D.3.4,

• Control Room Babitabilityo* dated February 24, 1962, states in parts that

"...the o

design meets the criteria identified in Item III.D.3.4 of NORRG-0737 and is acceptable.

" Contrary to the above, from Otsne 1989 until April 20, 1994, the Control Room Rmergency Filter system was not operable at -all times.when contain=nt integrity was required in that testing failed to demonstrate that a positive pressure could be malatsined in the control room during the periodic performance of the control roonn envelope pressurisation i

test.*

l Admission or Denial to violation l

The District admits the violation.

Reasons for Violat'on The circumstances surrounding the prolonged inoperability of the Control Room Emergency Filter System (CREFS) were provided to the NRC in LER 94-006.

As discussed in this LER, the unrecognized inoperability of CREFS was primarily the result of a plant culture that did not approach operability issues with a self-critical and questioning attitude.

Also contributing to the deficiency was an incomplete understanding of the original system design criteria, which led to unsubstantiated reliance on the adequacy of the perceived licensing basis.

Several opportuniti.?s occurred to address identified deficiencies in the both the design e 3 performance of the system.

These opportunities were missed because of a design basis that was not well defined, inadequate testing, a compliance-based approach to operability, and a failure to implement adequate corrective accion even though the pressurization test results were marginal. As a result of these collective deficiencies, the system should not have been considered operable.

Corrective stens Taken and the Results Achieved As discussed in LER 94-006, corrective actions have been taken that have restored CREFS to operability.

Specifically, door seals in the Control Room envelope were repaired, penetrations were sealed and the adjacent building ventilation control systems inspected and repaired. Testing was

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Attac$ ment 1 to NLS950028 Page 12 of 16 performed that confirmed positive pressurization between the range of j

+0.04" to +0.05" wg with respect to atmospheric pressure.

As discussed in the District's July 28, 1994, letter to the NRC regarding CREFS commitments, the following additional corrective actions have been

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taken prevent recurrence:

(1)

The worst case design basis conditions for Control Room dose has been reassessed, and specific CREFS performance criteria with respect to this scenario has been established and documented.

(2)

An operability limit of 2 +0.03" wg with respect to atmosphere for the Control Room envelope has been established, together with an administrative limit of 2 +0.04" wg (in contrast to the previous 2

+0.01" wg acceptance criteria.)

Surveillance testing to these.

limits is being conducted monthly. In the event the administrative limit is not met, the testing frequency would be increased to once every two weeks.

The design basis for the control room envelope continues to be " positive pressure."

(3)

A design change has been implemented that will eliminate the problem of Control Room and Cable Spreading Room pressure balance.

(4)

To provide additional margin to its established design basis, CREFS has been modified to increase ventilation flow and pressurization.

Currently, the field work has been completed.

Final design change closure is awaiting acceptance testing and NRC approval of the District's proposed Technical Specification Amendment for CREFS.

f Corrective Steos That Will Be Taken to Avoid Further Violations Upon NRC approval of the Technical Specification Amendment concerning

CREFS, the operability and administrative limits for Control Room pressurization will be increased, Date When Full Comoliance Will Be Achieved j

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The installation of the CREFS modification has greatly increased the Control Room pressurization capability.

However, CREFS flow is now well in excess of the Technical Specification band of 341 CFM 110%. The system will be returned to operability and full compliance achieved upon NRC l

approval the higher band proposed by the District as an amendment to the l

CREFS Technical Specification.

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Violation C.2 l

l Violation C.2 contained in Reference 1 cites the following:

"10 CTR Part 50, Appendix B, Criterion XIe

• Test Controlo* states, in part, that "[a] test program shall be established to assure that all testing required to demonstrate that structures sys teans, and ccanponents a

will perforan satisfactorily in service is identified and performed in t

accordance with written cost procedures which incorporate the requirements and acceptance limits contained in applicable design documents.

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• CNS Quality Assurance Program for Operation Policy Directive, Revision 10, Section 2.11, written to ingpiement the requirements of 10 CFR Part 50, Appendix Bo Criterion XI, requires that each type of test program

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performed will be defined by written procedures and instructions, and it l

requires that acceptance tests will be developed for stzuctures, systems, l

and connponents to demonstrate their capability to perfozu satisfactorily

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following repairs or modification.

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= contrary to the above, the licensee did not assure that all testing was l

identified and performed in accordance with written test procedures which I

incorporated the requirements and acceptance limits in applicable design I

documents.

Specifically, from atxne 1989 until June 1994, Burveillance-Procedure 6.3.17.18,

• Control Room Rnvelope Pressurisation Test, " Revision 4, was not aufficiently detailed in that it did not incorporate acceptance 1Laits to assure that the control Jtoon meergency riiter system would perform satisfactorily in service and because the procedure did not prohibit the inapproprince manipulation of pressures in the adjoining l

buildings as a precondition for conduction the test."

l Admission or Denial to Violation The District admits the violation.

Reasons for Violation The root cause of this violation was the CNS failure to view existing programs and methods with a self-critical and questioning attitude. With respect to the subject of this violation, a surveillance procedure resulted in inadequate guida,ce and acceptance criteria for CREFS operability.

In addition to the cultural issues that provided the general clim e r

this violation to occur, the CNS design and regulatory history of CREFS resulted in an inconsistent understanding of what the exact relationship was between positive Control Room pressurization and CREFS operability.

Pressurization was part of the original system licensing basis (albeit vaguely defined), but had not been specifically included as a surveillance requirement in the CNS Technical Specifications.

Given the compliance-oriented focus that was prevalent at the time, this ambiguity resulted in a surveillance procedure that required only nominal pressurization.

l Moreover, testing conditions ware ill-defined, in that, the required l

pressures of areas outside the L0ntrol Room envelope during the test were not specific.

Corrective Stecs Taken and the Results Achieved I

l The District previously addressed several corrective actions in the CAL response dated July 28, 1994. Also as dirr"ssed in the corrective actions f or Violation C.1, the worst case derign basis conditions for Control Room dose has been reassessed, and specAfic CREFS performance criteria with respect to this scenario % M an established and documented.

Surveillance Procedure 6.3.17.18 has been revised to define acceptance I

criteria reflecting the design basis performance requirements, and to l

specify the testing conditions required for areas bounding the Control Room envelope.

j An amendment to the CNS Technical Specifications has been submitted to the NRC to include demonstration of positive Control Room pressurization to the surveillance requirements of CREFS.

Corrective Steos That Will Be Taken to Avoid Further Violations This violation represents a deficiency in the surveillance Testing Program.

As discussed in Violation B.2, a comprehensive effort has been undertaken to verify that the surveillance requirements of the CNS l

Technical Specifications (as well as other license requirements that impact operability) have been adequately translated into surveillance

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procedures.

Date When Full comoliance Will Be Achieved As discussed in Violation B.2, verification that the key safety system surveillance requirements are adequately described by written procedures i

will be completed prior to startup. Verification of full compliance with I

having written procedures that encompass all surveillance requirements I

will be completed by July 31, 1995.

Violations Not Assessed A Civil Penaltv Section II.A contained in Reference 1 states the following:

I

• 10 CFR Part 50 Appendix Bo Criterion Vr "Xnstructions Procedures and Drawings

• states, in parts that "faictivities affecting quality shall be r

prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be acccalplished in accordance with these instructions, procedures, or drawings.

" Engineering Procedure 3.So " Drawing Control Procedure

• Revision 7, written in s

s part, to iJnplement 10 CFR Part 50, Appendix B, Criterion V, requires that safety-related drawings be included on the safety-related drawing list. "

A.

Violation i

" Contrary to the above, during an NRC inspection conducted June 13o 1994, through August 12, 1994, it was ' detezmined that safety-related Flow Diagram No. 2028, " Reactor Building and Drywell Bqui.nment Drain Bysten,"

i Jtevision M27, was not included on the safety-related drawing list. As a result of this determination, the licensee subsequently identified 13 L

other drawings containing safety-related components that were not included on the safety-related drawing 11at.

• Admission or Denial to Violation The District admits that deficiencies existed in the safety-related list and that 14 drawings were identified that did not have appropriate safety-related components identified.

i Reasons For The violation 1

The " safety-related list' was initially developed in 1985, according to the premise that it would include the drawings of those systems that had recognized safety and plant availability functions. The purpose of this was to ensure that quality-affecting activities would only be performed with reference to final Status 1 drawings (As-built, certified as constructed, certified, or certified final by vendor and signed), as opposed to Archival (Status 2) or Construction (Status 3) drawings. The list was established as an interim measure until more programmatic changes were completed.

Accordingly, Procedure 3.8 was revised to define the three Status Categories, and to proceduralize the requirement that the user ensure that safety-related activities involve only Status 1 drawings.

After this was accomplished, the safety-related list had no quality function with regard to this deficiency.

In 1986, a drawing verification project was initiated to validate the as-built status of selected Control Room drawings. The scope of this effort was initially confined to drawings contained on the previously identified safety-related list. Between 1986 and 1988, the list was revised numerous times as additional safety-related components were identified, which

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likewise affected the drawing verification project scope.

In 1989, a step was added to Procedure 3.8 to provide a formal mechanism l

for making additions or deletions to the list, which would in turn signal l

the Configuration Management Group of additional drawings that should be i

screened for as-built verification.

During the above processes, information was not completely transmitted between lists and drawings.

Corrective Actions Taken and the Results Achieved As discussed above, the Safety-Related List currently serves only as a scoping document for as-built drawing verification efforts.

This -. is a function better served by adequate project scoping instructions than by establishment in the CNS procedures. Accordingly, reference to this set of drawings has been removed from the Procedure 3.8.

The additional 3

drawings that were identified as containing safety-related components are being separately assessed for inclusion in the as-built verification project.

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Cgrrective Stens That Will Be Taken to Avoid Further Violations There are no further corrective steps being planned to address this issue.

Date When Full Comoliance Will Be Achieved l

CNS is in full compliance with the requirement that activities affecting quality be appropriately prescribed by procedures, with respect to the activities described by CNS Procedure 3.8.

i B.

Violation 2 l

"Contrkry to the above, during an RDtc inspection conducted Nay 23, 1994, l

through August 12, 1994, Maintenance Procedure 7.3.2.1,

• DB-25 anot M -50 Circuit Breakers - Betting, Testing, and Maintenance (With Amptectors), "

Jtevision 3, was determined to be inappropriate to the circuanstances in that the procedure did not contain a requirement to rearove tie-wraps frona j

the subject breakers following preventive maintenance, nor did the procedure provide for comprehensive post-maintenance testing of all circuit breaker functions following the conviecion of preventive maintenance."

Arbi ssion or Denial to Violation The District admits the violation.

Reasons for Violation This violation resulted from the discovery on May 16, 1994 that a tie-wrap was installed on the undervoltage trip device of the feeder breaker to MCC-N.

Subsequent investigations revealed that the tie-wrap was installed as allowed during the perf ormance of Maintenance Procedure 7.3.2.1.

This procedure was found to have no explicit requirements for removing the tie-wrap, or for post-maintenance testing that would identify such discrepant conditions.

As stated in the District's July 28, 1994 response to Confirmatory Action Letter 4-94-06b, the root cause of the event was the f ailure of management I

to ensure that requirements for configuration control were adequately implemented into the maintenance procedure.

Management's expectations l

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A contributing factor to the procedural omission was the inappropriate assumption that such restoration steps were within the " skill of the craf t,' and as such, did not require specific articulation. In this case, it is clear that restoration steps should have been provided.

Corrective Stens Taken and the Results Achieved As previously discussed in the CAL response dated July 28, 1994, the following steps have been taken to correct the immediate condition:

(1)

Plant walkdowns have been performed that have verified that no other tie-wraps or other blocking devices were installed on any of the 480 volt breakers on 480 volt busses lA, 1B, lE, 1F, and 1G.

(2)

A review was conducted of station procedures covering electrical and mechanical maintenance to determine if similar ambiguitie. existed with regard to blocking device removal.

This resulted in 18 procedure changes.

A similar review was performed for procedures controlled by the Operations, I&C, Engineering, and Radiological Departments, which likewise resulted in a number of procedure changes.

(3)

A revision has been made to Maintenance Work Practice (MWP) 5.0.4 to add guidance that any impairments, changes or blocking devices installed during the performance of maintenance activities have been removed prior to completion of the procedure.

(4)

Maintenance supervision has communicated to their departments the need for procedural compliance and immediate correction of procedural problems and/or incomplete understanding of procedure requirements, corrective Stens That Will Be Taken to Avoid Further Violations i

The District is committed to broad-based action to achieve excellence in Configuration Management, as discussed in the NPG Performance Improvement i

Plans.

These actions, in addition to the corrective actions described above will prevent further similar violations.

Date When Full comoliance Will Be Achieved j

CNS is now in full compliance with the requirement that activities affecting quality be appropriately prescribed by procedures, with respect i

to installation and removal of temporary blocking devices.

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