ML20082D884

Responds to NRC Re Violations Noted in Insp Repts 50-298/94-14,50-298/94-16 & 50-298/94-19 on 950118. Corrective Actions:Walkdowns of Primary Containment Penetrations Have Been Performed

ML20082D884
Person / Time
Site:	Cooper
Issue date:	03/16/1995
From:	Horn G NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20082D860	List: ML20082D884
References
NLS950069, NUDOCS 9504110023
Download: ML20082D884 (22)

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GENERAL OFFICE

" P o, box 499. COLUM8US. NE8RASKA 686024499

. TELEPHONE (402) 564-8561 xy Nebraska Public Power District m_ _.__.__ m._ _.

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, GUY R. IlORN Vice-President, Nuclear (402) 5615518 NLS950069 e ,

Match 16, 1995 .,

WI WR l )

Director, Office of Enforcement - //

U. S. Nuclear Regulatory Commission ,'N Attention: Document Control Desk gGIONItr f Washington, D.C. 20555 j Gentlemen:

Subject:

Resubmittal of a Reply to a Notice of Violation and Proposed Imposition of 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. R. Gray (USNRC) to Mr. G. R. Horn (NPPD) , dated February 16, 1995, regarding the NPPD Reply to a 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).

In your letter of February 16, 1995 you requested that the Nebraska Public Power District (the District) review three specific examples of what appeared to be unclear or inconsistent statements provided in our Reply to a Notice of Violation (Reply) dated January 18, 1995. You also requested that other portions of the Reply be reviewed and that the Reply be resubmitted. The resubmittal was being required to clarify the information currently available on the docket. The District has performed a review of the Reply and requests that Attachment 1 of the January 18 Reply be superseded in its entirety by Attachment 1 of this correspondence. Each example unclear or apparently inconsistent information is assessed in Attachment 2.

The District sincerely regrets that all of the information provided in the January 18 submittal was not consistent and clear. The District is taking-action to help assure this type of discrepancy does not occur in future NRC correspondence.

Should you have any questions concerning this matter, please contact my offi e.

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. R. Horn

'ce- resident, Nuclear l l

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NRC Resident Inspector w/ attachments  !

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Director, Office of Enforcement U. S. Nuclear Regulatory Commission March 16, 1995 Page 2 of 2 STATE OF NEBRASKA)

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PLATTE COUNTY )

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 statemente contained herein are true to the best of his knowledge and belief.

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Attachment.1 to NLS950069 Page 1 of 16 REFLY TO DECEMBER 12, 1994 NOTICE OF VIOLATION AND PROPOSED IMPOSITION OF CIVIL PENALTIES - EA NOS.94-164, 94-165,94-166 COOPER NUCLEAR STATION NRC DOCKET NO. 50-298, LICENSE DPR-46.

DOring NRC inspections conducted from May 23 through August 12, 1994, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions,

• 10 CFR Part 2, Appendix 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 2.205.

For-each of the violations and " problems," inadequate management 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, these corrective actions are not restated in each violation response.

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

PROBLEM AREA A- Primary Containment Intecrity Violations I. Violation A.1 Violation A.1 contained in Reference 1 cites the following:

= Technical specification 3. 7. A.2.a, " containment Integrity," states, in parte that

• primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water tennperature is above 212*r and when fuel is in the reactor vesse1. . . "

" Technical 8pecification Burveillance Requirement 4. 7.A.2.f.lo " Leak Rate Testing, " stateso in part, that *. . . local leak rate tests (LLRTs) shall be performed on the primary containment testable penetrations and isolation valves at a pressure of 58 psig during each reactor shutdown for refuelingo or other convenient intervalso but in no case at intervals no greater than two years.... The total acceptable leakage rate for all valves and penetrations other than the N8IVs (main steam isolation valves) is 0.60 La."

• Technical Specification 1.Yo

• Surveillance frequencyo* states o in parto that " performance of a Surveillance Requirement within the specifled time interval aball constitute connpilance with operability requirements for an Lc0 (limiting condition for operation) unless otherwise required by the specification."

= contrary to the above, from January 18, 1974, until May 27, 1994, primary containment integrity was not maintained at all t.ines when the reactor was critical or when the reactor water tenversture was above 212*r and fuel was in the reactor vessel in that the surveillance Requirement for the local leak rate testing of 82 convonents had never been langlemented at an interval not to exceed two years : As the result of testing conducted on June 23, 1994, Isolation Valve IA-65CV (one of the 82 coanyonents) failed the LLRT, resulting in a total leakage value that significantly. exceeded the 0.60 La limit. The 0.60 La limit corresponds to a leakage rate of 5.37 scmh (189. 60 scth) . The LLRT failure of Valve IA-65CV resulted in a total leakage rate that exceeded 17.66 scmh (623.57 scfh).*

E. ]

[j Attachment 1 >

to NLS950069.

Page 2 of 16 Admission or Denial to Violation The District admits the violation.

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 '; hat 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.

When the CNS Operating License was issued (1974), the specific testable Primary Containment penetrations and Primary Containment Isolation Valves (PCIVs) were listed in the CNS Technical Specifications. Eventually, this  ;

list was shifted to the Updated Safety Analysis Report (USAR) through a i License Amendment. The original licensing basis was that performance of  !

these specific tests constituted a method for Appendix J compliance that 1 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 Appendix J for public comment, adherence to the Technical l Specification /USAR list (along with other specific commitments that might i be made) was all that was required, despite later regulatory positions that contradicted this approach. Although these deficiencies were self-identified as a result of broad programmatic action that was accelerated in response to Inspection Report 93-17, the District duly acknowledges I that this condition should have been recognized and corrected long ago.

1 Corrective Steos Taken and the Results 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 containment penetrations have been performed. This activity contributed 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 was determined to be immediately practicable. Those that were not tested were either modified and then tested, or were designated as candidates for Appendix J exemption. The total as-found leak rate for the testable penetrations, except X-22 which contained drywell pneumatic supply check valve IA-CV-65CV, was 26 SCFH. With regard to penetration X-22, leakage through IA-CV-65CV was significant, preventing pressurization of the penetration using normal leak rate i 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-left leak rate is less than the Technical Specification limit.

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Attachment 1 to NLS950069 Page 3 of 16 (2) Design changes were implemented, which included addition of test connections, installation of welded caps on spare penetrations, complete redesign of some containment . isolation barriers, and 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 e test in accordance with the requirements of Appendix J, NRC exemptions have been obtained.

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

" program owner" responsibilities. These responsibilities require the primary duty to be ensuring the integrity of the overall program, rather than merely functioning as a testing facilitator. To assist in this objective, the 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 CNS is now in full compliance with the testing require.'ents necessary to demonstrate Primary Containment Integrity.

II. Violation A.2 ,

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

"10 CFR Part 50, Appendix ar Criterion x1, = rest Contro1," states, in part, that *[a] test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perfom satisfactorlly in service is identified and perfomed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents."  ?

• CNB Quality Assurance Program for Operation Policy Directive, Revision 10, Bection 2.11, written to invienent the requirements of 10 CFR Part 50, Appendix B, Criterion XI, requires that each type of teet program performed will be defined by written procedures and instructions, and it requires that acceptance tests will be developed for structures, systems, and components to demonstrate their capability to perform satisfactorily following repairs or modi 21 cation. l

" Contrary to the above, the licensee did not assure that all testing was identitled and performed in accordance with written test procedures which incorporated the requirements and acceptable limits. Brecifically as of Nay 14, 1994, 68 conponents passing through 54 primary containment penetrations, each required to be local leak rate tested in accordance j with the requirements of Technical specification surveillance Requirement l

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

• Leak Rate Testing, " had not been identified in a procedure j as requiring local leak rate testing, as required by CNS Quality Assurance i Program for Operation Policy Directive, Revision 10, Section 2.11. These components had never been local leak rate tested. \

" Contrary to the above, the licensee did not assure that all testing was identified and performed in accordance with written test procedures which incorporated the requirements and acceptable limits. As of . Tune 21, 1994, I

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Attachment 1 to NLS950069 Page 4 of 16 instrument pressure switches PC-PS-12A, Bo Co and DJ PC-PS-101A, B, C, and DJ PC-PS-119Ao B, C, and DJ PC-PS-161 and PC-PT-512A and Bo each required to be local leak rate tested in accordance with the requirements of Technical Specification Surveillance 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 Directiver Revision 10, Section 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.1. A lack of rigor in the Appendix J program (caused by a compliance based management philosophy)' rnd ted in not identifying all of the components for which Type C LLRTs werc 6racticable, and in isolating components which should have been lef t unisolated for inclusion in the Type A Integrated Leak Rate Testing 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 Stens That Will Be Taken to Avoid Further Violations As discussed in the District's response to Violation A.1, a testing basis document is being prepared. This 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.

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

"10 CrR Part 50, Appendix B, Criterion III,

• Design Controlo" states in part, that *(m]easures shall be established to assure that...the design basis...[is) correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled. "

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Page 5 of 16 ,

" Draft General Design Criterion 53, a measure written to comply with'the requirements of 10 CFR part 50, Appendix B, Criterion 111, as comunitted to in Appendix F of the Upds.:ed Safety Analysis Report (USAR) states that

=fplenetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus. "

• Draft General Design Criterion lo as comunitted to in Appendix F of the UBAR, states that *[t] hose systems and components of reactor facilities which are essential to ' the prevention of accidents which could offact the pubilc health and safety or mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the invottance of the safety function to be performed. "  :

" General Electric Design Specification No. 22A1153 .

• Codes and Industrial l Btandards

• Revision 1, stateso in Note '3 of the Appendix, that *[p]spings which is an integral part of the primary containment for isolation  ;

purposes, shall have at least the same quality and levels of assurance as the primary containment." u

" Contrary to Criterion III, the licensee did not assure that the above design bases were correctly translated into specifications and instructions and did not assure that deviations from quality standards ,

were controlled. Bpecificallys

a. As of Nay 14,'1994, numerous primary. containment penetrations had no redundant valving. These penetrations included, but were not limited to, Penetrations X-21, X-22, X-25, X-29E, - X-30E/F, X-33E/F, X-209A/B/C/Do and X-218.

b. As of February 22, 1994, 10 penetrations consisting of manus 11y '

operated vents, drains, and test connections and requiring closure for the containment functian o had a single manual isolation valve ' .;

for containment isolation as opposed to the required ' redundant valving and associated apparatus. }

l C. During an NRC inspection conducted atste 13, 1994, through August 12, 1994, it was determined that approximately 300 containment penetrations had not been designed, fab ~aated, or installed to the l same standards as the primary contro mnL because these components had not been correctly classified an 6 tveui:La1. As a resulto these penetrations had not been designea, fabricatedo 'or erected to quality standards that reflect the importance of the safety function to be perfommed (e.g., some welds were not nondestructively tested, l some penetrations were not local leak rate tested, and the penetrations were not treated as safety-related by the licensee *s quality assurance program)."

Admission or Denial to violation The District admits the violation.

Reasons for Violation As noted in this Violation, there were two areas of non-conforming Primary Containment penetration design (1) the lack of Primary Containment penetration barrier redundancy for all process lines passing through the Primary Containment, and (2) the improper classification and maintenance  !

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

detail below. 4

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Attachment 1 to NLS950069 Page 6 of 16 (1) . Barrier Redundancy- The cited Primary Containment penetration barrier redundancy discrepancies involve manual valve configurations on penetrations designated for process lines. As noted by the NRC ,

in Enclosure 2 of the NOV, these configurations were established during original plant construction. The configurations were built to be in conformance with the 1967 Draf t AEC General Design Criteria (GDC). These draft criteria did not explicitly require redundancy for process lines isolated by manual valves. The response to Final Safety Analysis Report (FSAR) Question 5.5 identified the District's position with respect to GDC 55 and 56, and Safety Guide 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 and the GDC still provides for the acceptability of containment 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 criteria, within which, Criterion 55 provides more prescriptive redundancy requirements. It is unclear to what degree the AEC 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 barrier configurations, both in the original design, and af ter various design changes.

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

(2) Eenetration classification- The penetration misclassifications resulted from an original design error, in that, the requirements of General Electric Specification 22A1153 for piping passing through the Primary Containment were not correctly - translated into the piping specifications. The passive function of helping to maintain Primary Containment integrity was not recognized as a safety function for otherwise non-essential piping. As a result, piping segments were inappropriately procured, fabricated and maintained to the requirements of USAS B31.1, rather than to a level of quality commensurate to that applied for the Primary Containment (as in USAS B31.7). The impacts of this discrepancy were that in certain 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 that were accelerated 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 ,

identify and correct these deficiencies was due in part to a compliance- I based focus, such that, undue reliance was placed in the continuing )

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.Page 7 of 16-adequacy ' of the original' plant design, based on AEC approval during -

c initial licensing. l t

Corrective'Stens Taken and the Results Achieved ,

i

, The District'. addressed actions regarding barrier redundancy .and penetration classifications in the CAL response dated July 28, 1994. The following is a summary of corrective actions noted in that letter, as well.

as additional responsive activities.

(1) Barrier Redundancy- 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, actions have been taken to control vent, drain, and test line barrier configurations (see also the District Reply to a Notice of Violation, Inspection Report 50-298/94-03, dated May 31, 1994).

(2) Penetration Classification- A document review was performed for 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 Conteinment, and a reconciliation has been performed between i USAS B31.1 and B31.7 for these segments. .These penetrations and components are now treated as Essential IIN, and will be maintained under the District's ASME Section XI Program.

Corrective Stens That Will Be Taken to Avoid Further Violations No further directly related corrective steps are planned. However, as stated during the Enforcement Conference, the two ongoing actions-described below help to prevent recurrence of similar types of violations.

(1) To help identify other licensing basis issues stemming from the original design, the design basis reconstitution ef fort is being expedited.

(2) The ASME Section XI boundaries are being reviewed to identify and 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 Comollance Will Be Achieved 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 acceptable to the NRC.

r

g Attachment 1 '

'to NLS950069 Page 8 of 16' {

PROBLEM AREA B- Onerability of the 480 Volt and 4160 volt Buses I. Violation B.1 Violation B.1 contained in Reference 1 cites the following:

" Technical specification 3.9.A.1.c a " Auxiliary slectrical Equipment," "

requires, in parts that the reactor shall not be made critical from a cold shutdown Condition unless the 4160 vole critical buses 1r and 10 and the '

480 voit critical buses 17 and la are energised, and the undervoltage and ,

loss of voltage relays, as well as their auxiliary relays, are operable.

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" Technical Specification Burveillance Requirement- 4.9.A.1.a r "Rmergency Buses Undervoltage relaysa

• states that "Once every 18 monthe r loss of voltage on emergency buses is simulated to demonstrate the load shedding from emergency buses and the automatic start of diesel generators." USAR Bection 2.2.7.2.1.a r *Btandby A-C Power (Diesel Generators) Test

• Capabilityo* defines the function of the protective scheme as providing for the clearing the buses of all motor loads excepting supply to the 480  ;

voit critical unit substation.

" Technical specification 1.Y, = surveillance frequency," states, in part, e that "portosmance of a surveillance Requirement within the specified time '

interval shall constitute conv11ance with operability requirements for an

.Lc0 (11miting condition for operation) unless otherwise required by the specification."

" Contrary to the abore s from January 18, 1974, until Nay 25,.1994, the reactor had been made critical without 4160 voit critical buses 17 and 1G,

• and 480 voit critical buses 1r and 10 being operable in that the undervoltage relays associated with several of the electrical loads supplied by thase buses had never been tested to demonstrate . their operability or upon testing, failed to perfazu their intended function of shedding their respective electrical loads from these buses.*

Admission or Denial to Violation '!

l 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 (zuestioning 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 j path from the sensing of undervoltage on the 4160 volt and 480 volt busses i through the opening of the ,aspective circuit breakers on undervoltage, i The requirement to test this function stemmed from a 1978 License l Amendment that specifically incorporated this Surveillance Requirement into the CNS Technical Specifications. The reconciliation performed 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 system functional testing (LSFT) associated with actuation of the undervoltage relays was not comprehensive.

Numerous opportunities occurred for earlier identification. Most notable of these opportunities were the design basis reconstitution effort, NRC information to the industry on Westinghouse DB-50 circuit breakers and deficient safety-related LSFTs, and an Electrical Distribution System

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. Attachment l' i to NLS950069 f

Page 9-of 16 Functiorial Inspection - (EDSFI) performed 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 Steos Taken and the Results Achieved Several steps were taken to address this-violation:

.i (1) LSFTs for 4160 volt buses 1F and 1G, and 480 volt buses 1F and 1G were satisfactorily performed. _

t (2) The applicable surveillance procedures were revised to verify that the load shedding function occurs as required.

(3) An electrical calculation was performed which demonstrated that even -

if load shedding of all of the non-essential 480 volt loads failed i

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 breekers, a Design Change was implemented which resulted in replacement of the undervoltage trip devices with shunt trip devices within those safety-related circuit i

breakers that are credited with shedding a non-essential load. Industry experience has shown this to be an effective configuration.

Corrective Stens That Will Be Taken to Avoid Further Violations This violation hac prompted a broader inquiry into the adequacy of the CNS Surveillance Testing Program. The intent of this ef fort is to verify that all of the surveillance testing requirements have been - correctly translated into the surveillance procedures. This effort is more fully discussed in the corrective action'for Violation B.2.

Date When Full Comnliance Will Be Achieved ')

CNS is now in full compliance with the requirements of Technical l Specifications 3.9.A.1.c and 4.9.A.1.a.

II Violation B.2 Violation B.2 contained in Reference 1 cites the following:

"10 crR rart 50, Appendix a, criterion 11, " Test controz," states, in parta that *[a] test program shall be established to assure that all testing required to demonstrate that structures, systesso and components will periosa satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptable limits contained in applicable design documents." \

• contrary to the above, the licensee did not assure that all testing was identified and performed in accordance with written test procedures which incorporated the requirements and acceptable limits. Specifica11ys I

a. During an NRC inspection conducted Nay 23o 1994, through August 12e 1994, Procedure 6.3.4.3, "Bequential Loading of Emergency Diesel Generators, " Revision 31, which is performed to satisfy Technical Specification Burveillance Requirement 4.9.A.1.a r " Loss of Voltage }

Relays," was determined to be inadequate because it did not assure l that the emergency diesel generators and critical buses would l

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perform ' satisfactorily in service in'that the procedure did not 'l contain requirements to verify that the 480-volt supply breakers for safety-related and nonsafety-related loads would shed from.their.

electrical buses within a specifled time r nor did the procedure -

identify that ; the control rod drive pump motors and ' 's ta tion air i compressors were required to be shed from the electrical bus.' i

b. During an inspection conducted Nay 23, 1994r through August 12, .

.1994,' the mtC identified that Procedure 6.3.20.1r *RER 8ervice Water Booster Pualp Flow Test and Valve Operability Test, " Revision 27a. did n not provide for the testing of the load shedding feature'of the; .

supply breakers associated with the 4160 volt residual' heat removal I service water booster pualps. "

Admission or Denial to Violation The District admits the violation.

Reasons for Violation ,

The root cause for not establishing written test. procedures that adequately reflect the Technical Specification requirements is r 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:

(1) Not ensuring the incorporation of all load' shedding verifications-

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

(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' experience with respect to Westinghouse DB-50 circuit breakers and -

inadequate LSFTs, and a previous EDSFI.

Corrective Steos Taken and the Results Achieved The surveillance procedures cited in this violation have been revised to reflect appropriate load shed testing. Additionally, as discussed in the' .,

District's August 8, 1994, response to an NRC Request for Additional i Information, the investigation into the def'iciencies of Surveillance Procedure 6.3.20.1 prompted a review of the LSFTs performed for several key safety systems. This review identified significant testing omissions. 1 LER 94-009, Revision 1, discusses the corrective actions = taken in response to the LSFT discrepancies.

Corrective Steos That Will Be Taken to Avoid Further Violations 1

CNS is currently verifying that all surveillance requirements contained in l the CNS Technical Specifications have been adequately translated into l surveillance procedures. 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 1 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 j the Technical Specification surveillance requirements. j l

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. Attachment 1-to NLS950069 Page 11.of 16 Date When Full Comnliance Will Be Achieved Verification. that the key safety system surveillance requirements ' are adequately- described by written' procedures has. been completed. As discussed in the CNS Phase 1 Plan, these key systems were 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 Emeraency Filter System .

I. Violation C.1 Violation C.1 contained in Reference 1 cites the following:

• Technical specification 3.12.R.1, = Control Room Emergency Filter system a "

states, in parta that ". . . the Control Room Reorgency Filter system. . .shall be operable at all times when containment integrity is required. "

• The Order Confirming Licensee Comunitments on Post-TNZ Related Zssueco dated July lo s'1981, confinam NPPD*m comanitment to con lplete NUREG-0737 .

" Clarification of TNZ Action Plan Requirementa r

• Xtem ZZZ.D.3.4, " Control. ,

Room Esbitability." Item ZZZ.D.3.4 involves the review of facility design requirements against the Standard Review Plan. The NPPD response to Generic Letter 80-90, dated December 30, 1980, submitted the control room habitability evaluation s which statedo in part, "the CNB control room _ .,

ventilation system is designed to maintain the control room at about 'l 1/4 in. H,0 (0.031 kPa) positive pressure by supplying air at a high enough l pressure that even when system ~ losses and the booster exhaust fan pressures are accounted for e the control room pressure is still positive..." i "A Batety Evaluation Report for the Cooper 8tation from the Accident Evaluation Branch on NOREG-0737: Zten No. ZZZ.D 3.4,

• Control Room Habitabilityo* dated February 24 1982, states, in parto that " ...the design meets the criteria identified in Item ZZZ.D.3.4 of NUREG-0737 and ia acc.,eaba.."

" Contrary to the above, from June 1989 until April 28, 1994 the Control Room Raergency Filter system was not operable at all times when l containment integrity was required in that testing failed to demonstrate  !

that a positive pressure could be maintained in the control room during the periodic performance of the control room envelope pressurization test."

Admission or Denial to Violation The District admits the violation.

. Reasons for Violation The circumstances surrounding the prolonged inoperability of the control Room Emergency Filter System (CREFS) were provided to the NRC in LER 94-006, Revision 1. 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 incemplete understanding of the original system

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Attachment 1-to NLS950069 W Page 12'of.16 , .

design criteria. This led to unsubstantiated reliance on the adequacy'of ,

the perceived licensing basis. Several opportunities occurred to address 'i identified deficiencies in both the design and 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 action even

• though' the pressurization test results were marginal. As a result of.

these collective deficiencics, the system should not have been considered  !

operable.  !

Corrective Steos Taken and the Results Achieved' Corrective actions, have been :taken that have restored CREFS to' I operability. Door seals in the Control Room envelope were repaired, '

penetrations were sealed, and the adjacent building ventilation control-systems inspected and repaired. Testing was performed that confirmed positive pressurization between the range of +0.04" to +0.05" wg with respect to atmospheric pressure. 't The following additional actions have been taken to prevent ~ recurrence:

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

The CNS Technical Specifications have been amended to reflect this increased system capability.

(3) A design change has been completed that has eliminated the problem of Control Room and Cable Spreading Room pressure balance. ,

(4) An operability limit of a +0.04" wg with respect to atmosphere for the Control Room envelope has been estehlished, together with an e administrative limit of. a +0.05" wg (in contrast to the . 2 +0.01" wg acceptance criteria that existed before corrective actions had been implemented.) 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. ,

corrective Steos That Will Be Taken to Avoid Further Violations CNS has completed the corrective actions that are necessary to prevent recurrence of this violation.

Date When Full Comoliance Will Be Achieved CNS is now in full compliance with the requirements for CREFS operability with respect to the ability to maintain positive pressure in the Control Room.

II. Violation C.2 L

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

'" 10 CFR Part 50, Appendix B, Criterion XI, " Test Controir* states, in part, that *(a) test' program shall be established to assure that all testing required to demonstrate that structures, systems, and contponents will , perform satisfactorily in service is identified and performed in

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Attachment 1 to NLS950069 Page 13 of 16 accordance with written cent procedures which incorporate the requirements and acceptance lintits contained in applicable design documents. "

"CNS . Quality Assurance Program for Operation Policy Directive r Revision 10o Bection 2.11, written to ingplement the requirements of 10 CPR Part 50, Appendix Bo Criterion XI, requires that each type of test program performed will be defined by written procedures and instructionso and it requires that acceptance tests will be developed for structures o systems, and congponents to demonstrate their capability to perform satisfactorily following repairs or modification.

" Contrary to the above r the licensee did not assure that all testing was identified and performed in accordance with written test procedures which incorporated the requirements and acceptance limits in applicable design documents. Specifica11yo froas June 1989 until June 1994 Burveillance Procedure 6.3.17.18 " Control Room Envelope Pressurization Testo*

Revision to was not aufficiently detailed in that it did not incorporate acceptance limits to assure that the Control Room Emergency rister system would perform satisfactorily in service and because the procedure did not prohibit the inappropriate manipulation of pressures in the adjoining buildings as a precondition for conduction the test."

Admission or Denial to Violation The District admits the violation.

Reasons for Violation The root cause of this violation was the 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 guidance and acceptance criteria for CREFS operability.

In addition to the cultural issues that provided the general climate for 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.

Moreover, testing conditions were ill-defined, in that, the required pressures of areas outside the Control Room envelope during the test were not specific.

Corrective Steos Taken and the Results Achieved The District previously addressed several corrective actions in the CAL response dated July 28, 1994. Also as discussed in the corrective actions for Violation C.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.

Surveillance Procedure 6.3.17.18 has been revised to define acceptance criteria reflecting the design basis performance requirements and to specify the testing conditions required for areas bounding the Control Room envelope.  ;

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• Attachment 1 i to NLS950069  !

Page 14 of 16 -j j

An amendment to the CNS Technical Spe::ifications was approved by the NRC l which included a demonstration of positive control Room pressurization for j the surveillance requirements'of CREFS.

Corrective Steos That Will Be Taken to Avoid Further Violations i i

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 ,

Technical Specifications (as well as other license requirements that  ;

impact operability) have been adequately . translated into surveillance. '

procedures. l Date When Full Comollance Will Be Achieved As discussed in Violation B.2, verification that the' key safety system $

surveillance requirements are adequately described by written procedures has been completed. Verification of full compliance with having written procedures that encompass all surveillance requirements will be completed by July 31, 1995. ,

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

• 10 CFR Part 50 Appendix Bo Criterion Vr " Instructions, , Procedures and Drawings r

• states, in part, that "[a]CtlYities affecting quality shall be prescribed by documented instructions, procedures s or drawings o of a type appropriate to the circumstances and shall be accoagplished -La accordance with these instructions, procedures, or drawings.

" Engineering Procedure 3.8s " Drawing Control Procedure r

• Revision 7, written o in parto to ingpienent 10 CYR Part 50, Appendix Bo Criterion V, requires that safety-related drawings be included on the safety-related drawing list. "

A. Violation 1

• Contrary to the above, during an NRC inspection conducted June 13, 1994, through August 12, 1994, it was determined that safety-related Flow Diagram No. 2028, " Reactor Building and Drywell Equipment Drain Rystem o "

Revision N27, was not included on the safety-related drawing list. As a result of this determination, the licensee subsequently identified 13 other drawings containing safety-related con lponents that were not included on the safety-related drawing 1Lat. "

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

Reasons For The Violation 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

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- Page 15-of 16 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 likewise affected the drawing verification project scope. >

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

for making additions or deletions to the list, which would in turn signal the Configuration Management Group of additional drawings that should be screened for as-built verification.

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

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

I Corrective Steos 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 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.

B. Violation 2  ;

" Contrary to the above r during an NRC inspection conducted May 23, 1994, through August 12, 1994 Naintenance Procedure 7.3.2.1, *DB-25 and DB-30 circuit Breakers - Bettingo Testing, and Maintenance (with Annotectors), " i Revision 3, was determined to be inappropriate to the circumstances in that the procedure did not contain a requirement to remove tie-wraps from the subject breakers following preventive maintenance, nor did the procedure provide for connprehensive post-maintenance testing of all circuit breaker functions . following the ~ contpletion of preventive

• maintenance."

Admission or Denial to Violation The District admits the violation.

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Attachment 1 to NLS950069 '

Page 16 of 16-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- l to - MCC-N. Subsequent- investigations revealed that . the tie-wrap was installed as allowed during the performance of Maintenance ' Procedure 1  ;

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.  ;

-e As stated in the District's July 28, 1994, response to Confirmatory Act M Letter. 4-94-06b, the root cause of the event was the failure of management i to ensure that ' requirements for configuration control were adequately t implemented into the maintenance procedure. Management's expectations .!

were not clearly communicated and effected through the procedure review-and approval process. As a' result, a requirement to remove the tie-wrap was not included at the conclusion of the procedural section.

l A contributing factor to the procedural omission was the inappropriate assumption that such restoration steps - were within the " skill of E 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 Steos Taken and the Results Achieved i

The following steps have been taken to correct the immediate condition:

l (1) Plant walkdowns have been performed that have verified that no' other l tie-wraps or other blocking devices'were installed on any of the "

480 volt breakers on 480 volt busses 1A, 1B, 1E, 1F, and 1G.

(2) A review was conducted of station procedures covering electrical and -

mechanical maintenance to determine if similar ambiguities existed with regard to blocking device removal. This resulted in 18,  ;

procedure changes. A similar review was performed for surveillance 4 procedures under the cognizance of CNS Operations, I&C, Engineering, ,

and Radiological, 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 proeddure.

(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 i requirements. ,

Corrective Steos That Will Be Taken to Avoid Further Violations The District is committed to broad-based action to achieve excellence in Configuration Management, as discussed in the NPG Performance Improvement Plans. These actions, in addition to the corrective actions described above will prevent further similar violations.

Date When Full Comnliance Will Be Achieved I 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, i

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'e Attachment 2 to NLS950069 Page.1 of 3' REASONS FOR UNCLEAR OR APPARENTLY INCONSISTENT INFORMATION PROVIDED IN THE JANUARY 18, 1995 RESPONSE TO NOTICE OF VIOLATION AND ,

PROPOSED IMPOSITION OF CIVIL PENALTIES Overview The NRC Notice of Violation and Proposed Imposition of Civil Penalties dated December 12, 1994, contained cited violations that required nine separate responsive sections in the District's reply (Violation I. A.1, I.A.2, I.A.3, I.B.1, I.B.2, I.C.1, I.C.2, II.A.1, and II.A.2). The violations pertained'to issues that had . already been substantively explored in previous . District correspondence to the NRC: the July 28 and August 8, 1994, responses to the NRC Confirmatory Actions Letters; Licensee Event Reports94-006 Rev. 1,94-009 Rev. O, and 94-011 Rev. 1; the July 28, 1994, letter regarding Updated Control Room Emergency Filter System commitments; and the November 7, 1994, letter regarding Progress on Improvements at CNS.

During the initial planning phase of the Reply, the District decided that the ,

document should be developed with reference to previous District correspondence on the docket. This would serve both as a guide to the NRC in obtaining additional related information as desired and as a tool for integrating the corrective actions with previous existing commitments on the docket. This ,

philosophy was purposely a departure from the District's standard practice of l providing essentially " stand-alone" submittals to the NRC. '

District submittals are authored by individuals with the technical experience to '

fully understand the issues- and information being described. Additionally, reviews by the cognizant disciplines and management help ensure that NRC submittals are complete and accurate in all material respects. The District, during the preparation and review of the Reply did not place sufficient attention to the references or content to ensure it was clear and consistent. The l following is a restatement of the examples and the District's response.

Soecific Eynmnles

1. "On page 6 of 16, under corrective staan Taken and the Results Achieved, it states, "The District addressed actions regarding barrier redundancy and penetration classifications in a letter dated August 8, 1994." From our review, your letter of August 8, 1994, does not appear to discuss barrier redundancy and penetration classifications. "

The NRC is correct in pointing out that the District's August 8, 1994, '

response to the CAL Request for Additional Information does not discuss barrier redundancy and penetration classifications. The correct reference should have been the July 28, 1994, CAL response, rather than the August 8 letter which is properly considered as supplemental to the initial CAL {

response. The reason for this inconsistency was human error on the part of the author that was not discovered during the review cycle prior to i submittal. Accordingly, the District is assessing the submittal review  ;

and approval process to help prevent recurrence of this type of i discrepancy.

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• On page 7 of 16, under itesa (1) Barrier Redundancv, it ^ s ta tes,'

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2. "Also, programmatic enhancennents have been made to control vent, drain, and test line barrier configurations as discussed in the District 's response ' to ,

Inspection ; Report 94-03." The referenced ^ response : to ' the inspection )

-report did not appear to discuss progranunatic enhancements to barrier configura tions. "

The. intent of ' this reference was to point out that previous District

. correspondence. existed- on the docket which discussed programmatic-enhancements ' for the issue of barrier redundancy discrepancies. The bottom of page 2 of the May 31, 1994,. Reply to a Notice of Violation, Inspection Report 50-298/94-03, discusses the programmatic enhancements being performed:

In addition to the above violationi the NRC inspection identified an Unresolved Item (298/9403-01) concerning - the use of a ' single, unlocked, manual valve for a containment isolation function. The District wishes to advise it is presently reconstituting the design basis for the Primary Containment system and Will evaluate. this issue within that task. Likewise, the CNS Plant Engineering Department is pursuing ef forts to resolve NRC concerns involving the identification and control of manual primary containment isolation valves. The District plans to complete these efforts by August 1994.

In Inspection Report 94-14, the NRC closed the Unresolved Item having determined that the identified-barrier redundancy discrepancies were an apparent violation of 10 CFR Part. 50, Appendix B, Criterion III. In Section 2.4 of that inspection report, the NRC stated its understanding of the information conveyed in the previous paragraph:

The licensee submitted its response to NRC Inspection Report 50-298/94-03 by letter dated May 31, 1994. The response stated that the licensee ' was reconstituting the design- basis for the primary containment system and would evaluate the issue within that task.

In addition,'the licensee advised that it was pursuing efforts to resolve NRC concerns involving the . identification and control of manual primary containmentLisolation valves, or more appropriately the administrative control of'the valve and cap / plug combination.

The licensee ' stated that it planned to complete this effort by August 1994.

The programmatic enhancements committed to in addressing barrier redundancy (design basis reconstitution and Plant Engineering pursuit of additional efforts to resolve the concern) encompass . the issues of-controlling vent, drain, and test line barrier configurations. What appears to be the point at issue is that the wording may suggest that the District's Inspection Report 94-03 response discusses more detailed and specific programmatic enhancements, and that excessive credit is being taken for this in the Reply. This was not the District's intention.

Since the NRC has indicated that they perceived' the reference to be unclear, the wording has been. revised in Attachment 1. This issue will be taken into consideration in future correspondence.

3. "On page 10 of 16, under corrective stens Taken and the Results Achieved, it states, "This review has identified significant testing omissions that are being addressed as documented in Licensee Event Report 94-009." The referenced licensee event report did not appear to discuss significant testing omissions.*

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~Page 3.of 3 In this' reference,-the District was attempting to convey that.LER 94-009' '!

Rev. O contained the corrective actions that were responsive to the. issue  !

of significant Logic System' . Functional _ Testing discrepancies.. . . .The .

preceding sentence to the above example sets the context.with which.the' reference is made:

. . . .the investigation into the deficiencies of Surveillance Procedure -

6.3.20.1 prompted a review of the Logic System Functional Testing,

, performed for several key safety systems.

In LER 94-009 Rev. 0, page 2 of 4 in the second paragraph,'the discrepancy  !

with Surveillance Procedure 6.3.20.1 is' discussed:

The continuing investigation into this condition [ inadequate load shed testing) also revealed that a contact in the load shedding j circuit of the Service Water Booster Pumps, 4160 VAC loads, was bypassed during the load shed testing; On'page 3 of 4 first paragraph, the LER further discussed the significant testing omissions.that were discovered:

It was determined that there were contacts in' additional systems.

which were not tested in accordance with Technical Specification requirements for Logic System Functional Testing. Based upon an-analysis of systems required to be operable for existing plant  ;

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conditions, Service Water, Reactor Equipment Cooling, 4160 VAC. Bus +

1F, and 4160 VAC Bus 1G were declared. inoperable on June 11 at 2:30 pm.

Lastly, on page 4 second paragraph, the LER discusses the following corrective actions:

Procedures were written or revised to test the load shed logic and the components not previously tested as required by.the Technical

' Specifications for Logic System Functional Testing. These tests are n presently (as of June 20) being performed.

Further investigations are being conducted to determine the cause of-the failures and actions to prevent recurrence, and will be reported in a supplement to this LER.

In summary, the language in LER 94-009 Rev. O appears to be clear in addressing the subject of Logic System Punctional Testing (LSPT) discrepancies. The point of confusion appears to be that the LSFT discrepancies in the LER were not specifically characterized as "significant testing omissions" as could be inferred from the wording in the Reply. The wording has been revised accordingly in' Attachment 1.

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o l LIST OF NRC COMMITMENTS l ATTACHMENT 3 l Correspondence No: NLS950069 The following table identifies those actions committed to by the District in j this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE NO ADDITIONAL COMMITMENTS TO THOSE PREVIOUSLY N/A COMMUNICATED IN THE 1/18/95 REPLY.

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l l PROCEDURE NUMBER 0.42 l REVISION NUMBER 0 l PAGE 10 OF 16 l l