IR 05000213/1986001

Insp Rept 50-213/86-01 on 860109-0210.Violations Noted:Two Design Change Notices for Plant Design Change Request PDCR-787 & Eight Notices for PDCR-789 Not Numbered Using Project Assignment Followed by Sequential Number

ML20140J514
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/26/1986
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20140J482	List: IR 05000213/1986001 ML20140J478 ML20140J495
References
50-213-86-01, 50-213-86-1, NUDOCS 8604040348
Download: ML20140J514 (12)

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U.S. NUCLEAR REGULATORY COMMISSION REGION I DCS Nos. 50/213-86-01-04 50/213-86-01-05 Report N /86-01 50/213-86-01-07 50/213-86-01-15 Docket N /213-86-01-27 License N DPR-61 Licensee: Connecticut Yankee Atomic Power Company P.- 0. Box 270 Hartford, CT 06101

. Facility: Haddam Neck Plant, Haddam, Connecticut

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Inspection at: Haddam Neck Plant Inspection conducted: January 9 - February 10, 1986 Inspectors: Stephen Pindale, Resident Inspector Paul D. Swetland, Senior Resident Inspector Approved by: ._- b --- a b E. C. McCabe, Chief, Reactot+ Projects Section 3B-66/s Date Summary:

Areas Inspected: This was a routine safety inspection of refueling outage activi-ties involving 206 inspection hours by two resident inspectors. The following ac-tivities were inspected: refueling preparations, plant operation, radiation pro-tection, physical security, fire protection, maintenance and surveillance testing, pl. ant modifications and followup on licensee event Results: Licensee preparation, control, and implementation of various refueling outage activities were found to be generally satisfactory. Several administrative errors in the implementation, testing and turnover of plant modifications were cited as violations (Detail 2.1). NRC open items were established to follow lic-ensee corrective actions related to recurrent failures of containment local leak rate tests (Detail 3.1) and licensee disposition of steam generator eddy current test results (Detail 5.2). Two previous NRC followup items remained open pending further licensee action and NRC review (Detail 4).

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TABLE OF CONTENTS P_ay 1.- Summary of _ Facility Activities 1 Review of Outage Activities l' Observation of Maintenance and Surveillance Testing 5 Followup on Previous Inspection Findings 7 Followup on Events Occurring During the Inspection 8

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DETAILS Summary of Facility Activities The ple.nt remained in a cold shutdown condition throughout this inspection period. The inspection covered days 5 - 38 of the planned 8-week refueling /

maintenance outag During this period, outage activities included onsite review and preparation for.several plant modifications, chemical decontamina-tion of the steam generator (SG) primary channelheads, SG eddy-current test-ing,. installation of a new refueling cavity seal and routine refueling inter-

val maintenance and surveillance. testing. . By the end of the inspection period,

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installation / fit-up problems for both the SG decontamination and reactor L'

cavity seal jobs had resulted in a 12 day delay in the projected outage schedul . Review of Outage Activities The inspector observed' outage activities during regular tours of the following

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Control Room --

Security Building

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Primary' Auxi.liary Building --

Fence Line (Protected Area)

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Vital Switchgear Room --

Yard Areas

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Diesel Generator Rooms --

Turbine Building

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Control Point --

Intake Structure and Pump

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Containment Building Building Control room instruments were observed for correlation between channels and for conformance with Technical Specification requirements. The. inspector ob-served various alarm conditions which had been received and acknowledge perator awareness and response to these off normal conditions were reviewe Control room and shift manning were compared to regulatory requirements.

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l Posting and control of radiation and high radiation areas was inspecte Compliance with Radiation Work Permits and use of appropriate personnel moni-toring devices were checked. Plant housekeeping controls were observed, in-cluding control and storage of flammable-material and other potential safety

hazards. The inspector also examined the condition of various fire protection systems. During plant tours, logs and records were reviewed to determine if entries were properly made and communicated equipment status / deficiencies accurately. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, and Plant Information Report The inspector observed selected aspects of plant security including access control, physical barriers, and personnel monitoring. Except where.noted otherwise in the following, conditions were found acceptable.

i ~ 2.~1 Implementation of Plant Modifications

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Plant modification activities were reviewed against the licensee's ad-ministrative control / quality assurance procedures referenced below.

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Procedure ACP 1.2-3.1, Preparation, Review and Disposition of Plant

' Design Change Requests (PDCRs), Revision 18

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Procedure ACP 1.2-3.4, Preparation, Re~ view, Approval, Revision and Control of Specifications, Revision 0

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Procedure ENG 1.7-19, Design Inputs and Design Verification, Revi-sion 0

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Procedure ENG 1.7-13, Construction Implementation of Operating Plant Modifications, Revision 0

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Procedure ENG 1.7-17,. Design Change Notices for Design Documents, Revision 1

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Procedure ACP'1.2-6.11, Drawing Change / Submittal Requests, Revision

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Procedure ENG 1.7-11, Preoperational Testing of Plant Modifications, Revision 0

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Procedure Eng 1.7-12, Turnover of Systems, Components, and Struc-tures, Revision 0

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Procedure GE&C 5.06, Installation of Non-conforming Material, Equipment and Parts, Revision 1 DuringLthe inspection, many plant modifications were in progress. . The in-spectors reviewed the implementation of selected modification packages to verify that: The modification was properly reviewed and approved prior to implementa-tion (including evaluation of work packages risk-released for construc-

.! tion prior to final approval).

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' The plant work scope was properly documented and controlled as necessary using work orders, procedures, drawings and specifications.

Appropriate in process control of field changes was documented and ac-

complishe The modification was completed in accordance with the design documents.

l Appropriate construction and pre-operational testing was satisfactorily j completed.

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3 Appropriate quality assurance documentation was completed and reviewed prior to system turnove Appropriate training and procedures were completed prior to system turnove The following discrepancies were identifie . Plant Design Change Request (PDCR) 787 controlled installation of two_new containment penetrations in the hemispherical steel con-tainment equipment. hatch. The purpose of the modification is to provide, for use during refueling operations, sealable penetrations which can be closed and. tested to containment integrity specifica-tions during plant operation. The penetrations allow temporary hoses and wires to enter containment to support outage work without fouling the personnel access hatch. The inspector reviewed the installation of these penetrations including the documentation of receipt inspections, welding, non-destructive testing (NDT) and personnel qualifications. Construction work on this project was satisfactorily performed in accordance with the project specifica-tion and associated work procedures.' During the review of completed modification records, the inspector noted that two field changes to the project, documented on Design Change Notices (DCNs) 245-85 and 009-86, were not serialized in accordance with step 6.2.1.7 of

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administrative control procedure ENG 1.7-17. The licensee stated that each discipline engineering group had established their own numbering systems for DCNs, and this departure from procedure ENG 1.7-17 had not been reconciled or corrected. While these field changes correctly revised project documents, drawings, or the as-built system configuration, the use of several different numbering systems for DCNs compromises the ability to establish a sequential file of field changes on a project specific basis. The licensee's failure to adhere to the requirements of procedure 1.7-17 violates Technical Specification 6.8. Other examples of this violation are cited in naragraph 2.1.3 of this repor (VIO 213/86-01-01)

The inspector also noted that, upon turnover of the completed pene-trations for testing, the licensee had identified that the-Better-ment Construction group installed non qualified rubber 0-rings in the penetration door seals during construction. These 0-rings re-mained in at least one penetration at turnover. The licensee in-itiated a plant nonconformance report (NCR #86-009) to assure that qualified 0 rings were installed prior to establishment of contain-ment integrity. The NCR did not identify how the system was turned over with nonconforming materials installed, nor did the licensee's corrective action program identify measures to prevent recurrence of this deficiency. Procedure GE&C 5.06 requires documented ap-proval, control and tracking of the installation of nonconforming materials in safety-related components. These controls had not been exercised for the installation of the 0-rings during the PDCR 787

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project. As a result of inspector discussions with plant engineer-ing and construction personnel on January 14,1%6, the licensee determined that a general distribution of the d3 tails of this prob-lem to Betterment and engineering personnel / supervisors, with a caution to assure conformance to quality assurance procedures, was appropriate.- This action was completed on January 21, 1986. Based on the licensee's identification of the nonconforming condition and final corrective action to prevent recurrence, this item constitutes a licensee-identified and corrected failure to follow procedure GE&C 5.01. No further discrepancies were identifie . PDCR-785 controlled the installation of air filters in the control air lines to the Emergency Diesel Generator (EDG) air-start solenoi valves. This modification was performed as a preventive measure to assure.that foreign material in the EDG air start system would not clog or stick the air-start solenoid valve Inspector review of the approved modification package identified that the post-installation testing specified in the PDCR test plan required only a routine monthly surveillance test run of the EDG prior to return-ing the system to operation. The redundant design of the air-start system operates such that 2 of 4 air-start motors spin the engine firs If the EDG does not start, then all 4 motors spin; followed by only the latter 2 motors spinning. During routine surveillance, the EDGs usually start on the initial spin and, therefore, two of the air-start motors are not normally tested. Therefore, the ap-proved retest would not have ver.ified the full reinstatement of the EDG air-start syste The licensee's preventive maintenance program for EDGs includes an air-start system redundancy test which verifies all aspects of system operation. This test was scheduled for com-pletion during the EDG outage in which PDCR-785 was scheduled for implementation. The inspector discussed the inadequacy of the specified post-installation testing with licensee engineering per-sonnel. The project engineer indicated his intent that the sur-veillance test specified in the modification package would be the final acceptance test in a series of modification and preventive maintenance activities scheduled for this outage. The licensee stated that the combination of all these tests would verify the complete operability of the EDGs. The inspector determined that this intended sequence of testing would verify EDG system oper-ability, however,.there were no administrative controls to insure that all the appropriate maintenance and testing was satisfactorily performed, particularly if the installation of PDCR-785 had been delayed to another tim Procedures 1.2-3.1 and 1.7-11 require each-i

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modification package to stipulate, in the PDCR, the test activities necessary to verify the correct completion of the modification, and

, restoration of all system functional capabilities. The test plan reviewed and approved with PDCR-785 was inadequate because it failed to specify all the tests necessary to verify that the redundant

. aspects of the EDG air-start system were satisfactorily restored.

! This is a violatio (VIO 213/86-01-02)

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2. PDCR-789 controlled the replacement and seismic upgrade of a portion of reactor cavity drain piping. The old, non-seismic pipe run was a potential cavity drain path which could also drain the spent fuel pool. This modification provided seismically qualified piping, valves and support out to the first cavity isolation boundar Inspector review of the piping installation and testing identified no unacceptable conditions. Eight field changes (DCNs) to this project were documented in accordance with procedure 1.7-17 '.o re-vise the project drawings and specifications. Upon comple an of the modification, these DCNs were reviewed by the Plant Operations Review Committee (PORC) on January 22, 1986. During NRC review of the completed project documentation, the inspector identified that the eight DCNs (C24-86, C7-86, C29-86, C32-86, C4-86, M8-86, MS-86 and M252-85) were not serialized in accordance with step 6.2. of procedure 1.7-17. The failure to adhere to this requirement compromises the sequential file of field changes on a project specific basis. Failure to follow plant procedures violates Tech-nical Specification 6.8. A second example of this violation is cited in paragraph 2.1.1 of this repor (VIO 213/86-01-01)

Procedure 1.7-17 also requires the licensee to evaluate each DCN to determine the effect of the change on the' scope and safety evaluation of'the PDCR. Those DCNs which are determined to be significant'must be incorporated into a revision to the PDCR. Of the eight DCNs written on this project, four DCNs were identified as significant changes, and three others did not have their sig-nificance determination documented on the DCN as required by step 6.2.8 of procedure 1.7-17. The PORC review of these DCNs did not recognize and correct these deviations from procedure 1.7-17 or require that the four DCNs labeled as significant changes be in-corporated in a revision to PDCR-789 in accordance with procedure 1.2-3.1. The inspector brought these discrepancies to the licen-see's attention on January 28, 1986. The licensee determined that none of'these field changes affected the conclusions of the safety evaluation for this modification. Therefore, the DCNs should have been determined to have been insignificant, with no revision to PDCR-789 now require However, the licensee's failure to document the significance determination for DCNs C007-86, C029-86 and C032-86, and the failure to initiate a PDCR revision based on the posi-r

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tive determination of significance for DCNs CO24-86, M008-86, M005-86 and M252-85 violated TS 6.8 and procedure 1.7-17. (VIO 213/86-

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01-03) In that these violations occurred in 7 of the 8 field changes for a PDCR, they indicate-a field personnel and site management ac-ceptance of non-adherence to procedures. That acceptance transcends their otherwise minor safety significance in this case.

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3. Observation of Maintenance and Surveillance Testing l The inspector observed various maintenance and problem investigation activi-t ties for compliance with requirements and applicable codes and standards,

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QA/QC involvement, safety tags, equipment alignment and use of jumpers, per-t

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sonnel qualifications, radiological controls, fire protection, retest, and reportability. Also, the inspector witnessed selected surveillance tests to determine whether properly approved procedures were in use, test instrumenta-tion was properly calibrated and used, technical specifications were satis-fied, testing was performed by qualified personnel, procedure details were adequate, and test results satisfied acceptance criteria or were properly dispositioned. The following activities.were reviewed:

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Temporary Power to Containment Via Containment Air Recirculation Fans (Work Order 86-00144)

Local Leak Rate Test of CC-CV-721 (Surveillance 5.7-51)

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Local Leak Rate Test of CC-CV-885 (Surveillance 5.7-93)

Maintenance of Freeze Protection (Work Order 85-02002)

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3.1 On January 14 and 15, 1986, the inspector witnessed Type C Local Leak Rate Tests (LLRTs) for two containment penetrations in the Component Cooling System (CCS). The CCS is a non-seismic /non-missile protected system which provides heat removal to reactor system components inside and outside of containment. CCS containment penetrations for flow paths into containment employ a si.ngle check valve (inside containment) as the sole isolation boundary. Survrillance procedure SUR 5.7-51 measured leakage past CC-CV-721, the isolation check valve for CCS flow to the reactor coolant pump thermal barriers. Surveillance procedure SUR 5.7-93 measured leakage past CC-CV-885, the isolation check valve for CCS-flow to the reactor neutron shield tank cooler. These two valves failed their last two LLRTs. These tests are performed by isolating the CCS piping downstream of the check valves and pressurizing that portion of piping between the check valve and the downstream boundary with nitrogen at 44 psig. Leakage past each isolation check valve is collected from a drain valve upstream of the containment isolation check valve using a graduated cylinde The leakage measured for CC-CV-721 and CC-CV-885 were greater, respectively, than 1 gpm and 1000 cc in 3 seconds. The acceptance cri-terion for each test was a measured leakage less than 150 cc in 15 minutes. Each test individually exceeded the maximum allowable combined leakage rate for all Type B and C LLRT The responsible group for each LLRT is the Inservice Inspection Engineer-ing Group (ISI). However, the tests are performed by Operations Depart-ment personnel who turn over the test results to ISI within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of completion of the surveillance test. ISI is responsible for corrective actions. A Plant Information Report (PIR) is generated for test failures, to track equipment status, corrective action, reportability, and manage-ment review and followu Both procedures SUR 5.7-51 and SUR 5.7-93 instruct the tester to immediately notify ISI of a test failur Opera-tions telephoned ISI immediately following each of the observed test failures. However, ISI waited to take action upon the test results until they had received the completed surveillance procedures. Consequently,

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SUR 5.7-51, performed on January 14, 1986, at approximately 1:30 and SUR 5.7-93, performed on January 15, 1986, at 2:00 p.m., were not determined to be reportable until' January 16, 1986 at 5:45 p.m. In these cases, four-hour reports were applicable per 10 CFR 50.72 (b)(2)(iii)

because the gross leakage past these single containment boundaries com-promised the design basis containment leak rate assumed in plant safety analyses. The licensee reported these events to NRC at 6:15 p.m.' January 16, 1986 af ter self-identificatio The inspector identified the fol-lowing concerns related to the conduct and reporting of these LLRT a) Immediate test failure notification from Operations to ISI did not initiate prompt ISI action,.in that ISI evaluation was delayed until procedure turnover.from Operations (up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> later). Such treatment of surveillance test failure data could result in plant operation outside of Limiting Conditions for Operation (LCOs) de-lineated in Technical Specifications (TS). This condition could exist if LLRTs must be conducted during plant operation or if other surveillance tests such as SUR 5.7-111, Core Deluge Check Valves Leak Test, are conducted in an operating mode which requires the tested components to be operable. The plant had relaxed containment integrity prior to conducting the observed LLRTs, and the delay in evaluation did not result in exceeding an LC0 condition or cause a safety-significant delay in the NRC response on this item. How-ever, the initiation of the PIR event reporting controls was un-necessarily delaye b) Some systems / components have separate LCO and ISI acceptance cri-teria. The operator performing ISI surveillance tests on these systems.may not recognize that test results exceeding the LCO cri-terion render the system inoperabl For the short term, the licensee plans to improve the method and timeli-ness by which the ISI Group will investigate failed test data, including immediate initiation of their evaluation process upon receipt of a tele-phone notification from operations. This evaluation will determine the resulting' condition of the system and the need for initiation of a PI For the long term, ISI is reviewing all of their surveillance procedures to identify those procedures satisfying both ISI and LC0 requirement The licensee plans to revise those procedures to clarify the bases for the acceptance criteria, such that test reviewers can easily discern when an equipmento 'perability limit has been exceeded. The inspector found l

the planned licensee actions on this licensee identified item acceptable and will' follow the implementation of these changes during a subsequent j inspection. (IFI 213/86-01-04)

4. Followup on Previous Inspection Findings

, During the course of the inspection, two NRC'open items were reviewed. Since

further licensee action and NRC followup were identified, these items remain i

ope Details follow:

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4.1 (0 pen) Followup Item (213/84-28-03) The licensee was to implement a revised leak test methodology for emergency core cooling system check valves. On January 10, 1986, the licensee approved surveillance proce-dure 5.7-111,. Core Deluge Check Valves Leak Test. Inspector review of this procedure identified inadequacies in the required format of the test acceptance criteria and prompt processing of failed test data. These

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problems were corrected in Revision 1 to procedure 5.7-111 dated February 26, 1986. This item will remain open pending satisfactory completion of this test (scheduled in March 1986).

4.2 (0 pen) Unresolved Item (213/85-21-09) Verification of the correct operation of the low pressure over pressure (LPOP) relief valves during the plant cooldown for the 1986 outage was planned for close out of NRC review of a malfunction of these safety valves during the previous re-fueling outage. On January 5,1985, the LPOP system was placed in ser-vice with reactor coolant system temperature and pressure at about 330F and 340 psig, respectively. The LPOP safety valves remained seated as

designed, and the system remained functional throughout the cooldown process. However, the licensee experienced difficulty in placing the LPOP system in service due to problems with pressur'e and temperature interlocks which prevented opening the LPOP system isolation valve After multiple verifications of the correct plant conditions (340F, 340 psig) for system operation, the licensee jumpered the LPOP isolation valve interlocks and placed the system in operation. Subsequent inves-tigation determined that an out-of-calibration ' temperature interlock setpoint in one train and an inoperable pressure interlock bistable in the other train prevented opening the isolation valves for either LPOP train until the interlocks were bypassed. The licensee repaired the discrepant interlock circuits to restore the system to full operabilit In addition, the licensee is reviewing the interlock design and system specifications to determine whether further actions are required to in-sure the proper functioning of the LPOP system on deman The inspector noted that interlock circuit checks may need to be performed more fre-quently and prior to plant cooldowns. Based on the satisfactory per-

'formance of the LPOP system in service, licensee corrective action in response to the previous premature operation of LPOP safety valves'was

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found to be adequat This item remains open pending NRC review of cor-l'

rective actions to assure proper operation of the LPOP isolation valve interlocks.

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5. Followup on Events Occurring During the Inspection 5.1 Licensee Event Reports (LERs)

The following LERs were reviewed for clarity, accuracy of the description

of cause, and adequacy of corrective action. The inspector determined

, whether further information was required and whether there were generic 3 implications. The inspector also verified that the reporting require-l ments of 10 CFR 50.73 and Station Administrative and Operating Procedures

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had been met,.that appropriate corrective action had been taken, and that the continued operation of the facility was conducted within Technical Specification Limit Inoperable Fire Doors 86-02 -- Mai r. Steam Safety Valve Test Failures - Event detailed in NRC h spection Report 50-213/85-2 ' Steam Generator #2 Eddy Current Test Result . Category C-3 86-04 -- Low' Pressure Over pressure Protection System Interlock Mal-function - Event detailed in paragraph 4.2. of this report 86-05 -- Switchgear Room Halon System Inoperable during Extended Main-tenance 5.2 Steam Generator (SG) Eddy Current Testing (ECT) -- LER 86-03 On January 19, 1986, the licensee SG ECT began on the No. 2 SG. By Janu-ary 27, 1986, 3610 tubes had been inspected and evaluated in this SG with a total of 367 degraded tubes (indications greater than 20% wall thick-ness) and 19 defective tubes (indications greater than 50% wall thick-ness). The licensee reported the No. 2 SG as a category C-3 SG (greater than 10% degraded tubes). ECT of the other 3 SGs proceeded with the in-tent of inspecting all tubes in each SG. The inspector followed the progress of ECT in SGs 3 and 4. After completing evaluations of the tests of a11' tubes, except 4 tubes in SG #4 (which were not tested due to difficulties getting the test apparatus into these locations), SGs 3 and 4-were declared Category C-2 based on about 8% degraded tubes per SG and less than 1% defective tubes. At the end of the inspection period, ECT of #1 SG was still in progress. These ECT results, the dis-position of the 4 untested tubes in #4 SG, and the completion of defec-tive tube plugging or sleeving will be reviewed during a subsequent in-spection (IFI 213/86-01-05). In regard to the untested tubes in #4 SG, site management was advised that relief from Technical Specification requirements for testing those tubes should be sought expeditiously in order to minimize the potential for delay of plant operation . Review of Periodic and Special Reports Upon receipt, periodic and special reports submitted pursuant to Technical Specification 6.9 were reviewed. This review verified that the reported in-formation was valid and included the NRC required data; that test results and supporting information were consistent with design predictions and performance specifications; and that planned corrective actions were adequate for resolu-tion of the problem. The inspector also ascertained whether any reported in-formation should be classified as an abnormal occurrence. The following periodic reports were reviewed:

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Monthly.0perating Reports 85-12 and 86-01 These reports cover plant operations from December 1,1985 - January 31, 198 No unacceptable conditions were identifie . Unresolved Items Unresolved items are matters about which more information is required in order to determine whether they are acceptable items or violations. Unresolved items identified during this inspection are discussed in Paragraph . Exit Interview During this inspection,' meetings were held with plant management to discuss the finding No proprietary information related to this inspection was identified.