IR 05000293/1986034

Partially Withheld Insp Rept 50-293/86-34 on 860916-1020 (Ref 10CFR73.21).Violation Noted:Implementation of Security Measures.Concerns Re Various Security & Fire Protection Issues Noted

ML20212B339
Person / Time
Site:	Pilgrim
Issue date:	11/06/1986
From:	Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20212B307	List: IR 05000293/1986034 ML20212B301 ML20212B314
References
50-293-86-34, NUDOCS 8612290224
Download: ML20212B339 (22)

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~l U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket / Report No. 50-293/86-34 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: September 16, 1986 - October 20, 1986 Inspectors: M. McBride, Senior Resident Inspector J. Lyash, Resident Inspector L. Doerflien, Project Engineer Approved by: #d/B4

[y.Strosnider, Chief,ReactorProjects Date wection 18 Areas Inspected: Routine resident inspection of plant operations, radiation protection, physical security, plant events, maintenance, surveillance, outage activities, and reports to the NR Results: One violation was identified regarding implementation of security measure Concern about the response of security management to identified problems is expressed in section 3.d. Licensee commitments to revise the MSIV maintenance procedure, and to submit a clarification memorandum to NRR concerning TAP Item II.D.3, are described in section 2.0. Also discussed in section 2.0 is a concern regarding failure to act on identified, recurring component failures in the RKR syste Concern about the effectiveness of the licensee's evaluation of the corrosion in the salt service water system is discussed in section 3.b. The unrecognized entry into a limiting condition for operation (LCO) is discussed in section 3.c. Review of the refueling bridge preoperational test and concerns regarding the quality designation of the refueling interlocks are discussed in section 3.c. Continued problems in the area of fire protection are detailed in section Section 4.b describes the discovery of a disconnected reactor protection system wire. Weld flaw indica-tions and a significant number of piping support inspection failures identified by the licensee's ISI program are discussed in section 4.c. An apparent weakness in the licensee's LER tracking and trending ability is identified in section 5.0. Section 6.0 discusses the licensee's request for extension of certain established commitment date PDR ADOCK 05000293 o PDR

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TABLE OF CONTENTS

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P_ag Summary of Facility Activities ........................ 2 Followup on Previous Inspection Findings, . . . . . . . . . . . . . . 2 Violations, Unresolved Items, Inspector Follow Items, TMI Action Plan Items Routine Periodic Inspections .......................... 8 System Alignment Inspection .......................... 9 Plant Maintenance and Outage Activities .............. 9 Surveil lance and Preoperational Testing . . . . . . . . . . . . . . 10 Refueling Mode Surveillance Testing Refueling Bridge Preoperational Test Review Physical Security .................................... 12 Fire Protection ...................................... 14 Radiation Protection ................................. 16 Review of Shutdown Cooling Mode of Operation ......... 16 Review of Plant Events .................................... 17 Residual Heat Removal System Isolation ............... 17 Loose Reactor Protection System Wire ................. 17 Inservice Inspection Program Results ................. 17 Review of Licensee Event Reports ( LERs) . . . . . . . . . . . . . . . . . . 18 Review of Selected Licensee Commitments . . . . . . . . . . . . . . . . . . . 19 Management Meetings ....................................... 20 Attachment I - Persons Contacted Attachment II - Licensee Evaluation of the Refueling Bridge Interlocks

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A DETAILS 1.0 Summary of Facility Activities The plant was shutdown on April 12, 1986 for refueling and completion of certain modifications. During the report period residual heat removal (RHR) pumps A and 0, and core spray pump A, were disassembled and inspected. Distortion of the A RHR pump casing and overheating of the impeller were identifie Final tie-in of Appendix R modifications for emergency diesel generator "A" was begun. The licensee inservice inspection program has identified surface indications on main steam, feedwater, standby liquid control and the residual heat removal system welds. Fire barrier walkdowns conducted by the licensee have identified a number of discrepancie The licensee, in conjunction with the Federal Emergency Management Agency, conducted a full scale emergency siren test on September 29, 1986. Boston Edison and the NRC announced a postponement of the 1986 annual emergency preparedness exercise to allow other agencies the opportunity to participate in the drill. The exercise had been scheduled for October 23, 1986, but will be delayed until December 10, 198 The SALP period for Pilgrim has been extended. The period, which started on November 1, 1985, will be extended from October 31, 1985, to April 31, 198 .0 Followup on previous Inspection Findings Violations (Closed) Violation (84-04-01), procedure valve lineup specified two valves to be closed which if implemented, would have made the containment atmosphere dilution system inoperable. The inspector reviewed procedure 2.2.70, Primary Containment Atmosphere Control System, Revision 30. The licensee has revised the procedure to correctly specify the normal position of the two blocking valves. As stated in the licensee's response, word processing activities have been transferred to site, and revision review responsibilities assigned. Since this violation the licensee has completed the Procedure Update Program which verified the accuracy of safety-related procedures. The inspector also compared the procedure valve lineups with plant drawings for two other systems and did not identify any similar discrepancies. The accuracy of valve lineup procedures and plant drawings are reviewed during bimonthly Engineered Safety Feature system walkdowns by the inspectors. The inspector had no further questions and considers this item close _ _ - - _ . . - . - _ _ _ _ _ . - _ -

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(Closed) Violation (86-01-08): Failure to log disabled control room annunciators. The inspector reviewed procedure no. 2.3.1, " General Action (Alarm Procedures)", and discussed disabled annunciators with an on-shift licensed operator. The inspector also reviewed the control room disabled annunciator log to verify all disabled annunciators were entered-in the log. The inspector determined that the information contained in the log was in accordance with the procedural requirements and that the operator was knowledgeable in maintenance of the log. The inspector had no further questions and considers this item close (Closed) Violation (86-01-09): Failure to follow procedures for completing a post trip review. In response to this violation the licensee issued an instructional memorandum, to the individuals involved in the preparation of post trip reviews, reinforcing the need for procedural adherence. The inspector reviewed procedure no. 1.3.37, " Post Trip Reviews," and two subsequent completed scram reports for adequac No additional deficiencies were noted. The inspector had no further questions and considers this item close (Closed) Violation (86-06-07), failure to perform insitu testing of the standby liquid control system (SLCS) explosive squib valve charge The inspector reviewed procedure 8.4.6, Manual Initiation Test of the SLC System, Revision 11. The licensee has revised the procedure to clearly require that the explosive squib valve charges be fired in place during the manual initiation test of the SLCS and that the replacement squibs have the same batch number as the fired charge. The inspector had no further questions on this ite Unresolved Items (Closed) Urresolved Item (84-18-02), provide acceptance criteria for snubber piston extensio Station procedure 3.M.4-28, Revision 11, for in-service inspection of snubbers listed expected settings for piston rod extensions with the system in the cold conditio Piston rod extension acceptance criteria, however, were not included The inspector reviewed procedure change notice (PCN)84-983 and the current revision of procedure 3.M.4-28 and verified that acceptance criteria had been adde Attachment B to the procedure states that "as-measured" piston extensions of plus or minus one quarter inch of the specified cold settings are acceptabl In support of this acceptance value PCN 84-983 references i nuclear engineering department memorandum NED 84-59 This memorandum documents engineering acceptance of the specified value. This item is close (Closed) Unresolved Item (84-40-01), licensee to evaluate impact on operation with control rod 42-39 inoperabl The inspector noted that the licensee operated the entire cycle 7 with control rod 42-39 and three other symmetrical rods fully inserted. The inspector reviewed a letter from the fuel vendor, dated January 16, 1985, which indicated that the inserted control rods would not adversely affect the cold shutdown margin or the transient analysis result The licensee has concluded that the

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inability to fully withdraw control rod 42-39 was due to bent spud fingers. The inspector noted that the licensee plans to remove this control rod drive mechanism during the current refueling outage to replace the spud. The inspector had no further questions on this ite Inspector Follow Items (Closed) Licensee Event Report (LER) Follow Item (81-LO-57), limitorque operator mounting bolts loose. This item was last updated in inspection report 82-16. It was originated in response to LER 81-57 describing loose valve motor operator mounting bolts for MOV-1001-36A and 1001-3 Problems with loose operator and valve bolts appear to be generic to other RHR valves. This problem is discussed in more detail under inspector follow item 85-28-03. Based on the update of item 85-28-03 contained in this report, this item is administratively close (Closed) Inspector Follow Item (84-23-02), review licensee relabeling of valves. The licensee is currently implementing an extensive valve relabeling program. This program was reviewed and discussed with the licensee during inspection 86-06. Adequacy and implementation of the effort is the subject of inspector follow item (86-06-03). Based on the above this item is administratively close (Closed) Inspector Followup Item (84-28-03), the high pressure coolant injection (HPCI) system preoperational test procedure lacked independent verification of system restoration. The inspector reviewed the completed data sheets for procedure TP-83-4, Pre-operational Testing of Valves Replaced in the HPCI System. The licensee revised the procedure to include independent verification of system restoration and this verification was properly implemented. The inspector also reviewed procedure 2.2.21, HPCI System, Revision 26, to verify that the system spectacle flange was included in the operating procedure valve lineup. This item is closed. As documented in inspection reports 86-06 and 86-25, the acceptability of the licensee's overall program for independent verification is still under revie (Closed) Inspector Followup Item (84-36-02), review adequacy of shift turnovers and noting off normal conditions. The inspector reviewed procedure 1.3.34, Conduct of Operations, Revision 10, and determined that the licensee's administrative controls for watch relief were adequat The inspector noted that the recent diagnostic team inspection (report no. 50-293/86-06) specifically reviewed implementation of the shift turnover process and found it adequate. The inspection team also concluded that licensed operator response to annunciators and off normal conditions was generally proper and thorough. The inspector had no further questions on this ite (Closed) Inspector Follow Item (86-14-02), cracked MSIV return spring This item was last updated in inspection report 86-21. General Electric Service Information Letter (SIL) 442 notified licensee's of potential cracking of main steam isolation valve (MSIV) return springs in valves

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supplied by Atwood & Morrill Compan SIL 442 recommends visual inspection of accessible spring surfaces, replacement of any cracked springs, and a ten cycle overload test of all new springs. Subsequently, IE Information Notice (IEN) 86-81 was issued addressing the potential cracking problem. IEN 86-81 states that the inspection and testing procedure described in GE SIL 442 is acceptable for affected BWR The licensee performed an informal inspection of the springs on June 19, 1986 using maintenance request 86-1-34. In addition, a visual inspection was performed by the licensee Quality Control (QC) group on July 10, 1986. This inspection is documented in QC IR 86-1-45. Prior to these inspections all eight valves had been disassembled to allow correction of a pilot poppet design problem. No discrepancies were identified. The licensee had revised the MSIV maintenance procedure, 3.M.4-8, to add a prerequisite that all new springs pass a five percent overload tes However, no reference to the test method prescribed by SIL 442 was included. In response the licensee initiated a procedure change notice specifying that the overload test be conducted in accordance with SIL 442. The inspector had no further question (Update) Inspector Follow Item (85-28-03), review cause for MO-1001-36A valve failure to open on November 4, 1985. This item was last updated in inspection report 86-31. Valve 1001-36A is a twelve inch motor operated globe valve in the A loop of RHR torus cooling. Licensee investigation identified two contributors to the valve failure. A loose operator control wire termination and loose valve yoke to operator mounting bolts were foun The wire was torqued to a value consistent with industry standards, and all other terminations checked. The licensee has approved a motor operator preventive maintenance procedure containing specific guidance on torque values to be used for various terminations. This procedure will be implemented for all environmentally qualified operators during this outage. The loose valve yoke to operator mounting bolts were retorque A brief review of past Licensee Event Reports (LER) and Failure and Malfunction Reports (F&MR) by the inspector identified a number of valve bolting and weld problems in the residual heat removal syste LER 79-028 6/27/79 Loose valve operator mounting bolts on MO-1001-36A LER 81-051 9/17/81 RHR test connection weld failure LER 81-57 10/9/81 Loose and sheared valve operator mounting bolts on MD-1001-36A and MD-1001-32 F&MR 85-231 11/4/85 Loose valve operator mounting bolts on MO-1001-36A

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F&MR 86-005 1/4/86 RHR test connection weld failure F&MR 86-091 4/14/86 Loose valve body to bonnet bolts on MD-1001-298 F&MR 86-294 10/14/86 Loose valve body to bonnet bolts on MD-1001-51 F&MR 86-222 8/86 Five RHR instrument line failures 233 (see IFI 86-29-02)

239 244 245 In addition, loosening of operator mounting bolts on core spray valves M0-1400-4A and 1400-4B has been a recurring problem (see inspection report 86-14).

In October, 1983 the station issued a Management Corrective Action Plan Request (MCAR) identifying repeated vibration induced failures in the residual heat removal, core spray and reactor water cleanup system This document also references industry information discussing similar problems; INPO SER 64-83 and NRC IE Information Notice 83-7 Failure and Malfunction Report Trend Analysis reports compiled by the licensee's station technical engineering staff also identified this proble Reports issued in 1983, 1984, 1985 and 1986 cite a pattern of valve bolting and small diameter pipe weld failures in low pressure ECCS, and list a large number of specific failure During the inspection period the licensee produced ne indication that action had been taken to address the generic concerns identified in the MCAR and four Failure and Malfunction Report Trend Analyse This apparent failure to followup clearly documented adverse trends is considered a significant weakness. The inspector noted that Quality Assurance Deficiency Report 1550 also identified that adequate corrective actions were not being taken in response to the F&MR trend analysi Licensee engineering stated that an evaluation had been initiated in response to the RHR test connection weld failure experienced in January, 1986. A second investigation was begun in response to the recent w?id failures. In responsa to the inspector's questions the licensee's technical engineering group initiated an engineering service request to address the possible generic problem. The inspector will evaluate the licensee s process for followup of identified adverse trends and plans for resolution of this issue during a future inspectio . .

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(Update) Inspector Follow Item (86-14-07), review licensee evaluation of loose RPS ground wiring. An additional example of loose RPS wiring was identified by the licensee on October 14, 1986. Details of this most recent problem are included in section 4.b of this report. This item is updated to include followup of this inciden (Update) Inspector Follow Item (86-29-05), review licensee analysis of 1400-48 yoke indications and 1400-4A examination results. These valves were designed using an adapter plate between the valve operator and yok While investigating recurring yoke to adapter plate capscrew failures the licensee identified excessively high torque switch settings. The valve manufacturer, Anchor / Darling, was requested to perform a detailed stress analysis and identify the impact on valve components. This analysis indic.ated possible damage or failure of the valve yoke in the area of the yoke flange. Anchor /Dorling recommended a magnetic particle examination of all yoke surfaces. Examination of the 1400-4B valve yoke revealed a 7/16 inch long crack. This crack was repaired. Examination of the 1400-4A valve found no indication The stress analysis performed by Anchor / Darling was based on the design loads, including the 8200 lbs thrust generated at the design torque switch settin Under these conditions, the yoke flange area was subjected to stress in excess of twice that allowed by ASME code requirement The Anchor / Darling supplied valve yoke did not meet design requirement Actual as found torque switch settings produced thrust values of approximately 16,000 lbs, adding to the calculated yoke stres On October 14, 1986, the inspector questioned licensee engineering personnel regarding: 1) possible damage to other valve components, 2) :

the effect of the cyclic overload on the valve yoke, 3) the presence of any valves utilizing a similar adaptor plate design and 4) the generic implications of the Anchor / Darling design erro The licensee stated that the design analysis indicates no other component damage. The two yokes in question will be replaced during the current outage. Review of station valves identified one additional valve utilizing the adapter plate design; MOV-1201-80. The licensee stated that a review of this valve's stress analysis would be conducted. On October 16, 1986, the inspector was informed that the vendor was evaluating the design deficiency for reportability under Part 2 This item will remain open pending 1) completion of review of the stress analysis for the valves in question and 2) review of the Anchor / Darling evaluation and its impact on other valve design TMI Action plan Items (Update) TAP Item II.D.3, Safety Relief Valve Position Indication. This item was last updated in inspection report 50-293/86-25. The licensee has made conflicting commitments under the TAP item and Regulatory Guide 1.97 regarding the qualification of this instrumentatio Licensee

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engineering personnel stated that a letter would be issued to the NRC:NRR clarifying their position on safety relief valve position indication equipment qualificatio .0 Routine Periodic Inspections The inspectors routinely toured the facility to assess general plant and equipment conditions, housekeeping, and adherence to fire protection, security and radiological control measures. Ongoing work activities were monitored to verify that they were conducted in accordance with approved administrative and technical procedures, and that proper communications with the control room staff had been established. The inspector observed valve, instrument and electrical equipment lineups in the field to ensure that they were consistent with system operability requirements and operating procedure During tours of the control room the inspectors verified proper staffing, access control and operator attentiveness. Adherence to procedures and limiting conditions for operations were evaluated. The inspectors examined equipment lineup and operability, instrument traces and status of centrol room annunciators. Various control room logs and other available licensee documentation were reviewe In addition to routine equipment operability confirmation the inspectors performed independent walkdowns of seiected safety system Confirmation of the as-built system configuration, identification of any degraded conditions and procedure adequacy were evaluate The inspector reviewed outage, maintenance and problem investigation activities to verify compliance with regulations, procedures, codes and standard Involvement of QA/AC, safety tag use, personnel qualifications, fire protection precautions, retest reqairements, and reportability were assesse The inspector observed tests to verify performance in accordance with approved procedures and LC0's, collection of valid test results, removal and restoration of equipment, and deficiency resolutio Radiological controls were observed on a routine basis during the reporting period. Standard industry radiological work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were observed. Independent surveys of radiological boundaries and random surveys of nonradiological points throughout the facility were taken by the inspecto Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those checks included security staffing, protected and vital area barriers, access control, badging, and compensatory measures when required.

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. System Alignment Modifications to the plant electrical distribution system to satisfy 10 CFR 50, Appendix R requirements continued throughout the inspection period. The "A" emergency diesel generator was removed from service to facilitate tie-in of newly installed power and control cable and installation of improved safe shutdown control panel Certain portions of the associated 4160 VAC and 480 VAC distribution systems were also removed from service to allow similar appendix R tie-ins. In ptrallel with this effort the licensee continued inspection of the A and C RHR and A core spray pumps, as described in section 4.0 of this repor The inspectors evaluated the impact of this ongoing work on decay heat remosal and other system operability. Electrical and mechanical system lineups implemented to minimize the impact on plant operations were examined. Limitations on allowable plant activities imposed because of equipment status were reviewed. No problems were identifie Plant Maintenance and Outage Activities

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The inspector observed activities associated with the Appendix R modifications program. Cable pulling, splicing and termination work appeared to be well controlled, and conducted in a technically acceptable manner. Post work tests, such as continuity and insulation tests were properly performed and documente On October 14, 1986, the licensee identified a leak in the station salt service water (SSW) system, at the outlet of the

"A" reactor building closed cooling water (RBCCW) heat exchanger. The leak was caused by a one eighth inch diameter hole in a section of eighteen inch diameter pipin Inspections conducted by the licensee's quality control group indicate that pipe wall thickness in the local area surrounding the hole is significantly less than expected. Preliminary licensee evaluation suggests the cause to be localized corrosion. This piping is lined, with rubber coating to prevent the salt water from attacking the carbon steel. The "A" SSW loop has been isolated and drained. The licensee plans to remove the affected section of pipe for inspection and repai The effectiveness of the licensee's evaluation, and control of salt water corrosion in the salt service water system is unresolved, pending further inspector review (86-34-01).

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The A and C residual heat removal (RHR) and A core spray pumps and motors were disassembled during this report perio Inspection of the A RHR punp impeller and shaft showed significant discoloration and shaft distortion due to overheating. Pitting of the pump impeller, probably due to cavitation, was also identified. The pump casing was found to be severely distorted. Static clearance between the stuffing box and casing, and running clearances between the pump wear rings were unacceptable. The licensee will replace the pump shaft and impeller because of the overheating and damage caused during disassembly. General Electric, in conjunction with the pump vendor, has proposed a machining repair to restore the original casing tolerances. The D RHR pump previously showed less severe signs of overheating. Licensee investigation of the distortion and overheating is in prog ess. The inspectors will continue to closely monitor the progress and results of these activitie During the period the inspector was informed that Pennsylvania Power & Light Company had discovered terminal blocks utilized in 480 VAC applications at their Susquehanna station which were not environmentally qualified. The inspector discussed this problem with Boston Edison engineering personnel. The licensee informed the inspector that several types of terminal blocks

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had been qualified for use at Pilgrim, and referenced applicable qualification test results. The inspector had no further question c. Surveillance and preoperational Testing Refueling Mode Surveillance Testing Pilgrim technical specifications require operability of certain reactor protection system (RPS) trip functions with the mode switch in refuel position. On October 3, 1986 the inspector noted that the reactor mode switch had been placed in the refuel positio Surveillance testing to demonstrate operability of two of the required RPS trips, IRM and ApRM high flux, had not been performe Licensee personnel were not aware that the plant was in an LCO at the time the inspectors questioned the apparently missed surveillance tests. The situation was not clearly defined and the potential for violation of the technical specification existed. The licensee placed the mode switch to shutdown when informed of this situatio The technical specification action prescribed if this RPS operability requirement can not be met is to complete insertion of all control rods within four hours. The inspector verified that all control rods were inserte Review by the licensee determined that no rod had been withdrawn since the time the LCO had been entere Based on this information it appears that the licensee was not in

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violation of the technical specifications. The licensee

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subsequently returned the mode switch to refuel and placed an administrative caution tag on the switch stating that no rod withdrawal was permitte The surveillance tests were not performed because no test due date had been specified on the Master Surveillance Tracking Program (MSTP). Instead the due date in the MSTP was replaced with a RF0VII designation. Because the MSTP cannot relate tests to plant modes of operation, the tests were not scheduled. Similar MSTP problems have been previously identified (See IFI 86-21-08). At the exit meeting, the inspector expressed concern that the LCO was not recognized and that the MSTP was not able to properly schedule tests which depend on plant mod Refuel Bridge Preoperational Test Review On September 25, 1986, the inspector noted that the preop test procedure for the newly installed refueling bridge Temporary Procedure, TP-86-127 did not include independent verification of those procedure steps that checked the refueling interlocks. In addition, procedure prerequisite step VI, G (Q.C. hold and witness points) was marked "No Q.C. requirements, non-Q", and signed by a Q.C. representative. The Plant Design Change, P.D.C. 85-58,

" Refueling Bridge Replacement", that was used this outage to install a new bridge (including portions of the interlocks) was also marked non- The inspector questioned the lack of independent verification and the non-Q designation for the refueling interlocks. In response, the licensee inserted verification steps into the procedure and added Q.C. nold and witness points. However, a licensee engineering evaluation, dated October 15, 1986 indicated that the interlocks had been properly classified as non-Q. The evaluatior indicated that procedural restrictions and technical specification limits (not the interlocks) form the primary means of preventing an inadvertent criticality during refueling. The evaluation referred to Attachment G to the FSAR, General Electric Service Information Letter (SIL) 372, and other information from General Electric to support the non-Q determinatio However, the Pilgrim FSAR Section 7.6, "Rcfueling Interlocks",

contains two safety design bases for the interlocks. Section 7.6.4,

" Safety Evaluation", states that the reft.eling interlocks are designed to prevent criticality during refueling operation Procedural controls are described as a fourth level of backup to the interlocks. Section 7.6.6 in the FSAR states that the interlocks

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are required, along with certain refueling operation restrictions to provide assurance that operations remain within the envelope of conditions considered by the station safety analysis. The station safety analysis, FSAR section 14.5.5, " Fuel Handling Accident"

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states that the refueling interlocks impose restrictions which prevent an inadvertent criticality during refueling operation Procedural controls are not discussed in this section of the safety analysi This item is unresolved, pending further inspector review and NRR review (86-34-02). Security On October 9, 1986, an article in a local news publication quoted local authorities as stating that no additional special police appointments for station security guards would be made until applicant screening practices are reviewed. The inspectors questioned the affect of this on the licensee security program. It was determined through discussion with NRC security specialists that while special police powers are an item discussed in the security plan, they are beyond the licensee's control. In the event that the licensee cannot comply with the provisions of the plan, a change may be necessary in accordance with the provisions of 10 CFR 50.54(p)

within sixty days. Steps taken to investigate the impact on the security program were slow and appeared to be conducted in response to prompting by the inspecto THIS PARAGRAPH CONTAINS AND IS NOT FOR PUBLIC DISCLOSURE, IT IS INTENTIONALLY LEFT BLAN .

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THIS PAGE CONTAINS SAFEGUARDS INFORETION AND IS NOT FOR PUBLIC DISCLOSURE, IT IS INTENTIONALLY LEFT BLAN :

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THIS PARAGRAPH CONTAINS AND IS NOT FOR PUBLIC D.ISCLOSURE, IT IS INTENTIONALLY LEFT BLANK.

. Fire Protection Recently during review of the once per cycle fire barrier surveillance procedures the licensee identified approximately 25 walls which were not included for inspection during RF0 #6 as required by existing commitments. At least four of these 25 walls are still required to be fire barriers under the current technical specifications. These missed walls indicate weakness in the licensee's program for identifying and testing fire barrier The Itcensee has completed approximately twenty five percent of the walkdowns required by the existing fire penetration surveillance procedures. Approximately 75 maintenance requests (MR) have been initiated in response to barrier penetration deficiencies. The licensee believes that few deficiencies constitute actual penetration degradation. However, at least one missing penetration seal and several degraded seals have been identified. The 75 MRs written to date could be classified as:

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Discrepancies identified in barriers no longer required by technical specifications

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Discrepancies resulting from unclear acceptance criteria which may not constitute seal degradation

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Actual degraded or missing seals Because of the number of identified discrepancies the licensee determined that the. integrity of all uninspected barriers was

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Based on this concern the licensee took prompt action and established fire watches in uninspected areas. The method used to

establish these watches, however, was not in accordance with the licensee's approved fire protection progra No specific firewatch postings were made. Instead personnel were permitted to make the required area tours and document successful inspection of these areas by marking a clipboard carried with them. In light of the large number of missed fire watch postings identified and described below this relaxation of fire watch administrative controls does not seem prudent. The demonstrated unreliability of fire watch i m m-.-m-r --rw,- ,-e,,-, ~-~---r---,,,e-,n-- , , a__,-n m--- .g,, ,.,c-, , , - - - - - , , , - - - , - - - - - - - - - . - , - - - - - , - . - - - - - ~ , , - - - - ~

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implementation indicates that more stringent monitoring is warrante The Operations Review Committee was not given the opportunity to review this deviation from normal station practic In summary the points listed below appear to be real or potential fire protection problems:

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Fire barriers required by technical specifications have not been accurately identifie Penetration inspection acceptance criteria are not clearly define At least four technical specification required barriers were not inspected during RF0 # Missing or significantly degraded barrier penetration seals have been identifie Weak compensatory action was taken for suspected fire barrier deficiencie These items will remain unresolved pending inspector followup and completion of the licensee's evaluation (86-34-04).

Licensee review on October 1, 1986 noted 17 instances in a twenty-four hour period where hourly fire watches were not properly documented and probably missed. The licensee reported the failure to perform the requirea watches via ENS. Licensee followup to this incident was limited. A day after the condition was identified and reported to the NRC, the licensee had not identified the cause of the missed watches and had not taken action to prevent recurrenc Steps to address these concerns were taken after questioning by the inspectors. This slow response was discussed with plant management during the period and at the exit intervie The inspector discussed control of combustible material near the licensee's gas bottle storage facility with the licensee prior to and following the NRC management meeting on September 9, 198 According to signs posted on the facility, combustible material should not be stored within 50 feet of the building. However, combustible material was found near the building on several occasions. Lack of aggressive licensee corrective actions for this problem was discussed in NRC inspection 50-293/86-29. Additional combustible material (wooden boards) was noted in a pile adjacent to the building on October 17, 1986. The inspector discussed the continuing problem with the plant Manager. The material was promptly removed. The control of combustible materials in the plant will be reviewed during future routine plant tours. The inspector had no further questions at this tim .

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. Radiation Protection On October 16, 1986 the inspector checked radiation levels at several dirt piles located in and near the licensee's contractor parking area. No levels above area background were detected. A licensee health physics technician accompanied the inspector during

~the survey. Survey instruments included an RM-14 counter with a pancake GM probe, a micro-R meter, and a small GM survey meter. The licensee indicated that a portion of the dirt had originated in areas inside the protected area security fence. The inspector had no further questions about this material at this tim Review of the Shutdown Cooling Mode of Operation The inspector reviewed procedure no. 2.2.86, "Resioual Heat Removal," Revision 25; held discussions with licensee personnel; and observed the Shutdown Cooling (SDC) lineup in operation during the inspection to determine the adequacy of the licensee's controls to prevent inadvertent draining of the reactor vessel to the suppression pool (torus) through the Residual Heat Removal (RHR)

syste The inspector compared the licensee's controls to the conclusions and recommendations of the NRC Office of Analysis and Evaluation of Operational Data (AE00) engineering evaluation report

" Inadvertent Draining of Reactor Vessel During Shutdown Cooling Operation," dated August 1986. The inspector noted that the procedure contains caution statements to alert the operator of the potential for draining the reactor vessel when placing SDC in operation. The procedure also requires that the RHR pump suction, minimum flow, crossconnect, containment spray, and torus cooling / spray valves be shut and tagged with Nuclear Water Engineer tags when placing SDC in operation. In addition, the Technical Specifications require that the reactor vessel low level and high pressure isolation functions (for the SDC suction valves and RHR injection valve) be operable during SDC operation. Although no interlocks exist between the RHR pump, SDC and torus suction valves, the inspector determined that the licensee's controls to prevent inadvertent draining of the reactor vessel through SDC appeared adequate and were in general in agreement with the recommendations of the AE00 repor The inspector also noted that the licensee placed IE Information Notice 84-81, " Inadvertent Reduction in primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup" in the licensed operator required reading to provide this operating experience feedback to all operators. The inspector had no further questions in this are .

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4.0 Review of Plant Events The inspectors followed up on events occurring during the period to determine if licensee response was thorough and effective. Independent reviews of the events were conducted to verify the accuracy and >-

completeness of licensee informatio Residual Heat Removal System Isolation . . -

On September 17, 1986, a spurious instrument signal caused the RHR pump suction valves to automatically close. The RHR system was aligned in shutdown cooling at the time. The isolation signal was quickly cleared and the valves reopened. The temporary loss of cooling flow was of no safety concern since the reactor residual heat load is low and other systen.s are available _ to control reactor terperature. The licensee informed the NRC, via the ENS, of the automatic isolation. Licensee investigation to identify the source of the spurious signal is in progres .

g ,; s Loose Reactor protection System Wire On October 17, 1986, the licensee notified the NRC via ENS that a disconnected reactor protection system (RPS) wire had been identi-fied. The lead had been pulled completely out of the termination lug, which remained attached to the terminal strip. The wire is part of the circuitry providing an input to RPS when either main steam line isolation valve (MSIV) in the 0 main steam line is close It also drives a control room annunciator and computer point under these circumstances. The loose wire may have gone ur. detected because the MSIVs have been closed since plant shutdown in April, deenergizing this portion of the RPS circui This wiring is located in a main control room back panel. Extensive modification work for installa-tion of the new process computer and analog trip system is being performed in these panels. The licensee is investigating the cause of the loose wire. The inspectors will monitor this investigation under a previous inspector follow item (86-14-07). Inservice Inspection (ISI) Program Results Boston Edison began the second ten year inservice inspection program at Pilgrim on December 8, 1982. This program is to be conducted to meet the requirements of ASME Section XI 1980 edition, including the 1980 winter addendum. Initial inspections were conducted during the 1983/1984 outage. Inspections planned during the present refueling outage will complete the requirements for the first inspection period of the second 10 year interva .

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Under the version of the code used during previous inspection programs only volumetric examination of class 1 and 2 pressure retaining piping welds (Category B-J and C-F) was required. The 1980 edition however, requires a surface examination in addition to volumetric examination of the lower one third of the weld. In -

submitting its second ten year plan Boston Edison requested relief from the surface examination requirement. Inspections conducted during the 1983/1984 outage were performed based on the assumption that this relief would be granted. The relief request, however, was denie During this report period the licensee began performance of surface examinations which, based on the pending relief request, were not conducted during the 1983/1984 outage. A total of 232 surface examinations are schedule Five main steam line (MSL), two main feedwater line and one residual heat removal suction line weld flaw indications have been identified. All indications are inside the drywell, linear, and less than one inch in length. The licensee reported discovery of the MSL and RHR indications to the NRC via ENS. A linear indication on a standby liquid control system socket weld in the drywell was also identified. This indication was on a one and one half inch line which had previously been exempt from surface examinatio Since no prior surface examinations have been conducted in the areas, shese flaws may have existed since initial construction. The licarsee's engineering staff is evaluating the significance of the indications and possible repair options. The inspectors will continue to follow licensee ISI activitios and resolution of indica-tions (86-34-05).

On October 3, 1986 the licensee reported via ENS that two damaged hydraulic snubbers for the recirculation system had been identified during ISI inspection. The damage consisted of welders arc strike and tool mark Brief review of recent ISI inspection results indicates that at least seven piping supports with icose fasteners have been noted. These supports were last inspected during the 1984 outage. The number of loose support bolts developed in less than a full operating cycle suggests an underlying common problem. The licensee had recognized this as a potential generic problem and a management corrective action request had been initiated. The inspector will review resolution of this issue during a future inspection (86-34-06).

5.0 Review of LER's LER's submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further l

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information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LER's were reviewed:

LER N Event Date Report Date Subject 86-20 8/20/86 9/22/86 Unidentified fire barrier walls and penetrations 86-21 8/27/86 9/29/86 Standby gas treatment system single failure mode 86-22 8/29/86 9/29/86 Missed technical specification source leak check 86-23 9/12/86 10/14/86 Missed fire watch and fire watch patrnis

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LER 86-20 states that a previous event of a similar nature was reported in LER 84-007. A brief review of LER history by the inspector identified at least two additional instances, LERs 83-08 and 83-42, which also identified similar events. The generation of at least four LERs for a common problem indicates a program problem (see unresolved item 83-06-01). The licensee stated that this information would be included in an updated LER. The licensee does not currently have an adequate system in place to trend and crossreference LERs to identify recurring problems. An additional example of this was the licensee's inability to recognize the failure of MD-1001-36A valve in November 1985 as a recurring problem. An LER was not written even though at least two LERs had previously been written on the same valve for similar reasons (see section 2, IFI 85-28-03). This concern was discussed at the inspector's exit meetin .0 Rip tew of Selected Licensee Commitments In response to inspection report 50-293/81-22, and the subsequent imposition of civil penalty, the licensee committed to completion of a series of electrical walkdowns and procedure revisions. These walkdowns were directed at ensuring accurate information concerning 120 VAC and 125 VDC electrical distribution was included in station procedures. During inspection 50-293/85-14 the licensee stated that the walkdowns and procedure revisions would be completed by 1995. This date was determined to be unacceptable, and the licensee revised the completion date to September 30, 1986. During this inspection period the licensee informed l the inspector that the September 30, 1986 date would not be met and delayed this completion dat k A \ l q ,

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NRC inspection report 50-293/85-17, dated August 6, 1985, contained a notice of deviation. The deviation detailed the licensee's fa11bre to i-properly implement inservice test (IST) program commitments. By letter 3 dated September,4, 1985, the licensee transmitted the response to this deviation. The' response contained a commitment to complete the \

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corrective actions described by February 24, 1986. Subsequently, on four, N separate occasions, the licensee has found it necessary to postpone the '

commitment date for completion of these corrective action g S ,' > -

Extensive licensee commitment The impact of these efforts,owever,b ~

h,controi progr_amsipp,rovegants 1: not,9Et evident as demanstratedare underwa by these event ,w - ,( '

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Post Trip Review Program -

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l Based on the licensee's responses dated November 7;-1983 and' Augustg 13,

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1985, to Generic Letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events," NRR issued safety evaluations for the licensee's post trip review program descriptions / procedure and data /

( information capab,ility on September 11, 1985, and July 31, 1986,

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respectively. The ipspector reviewed procedure no.1.3.37, " Post Trip Reviews," and determined that the procedure was adequate and that tha ~

5 program was as described in the licensta's responses to' Generic._Lehtes r 83-28 and associated safety evaluation. The inspecior also reviewed'the s "

last two completed p'ost trip reviews and determined tnt'. implementation -

of the program appeared adequate.,

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7.0 Management Meetings

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,At periodic intervals during the course of the inspection period,

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imeetings were held with senior facility management to discuss the

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inspection scope and preliminary findings of the resident inspectors. No 'r written material was'given to the:1,1censee that was not previously '

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Attachment I to Inspection Report 50-293/86-34 Persons Contacted L. Oxsen, Vice President, Nuclear Operations --

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• A. Pederson, Nuclear Operations Manager

• K. Roberts, Director Outage Management D. Swanson, Nuclear Engineering Department Manager N. Brosee, Maintenance Section Head T. Sowdon, Radiological Section Head J. Seery, Technical Section Head E. Ziemianski, Management Services Section Head P. Mastrangelo, Chief Operating Engineer B. Eldridge, Acting Chief Radiological Engineer R. Sherry, Chief Maintenance Engineer J. McEachern, Resource Protection and Control Group Leader E. Graham, Compliance and Administrative Group Leader

• Senior licensee representative present at the exit meetin s

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