IR 05000293/1997004
| ML20217K876 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/07/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20217K867 | List: |
| References | |
| 50-293-97-04, 50-293-97-4, NUDOCS 9708180229 | |
| Download: ML20217K876 (19) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION l l
Docket No:
50 293 Report No:
97 04 i
Licensee:
Boston Edison Company
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Pilgrim Nuclear Power Station
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Location:
Plymouth, Massachusetts Dates:
May 5 9,1997; May 19 23,1997 Inspectors:
S. K. Chaudhary, Sr. Reactor Engineer, CMMEB R. Bhatia, Reactor Engineer, EEB D. Moy, Reactor Engineer, SEB J. Noggle, Sr. Radiation Specialist, RSB Approved by:
William J. Lazarus, Chief Civil, Mechanical and Materials Engineering Branch Division of Reactor Safety
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K 05 293 e
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EXECUTIVE SUMMARY This team inspection was conducted to determine the effectiveness of the licensee's efforts in maintaining the design basis of safety related systems the adequacy and acceptability of the design changes and modification to the core spray system; the readiness of the core spray system to fulfillits intended design function; the effectiveness of the licensee engineering controls and efforts in minimizing radiation exposure to workers; and general management controls to assure high quality engineering activities.
Core Sorav System The inspection team concluded that the current design configuration of the system was consistent with the design and licensing basis of the system, and the system readiness and operability was assured through effective surveillances and good functional checks.
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honram, Procedures and Manaaement Controls The inspection team concluded that the licensee's problem report and corrective action program effectively addressed the specific issues reviewed, including station wide management and support issues and concerns. The root cause analysis, safety evaluations, and operability determinations were of good quality, indicating appropriate management emphasis on safety.
Quality Assurance The team concluded that the audits and surveillance program established by the licensee were comprehensive and effective.
Radioloaical Review of Core Sorav Svstem Enoineerina The team ct.'cluded that the interdisciplinary effort between radiation protection (RP) and engineering during the design and development phase of the emergency core cooling system (ECCS) strainer replacement project was not effective. Engineering did not assure an adequate interface with RP during the design development stage to alert RP of the design concept, and RP did not actively solicit engineering infortnation during the development stage and waited for engineering to supply detailed Information. Therefore, during the ECCS strainer design, the multidiscipline team approach was not effective in providing the RP/ALARA group with the time needed to optimiza dose reduction measures =
resulting in higher than reasonably achievable radiation exposure levels. The civil / structural engineering interface with Radiation Protection /ALARA was very effective in administering the station's permanent shielding program. However, the system engineers were not aware of the doses associated with their systems, il
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TA8LE OF CONTENTS PAGE N0.
E1 Conduct of Engineering
........................................ 1 E1.1 Core Spray System Review................................ 1 E3
- Programs, Procedures and Management Controls...................... 5
E7 Qu ality A s s ur a nc e............................................ 7 E7.1 Audits and Curveillances
.................................. 7 E7.2 Strainer Replacement Project
............................... 7 E8 Miscellaneous Engineering issues (37551, 92903)..................... 8 E8.1 Review of the Bares for Technical Specifications Requirements for the Auxiliary Power Supply System
............................. 8 E8.2 Radiological Review of Core Spray System Engineering............. 8 E8.3 (Closed) Unresolved item 50 2 9 3/9 5 15 01.................... 12 E8.4 (Closed) Inspector Followup Item 50 293/96 10-01
.............. - 14 E8.5 (Closed) EA96-07 Violation 50 293/96-07 01
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r XI E xi t M e e t i ng............................................... 17 PARTIAL LIST OF PERSONS CONTACTED............................... 18
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Report Details
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E1 conduct of Engineering E1,1 Core Sorav System Review a.
inspection Scops The team reviewed the core spray system (CSS) design and its design basis documentation, such as the licensee's established system design drawings, calculations, analyses, and other engineering documents that are used to support system performance during normal and accident or abnormal conditions; and to assess the adequacy of the core spray system to fulfillits intended function, b.
Findinas and Observations Core Sorav Svstem Deslan The core spray system at Pilgrim Station, as described in the station Updated Final Safety Analysis Report (UFSAR), consists of two independent spray loops. Each loop is capable of supplying sufficient cooling water to the reactor vessel to cool the core in the event of a design basis loss of coolant accident (LOCA). The two spray loops are physically and electrically separated to preclude a single physical event rendering both loops inoperable. Each loop includes one pump, associated valves, and piping to route water from the suppression pool to the reactor vessel.
Deslan Basis Documentation Review The team's review of the design basis documentation indicated that the core spray system electrical and control design included the necessary sensors, relays, wiring, and valve operating mechanisms to initiate its normal and abnormal starting, operation, and testing of the system. System design was consistent with the UFSAR requirements.
The team verified the selected portion of the CSS components during a walkdown and found that the applicable cables and logic control circuits (from the sensors to control room panels) were appropriately routed up to the main control room panels via the cable spreading room. The review of the electrical power distribution design revealed that each core spray pump was powered from a separate safety-related ac bus that was capable of supplying standby ac power from the respective safety-related emergency diesel generator in the event of a loss of offsite power (LOOP).
The ao and de voltage and power supply requirements for the power and controls of core spray systems components such as ac motors, breakers, system valves and i
sensors, were also found adequately designed for automatic systems operation in normal and abnormal operating conditions.
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Overall, the CSS was fot.nd to be installed and operational consistent with the design requirements as described in the UFSAR.
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l Core Sorav System Related Modifications Reviewed i
The team reviewed the following design changes implemented on the CSS including the design input assumptions, methods of analyses, and safety evaluations:
PDC 9418D (minr***
Modification to the core spray system Valves MO 1400-25B (Inboard injection isolation valve) for G.L. 8910 (SE 2817 dated 4/6/94).
j PDC 96 07:
Installation of permanent vibration monitoring equipment on core spray motors P215A&B.
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Modification to the valves and motor operators of core spray system test valves MO 1400 4A & B (SE 2994 dated 6/27/96).
PDC 9610E (Minor):
Modificatien to miscellaneous system MOVs including the CSS system MOVs, MO 1400 3A,38,24A and 24B etc. This design change included the design to automatically close the valves on an isolation signal if these jog valves have been opened to less than 4%.
FRN95 02 89:
Replace thermal overload heater and relay in breaker B1843 for valve MO 1400 48.
PDC 95 29 (Minor);
Rewire valve breaker for "LS 8" modification on Valves MO 1400 4A/4B to enhance the valve position indication.
FRN95 02 82/105; Replace *SBM" control switches in various core spray valves.
PDC 96 08C Modification of MO 1400-4A and MO 1400-48.
PDC 96 20 Modification to trend core spray leakage.
PDC 96 32 Residual heat removal & core spray pump suction strainer replace.
PDC 9418D Modification to M01400 258 for GL 8910.
PDC 96-07 Installation of vibration sensors on RITR and core spray pump *
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Core Sorav System Related Calculations Reviewed The team reviewed the following design calculations to assure that system changes were appropriately being evaluated to reflect the operability of the CSS under
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normal and abnormal operating conoitions:
PS 79, Rev. 4, dated 11/19/96...EDO loading calculation, i
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PS 133, Rev. 2, dated 5/23/95... Stroke time and voltage concerns of various MOVs.
PS 135, Rev. O, dated 8/13/96... Electrical performance and stroke time evaluation of priority 5 ac MOVs.
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PS 141, Rev.1, dated 10/9/96... Thermal overload heater sizing for MOVs.
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PS 152, Rev. O, dated 6/17/94...AC MCC voltages per G.L. 8910 MOV Evaluation.
N110, Aev. O, dated 1/21/91, "OPL 4 (ECCS parameters) for SAFER /GESTR LOCA analysis of PNPS.
.G.qLe Sorav Svstem Related Loalc and Other Systems Loalc Functional Teg Proceduresand Surveillance Tests Reviewed The team reviewed the CSS system post modification test proceduies and surveillance test procedures to assure their adequacy to meet the administrative control and system design basis acceptance requirements.
The review of completed surveillance test procedures included the core spray system "A" loop functional tests, drywell high pressure auto initiation functional test, an automatic start of core spray pump "A" logic system, core spray system
valves functional tests, and emergency safety bus load shedding and safety loading sequencing functional tests, as required by the Technical Specification (TS) and design basis requirements. The review indicated that the acceptance criteria were
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well defined, and were consistent with the system design basis and operational requirements.
The iRm noted that the licensee had reviewed the logic system functional test (LGF1) procedures associated with the CSS and automatic depressurization system (ADS)in response to NRC Generic Letter 96 01 to assure that safety system LSFTs were adequate. The team also independently reviewed the CSS LSFT procedures and noted that these procedures had appropriately included the functional testing
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requirements of all system components including the analog, relay contacts, and
Interlocks devices, as required by the design basis documentation.
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4 The team determined that the post modification and surveillance test procedures demonstrated the adeqt:acy of the CSS and other supporting systems such as emergency diesel generator component functionality and operability. The team concluded that the licensee hcd a comprehensive program in place to assure the functionality of the CSS.
Based on the above review, the team determined that the core spray system was I
functional and was installed and operated in a manner consistent with the design
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I bases specified in the licensing documents. The operability of the system was assured through the post modification, surveillance, and functional checks. The
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design changes performed did not affect the required design requirements and overall function and performance of the CSS. The supporting design cnalyses and mechanical and electrical calculations performed associated with these modifications were found adequate to justify the specified design changes of this system.
Core Sorav Inlection Valve Review The team noted that the closing time design value for core spray injection valve (1400 25A/B normally closed inboard value) was indicated as 20 seconds in the UFSAR licensing basis. In 1993, Boston Edison completed a calculation (N110, Rev. O, "OPL 4 ECCS parameters for SAFER /GE STR LOCA Analysis of PNPS) to increase the closing time from 20 to 22 seconds to gain more closing time margin for those valves. The inspector verified that the UFSAR was updated to indicate this change in the 1994 revision. However, this maximum closing time was shifted back to 20 seconds in the next revision of the UFSAR for an unknown reason. At the time of this inspection, the UFSAR indicated the licensing basis to be 20 seconds, however, it was not supported by the General Electric's LOCA (NEDC-31852P, power station SAFER /GESTRA LOCA Analysis."
The team also reviewed Safety Evaluation 2730 for the outboard isolation injection valves. In addition to normally cissed core spray isolation valve, the safety evalaation described the normally open (outboard) valves. However, it did not document the bases for changing the UFSAR value from 18 to 22 seconds. By increasing the closing, time from 18 to 22 seconds, the valve will respond slower and therefore be nonconservativ.
The team reviewed the surveillance test and procedures for past several years for both of these inboard and outboard valves and determined that the administrative limits on these valves were set at 16 and 18 seconds, respectively. The team concluded that there were no operational safety concerns regarding the closing time of these inboard and outboard valvea, either currently or in the past.
All analyses were technically acceptable. Na safety concern or operability issues related to the LOCA or containment analyses was identified by the tea __-_-__-__- - -- - -
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The licensee has identified these discrepancies and intends to correct the error during the next UFSAR revision (June 1997). The team verified that the LOCA analysis was based on the 22 seconds closing time for the core spray injection valve (1400-25A/B).
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Conclusion The team concluded that the current design configuration of the core spray system was consistent with the design basis and the licensing conditions. Despite the l
discrepancies identified in the UFSAR descriptions, the operability of the system
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was assured through effective surveillance and appropriate functional checks.
I E3 Programs, Procedures and Management Controls
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Insoection Scone I
l The team reviewed the licensee's program and procedures for identification, documentation, and resolution of problems / concerns to determine the adequacy and effectiveness of: 1) the problem reporting program; 2) safety evaluations with associated engineering analyses; 3) root cause analyses; and 4) operability determinations, b.
Findinas and Observation The licensee's problem reporting program is described in procedure No.1.3.121
" PROBLEM REPORT PROGRAM," Rev. O. The team noted that the problem report procedure was clear and detailed and established an effective identification, documentation, and resolution program. The program covered plant operation, ma'ntenance, c,ngineering, and plant support, including establishing detailed requirements for root cause analyses and evaluations.
Tlie tsam reviewed several problem reports of various safety related systems to aesure that the licensee's response to conditions adverse to quality was appropriate, and that recurring component failure issues and root cause analysis of such component failures was being appropriately performed to re solve the issue.
The team reviawed the licensee's corrective actions regarding approximately forty problem reports associated with electrical and control system related issues.
Approximately twenty additional reports of various safety related systems were also reviewed. Overall, the team found them acceptable.
The review of root cause evaluations was performed in conjunction with the review of the problem reports. The team noted that the root cause determinations were of good quality, they contained thornugh technical analyses, and engineering judgements to identify the root cause.
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The team reviewed approximately eighteen safety evaluations (10 CFR 50.59.a) to
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assess the adequacy and technical validity of the evaluations. The team observed that the evaluations included detailed descriptions of the safety significance of the
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problem; adequate technical analyses; and were supported by engineering analyses, calculations, and judgements.
The team reviewed approximately twenty operability determinations to assess the adequacy of the evaluations. The team observed that the operability evaluations
were satisfactory. They were supported by adequate analyses, evaluations, and
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engineering judgements based on valid technical premises. The detailin each evaluation was commensurate with the safety significance of the affected system, and the complexity of the identified problem, c.
ConclusioD Based on the problem reports reviewed, the team concluded that the licensee was effectively addressing plant emergent issues. In addition, the root cause analyses reviewed were good. The safety evaluations and operability deterrninations, reviewed were supported by appropriate engineering analyses, calculations, and judgements.
E7 Quality Assurance E7.1 Audits and Surveillances
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'he team reviewed the quality assurance (QA) audits and surveillances to assess
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the adequacy of the program, and the validity of observations and findings.
The audits and surveillances were of high quality and included in-depth assessments of technical content of the engineering efforts. QA audit 95 07 of the core spray system was a good example of the depth and breadth of the audit scope. The findings of this audit were consistent with the team's observation of the status of the core spray system design and functionality. The team determined that adequate technical, administrative, and management resources were applied to the audit and surveillance program. The team considered the audit and surveillance program effective.
E7.2 Strainer Replacement Prolect The team reviewed the engineering efforts and the project management of the ECCS strainer replacement project. Strong management focus was evident in the
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engineering and QA area, however, the torus strainer enhancement and replacement project revealed an ineffective interface between engineering, installation personnel, and the radiation protection group. The engineering output documents did not provide adequate and clear information to support the installation, nor was radiation exposure to workers adequately considered. The radiological group did not
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aggressively follow up the fast evolving engineering design, and the installation / construction group failed to identify and resolve the unclear Infonnation regarding design tolerances during construction. The above weak interfaces amongst these organizations resulted in unnecessary radiation exposure to workers due to rework.
E8 Miscellaneous Engineering lesues (37651,92903)
E8.1 Review of the Bases for Technical Spaglficelons (TSI Reautrements for the Auxillary Pgwer Sunolv System The team reviewed the auxillary power supply requirements and inoperable equipment conditions in the TS. TS Section 3.9.B.2 allows the reactor to operate up to 25% power without electrical power available from both startup and shutdown transformers; providt.d, i:oth diesel generaters and associated emergency buses are available. However, the TS basis section of Auxiliary Electrical System, Section 3.9, does not clearly address this situation. A review of the FSAR revealed no additional Information regarding these operating conditions. The licensee
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indicated that they planned to revise the basis section to eliminate this ambiguity.
The adequacy of this TS is still under review by NRC. (lFl 97 04 01)
E8.2 Radioloalcal Review of Core Sorav Svetem Enoineerino a.
Scone (37550)
The inspector reviewed the radiological environment for maintenance of the core spray equipment and applicable engineering shielding evaluations. The installation of the ECCS strainers during the spring 1997 outage was reviewed with respect to radiological engineering performance.
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Observations and Findinas The core spray pumps and principal valves are located in the reactor building basement corner rooms and are in close proximity to the RHR system piping and components. RHR heat exchanger and piping generate high radiation areas for the core spray equipment areas. The civil / structural engineering group has interfaced effectively with the ALARA/RP group to begin improving the radiological environment. Permanent shielding has been installed on half of the RHR piping in the RHR corner rooms and on the bottom head of the RHR heat exchangers.
Average corner room dose rates were reduced from approximately 50 mR/hr to.
40 mR/hr. Direct pipe loading for shielding application was maximized, while half of
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the piping in the room will require a structure loaded shielding approach. Tnc inspectors reviewed the following engineering packages and referenced calculations.
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FRN 90 0174 RHR pump discharge and suction piping and RHR piping shielding FRN 97 01-05 RHH heat exchanger bottom head shielding PDC 97 05 Permanent shielding in drywell PDC 96 07 Vibration sensors for RHR and core spray pumps j
i The reviewed engineering packages were found to be complete and thorough
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engineering evaluations. PDC 97 05 provided an effective evaluation and l
commercial dedication of permanent lead blanket materials and provided appropriate l
quality control requirements for procurement purposes. The inspectors noted the installation of remote vibration and temperature monitoring systems for the RHR and
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core spray pumps to allow for performance monitoring for prodletive maintenance.
This modification was designed to reduce exposure while collecting equipment performance data.
ECCS Strainer Modification NRC issues /Backaround There were several related regulatory design basis considerations that impacted ECCS strainer changes, in May 1996, the NRC issued Bulletin 90 03, which required BWR licensees to address an ECCS strainer plugging concern with a written response within 180 days and to make any necessary plant modifications during the next refuo'ing outage. Due to industry design criteria development, BWR licensees requested deferrals. Pilgrim Station's licensing basis did not take credit for the expected overpressure conditions during a design basis accident that would aid in assuring adequate net positive suction head was maintained for ECCS system torus water supply. NRR determined that this issue would have to be resolved prior to restart from the spring 1997 outage as this represented an unreviewed safety question. Additionally, Pilgrim Station's ultimate heat sink temperature was based on 65 degrees F. However, during the summer season, intake water temperatures are likely to exceed this value. The licensee submitted a November 1,1996 response to IE Bulletin 96-03, indicating that they intended to install the larger suction strainers. The large beam stralner assemblies were installed during the spring 1997 outage.
ECCS Strainer Prolect Timeline Thc engineering organization began working on a new ECCS strainer design in January 1996. By April 1996, a Torus suction strainer design was being pursued that was structurally supported by the existing Torus suction nozzles. An initial exposure estimate of 17 person-rem for installation of the modification was established. in May 1996, NRC IE Bulletin 96-03 was issued requiring all BWR licensees to address Torus / suppression pool strainer fouling, with an expectation that the required plant modifications would be completed during the next scheduled outage. The BWR licensees appealed to the NRC for time extensions due to continued development of insulation transport mechanism analyses and prototype strainer testing research that was ongoing. By September 1996, the engineering department determined that a nozzle-supported strainer design might not be larga
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enough to accommodato expected strainer fouling and switched to a large stralner beam design that spans a Torus bay and would be supported by the existing Torus ring girders, in October 1996, engineering contract services were obtained, and by mid November 1996, the basic strainer conceptual design was completed. This included a three piece beam design with flanges and pipe connections that required in field welding; included an end bracket design with reinforcement of the ring girders; and included parallel strainer installation by two crews instead of sequential strainer installation by one crew. In mid November 1996, a commitment was made
to go forward with the stralner installation during the spring 1997 outage. At that time, the new design was not toevaluated for radiation exposure impact and the original 17 person-tem estimate was still being carried forward. A
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construction / implementation team was formed that included an ALARA specialist that met frequently to review construction details es they were provided to construction by engineering. On January 17,1997, the PDC package details were completed and this package contained a completed Radiological Technical Support
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l Division (RTSD) sign off by the System and Safety Analysis (S&SA) group, which required that an ALARA design review be completed. On January 27,1997, the construction / implementation team provided detailed person hour estimates to the ALARA proup. On February 6,1997, the ALARA design review was completed indicating 21,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of work, which represented approximately 7 times more
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l hours than originally estimated during the summer of 1996. The exposure estimate increased from 17 person-rem to 100 person-rem less than 3 weeks before the outage. The actual exposure cost of the modification resulted in approximately 95 person rem.
ALARA measures that were incorporated into the engineering design included:
torus de-sludging (previously scheduled for Torus coating inspection), minimization of welding, and the use of bolted componeno. The RP/ALARA group reviewed radiation surveys and determined that significant shielding would not be practicalin the time allotted and instead focused on work force occupancy economizing factors, principally utilizing mockup training to improve performance inside the Torus.
The inspector was informed by the radiation protection manager (RPM) that the excessively high drywell dose rates was a priority and attention we directed to solving that problem. The inspector was also informed that the three ALARA specialists were each involved in five different construction / implementation teams for the outage. The strainer modification project was an atypical accelerated engineering installation that required a team approach on parallel path scheduling of engineering design, material procurement, and construction activities to meet a very short schedule. The parallel approach did not, however, include RP/ALARA during the engineering design phase and did not provide sufficient lead time to adequately evaluate the exposure reduction techniques for this project. For the ECCS strainer modification, reasonable planning time was not made available to allow for a review and determination of Torus work area s7urce terms and for a comprehensive evaluation of dose reduction options (e.g., RHR system chemical decontamination, RHR piping shielding, use of miniature submarine robot, etc). The RPM indicated that more dose could have been saved, given more tim _____ - - _ _ - _ - _ _ _ _ _ _ _ _ _
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fnaineerina Pronram Processes As was stated earlier, the strainer installation was an atypical parallel path engineering / construction task with a very aggressive schedule. Although the
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primary objective of installing the strainers was achieved, the multi-discipline team approach was not effective in providing the time needed to optimize dose reduction L
measures and resulted in high exposure costs.
The engineering processes were reviewed to determine whether effective program elements are in place to be successful in handling normal sequential project development (i.e., engineering, uonstruction, Ro/ALARA). The inspector determined that the engineering program includes a Design Criteria loput document that can
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involve outside engineering interests early in the engineering design stage. A design L
criteria input document was developed for the ECCS strainer project and a copy was provided to RP/ALARA in mid-December 1996, however, neither engineering nor RP/ALARA made effective use of this provision. Later in the engineering design
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phase, an ALARA Design Review may be specified by the S&SA group which requires engineering to provide a detailed design review against a standard !ist of
construction exposure reduction techniques. This ALARA Design Review is
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effective for new construction designs to minimize radiation effects during later maintenance of operation activities, however, this program element does noi address radiation source term reduction, or other dose rate reduction techniques such as robotic or remote construction techniques, or shielding during modification installations. For the ECCS strainer project, the ALARA Design Review was completed on February 6,1997, and did not provide any significant contribution to exposure reduction for the project. In review, the engineering process does provide for design input erly in the engineering design stage which, if properly utilized, could provide for effective involvement of the RP/ALARA group.
The inspector reviewed 0V n /SD reviews perf tamed by the S&SA group and the RP/ALARA involvement a d,e engineering desigr. process. Approximetely 3 years ago, 3 engineers were transferred from the RP Department's RTSD to the S&SA group. While working for RP, thcse engineers provided ALARA Design Reviews for the engineering department utilizing the resources available from the RP/ALARA organization. 'nce these engineers joined the S&SA group, their new responsibilities include reviewing engineering packages for the following considerations: fire protection, emergency preparedness, security, technical specification /UFSAR requi iments, and ALARA design reviews. Since leaving the RP department, S&SA ha been providing the RTSD signoff reviews although they no longer draw on the resources of the RP/ALARA group. Since the engineers were transferred from RP, the RP/ALARA group has attempted to compensate by becoming involved earlier in the process, specifically by becoming involved in construction implementation planning meetings. The result of these developments has been S&SA engineering providing the ALARA design reviews without sufficient RP input, and RP/ALARA becoming involved only after the engineering design phase has been completed.
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Conclusions The civil / structural engineering interface with RP/ALARA has been very etfactive in the station's permanent shielding program. However, the interdisciplinary effort between RP and engineering during the design and development phase of the ECCS strainer replacement project was not effective. Engineering did not provide an adequate interface with RP during the design development stage to alert RP of the design concept, and RP did not actively solicit engineering information during the development stage. Thus, during the ECCS strainer design, the multi-discipline
team approach was not effective in providing the RP/ALARA group with the time needed to optimize dose reduction measures resulting in high exposure costs.
The team noted that the licensee did not adequately determine the external Torus piping radiation sources and did not provide a comprehensive evaluation of exposure reduction solutions (e.g., RHR system chemical decontamination, permanent shielding, or robotic submarine video surveillance, etc); and the RP/ALARA group did not supply sufficient staff resources to adequately address the demands of the outago that included excessive drywell dose rates and the ECCS strainer modification. The RP/ALARA group also did not become involved in the engineering design process when alerted to the new strainer design by the design criteria input review.
Eemcering continued to provide radiological technical support input, although they hM E :t the RP department several years previously. RP/ALARA was involved at it, umstruction planning level, but not during the engineering design stage. This
.:A ted in a gap in the engineering /RP/ALARA interface.
System engineers were not aware of the dose associated with their systems and not accountable for this parameter.
The Vice President station director indicated that a more multidiscipline engineering approach was needed and would be evaluated. The routine engineering inspection program will follow these developments.
E8.3 (Closed' Unresolved item 50-293/95-15-01: Valve wiring 06screpancies.
In the 1993 NRC inspection 50 293/9315, three significant field wiring deficiencies were identified by the licensee in safety-related motor-operated valves (MOVs). The field wiring problems were appropriately corrected at that time. However, the NRC left this issue open as a general plant wiring concern pending further corrective actions; i.e., completion of the root cause analysis; a design evaluation of other MOVs; and other appropriate corrective actions applicable to address the wiring concern in the station in general.
To address the above general wiring discrepancy concern, the licensee reviewed all safety-related MOV design documentation to evaluate similar wiring problems (seal-in valve circuit as-found in the open condition), but found no similar wiring problems in the core spray and the reactor water cleanup system,.
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The review of the root cause analyses of the above three MOV wiring discrepancies indicated that they were caused by personnel errors during the installation of modification (PDC 92 39)in 1993. To assure that no other concerns existed in other MOV wiring, the licensee also reviewed the applicable maintenance work requests associated with PDC 92-39, and found no discrepancies.
The tearn noted that the licensee had also reviewed the wiring discrepancies in problem reports dating back to 1989. The licensee's evaluation determined that most of the wiring, despite the discrepancies, was still electrically equivalent to the design configuration. The licensee noted that the other common wiring errors in MOV operators and in the MCC cubicle wiring occurred during the implementation of the plant modifications or plant maintenance work orders.
The team noted that the licensee had taken the following additional corrective actions to address the genere! design documentation, testing and other concerns associated with this issue:
Ve.-ified approximately 33% of MOV MCC wiring in various safety-related
systems and found no additional wiring concerns.
- Prepared a new internal wiring diagram for all safety-related ac and de MCC cubicle wiring diagram of valves tc reflect an individual wiring diagram in cases where a composite drawing existed to prevent recurrence of a future wiring interpretation error.
- Revised the above three unique MOV wiring configurations in the CSS and reactor water cleanup system and implemented it (PDCR 95 29) in 1997 refueling outage to eliminate the possibikty of similar wiring and testing concerns on these MOVs. The new design uses a common control conductor for the operation of manual and automatic closure of these MOVs (seal-in circuit).
- Adequate training was provided to tte engineering and the maintenance staff to preclude such wicing errors in the future.
- la addition, the 480 Vac motor control center testing and maintenance Procedure No. 8.Q.3-3 was revised on January 5,1997, for verification and visual inspection of wiring configuration in other safety-related MOV MCCs during the routine maintenance testing to address this generalissue.
The team reviewed the selected sample of design documentation and verified by inspection of the wiring of ten safety-related MOV circuits in MCC cubicles area and two control room panel control circuits fcr similar wiring discrepancies and identified no concerns.
The team concluded that the licensee nas adequately addressed this issue. This item is closed.
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E8.4 (Closed) Insoector Followuo-item 50 293/96 10-01: Pertaining to the Agastat relays qualified life and the replacement of normally-energized relays in the station.
During the routine surveillance and calibration activities in January 1996, the licensee observed an indication of overheating in the normally energized Agastat model GP relays in the reactor protective systum panel. The relay coil cores (bobbin) were found cracking and brittle. At that time, the licensee staff had appropriately initiated the problem reports (prs) to address this emergent problem.
Since these relays were re-qualified and their service life was extended from 10 to 22 years by the licensee's internal evaluation, the NRC inspector questioned the adequacy of this extension, and also other general concerns with the licensee's preventive maintenance program for this type of relay.
In the review of problem report PR 92 0627, issued on August 8,1993, the team determined that the licensee extended the service life of the Agastat GP relays from the manufacturer-recommended 10 years to 22 years. This change was made on the basis of the material analysis performed by Wyle Laboratories, and was documented in their Qualification and Verification Report No. 48687-REL l-0. The Wyle report determined that the Agastat GP relay expected life would be about 22 years for continuously normally energized relays, and 40 years for those relays that were normally-energized for only 1% of the time, respectively. Based on this analysis, the licensee concluded that Pilgrim Station relays (since the relays were installed in 1985) would not reach the end-of-life until about the year 2007, because no indication of any end of-life failure had occurred through 1992. At that time, the licensee decided to re-evaluate the performance of these relays in 1998, when additional plant data would become available.
In December 1996, a problem report PR96,9640 was issued. The licensee found that several relay coil cores (bobbin) in the normally energized relays of the reactor protection system were showing a sign of cracking and brittleness. These relays were being replaced as a result of their ongoing Agastat GP s, ries replacement program work to ensure the service life of 22 years is not exceeded. The licensee determined the root cause of this problem to be an elevated relay core temperature (approximately 199*F) in the normally energized condition. Based on the engineering evaluation performed on these relays and the as-found material condition, the licensee determined that this condition was of no immediate safety concern, because, the primary function of the bobbin material (Zytel) was used only as an insulator. Therefore, these relays were considered to be functional in this degraded condition. The licensee also reviewed the station failure database and found no functional failures related to the aging of Agastat relays.
The team noted that the licensee's recent re-evaluanon concluded that the service life of 7 years was conservative. This was based v. the station experience data available of as much as 15 years of continuously energized service condition and with no reported or detectacle anomalies (as documented in the PR 96.9640 dated May 23,1997). The team concluded that the licensee's re-evaluation appropriately L
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concluded the new limit on the service life was 7 to 10 years for all normally-energized safety related and critical to safety applications Agastat GP relays. The licensee believes this is a conservative approach even though plant relays have indicated longer than this performance life.
The inspector's review of the above documentation revealed no concerns. A visual inspection of several selected relays in the core spray system logic and other panels were found satisf actory. Per discussion with the licensee, the team determined that with the exception of two relays (16A-K18X6 and 16-K18X2), all safety-related and critical-to-safety relays have been replaced to satisfy the revised 7 years end-of-service guidelines. The licensee concluded that these two relays, even though they have exceeded the current conservatively established guidelines, would not reach at the end-of-service life prior to their replacement by October 31,1997, because the visual inspection of these relays performed by the licensee indicates no adverse signs and degradation. Also, they are located in a mild environment (control room),
and the similar relays performance test data have indicated no functional failure.
Based on the above, the inspector concluded that the licensee's corrective actions were appropriate. The replacement of all safety-related Agastat model GP relays before the end-of service life of 7 to 10 years in the energized epplication was appropriate. Since all the safety-related and critical-to-safety station relays had been replaced at this time with the exception of two, and these relays were being appropriately tracked for replacement by October 31,1997, as documented in MR19700064 and 76, the inspector concluded that the licensee had adequately resolved the service life and functionality issue of the Agastat model GP relays.
This issue is closed.
E8.5 (Closed) EA96-07 Violation 50-293/96-07-01: A lack of adequate protection settings were found on several of electrical penetration circuit breakers.
On October 21,1996, NRC issued a Notice of Violation (NOV) to the licensee for failure to mairitain primary containment integrity of electrical penetrations. Two electrical penetrations were not properly protected due to improper trip-settings of 12 electrical penetration breakers, in addition, the NRC was concerned that proper measures were not established in this case to assure conditions adverse to quality were promptly identified and corrected because this issue was identified by the licensee engineering in 1991, but was not adequately addressed in a timely manner.
The licensee, in their response to the NRC dated November 20,1996, agreed with this issue and stated that the following immediate and long-term corrective actions were being taken to address this NOV concern:
Corrective action taken to achieve immediate comoliance:
Upon its discovery, the senior shift supervisor was notified and the plant entered a Limiting Condition of Operation (LCO) as per the TS requirements established procedur m _.._
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The affected breakers trip setting penetrations were promptly changed and
breaker circuits were opened to protect the integrity of the penetrations.
A visual inspection was performed of the penetrations to assure their
integrity.
- Later, all the applicable circuit breakers were replaced with new circuit
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breakers (by April 23,1996) with better thermal-magnetic design feature breakers for better penetration protection.
- The Engineering Design Guide Standards, EB19 and EB20, were reviewed and updated to assure proper design considerations.
Engineering reviewed all applicable containment electrical penetration
calculations for similar design concerns and found no additional issues.
These calculations (PS 119 and PS-124) were updated as per the established procedures to reflect the revised breaker design changes.
Lona term corrective actions taken to avoid recurrence:
Nuclear engineering service group revised the design calculation, Procedure 3.05, on December 20,1996, to include the instructions that if a deficient condition is identified in the calculation, a problem report shall be initiated to track this issue.
- Nuclear organization reviewed and updated the long term plan Procedure, NOP 89A1, on December 30,1996, to include the additional guidance to improve the staff actions from a human factors perspective to reduce the chance of recurrence of similar issues.
- A special training session was held for engineering department staff personnel to emphasize the management expectations in regard to the timeliness of corrective actions, engineering procedures changes, and technical and administrative issues associated with the cited violation.
- In addition, formal special training was provided to staff personnel under the engineering and support personnel training program activities to broaden the awareness of the violation, the reasons for the violation, and the corrective actions taken and planned to avoid recurrence.
The team considered the above corrective actions acceptable. The team verified the corrective actions by staff interviews, review of the random applicable documentation, and a visual inspection of the selected circuit breakers components in the station. Based on the above corrective actions completed by the licensee, the team concluded that this item is close ~ - -.. -
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i V. Mananoment Meetinas XI Exit Meeting The inspection team met with the licensee representatives at the conclusion of the inspection on May 23,1997. The team loader summarized the scope of the inspection and j
inspection results at this meeting and the licensee acknowledged the inspection findings.
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PARTIAL LIST OF PERSONS CONTACTED
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Boston Edison Company a
i Kristin DiCroce, Sr. Regulatory Affairs, Engineering
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H. V. Oheime, General Manager - Technical Section William R. Kline, Group Manager - Nuclear Engineering Services
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i Leon J. Olivier, Vice President - Nuclear Organization
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R. Sheridon, inspection Team Leader - OC
Bruce Chenard, Department Manager Electrical Engineering Susan Landahl, Radiation Protection Manager T. A. Venkataraman, General Manager - QA F. J. Mogolesko, Project Mane.ger - NESG Nancy L. Desmond, Group Manager - Regulatory Relations C. S. Godard, Manager - NSG
J. J. McClellan, Sr. QA Engineer J. Keene, Department Manager - Regulatory Affairs R. O'Neil, Department Manager - Station Services J. Gerety, Deputy Group Manager - NESG l
Swapan Das, Sr. Electrical Engineer U. S. Nuclear Reaulatorv Commission
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R. Laura, Sr. Resident inspector i
B. Korona, Resident Inspector
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