IR 05000440/2003002
| ML030660492 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/07/2003 |
| From: | Dave Hills NRC/RGN-III/DRS/MEB |
| To: | Kanda W FirstEnergy Nuclear Operating Co |
| References | |
| IR-03-002 | |
| Download: ML030660492 (30) | |
Text
March 7, 2003
SUBJECT:
PERRY NUCLEAR POWER PLANT NRC INSPECTION REPORT 50-440/03-02
Dear Mr. Kanda:
On February 14, 2003, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Perry Nuclear Power Plant. The enclosed report documents the inspection findings which were discussed on February 14, 2003, with you and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The NRC inspection team reviewed selected procedures and records, observed activities, and interviewed personnel. Specifically, the inspection focused on permanent plant modifications and changes, tests, and experiments evaluated in accordance with 10 CFR 50.59 requirements.
Based on this inspection, the team identified a Severity Level IV violation of NRC requirements associated with two examples of failure to perform a review as required by 10 CFR 50.59.
Specifically, your staff failed to complete a documented safety evaluation for a change to the facility as described in your Updated Final Safety Analysis Report that involved: 1) the incorporation of new electrical standards affecting battery maintenance and acceptance criteria; and 2) changes to a plant drawing and procedure which reduced electrical separation criteria.
We determined that this issue is a violation of 10 CFR 50.59 requirements. Because the violation was non-willful and non-repetitive and because it has been entered into your corrective action program, the NRC is treating this issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRC's Enforcement Policy.
Additionally, two findings of very low safety significance (Green) were identified which involved violations of NRC requirements. However, because of their very low safety significance and because they have been entered into your corrective action program, the NRC is treating these issues as Non-Cited Violations, in accordance with Section VI.A.1 of the NRC Enforcement Policy. If you contest the Non-Cited Violations, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Perry Nuclear Power Plant.
In accordance with 10 CFR Part 2.790 of the NRCs "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
David E. Hills, Chief Mechanical Engineering Branch Division of Reactor Safety Docket No. 50-440 License No. NPF-58 Enclosure:
NRC Inspection Report 50-440/03-02(DRS)
cc w/encl:
B. Saunders, President - FENOC K. Ostrowski, Director, Nuclear Maintenance Department V. Higaki, Manager, Regulatory Affairs J. Messina, Director, Nuclear Services Department T. Lentz, Director, Nuclear Engineering Department T. Rausch, Plant Manager, Nuclear Power Plant Department Public Utilities Commission of Ohio Ohio State Liaison Officer R. Owen, Ohio Department of Health
SUMMARY OF FINDINGS
IR 05000440-03-02; First Energy Nuclear Operating Company; on 02/10-14/2003; Perry
Nuclear Power Plant. Permanent plant modification and changes, tests, and experiments.
The baseline inspection was conducted by resident and region-based inspectors. The inspectors identified two Green findings associated with Non-Cited Violations and one Severity Level IV Non-Cited Violation (NCV) associated with a failure to perform 10 CFR 50.59 safety evaluations.
The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3, dated July 2000, and can be found on the NRC website at: http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/index.html A.
Inspector Identified Findings
Cornerstone: Mitigating Systems
- NCV. The team identified a Severity Level IV Non-Cited Violation associated with the licensees failure to perform safety evaluations in accordance with 10 CFR 50.59 for changes made to the facility as described in the Updated Final Safety Analysis Report.
Specifically, the licensee failed to complete a documented safety evaluation for a change to the facility as described in the Updated Final Safety Analysis Report that involved: (1) the incorporation of new electrical standards affecting battery maintenance and acceptance criteria; and (2) changes to a plant drawing and procedure which reduced electrical separation criteria.
Because the Significance Determination Process (SDP) is not designed to assess the significance of violations that potentially impact or impede the regulatory process, this issue was dispositioned using the traditional enforcement process in accordance with Section IV of the NRC Enforcement Policy. However, the results of the violation, that is, the failure to evaluate changes in battery maintenance and acceptance criteria and electrical separation criteria, were assessed using the SDP.
The team considered this issue of more than minor significance, because if left uncorrected, the finding could become a more significant safety concern. The team was concerned that two examples of failure to perform a documented safety evaluation in the relatively small sample of screenings reviewed indicated a potential for a more significant safety concern if the underlying cause was not identified and corrected. For the examples identified, the team determined that the issue was of very low safety significance because: (1) other licensees had been successful in evaluating changes under 10 CFR 50.59 from the older IEEE 450 requirements to compliance with the newer IEEE 450-1995 requirements; and (2) the reduced electrical separation distance that the licensee had incorporated into their design drawing and configuration met the requirements in IEEE 384 which had been endorsed by the NRC in Regulatory Guide 1.75. Therefore, the results of the violation were determined to be of very low safety significance and the violation of 10 CFR 50.59 was classified as a Severity Level IV violation.
Because this non-willful violation was non-repetitive, and was captured in the licensees corrective action program, this issue is being treated as a Non-Cited Violation, consistent with the NRC Enforcement Policy (Section 1R02).
- Green.
The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, for the for the licensees inadequate design reviews associated with the installation of half-couplings on a B train 14 inch emergency service water elbow. The licensee installed half-couplings in response to a through-wall leak and an area of wall loss identified on a 14 inch emergency service water elbow.
However, the licensees design review was inadequate in that, it failed to include the requirements of Section XI of the American Society of Mechanical Engineers Code.
Specifically, the licensee failed to identify the cause of the flaw, failed to adequately characterize the dimensions of the flaw, nor was the potential growth of these flaws considered. Further, the repair design did not include flaw removal or component replacement.
This finding was more than minor based on the degradation in the plants design basis for the B train emergency service water elbow. This degraded elbow affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of the emergency service water system, because the non-Code repairs (half-couplings) and associated piping flaws could have resulted in premature failure of the elbow. The finding was determined to be a licensee performance deficiency of very low safety significance (Green) by the significance determination process because the finding was a design or qualification deficiency that did not result in a loss of system function (Section 1R17).
- Green.
The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, for the licensees inadequate design review associated with installation of a rupture disc in the exhaust piping of the division 3 high pressure core spray emergency diesel generator. This finding was self-revealed on October 25, 2000, after the diesel generator was placed in service following this modification, the rupture disc failed in less than 3 minutes due to pressure induced fatigue. The licensees design review for the rupture disc was inadequate because it did not adequately consider pressure induced fatigue loading.
This finding was more than minor based on the failed rupture disc impacting the mitigating systems objective of ensuring the availability, reliability, and capability of the division 3 high pressure core spray diesel generator in response to initiating events.
Specifically, the failed rupture disc created an opening which could have allowed foreign material to enter the exhaust system and cause premature failure of the division 3 diesel generator. The finding was determined to be a licensee performance deficiency of very low safety significance (Green) by the significance determination process because the finding was a design or qualification deficiency that did not result in a loss of system function (Section 1R17).
B.
Licensee Identified Findings No findings of significance were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments
Review of Evaluations and Screenings for Changes, Tests, or Experiments
a. Inspection Scope
From February 10, 2003, through February 14, 2003, the team performed an on-site (room 205 of the site administrative building) review of 10 safety evaluations required by 10 CFR 50.59 and 14 safety evaluation screenings or regulatory applicability determinations (RADs). The team reviewed these documents to ensure compliance with the requirements of 10 CFR 50.59. The team also referred to Nuclear Energy Institute 96-07, Guidelines of 50.59 Evaluations, Revision 1, to determine acceptability of the completed evaluations, and screenings. The NRC endorsed the Nuclear Energy Institute 96-07 in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, November 2000. The team also consulted Inspection Manual, Part 9900, 10 CFR Guidance: 50.59. The documents reviewed during this inspection are listed at the end of the report.
b. Findings
b.1 Failure to Perform 10 CFR 50.59 Safety Evaluations Introduction The team identified a Severity Level IV Non-Cited Violation associated with the licensees failure to perform safety evaluations in accordance with 10 CFR 50.59 for changes made to the facility as described in the Updated Final Safety Analysis Report (UFSAR).
Description On January 22, 2002, the licensee completed RAD 02-00055. In this RAD, the licensee evaluated the change to UFSAR Table 1.8-1 which identified the license basis commitments to Regulatory Guides. This UFSAR table identified compliance with Regulatory Guide (RG) 1.129, Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Nuclear Power Plants. Additionally, the UFSAR table identified compliance with IEEE Std. 450-1980, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations. The licensee had changed the UFSAR Table 1.8-1 to indicate compliance with IEEE 450-1995 instead of IEEE 450-1980. This change was substantial in that, the 1995 standard had reduced the required battery testing frequency and had changed the acceptance criteria for testing the battery. However, the licensee incorrectly concluded in RAD 02-00055 that a documented review in accordance with 10 CFR 50.59 for this change to the facility (battery testing and acceptance) as described in the UFSAR was not required.
On November 15, 2002, the licensee completed 10 CFR 50.59 screen 02-01534. The licensee performed this screen to evaluate a revision to drawing D-214-004 and procedure PAP-0204 which added electrical separation requirements for the installation of temporary cables. In one of these drawing changes, the licensee applied a separation criteria of 6 inches between a temporary cable with a barrier (such as temporary cable in conduit) and a Class 1E cable tray. The team noted that this condition was analogous to establishing a separation requirement of 6 inches between non-Class 1E cable in conduit and a Class 1E cable tray. The licensee concluded that a documented safety evaluation in accordance with 10 CFR 50.59 was not required based on, The separation criteria adapted for temporary cable installations as documented on drawing D-214-004 is consistent with that described in UFSAR, Section 8.3.1.4.1.7 and IEEE-384 for permanent plant cable installations. The team agreed that this change was consistent with IEEE-384, however it was not consistent with more restrictive electrical separation requirements identified in UFSAR Section 8.3.1.4.1.7. This UFSAR Section stated, Separation between conduit and tray system containing Class 1E circuits or non-Class 1E circuits is 3 foot horizontal and 5 foot vertical, except for the cable spreading area which is reduced to 1 foot horizontal and 3 foot vertical.
Consequently, the changes the licensee made to the plants design drawing and implementing procedure were not in accordance with the UFSAR Section 8.3.1.4.1.7.
Therefore, a documented safety evaluation was required in accordance with 10 CFR 50.59 for this change to the facility as described in the UFSAR.
Analysis Because violations of 10 CFR 50.59 are considered to be violations that could potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the SDP. However, the results of this violation are assessed using the SDP. In these cases, the licensees failure to perform safety evaluations in accordance with 10 CFR 50.59 resulted in inconsistencies between the UFSAR and the plants design basis and configuration for electrical separation requirements, and the maintenance and testing associated with the station batteries.
The team considered this issue of more than minor significance, because if left uncorrected, the issue could become a more significant safety concern. The team was concerned that two examples of failure to perform a documented safety evaluation in the relatively small sample of screenings reviewed indicated a potential for a more significant safety concern if the underlying cause was not identified and corrected. For the examples identified, the team determined that the issue was of very low safety significance. The team concluded this because; 1) other licensees had been successful in evaluating changes under 10 CFR 50.59 from the older IEEE 450 requirements to compliance with the newer IEEE 450-1995 requirements, and 2) the reduced electrical separation distance that the licensee had incorporated into their design drawing and configuration met the requirements in IEEE 384 which had been endorsed by the NRC in RG 1.75. The team considered that if the licensee had completed a safety evaluation in accordance with 10 CFR 50.59 for these examples, it would have likely been successful and therefore this issue was determined to be of very low safety significance.
Enforcement 10 CFR 50.59(d)(1) stated, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments. These records must include a written evaluation which provides for the determination that the change, test, or experiment does not require a license amendment. Contrary to these requirements, as of February 14, 2003, the licensee failed to perform two written safety evaluations for changes to the facility as described in the UFSAR documented in RAD 02-00055 completed January 22, 2002 and screen 02-01534 completed on November 15, 2002. Specifically, the RAD and screening did not provide a written evaluation which provides for the determination that the change, test, or experiment did not require a license amendment. The results of this violation were determined to be of very low safety significance; therefore, this violation of the requirements contained in 10 CFR 50.59 was classified as a Severity Level IV Violation. However, because this non-willful violation was non-repetitive, and was captured in the licensees corrective action program (condition report (CR) 03-00724), it is considered a Non-Cited Violation (NCV 50-440/03-02-01) consistent with Section VI.A.1 of the NRC Enforcement Policy.
b.2 Inadequate 10 CFR 50.59 Safety Evaluation Introduction The team identified an unresolved item associated with an inadequate safety evaluation.
Specifically, the licensee failed to assess the affects of noble metal chemical addition (NMCA) on the peak cladding temperature (PCT) under accident conditions due to potential catalytic affects on the Zirconium (Zr) metal-water reaction rate.
Description In February of 2001, the licensee approved a safety evaluation 01-0007, TXI-321 Noble Chemical Metal Addition, to document the safety evaluation for the addition of NMCA on the reactor primary coolant system components and fuel cladding. The NMCA process was implemented to reduce the susceptibility of components within the core and reactor coolant systems to stress corrosion cracking. The licensee concluded in evaluation 01-0007, that implementation of this activity would not require prior NRC approval based on meeting the criteria of 10 CFR 50.59. The licensee subsequently implemented the NMCA at the Perry site. However, the team identified that the licensee had failed to consider the potential affects of NMCA on PCT under accident conditions.
The noble metals used in the NMCA act as a catalyst to facilitate recombination of hydrogen and oxygen. Under normal operation, oxygen is produced in the core by radiological decomposition of water and is removed by combination with hydrogen which is added to the reactor coolant system. During a loss of coolant accident (LOCA)hydrogen can be produced as a byproduct of the Zr metal-water reaction and oxygen is present in the emergency injection water sources such as the condensate storage tank or the containment sump. Under LOCA conditions, the NMCA could cause the hydrogen to be removed at an accelerated rate from the product side of the Zr metal-water reaction (Zr + 2H2O -> ZrO2 + 2H2 + heat) which may tend to increase the reaction rate. This would occur because the NMCA has been demonstrated to increase the reaction rate for recombination of hydrogen and oxygen (2H2 +O2 -> 2H2O + heat).
Additionally, the NMCA may directly increase the reaction rate for the Zr metal-water reaction through catalytic action, by lowering the initiation temperature for this reaction which is normally about 1700 degrees Fahrenheit (e.g. lower the required activation energy). Thus, the addition of NMCA could increase the heat input to the fuel cladding due to the enhanced exothermic Zr metal-water and hydrogen-oxygen recombination reactions. The heat produced from the Zr metal-water reaction is considered in computer models used to evaluate the emergency core cooling system performance with respect to PCT (10 CFR 50.46). If the Zr metal-water and hydrogen-oxygen reaction rates are substantially increased by NMCA, the heat input to the cladding following a design basis LOCA could be increased above that accounted for in the computer models used to calculate PCT. If the increase in heat input to the cladding was substantial, the team was concerned that it could increase the likelihood of cladding failure by thermally induced failure mechanisms related to creep or creep-rupture in a post LOCA environment. Cladding failure would result in increased fission product activity released into the coolant which could potentially impact the on and off-site dose consequences previously considered for this accident. The licensee entered this potentially generic concern into the corrective action system (CR 03-0071) and completed an operability determination.
The licensee considered that this issue did not affect operability principally because their fuel reload analysis showed a substantial margin to the PCT limit (e.g., maximum PCT of 1560 degrees Fahrenheit for the limiting GE 14 fuel configuration, vice the 2200 degree Fahrenheit limit). Additionally, the licensees operability determination indicated that the steam environment (with highest PCT) which could exist for 2-3 minutes after a LOCA would not allow substantive amounts of oxygen to be produced through radiological decomposition and without a source of oxygen recombination of hydrogen and oxygen (facilitated by NMCA) can not occur. The team concluded that specific information on post LOCA PCT profiles in concert with potential sources and levels of oxygen and hydrogen may be needed to resolve this issue. The team considered this issue an unresolved item pending review of the revised 50.59 safety evaluation on the NMCA process (URI 50-440/03-02-02).
1R17 Permanent Plant Modifications
Review of Recent Permanent Plant Modifications
a. Inspection Scope
From February 10, 2003, through February 14, 2003, the team performed an on-site (room 205 of the site administrative building) review of 10 permanent plant modifications. Additionally, the team reviewed a sample of setpoint changes, equivalency evaluations and a commercial grade dedication that were performed by the licensee's engineering staff. The team reviewed these modifications and supporting calculations to verify that the completed design changes were in accordance with design and licensee requirements identified in the Perry UFSAR and Technical Specifications.
The teams also reviewed applicable industry design standards, such as the Institute of Electrical and Electronics Engineers or American Society of Mechanical Engineers Code (ASME) Sections XI and Section III, to evaluate acceptability of the modifications.
Additionally, the team reviewed applicable post-modification testing to verify that the system, and associated support systems, functioned properly and that the modification accomplished its intended function.
b. Findings
b.1 Inadequate Design Review for Emergency Service Water Elbow Modifications Introduction Green. The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees inadequate design reviews associated with the installation of half-couplings on a 14 inch emergency service water (ESW)elbow.
Description On November 6, 2000, the licensee identified (CR 00-3433) a through-wall leak on a 14 inch ESW elbow downstream of valve 1P45-F0541B at the outlet of the B ESW heat exchanger. The licensee did not identify a root cause for this degradation and used only an ultrasonic thickness gauge to determine the extent of the flaw. The licensee evaluated and approved the design of a 1-1/2 inch half-coupling with plug to cover up the through-wall hole in the 14 inch elbow in engineering change package (ECP) 00-8092. This modification was approved and installed on November 7, 2000.
On August 2, 2001, the licensee identified (CR 01-2974) a new area of wall loss on the same 14 inch ESW elbow with the through-wall leak as discussed above. The licensee did not identify a definitive cause for the degradation and again used only an ultrasonic thickness gauge to determine the extent of the flaw. On October 29, 2001, the licensee evaluated and approved ECP 01-8043 for the design of a 1-1/2 inch half-coupling with plug to cover up the new area of degradation. The degraded area had reduced the nominal 0.375 inch wall thickness to 0.148 inch. The licensee installed the half-coupling and plug over the degraded area on April 4, 2002.
The licensee had used an ultrasonic thickness gage on the degraded ESW elbow, which was not adequate to detect or characterize the extent of these flaws (hole/degraded areas). Specifically, the ultrasonic thickness gage utilized a 0 degree ultrasonic transducer, which was not capable of detecting planar flaws (e.g., cracks). To confirm that cracking (planar flaws) did not exist, the licensee should have performed an ultrasonic examination (UT) using an angle beam type transducer. The team was concerned that if undetected cracking (planar flaws) existed, the residual weld tensile stress induced in the pipe wall for these highly constrained repair weldments could have increased the rate of flaw growth and caused a premature failure of the elbow.
Therefore, the team questioned the operability of the pipe elbow and consequently the B ESW train. On February 12, 2003, the team's concern prompted the licensee to perform additional UT (using 0 degree dual element, 60 degree and 70 degree angle beam transducers) of the ESW elbow to characterize the nature of the base metal surrounding the repaired area. During this inspection, the licensee identified three new flaws, two of which were just under the attachment weld of the half-coupling installed in 2000. The deepest of these flaws reduced the 0.375 nominal wall thickness to 0.14 inches. However, due to the half-couplings installed over the flaw areas, the original flaws could not be measured.
On February 14, 2003, the team reviewed the licensees operability determination 03-00699. As part of this operability determination, the licensee identified a number of potential causes for the observed degradation in this ESW elbow including erosion/corrosion and microbiological corrosion. In this evaluation, the licensee evaluated a bounding sized planar and non-planar flaws, which were assumed to exist under the half-couplings. The licensee then compared the bounding sized flaws to the maximum allowable sized flaws calculated in accordance with the methodology discussed in Code Case N-513, Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 Piping (NRC approved methodology). However, the team identified that the licensee had not explicitly documented how the acceptance criteria of Section 4.0 of Code Case N-513 were met. The licensee subsequently revised their operability determination and demonstrated that the flaws left within the ESW elbow did not reduce the structural margins below the acceptance criteria for planar flaws in Section 4.0 of Code Case N-513. Therefore, the licensee considered the ESW elbow operable until the upcoming refueling outage. Because the licensee determined that the affected component was operable, the team did not have an immediate safety concern.
The teams concern for the B loop elbow, appeared to prompt the licensee to perform UT of the same elbow in the A loop of the ESW system prior to the normally scheduled monthly UT checks. During this UT, the licensee identified a degraded area (0.109 inch thick) which was below the required minimum wall thickness (0.116 inch) (CR 03-0833).
The licensee again applied Code Case N-513 to demonstrate operability of the degraded area until the upcoming refueling outage. Because the licensee determined that the affected component was operable, the team did not have an immediate safety concern.
The licensees design reviews for ECP 00-8092 and ECP 01-8043 were inadequate in that, the licensee failed to incorporate the applicable requirements of Section XI of the ASME Code. Specifically, the licensee had failed to identify a definitive cause for the flaws, failed to adequately characterize the dimensions of the flaws, nor was the potential growth of these flaws considered. Each of these actions was required by the ASME Code Section XI (paragraphs IWA -3300, IWA-4130, and IWD-3000). The licensee's repair design did not include removal of the flaw as required by paragraph IWA-4300 of Section XI. Additionally, the use of a half-coupling and plug to cover up the flaw was not a recognized repair method identified in paragraph IWA-4130 of Section XI. Further, the licensees repair method was not consistent with NRC approved alternative repair methods discussed in Code Cases N 513, and N 523 or Generic Letter 90-05 "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2 and 3 piping."
Analysis The team reviewed this finding against the guidance contained in Appendix B, Issue Dispositioning Screening, of IMC 0612, Power Reactor Inspection Reports. In particular, the team compared this finding to the findings identified in Appendix E, Examples of Minor Issues, of IMC 0612 to determine whether the finding was minor.
Following that review, the team concluded that none of the examples listed in Appendix E accurately represented this example. As a result, the team compared this performance deficiency to the minor questions contained in Section C, Minor Questions, to Appendix B of IMC 0612. The team concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002, because the finding:
- (1) involved the design control and equipment performance attributes of the mitigating systems cornerstone; and
- (2) affected the mitigating systems objective of ensuring the availability, reliability, and capability of the ESW system in response to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, these non-Code repairs represented a degradation in the plants design basis (ASME Code Sections III and XI) for the B train emergency service water elbow. This degraded elbow affected the mitigating systems cornerstone objective of ensuring the availability, reliability, and capability of the emergency service water system, because these non-Code repairs and associated piping flaws could have resulted in premature failure of the elbow.
The team evaluated the finding using Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding was a design or qualification deficiency that did not result in loss of component/system function.
Therefore, the team screened this finding to have very low safety significance (Green).
Enforcement 10 CFR 50, Appendix B, Criterion III, Design Control, required, in part, that applicable regulatory requirements and design basis are correctly translated into specifications, drawings, procedures and instructions. 10 CFR 50.55a(g)(4) required, in part, that throughout the service life of boiling or pressurized water cooled nuclear power facility, components, which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the requirements of Section XI (applicable regulatory requirement). The 14 inch ESW elbow downstream of valve 1P45-F0541B was a Code Class 3 component.
Contrary to these requirements, in engineering change package 00-8092 approved on November 7, 2000, and 01-8043 approved on October 25, 2001, the licensee failed to incorporate the applicable regulatory requirements of the 1989 Edition of ASME Code Section XI, paragraphs (IWA-3300, IWA-4130, IWA-4300, IWD-3000) associated with flaw evaluation, flaw removal and component repair into applicable specifications, drawings, procedures and instructions for the 14 inch ESW elbow downstream of valve 1P45-F0541B. Consequently, the licensee had to conduct substantial additional non-destructive examinations and flaw growth analysis to confirm that the system was operable. Failure to incorporate the applicable regulatory requirements into these engineering change packages is a violation of 10 CFR 50 Appendix B, Criterion III. This violation is associated with an NRC identified finding that is characterized by the significance determination process as having very low risk significance (Green) and is being treated as a Non-Cited Violation of 10 CFR 50 Appendix B, Criterion III, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-440/03-02-03). The licensee entered this finding into the corrective action system (CR 03-0744).
b.2 Inadequate Design Review for Rupture Disc Modification Introduction Green. The team identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees inadequate design review associated with installation of a rupture disc in the exhaust piping of the division 3 high pressure core spray (HPCS) emergency diesel generator (EDG). This finding was self revealed on October 25, 2000, after the diesel generator was placed in service following this modification, the rupture disc failed in less than 3 minutes due to pressure induced fatigue.
Description
The team identified a self-revealing finding associated with ECP 00-6009, in which the licensee evaluated and approved (on August 22, 2000) the design of a rupture disc for installation into the exhaust piping for the division 3 HPCS EDG. The normal EDG exhaust vent piping was located on the roof of the EDG building and was susceptible to damage induced by tornado or high wind driven projectiles. The licensee installed a rupture disc in an exhaust branch line to serve as an alternate vent path in the event that the normal exhaust vent piping was restricted or damaged. The rupture disc was bolted into the terminal end of a vertical run of exhaust duct. If the rupture disc was actuated, it would allow EDG exhaust gases to vent into a partially enclosed concrete corridor with open doorways to the building roof.
October 25, 2000, after the EDG was placed in service following this modification, the rupture disc failed in less than 3 minutes. The licensee promptly identified the failed rupture disc during the initial EDG operation and implemented a modification to remove the rupture disc to allow the exhaust to vent directly through this branch. The licensee also added steel screens from the floor to the ceiling around the perimeter of the open vertical run of exhaust duct. This barrier served to protect the open duct from introduction of foreign materials. However, with the modified exhaust configuration, the high temperature EDG exhaust gases contacted the concrete ceiling of the building structure above the opening and caused concrete spalling (chips). If these concrete chips had fallen into the open vertical run of duct which used to contain the rupture disc, the EDG could have been rendered inoperable. Specifically, concrete debris could have entered the vertical exhaust duct and lodged in the turbocharger or other engine components causing failure of the EDG. Fortuitously, the licensee identified this condition before introduction of concrete chips and added a steel plate mounted to staging above the open exhaust vent to provide a barrier from falling concrete chips.
Because the licensee determined that the affected component was operable, the team did not have an immediate safety concern.
The licensee had a vendor perform an investigation of the failed rupture disc and provide a report Failure Analysis Report Diesel Generator Rupture Disc dated November 22, 2000. In this report, the vendor concluded that the rupture disc failed by fatigue induced by cyclic exhaust pressure transients during normal diesel engine operation. The licensees had not adequately considered the impact of normal cyclic pressure exhaust loads on the design service life of the rupture disc in ECP 00-6009.
Analysis The team reviewed this finding against the guidance contained in Appendix B, Issue Dispositioning Screening, of IMC 0612, Power Reactor Inspection Reports. In particular, the inspectors compared this finding to the findings identified in Appendix E, Examples of Minor Issues, of IMC 0612 to determine whether the finding was minor.
Following that review, the team concluded that none of the examples listed in Appendix E accurately represented this example. As a result, the team compared this performance deficiency to the minor questions contained in Section C, Minor Questions, to Appendix B of IMC 0612. The team concluded that the finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Disposition Screening, issued on April 29, 2002, because the finding:
- (1) involved the design control and equipment performance attributes of the mitigating systems cornerstone; and
- (2) affected the mitigating systems objective of ensuring the availability, reliability, and capability of the division 3 HPCS EDG in response to initiating events to prevent undesirable consequences (i.e., core damage).
When the licensee removed the failed rupture disc and declared the EDG operable, the EDG was in a less reliable configuration. Specifically, the removed rupture disc created an opening which could have allowed foreign material (e.g. concrete chips) to enter the exhaust system and cause premature failure of the division 3 HPCS EDG. The licensee subsequently installed foreign material exclusion barriers to ensure the continued reliability of the HPCS EDG.
The team evaluated the finding using Manual Chapter 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding was a design or qualification deficiency that did not result in loss of component/system function.
Therefore, the team screened this finding to have very low safety significance (Green).
Enforcement 10 CFR 50, Appendix B, Criterion III, Design Control, required, in part, that measures shall provide for verifying and checking the adequacy of design, such as by the performance of design reviews. Contrary to these requirements, in engineering change package 00-6009 approved on August 22, 2000, and installed on October 25, 2000, the licensee failed to consider the in-service cyclic pressure induced fatigue loads on the design life of the component. Consequently, the rupture disc failed within 3 minutes after commencing EDG operation. Failure to perform an adequate design review and identify the in-service cyclic pressure transients as a design parameter and incorporate this into the rupture disc design is a potential violation of the 10 CFR 50 Appendix B Criterion III, Design Control. This violation is associated with a self revealing finding that is characterized by the significance determination process as having very low risk significance (Green) and is being treated as a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion XVI, consistent with Section VI.A.1 of the NRC Enforcement Policy (NCV 50-440/03-02-04). The licensee entered this finding into the corrective action system (CR 00-3320 and CR 01-2283).
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
a. Inspection Scope
The team performed an on-site (room 205 of the site administrative building) review of a sample of permanent plant modifications and 10 CFR 50.59 program problems that were identified by the licensee and entered into the corrective action program in condition reports. The team reviewed these condition reports to confirm that the licensee had appropriately described the scope of the problems as documented in the condition reports. Additionally, the teams review included confirmation that the licensee had an appropriate threshold for identifying issues and had implemented effective corrective actions related to design issues. The specific corrective action documents that were reviewed by the team are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA6 Meetings
Exit Meeting The team presented the inspection results to Mr. W. Kanda and other members of licensee management at the conclusion of the inspection on February 14, 2003. The licensee acknowledged the information presented during this meeting. No proprietary information was identified.
KEY POINTS OF CONTACT Licensee W. Kanda, Vice President - Nuclear T. Rousch, Plant Manager T. Lentz, Engineering Director D. Haviland, Supervisor Structural Mechanical Design C. Angstadt, Engineering Assessment Board Chairman D. Gartner, I&C Lead Engineer K. Russel, Compliance Engineer D. Miller, Staff Consultant J. Zarea, Electrical Design Engineering NRC Ray Powell, Senior Resident Inspector LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-440/03-02-01 NCV Failure to perform a safety evaluation for changes to the plant as described in the UFSAR.
50-440/03-02-02 URI Increased heat input on PCT from the Zr metal-water and hydrogen-oxygen reactions facilitated by Noble Metals.
50-440/03-02-03 NCV Failure to perform adequate design reviews for installation of half-couplings on a B train emergency service water elbow.
50-440/03-02-04 NCV Failure to perform adequate design reviews for installation of a rupture disc in the exhaust piping of the division 3 high pressure core spray diesel generator.
Closed 50-440/03-02-01 NCV Failure to perform a safety evaluation for changes to the plant as described in the UFSAR.
50-440/03-02-03 NCV Failure to perform adequate design reviews for installation of half-couplings on a B train emergency service water elbow.
50-440/03-02-04 NCV Failure to perform adequate design reviews for installation of a rupture disc in the exhaust piping of the division 3 high pressure core spray diesel generator.
Discussed None LIST OF
DOCUMENTS REVIEWED
1R02
Evaluations of Changes, Tests, or Experiments
CFR 50.59 Safety Evaluations
00-0014
SMRF 99-5010, R/0 and associated USAR
Change Request (CRF)
April 3, 2000
00-0021
SMRF 99-5056 Rev. 0 and associated USAR
Change Request
March 31, 2000
00-0052
DCP 00-6009 Revision 0 and Associated
UFSAR Change Request
August 24, 2000
00-0084
Safety Evaluation of SMRF 00-5018
November 21,
2000
00-0089
Safety Evaluation for Temporary Modification
1-00-004
November 30,
2000
01-0001
TXI-321 Noble Chemical Metal Addition
January 30,
2001
01-0007
TXI-321 Noble Chemical Metal Addition
February 8,
2001
01-0018
DCN 5910 R/0 and UFSAR Change Request
March 18, 2001
01-0026
Temp Mod 1-01-002 Rev 0
April 16, 2001
01-0036
DCP 99-5020
October 23,
2001
CFR 50.59 Safety Evaluation Screenings or Regulatory Applicability Determinations
01-00384
Plant Data Book, Administrative Requirements
Section, Section 7.5.5, Diesel Generator
Maintenance Program procedure change
October 08,
2001
2-00055
Changed USAR Table 1.8-1 to change
commitment for compliance with IEEE 450-
1980Property "IEEE" (as page type) with input value "IEEE 450-</br></br>1980" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. to compliance with IEEE 450-1995
January 22,
2002
2-00384
ECP 02-00028 and USAR Change Request
2-034
July 16, 2002
2-00436
RPV Steam Dome and RHR Cut-in Permissive
High Pressure Channel B Calibration for 1B21-
N678B
Revision 0
2-00453
RPV Steam Dome and RHR Cut-in Permissive
April 22, 2002
2-00530
Installation of Jumpers to Defeat Faulty
Circuitry for the Extraction Steam to the Steam
Jet Air Ejector and Off-Gas Preheater Drain
Pot Level Switch N22-N273
Revision 0
2-00569
Replace Existing PSA Snubber for 1N25-
H0027 with a Lisega Snubber
Revision 0
2-00940
Emergency Closed Cooling System A loop
August 8, 2002
2-01210
Elimination of MSIV Stem Leakoff Piping
October 3, 2002
2-01301
Temporary Modification to Eliminate Locked in
High Oil Level Annunciator for Reactor
Recirculation Motor A
Revision 0
2-01534
Revision to Drawing D-214-004 (ECR-02-0303)
and Procedure PAP-0204 to add separation
requirements for the installation of temporary
cables.
November 15,
2002
2-01617
Augmented Visual Inservice
Inspection/Examination of Safety-Related
December 9,
2002
2-01669
ESW Pump A(B) and Valve Operability Test
December 11,
2002
2-01799
Scram Discharge Volume Vent and Drain
Valves Operability Test
December 23,
2002
Condition Reports
01-1332
RFO8 Need: Circ WTR SYS Steel to Concrete
Groundwater In-Leakage
May 12, 2001
01-3269
September 6,
2001
01-3683
Improperly prepared Regulatory Applicability
Determinations
October 17,
2001
01-3747
No 10CFR50.59 Evaluation Performed for
Changes to P45 SVIs with Capacitance
Sensing Instrumentation
October 25,
2001
01-4365
50.59 Self-Assessment Results
December 26,
2001
2-00250
Plant underdrain temperary modification 1-02-
001 50.59 not conservative
January 23,
2002
2-00272
Temporary Modification with No 10CFR50.59
Evaluation
January 28,
2002
2-00745
Temporary Modification Installed with no 10 cfr
50.59 Evaluation Performed
May 13, 2002
2-01332
M&TE for monitoring RWCU pump mechanical
seal cooling line for RWCU pumps
April 30, 2002
2-01422
Equivalent Changes Incorrectly Exempted from
cfr 50.59 Applicability
May 8, 2002
2-03487
Service water chlorination system
September 12,
2002
2-03788
DCP 96-044 safety evaluation 00-0007
October 10,
2002
Condition Reports Initiated as the Result of NRC Inspection
03-00478
50.59 database incomplete
January 30,
2003
03-00695
Potential failure to perform required 50.59
evaluation related to RAD 02-00055
February 11,
2003
03-00721
Nuclear fuel
February 12,
2003
03-00723
Raceways (conduits and cable trays)
February 12,
2003
03-00724
Electrical raceways (conduits and cable trays)
February 12,
2003
03-00743
NRC identified a number of issues with 50.59
documentation quality
February 13,
2003
Other Documents
Guidelines for 10 CFR 50.59 Implementation
Revision 1
License Amendment
No. 74
Amendment No. 74 to Facility Operating
License N0. NPF-58 - Perry Nuclear Power
Plant, Unit No. 1 (TAC No. M92190)
November 16,
1995
Change Request 01-
090
Replacement of Division 3 100Ah batteries
with 250 Ah batteries
March 6, 2000
Procedures
NOP-LP-4003
Evaluations of Changes, Tests and
Experiments
Revision 0
PAP-305
Safety Evaluations
Revision 8
PAP-520
Changes to the Updated Safety Analysis
Report and Other Licensing Documents
Revision 6
1R17
Permanent Plant Modifications
Calculations
R48-05
Overpressure Protection Rupture Disc
Revision 3
R48-21
Rupture Disc Qualification
Revision 0
P42-005
Emergency Closed Cooling Surge Tank Hi/Lo
Alarm 1P42N0131A,B
Revision 4
Qualification of Instrument Support 1P42-
H0183 Emergency Closed Cooling Surge tank
Elevation 665-0"
Revision 0
X-302
Surge Tank Instrument Support
Revision 1
X-506
I/F 1P42-A001A
Revision 1
R45-T04 R/7
Calc Adjusts the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Inventory and the 7
Day Inventory
November 12,
2001
Condition Reports
99-2439
Stuck Relay in ESW system caused by current
induction
October 9, 1999
99-3033
Fuel Oil Transfer Pump was inadvertently
started
December 5,
1999
00-2216
While attempting to declare the Unit 2 Div 3
Battery Operable, a DCP found Open
July 21, 2000
00-2495
Installation of the Isolation Transformer did not
Correct the Problem
July 17, 2000
00-3433
ESW B through wall pipe leak downstream of
November 6,
2000
00-3320
Exhaust system rupture on division 3 diesel
generator opened after modification
October 25,
2000
01-0230
Division 1,2, and 3 testable rupture disc
January 22,
2001
01-1277
1BN31 Accumulator Tank With Magnetrol
Level Switch
March 10, 2001
01-1485
Breaker F1F05 tripped
March 19, 2001
01-1715
ECC-B Surge tank Valve 1P42-F0668 Out of
Position
April 2, 2001
01-2181
FME, Metal Chunk that is not Part of the Valve
was removed from 1n27F0170
May 12, 2001
01-2283
Diesel generator testable rupture disc supply
fan dampers and core water pumps
May 21, 2001
01-2620
Premature Use of Calculation Results
July 2, 2001
01-2949
Auto Operation OD Damper Will Not Work
July 31, 2001
01-2994
Emergency service water loop B 14 inch piping
elbow degraded areas
August 2, 2001
01-3776
P42 Latent Issues ECC Hi/Lo Temp ARI
doesnt address temp Control Valve Failure
October 29,
2001
01-4073
Latent Issues Review-Procedural
Enhancements
November 27,
2001
01-4149
Design control audit
December 3,
2001
2-00723
Latent Issues ESW Valves do not Have Hot
Short Mod
March 12, 2002
2-01458
RFA to Evaluate Strain Relief Used During the
RC&IS (J50/P50) Cable Modification
May 14, 2002
2-01465
Error in the EQ Evaluation for ECP 02-0028
Rev. 0
May 14, 2002
2-01650
Emergency Closed Cooling Valves Differential
Pressure
May 28, 2002
2-01869
Safety Class Code Designators on Drawing D-
2-621
June 12, 2002
2-03099
1P45-F040A failed to close
September 5,
2002
2-03389
NQA finding during quarterly assessment
September 20,
2002
2-03604
Unit 1 Division 3 Battery
October 2, 2002
2-04210
Air block valve
October 22,
2002
Condition Reports Initiated as the Result of NRC Inspection
03-00699
Emergency service water elbow downstream
of 1P45F0541B
February 11,
2003
03-00733
Commercial grade dedication process
weakness
February 12,
2003
03-00739
Division 3 diesel generator exhaust rupture
disc
February 13,
2003
03-00744
Human performance concerns ASME Code
February 12,
2003
Drawings
D-214-004
Electrical Conduit and Tray Separation Criteria
Revision U
D-301-801
30" Pressure Relief Valve
Revision 0
D-301-832
30" Pressure Relief Valve Assembly Division 1
Revision 0
D-301-833
Testable Rupture Disc Latch Assembly
Revision 0
D-302-355
HPCS and Standby Diesel Generator Exhaust,
Intake and Crankcase
Revision R
D-302-0621-00000
Emergency Closed Cooling System
Revision HH
Other Documents
Failure Analysis Report Diesel Generator
Rupture Disc
November 22,
2000
J1103754SRLR
Supplemental Reload Licensing Report for
Perry Nuclear Power Plant Unit 1 Reload 8,
Cycle 9
Revision 1
CERF 1722
Equivalency Evaluation for Replacement of
Motor Feed Pump Impellers and Diffuser to
Reduce Vibration
Revision 0
VLI-P42
Emergency Closed Cooling System
Revision 6
ARI-H13-P601-20
RHR A
Revision 5
ARI-H14-P601-17
RHR B and C
Revision 5
PTI-P42-P0010
ECC A Loop Total Leakage Verification
Revision 2
1-99-1089
Time Delay for Short Time Function must be
changed from Max to Intermediate for proper
coordination
September 9,
1999
0-00-1001
Establish same CFA setting that has been
approved for Div 1 & 2 Battery Chargers
July 27, 2000
1-00-1039
100-PY-B(Y) Startup Supply to Bus L20 Brkr
L1001 Protection Change Pickup Setting from
5400A to 6000A
October 18,
2000
SCR 2-00-1002
Change Pickup Setting from 5400A (9A tap).
Change Time Dial setting from 7 to 4.
October 19,
2000
SCR 1-01-0039
Change to Improve the breaker settings to
eliminate nuisance trips.
September 7,
2001
Equivalent Change 00-
8062
Replace the existing obsolete flow
instrumentation 1P45N0270 and 1P45N0271
instrumentation with newer model instruments.
July 17, 2000
Modifications
99-5003
Add a 4-5 Minute Time Delay to Main &
Reactor Feed Pump Turbines For any RCIC
Initiations
February 12,
2001
99-8055
Replace DIV 1 DG Sonic Level Instrumentation
with Capacitance Sensing Instrumentation
November 12,
2001
00-5004
Reconfigure the Stop/Auto/Start Control
Switches
June 5, 2000
00-5018
Replace The Existing Magnetrol Level
Switches with Similar Magnetrol Level
Switches with Narrower Ranges to Provide
Nominal System Leakage Rate of 2.77 GPH
and Maintaining the 7 day Supply of Water
Revision 1
00-6009
Replace the existing hinged disc with a metal
rupture disc that will burst when exposed to
back pressure for the division 3 testable
rupture disc
August 22, 2000
00-8092
Repair ESW elbow leak at 3 Oclock position
looking east at the ECC B heat exchanger
ESW piping
November 7,
2000
01-8043
Install a 1-1/2 " 3000#, SA 105 half-coupling
October 29,
2001
2-0254
Relocate the 1P45K0027 relay from
1H51P0805, Compt. D (F1G04-D) for
hydrometer 1P45F0160
November 20,
2002
PRDC-0005
Replacement of Topaz Inverter and SOLA
Power Supply Units Resulting in Change to
Division 2 battery sizing calculation
June 1, 2000
PRDC-0014
Replacement of Topaz Inverter and SOLA
Power Supply Units Resulting in Change to
Division 1 battery sizing calculation
June 1, 2000
Purchase Orders
7046059
Class III piping components
November 6,
2000
7056588
Inconel X-750 valve spring
March 12, 2001
Procedures
NOP-CC-2003
Engineering Changes
Revision 1
NEI-0373
Processing Engineering Changes
Revision 6
PAP-1403
Control of Setpoints
Revision 9
NEI-0375
Equivalent Replacements
Revision 4
NEI-0420
Procurement Engineering
Revision 0
Procurement Engineering Technical Evaluation
for Parts Procurement/Inspection
Revision 2
NEI-0341
Calculations
Revision 9
NOP-CC-2004
Design Interface Reviews and Evaluations
Revision 0
PAP-1124
Pre-Maintenance and Post-
Maintenance/modification Test Program
Revision 0
FTI-F0036
PMT Program Matrix
Revision 2
Work Orders
00-009369
Implement a design change to install a half-
coupling and plug per at risk SMRF 00-8092
November 7,
2000
00-001032
Diesel generator exhaust relief device 30 inch
October 25,
2000
Supplemental Information Request
Document Type
Number(s)
Requesting
Inspector
Modifications (see Note 1)
00-5004, 99-5003, 99-8055
R. Winter
Modifications (see Note 1)
00-8062, 01-8040, 02-0020, 02-0254
R. Daley
Modifications (see Note 1)
00-5018
J. Ellegood
Modifications (see Note 1)
00-8092, 00-6009
M. Holmberg
Setpoint Changes
0-00-1001, 1-00-1039,1-99-1089
B. Winter
Setpoint
Changes
1-01-0039, 2-00-1002
R. Daley
Equivalency Evaluation
CERF 01722
J. Ellegood
PO. Number 7056588
S/N Number 91132595
M. Holmberg
Calculations
Supporting Modifications
R45-T04
R. Winter
Calculations
Supporting Modifications
PRDC-0005 R/4 DCC-5
PRDC-0014 R/0 DCC-1
R. Daley
Calculations
Supporting Modifications
All calculations supporting modification
00-5018.
J. Ellegood
Calculations
Supporting Modification
All calculations supporting
modifications
00-6009 and 00-8092.
M. Holmberg
Condition Reports (mod
related)
00-2216, 01-2620, 01-3776, 02-00723
R. Winter
Condition Report (mod
related)
01-2184, 02-03604
R. Daley
Condition Reports (mod
related)
01-2949, 02-01465
J. Ellegood
Condition Reports (mod
related)
01-0230, 02-03389, 02-04210
M. Holmberg
Document Type
Number(s)
Requesting
Inspector
Safety Evaluations
01-0026, 01-0036
R. Winter
Safety Evaluation
00-0089
J. Ellegood
Safety Evaluations
00-0014, 00-0021
R. Daley
Safety Evaluations01-001, 01-007,01-018
M. Holmberg
safety evaluation
screenings
2-01210, 02-00453, 02-00940
R. Winter
RAD and 2 safety
evaluation screenings
RAD 02-00055, Screens 01-00294, 01-
00384
R. Daley
safety evaluation
screenings and 1 RAD
2-00530; 02-00569; 02-01301; RAD
2-00436
J. Ellegood
safety evaluation
screenings and 3 RADs
Screening 02-00384 associated with
UFSAR change and CR 02-034.
Screening for Modifications 01-8034
and 02-0237.
RADs 02-01799, 02-01669, 02-01617
M. Holmberg
Condition Reports (related
to 50.59 evaluation)
2-00272, 01-3683, 01-3747
R. Winter
Condition Reports (related
to 50.59 evaluation)
01-3269, 01-4365
R. Daley
Condition Reports (related
to 50.59 evaluation)
2-01422; 01-01332; 02-00272
J. Ellegood
Condition Reports (related
to 50.59 evaluation)
2-03487, 02-01332, 02-00250
M. Holmberg
Special Test
TXI-0321-002 Noble Chemical Metals
Addition
M. Holmberg
Note 1 - Copy of the Modification Package to include; document describing the need for the
modification, condition reports associated with the modification or installation of the
modification, description of the modification, the safety evaluation or screening, material lists
and procurement records, completed modification acceptance tests, procedures and drawings
affected/changed due to modification, records confirming applicable environmental qualification
(e.g. steam, fire, flooding, seismic) of modified components and applicable UFSAR Sections
changed due to the modification. Also, provide a copy of the vendor technical manual
associated with the modified equipment.
LIST OF ACRONYMS USED
American Society of Mechanical Engineers
CFR
Code of Federal Regulations
CR
Condition Report
Division of Reactor Safety
Engineering Change Package
Emergency Service Water
IMC
Inspection Manual Chapter
Loss of Coolant Accident
Noble Metal Chemical Addition
Non-Cited Violation
NRC
Nuclear Regulatory Commission
Publicly Available Records
Peak Clad Temperature