ML20044H500
| ML20044H500 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 06/01/1993 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H501 | List: |
| References | |
| NUDOCS 9306090113 | |
| Download: ML20044H500 (27) | |
Text
{{#Wiki_filter:g~ f$pn REQy . f*,, UNITED STATES f j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 3 %....+/ DUQUESNE LIGHT COMPANY OHI0 EDIS0N COMPANY PENNSYLVANIA POWER COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.173 License No. OPR-66 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Duquesne Light Company, et al. (the licensee) dated December 30, 1992, as supplemented March 12, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; l C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. J l 9306090113 930601 ~ PDR ADOCK 05000334 p PDR
5 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to-this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-66 is hereby-l amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as' revised through Amendment No.173, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION - G,$ l bV Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical i Specifications Date of Issuance: June 1, 1993 f i i E b
7 i ATTACHMENT TO LICENSE AMENDMENT N0t.173 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following pages of Appendix A Technical Specifications, with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 3/4 4-8 3/4 4-8 3/4 4-9 3/4 4-9 3/4 4-10 3/4 4-10 3/4 4-10a 3/4 4-10a 3/4 4-10b 3/4 4-10b 3/4 4-10c 3/4 4-10c - l 3/4 4-10d 3/4 4-10d 3/4 4-10e B 3/4 4-1 B 3/4 4-1 i 8 3/4 4-la B 3/4 4-la B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-2a i l l l l 'm+, ~ --c--. ,,,-r-- ---wwq. - -,, - - ---,em.-- r.m y nn.-
~ ~ DPR-66 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,y above 200 F. SURVEILLANCE REQUIREMENTS 4.4.5.1 Steam Generator Samole Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Samole Selection and Insoection - The steam generator tube minimum sample
- size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubes shall be examined in accordance with Article 8 of Section V (" Eddy current Examination of Tubular Products") and Appendix IV to Section XI -(" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing") of the applicable year and addenda of the ASME Boiler and Pressure Vessel Code required by 10CFR50, Section 50.55a(g). When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3 percent of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50 percent of the tubes inspected shall be from these critical areas. l b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include: 1. All nonplugged tubes that previously had detectable wall penetrations greater than 20 percent, and BEAVER VALLEY - UNIT 1 3/4 4-8 Amendment No.173
~ - _ as ! i l DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 7. Tubes in those areas where experience has indicated potential problems, and 3. At least 3 percent of the total number of sleeved tubes in all three steam generators. A sample size ?ess than 3 percent is acceptable provided all the sleeved tubes in the steam generator (s) examined during the refueling outage are inspected. These inspections will include both the tube and the sleeve, and 4. A tube inspection pursuant to Specification 4.4.5.4.a.8. If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve incoection, this shall be recorded and an adjacent 1 tube shall be selected and subjected to a tube inspection. c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided: i 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and 2. The inspections include those portions of the tubes where imperfections were previously found. 1 The results of each sample inspection shall be classified into one of the following three categories: Cateaory Inspection Results C-1 Less than 5 percent of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1 percent of the total tubes inspected are defective, 'or between'5 percent and 10 percent of the total tubes inspected are degraded tubes. BEAVER VALLEY - UNIT 1 3/4 4-9 Amendment No.173 . ~ -.
DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ~ C-3 More than 10 percent of the total tubes inspected are degraded tubes or more than 1 percent of the inspected tubes are defective. Note: In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10 percent) further wall penetrations to be included in the above percentage calculations. 4.4.5.3 Inspection Frecuencies The above required inservice inspections of steam generator tubes shall be performed at the following frequencies: The first inservice inspection shall be performed after 6 a. Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including.the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months. b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, .the inspection frequency shall be reduced to at least once per 20 months. The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a j third sample inspection is not required. c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions: l j l 1. Primary-to-secondary tube leaks (not including leaks l originating from tube-to-tube sheet welds) in excess j of the limits of Specification 3.4.6.2, 2. A seismic occurrence greater than the Operating Basis Earthquake, j BEAVER VALLEY - UNIT 1 3/4 4-10 Amendment No.173 l
~7 ' DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 7. A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4. A main steam line or feedwater line break. 4.4.5.4 Acceptance Criteria a. As used in this Specification: 1. Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that l required by fabrication drawings or specifications. Eddy-current testing indications below 20 percent of the nominal tube wall thickness, if detectable, may be considered as imperfections. 2.
Dearadation means a service-induced cracking,
wastage,. wear or general corrosion occurring on either.inside or outside of a tube or sleeve. l 3. Dearaded Tube means a tube or sleeve containing l imperfections greater than or equal to 20 percent of the nominal wall thickness caused by degradation. 4. Percent Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by degradation. 5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube l containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube. 6. Pluccina or Repair Limit means the imperfection depth l at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection. The plugging or repair i limit imperfection depths are specified in percentage of nominal wall thickness as follows: a. Original tube wall 40% b. Babcock & Wilcox kinetic welded sleeve wall 40% BEAVER VALLEY - UNIT 1 3/4 4-10a Amendment No.173
- = - - - a..A ~ DPR-66 REACTOR COOLANT SYSTEM + SURVEILLANCE REQUIREMENTS (Continued) 7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above. 8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support to the cold leg. 9. Tube Repair refers to sleeving which is used to maintain a tube in-service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure. The following sleeve designs have been found acceptable: a) Babcock Wilcox kinetic welded
- sleeves, BAW-2094P, Revision 1 including kinetic sleeve i
" tooling" and installation process parameter changes. I b. The steam generator shall be determined OPERABLE after completing the corresponding actions (pP.7 or repair all tubes exceeding the plugging or repair liwit) required by Table 4.4-2. 4.4.5.5 Reports a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to l Specification 6.9.2. b. The complete results of the steam generator tube and sleeve I inservice inspection shall be submitted to the Commission l in a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include: 1. Number and extent of tubes and sleeves inspected. 1 2. Location and percent of wall-thickness penetration for each indication of an imperfection. 3. Identification of tubes plugged or repaired. I BEAVER VALLEY - UNIT 1 3/4 4-10b Amendment No.173
~7, DPR-66 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) c. Results of steam generator tube inspections which fall into Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence. BEAVER VALLEY - UNIT 1 3/4 4-10c Amendment No.173 l t
DPR-66 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION i Preservice Inspection N3 Yes No. of Steam Generators per Unit Two Three Four Two Three Four First Inservice Inspection All One Two Two Second & Subsequent Inservice Inspections One (1) One (1) One (2) One (3) Table Notation: (1) The inservice inspection may be limited to one steam generator on a rotating schedule encompassing 3 N percent of the tubes (where N is the number of steam generators in the plant) if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note-that under some circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions. (2) The other steam generator not inspected during the first inservice inspection shall be inspected. The third and subsequent inspections should follow the instructions described in (1) above. (3) Each of the other two steam generators.not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described in (1) above. BEAVER VALLEY - UNIT 1 3/4 4-10d Amendment No.17.3 l 1 I i
4 DPR-66 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G. C-2 Plug or repair C-1 None N/A N/A l defective tubes and inspect additional C-2 Plug or repair defective C-1 None l 2S tubes in this S.G. tubes and inspect additional 4S tubes in C-2 Plug or repair l this S.G. defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 N/A N/A result of first sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug or S.G.s are repair defective C-1 tubes and inrpect 2S tubes in each other S.G. Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second Notification to NRC additional sample g pursuant to S.G.s are Specification 6.6 C-3 Additional Inspect all tubes in S.G. is each S.G. and plug or N/A N/A C-3 repair defective tubes. l~ Notification to NRC i pursuant to Specifi-l cation 6.6. i 3N Where N is the number of steam generators in the unit, and n is the number of steam generators inspected 8* n during an inspection. BEAVER VALLEY - UNIT 1 3/4 4-lOe-Amendment No.173-I-
A 1 DPR-66 l 3/4.4 REACTOR COOLANT SYSTEM l BASES ~ 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design DNBR limit during all normal operations and anticipated transients. In Modes 1 and 2, with one reactor coolant loop not in operation, THERMAL POWER is restricted to less than or equal to 31 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset. Either action ensures that the DNBR will be maintained above the design DNBR limit. A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (31 percent of RATED THERMAL POWER). In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, due to the initial conditions assumed in the analysis for the control rod bank withdrawal from a subcritical condition, two operating coolant loops are required to meet the DNB design basis for this Condition II event. In MODES 4 and 5, a single reactor coolant loop or RHR subsystem provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE. The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting a Reactor Coolant Pump with one or more RCS cold legs less than or equal to 275'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water level in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 25'F above each of the RCS cold leg temperatures. Power is removed from the isolated loop stop valves (hot leg and cold leg) to ensure that no reactivity addition to the core can i BEAVER VALLEY - UNIT 1 B 3/4 4-1 Amendment No.173
- 2. A l t DPR-66 3/4.4 RE.AC_ TOR COOLANT SYSTEM BASES l 3/4.4.1 REACTOR COOLANT LOOPS, (continued) occur while the loop is isolated due to inadvertent opening of the isolated loop stop valves. Isolated loop startup is limited to Modes l 5 and 6 in accordance with the NRC SER on N-1 loop operation. l Verification of the isolated loop boron concentration prior to opening the isolated loop stop valves provides a reassurance of the I adequacy of the shutdown margin in the remainder of the system. Restoration of power to the hot leg stop valve allows opening this i l valve to complete the recirculation flowpath in conjunction with the relief line bypassing the cold leg stop valve and ensures adequate mixing in the isolated loop. This enables the temperature and boron concentration of the isolated loop to be brought to equilibrium with the remainder of the system. Limiting the temperature differential between the isolated loop and the remainder of the system prior to opening the cold leg stop valve prevents any significant reactivity effects due to cool water addition to the core. Startup of an idle loop will inject cool water from the loop into the core. The reactivity transient resulting from this cool wat ir injection is minimized by delaying isolated loop startup until its temperature is within 20'F of the operating loops. Making the reactor suberitical prior to loop startup prevents any power spike which could result from this cool water induced reactivity transient. 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 345,000 lbs. per hour of saturated steam at the valve set point. The relief capacity of a single safety valve is adequate to reliave any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves. BEAVER VALLEY - UNIT 1 B 3/4 4-la Amendment No.173
N DPR-66 REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3 /4. 4. 3 SAFETY VALVES (Continued) Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.4 PRESSURIZER i The requirement that (150)kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY. 1 3/4.4.5 STEAM GENERATORS One OPERABLE steam generator in a non-isolated reactor coolant loop provides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requirement for two OPEPABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RCS temperatures greater than 350*F if one steam generator becomes inoperable due to single failure considerations. Below 350*F, decay heat is removed by the RHR system. The Surveillance Requirements for inspection of the steam generator [ tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is 1 essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator-tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in strest corrosion cracking. The extent of cracking during plant BEAVER VALLEY - UNIT 1 B 3/4 4-2 Amendment No.173
.m 4 DPR-66 REACTOR COOLANT SYSTEM . BASES 3/4.4.5 STEAM GENERATORS (Continued) operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator). Cracks having a primary-to-secondary leakage less than this limit during' operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging or repair will l be required of all tubes with imperfections exceeding the plugging or repair limit. Degraded steam generator tubes may be repaired by the installation of sleeves which span the degraded tube section. A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded, therefore, the sleeve is considered a part of the tube. The surveillance requirements identify those sleeving methodologies approved for use. If an installed sleeve is found to have through wall penetration greater i than or equal to the plugging limit, the tube must be plugged. The plugging limit for the sleeve is derived from R.G. 1.121 analysis which utilizes a 20 percent allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional i degradation growth. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness. j Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. BEAVER VALLEY - UNIT 1 B 3/4 4-2a Amendment No.173
pa *Ec ?t UNITED STATES [; j NUCLEAR REGULATORY COMMISSION W 's WASHINGTON, D.C. 20G55-0001 \\..m [ DUOVESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 License No. NPF-73 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Duquesne Light Company, et al. (the licensee) dated December 30, 1992, as supplemented March 12, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
4 4 h 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-73 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 52, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. DLC0 shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION -{ dfalOSM Walter R. Butler, Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 1, 1993 i I
---A ATTACHMENT TO LICENSE AMENDMENT NO.52 FACILITY OPERATING LICENSE NO. NPF-73 ~ DOCKET NO. 50-412 Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. Remove Insert 3/4 4-11 3/4 4-11 3/4 4-12 3/4 4-12 3/4 4-13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-14a 3/4 4-14b 3/4 4-16 3/4 4-16 8 3/4 4-3 B 3/4 4-3 B 3/4 4-3a 9 i
..mJ : NPF-73 REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATORS LIMITING CQJiDITION FOR OPERATION i t 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,y above 200*F. SURVEILLANCE REQUIREMENTS 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting' and inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample
- size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2.
The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Steam generator tubes shall be examined in accordance with Article 8 of Section V (" Eddy Current Examination of Tubular Products") and Appendix IV to Section XI (" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing") of the applicable year and addenda of the ASME Boiler and Pressure Vessel Code required by 10CFR50, Section 50.55a(g). When applying the exceptions of 4.4.5.2.a through 4.4.5.2.c, previous defects or imperfections in the area repaired by sleeving are not considered an area requiring reinspection. The tubes selected for each inservice inspection shall include at least 3 percent of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50 percent of the tubes inspected shall be from these critical areas. b. The first sample of tubes selected for each inservice l inspection (subsequent to the preservice inspection) of each steam generator shall include: BEAVER VALLEY - UNIT 2 3/4 4-11 Amendment No. 52
..a NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 17 All nonplugged tubes that previously had detectable wall penetrations greater than 20 percent, and l 2. Tubes in those areas where experience has indicated potential problems, and 3. At 'least 3 percent of the total number of sleeved tubes in all three steam generators. A sample size less than 3 percent is acceptable provided all the sleeved tubes in the steam generator (s) examined during the refueling outage are inspected. These inspections will include both the tube and the sleeve, and 4. A tube inspection pursuant to Specification 4.4.5.4.a.8. If any selected tube does not permit the passage of the eddy current probe for a tube or sleeve inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection. c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided: 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and 2. The inspections include those portions of the tubes where imperfections were previously found. The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results C-1 Less than 5 percent of the total tubes inspected are degraded tubes and none of the inspected tubes are defective. C-2 One or more tubes, but not more than 1 percent of the total tubes inspected are defective, or between 5 percent and 10 percent of the total tubes inspected are degraded tubes. BEAVER VALLEY - UNIT 2 3/4 4-12 Amendment No.52
r "*w f NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) ~ C-3 More than.10 percent of the total tubes inspected are degraded tubes or more than 1 percent of ' the - 1 inspected tubes are defective. Note: In all inspections, previously degraded tubes or sleeves must exhibit significant (greater than 10 percent) further. wall penetrations to be included in the above percentage calculations. 4.4.5.3 Inspection Frecuencies The above required inservice inspections of steam generator tubes shall be performed at the following frequencies: a. The first inservice inspection shall be performed after 6 Ef fective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall' be performed at intervals of not less than-12 nor more'than 24 calendar months after the previous inspection. If two consecutive inspections following service under All Volatile Treatment (AVT) conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections-demonstrate that previously observed degradation has not: continued and no additional degradation'has. occurred, the inspection interval may be extended to a maximum of once per 40 months. b. If the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required. c. Additional, unscheduled inservice inspections shall-be performed on each steam generator in accordance with the first sample inspection.specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions: l 1. Primary-to-secondary tube leaks (not including leaks l originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, 2. A seismic occurrence greater than the Operating Basis i Earthquake, - 1 BEAVER VALLEY - UNIT 2 3/4 4-13 Amendtnent No.52
NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 37 A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4. A main steam line or feedwater line break. 4.4.5.4 Acceptance Criteria a. As used in this Specification: 1. Imperfection means an exception to the dimensions, finish or contour of a tube or sleeve from that l required by fabrication drawings or specifications. Eddy-current testing indications below 20 percent of the nominal tube wall thickness, if detectable, may be considered as imperfections. 2.
Dearadation means a service-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube or sleeve. l 3. Dearaded Tube means a tube or sleeve containing l imperfections greater than or equal to 20 percent of the nominal wall thickness caused by degradation. i 4. Percent Dearadation means the percentage of the tube or sleeve wall thickness affected or removed by t degradation. 5. Defect means an imperfection of such severity that it exceeds the plugging or repair limit. A tube l containing a defect is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube. 6. Pluacina or Repair Limit means the imperfection depth l at or beyond which the tube shall be removed from service by plugging or repaired by sleeving in the affected area because it may become unserviceable prior to the next inspection. The plugging or repair limit i imperfection depths are specified in percentage of nominal wall thickness as follows: a. Original tube wall 40% b. Babcock & Wilcox kinetic welded sleeve wall 40% BEAVER VALLEY - UNIT 2 3/4 4-14 Amendment No.52
= -a. NPF-73 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 7 '- Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above. 8. Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support to the cold leg. 9. Tube Repair refers to sleeving which is used to maintain a tube in-service or return a tube to service. This includes the removal of plugs that were installed as a corrective or preventive measure. The following sleeve designs have been found acceptable: a) Babcock Wilcox kinetic welded
- sleeves, BAW-2094P, Revision 1 including kinetic sleeve
" tooling" and instal]ation process parameter changes. b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or repair all tubes exceeding the plugging or repair limit) required by Table 4.4-2. 4.4.5.5 Reports a. Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged or repaired in each steam generator shall be reported to the Commission in a Special Report pursuant to l Specification 6.9.2. b. The complete results of the steam generator tube and sleeve inservice inspection shall be submitted to the Commission in l a Special Report pursuant to Specification 6.9.2 within 12 months following the completion of the inspection. This Special Report shall include: 1. Number and extent of tubes and sleeves inspected. i 2. Location and percent of wall-thickness penetration for each indication of an imperfection. 3. Identification of tubes plugged or repaired. I BEAVER VALLEY - UNIT 2 3/4 4-14a Amendment No.52
. _. _ = _ J. NPF-73 REACTOR COOLANT SYSTFJ t SURVEILLANCE REQUIREMENTS (Continued) c. Results of steam generator tube inspections which fall into i Category C-3 shall be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. The written report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence. 6 BEAVER VALLEY - UNIT 2 3/4 4-14b Amendment No. 52 ]
NPF-73 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTIOh 4 Sample. Size Result Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A N/A S Tubes per S.G. C-2 Plug or repair C-1 None N/A N/A l defective tubes and l inspect additional C-2 Plug or repair defective C-1 None 25 tubes in this S.G. tubes and inspect additional 4S tubes in C-2 Plug or repair l this S.G. defective tubes C-3 Perform action for C-3 result of first sample C-3 Perform action for C-3 N/A N/A result of first sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug or S.G.s are repair defective C-1 tubes and inspect 2S tubes in each other S.G. Son.e S. G. s Perform action for N/A N/A i C-2 but no C-2 result of second tification to NRC additional sample pursuant to 50.72 S.G.s are (b)(2) of 10 CFR C-3 Part 50 Additional Inspect all tubes in S.G. is each S.G. and plug or N/A N/A C-3 repair defective tubes. Notification to NRC pursuant to 50.72 (b) (2) of 10 CFR Part 50 2 s-n Where n is the number of steam generators inspected during an inspection. BEAVER VALLEY - UNIT 2 3/4 4-16 Amendment No. 62 (. bt 1
l NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continued) decay heat removal capabilities for RCS temperatures greater than 350*F if one steam generator becomes inoperable due to single failure considerations. Below 350*F, decay heat is removed by the RHR system. The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to
- design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken. The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage 500 gallons per day per steam = generator). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulate.d accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per ' steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged. Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging or repair will be required of all tubes with imperfections exceeding the plugging or repair limit. Degraded steam generator tubes may be BEAVER VALLEY - UNIT 2 B 3/4 4-3 Amendment No.52
_a NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (Continuedi repaired by the installation of sleeves which span the degraded tube section. A steam generator tube with a sleeve installed meets the structural requirements of tubes which are not degraded, therefore, the sleeve is considered a part of the tube. The surveillance requirements ideritify those sleeving methodologies approved for use. If an installed sleeve is found to have through wall penetration greater than or equal to the plugging limit, the tube must be plugged. The plugging limit for the sleeve is derived from R.G. 1.121 analysis which utilizes a 20 percent allowance for eddy current uncertainty in determining the depth of tube wall penetration and additional degradation growth. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20 percent of the original tube wall thickness. Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pursuant to Specification 6.6 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary. BEAVER VALLEY - UNIT 2 B 3/4 4-3a Amendment No.52 -}}