Text

LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Appendix a Through Appendix N
	ML081330088
Person / Time
Site:	LaSalle
Issue date:	04/14/2008
From:	; Exelon Generation Co, Exelon Nuclear
To:	; Office of Nuclear Reactor Regulation
References
	RS-08-045
	Download: ML081330088 (520)
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LSCS-UFSAR A.0-1 REV. 13 APPENDIX A GLOSSARY Appendix A of the FSAR contained a glossary defining key terminology and acronyms. This glossary is includ ed in the UFSAR as Section 1.2.4.

B.0-1 REV. 14, APRIL 2002 APPENDIX B CONFORMANCE TO REGULATORY GUIDES B.0 INTRODUCTION This Appendix to the UFSAR contains summary information on the Applicant's assessment of plant conformance to NRC Regulatory Guides 1.01 through 1-123. The following definitions are intended to assist the evaluator his interpretation of this Appendix:

The phrase Assessed Capability in the Design identifies those Regulatory Guide provisions for which a design assessment was performed by the Applicant, through the A-E, NSSS supplier, and/or equipment vendors to determine the capability of the previo usly approved design to meet the Regulatory Guide requirements.

The phrase Incorporated in the Design identifies those Regulatory Guide provisions which were included in th e LaSalle County Station (LSCS) Units 1 and 2 design commitments during the construction permit review, or those provisions of more recently published Regulatory Guides which were subsequently included in the LSCS design.

Conformance to the provisions of the Regulatory Guides is indicated under either of two general categories, full compliance or comp liance with the intent or objectives of UFSAR Regulatory Guide via the alternate approach cited in the UFSAR.

Phrases indicating compliance with guidance mean that the provisions of the Regulatory Guides are met by direct conformance or by the assessed capability of the design.

Phrases indicating compliance with the intent of Regulatory Guide provisions via an alternate approach, acknowledge the NRC Rules which state that "Regulatory Guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions different from those set in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission." Where this method is employed, the description and justification of an alternate approach is provided. Engineering calculations and analytic details are not included in the UFSAR but are part of the de sign backup for the LSCS plant.

NRC letter of July 1, 1975 from P.C. Young to Byron Lee, Jr. acknowledged that the cut-off date for addressing Regulatory Guides for the LSCS Units 1 and 2 FSAR was 31 December 1975 (8 months prior to the in itial tendering of the FSAR, which was 31 August 1976).

B.0-2 REV. 14, APRIL 2002 REGULATORY GUIDE 1.1 - REV. 0 NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS Regulatory Guide 1.1 requires that containment overpressure not be relied upon to provide adequate net positive suction head (NPSH) for emergency core cooling and containment heat removal system pumps.

The emergency core cooling system pumps, which include the containment heat removal system pumps take suction from the suppression pool. Calculations for determining the available net positive su ction head (NPSH) to these pumps are based on saturated water in the suppression pool, and will therefore be independent of containment pressure. The increase in containment pressures during a loss-of-coolant accident (LOCA) is not included in the determination of the available NPSH for these pumps. Available NPSB was determined by subtracting the suction line friction losses from the available static head. Pump performance tests verified that the available NPSH exceeds the required NPSH for rated flow conditions. Refer to Subsection 6.3.2.

Requirements for NPSH available at the centerline of the pump suction nozzles for each pump are given in Figures 6.3-1 (HPCS), 6.3-4 (LPCS), and 6.3-8 (LPCI).

Pump characteristics curves are given in Figures 6.3-3 (HPCS), 6.3-6 (LPCS), and 6.3-9 (LPCI).

The requirements of this guide are incorporated in the LSCS design.

REGULATORY GUIDE 1.2 - REV. 0 THERMAL SHOCK TO REAC TOR PRESSURE VESSELS Regulatory Guide 1.2 states that potential reactor pressure vessel brittle fracture which may result from emergency core cooling system operation need not be reviewed in individual cases if no significant changes in presently approved core and pressure vessel designs are proposed. Should it be concluded that the margin of safety against reactor pressure vessel brittle failure due to emergency core cooling system operation is unacceptable, an engineering solution, such as annealing, could be applied to assure adequate recovery of the fracture toughness properties of the vessel material. This regulatory guide requires that available engineering solutions be outlined and requires that it be demonstrated that the design does not preclude their use.

B.0-3 REV. 14, APRIL 2002 An investigation of the structural integrity of boiling water reactor pressure vessels during a design-basis accident (DBA) has been conducted (refer to NEDO-10029, "An Analytical Study on Brittle Fracture of GE-BWR Vessel Subject to the Design Basis Accident"). It has been determin ed, based upon methods of fracture mechanics, that no failure of the vessel by brittle fracture as a result of a DBA will occur. The investigation included:

a. a comprehensive thermal analysis considering the effect of blowdown and the low-pressure coolant injection (LPCI) system ref boding;

b. a stress analysis considering the effects of pressure, temperature, seismic load, jet load, dead weight, and residual stresses;

c. the radiation effect on material toughness (NDTT shift and critical stress intensity); and

d. methods for calculating crack tip stress intensity associated with a non-uniform stress field following the design-basis accident.

This analysis incorporated very conservative assumptions in all areas (particularly in the areas of heat transfer, stress analysis effects of radiation on material toughness, and crack tip stress intensity). Therefore, the results reported in NEDO-10029 provide an upper bound limit on brit tle fracture failure mode studies.

Because of the upper bound approach, it is concluded that catastrophic failure of the pressure vessel due to the DBA is shown to be impossible from a fracture mechanics point of view. In the case studies, even if an acute flaw does form on the vessel inner wall, it will not propagate as the result of the DBA.

The criteria of 10 CFR 50 Appendix G are interpreted as establishing the requirements for annealing. Paragraph IV C of Appendix G requires vessels to be designed for annealing of the beltline only where the predicted value of adjusted RT exceeds 200°F. This predicted value is no t exceeded, therefore design for annealing is not required.

The reactor pressure vessels utilized for LSCS Units 1 and 2 employ no significant core or vessel design changes from previously approved BWR Pressure Vessels such as all units of Browns Ferry.

Reference Subsection 5.2.3.3.1.

The requirements of this guide are incorporated in the LSCS design.

B.0-4 REV. 14, APRIL 2002 REGULATORY GUIDE 1.3 - REV. 2 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILIN G WATER REACTORS Regulatory Guide 1.3 provides acceptable assumptions that may be used in evaluating the radiological consequences of a loss-of-coolant accident for a boiling water reactor.

The assumptions used for evaluating the radiological consequences of a LOCA in this UFSAR are in conformance with the requ irements of this regulatory guide. For details see Chapter 2..0 and Section 15.6.

The key implementation assumptions used for LSCS were as follows:

a. Meteorology per Regulatory Position C.2h(4); yes.

b. Containment Leak Rate .635% per day.

c. SGTS Filter Efficiency 90% for all iodine forms.

d. Leakage to Unprocessed Area is essentially zero; yes;

e. Leakage to Processed and Mixed Area is 92% of the containment leak; no.

Most of the models and conditions that are prescribed in this guide are demonstrably inconsistent with actual physical phenomena. The attendant conservative bias impacts on design decisions throughout the total plant. For this reason additional analyses are provided in Subsection 15.6.5 which utilize both NRC and realistic assumptions to demonstrate the conservative bias in the regulatory guide requirements. In either case, the doses evaluated are within regulatory limits specified.

The requirements of this guide were in corporated in the evaluation of LSCS accidents.

B.0-5 REV. 14, APRIL 2002 REGULATORY GUIDE 1.4 - REV. 2 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS This regulatory guide is not applicable to the LSCS.

REGULATORY GUIDE 1.5 - REV. 0 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A STEAMLINE BREAK ACCIDENT FOR

_BOILING WATER REACTORS Regulatory Guide 1.5 provides acceptable assumptions that may be used in evaluating the radiological consequences of a steamline break accident for boiling water reactors.

The applicant incorporated the guidelines given in this Regulatory Guide 1.5, and assumed a 5.5 second closure time for the steamline isolation valves.

Other key implementation assumptions used in the analyses for LSCS were as follows: a. All Regulatory Position requirements included in the NRC solution.

b. Site Boundary as near as 509 m.

c. LPZ as near as 6400 in.

Some of the models and conditions that are prescribed are demonstrably inconsistent with actual physical phenomen

a. For this reason additional analyses are provided in Subsection 15.6.4 of th e UFSAR which utilize realistic assumptions to demonstrate the Conservative bias in the regulatory guide requirements. In either case, regardless of model used for evaluation, the resultant dose is less than the specified regulatory limits.

Estimated doses from steamline breaks fo r ranges b and c above are dominated by estimated consequences from LOCA analysis.

Appendix C provides the solution for high energy line breaks outside the containment.

The provisions of this guide were incorporated in the analyses of potential radiological consequences from this type of accident.

B.0-6 REV. 14, APRIL 2002 REGULATORY GUIDE 1.6 - REV. 0 INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE1 POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS Regulatory Guide 1.6 describes an acceptable degree of independence between redundant standby (onsite) power sources and between their distribution systems as required by GDC-17.

Conformance with Regulatory Guide 1.6 is described in the following subsections for each regulatory position of Paragraph D of that guide.

Independent Load Assignment Each Class 1E load (ac or dc) is assigned to a Division 1, 2, or 3 load group.

Assignment is determined by the nuclear sa fety functional redundance of the load.

Usually one (two at the most) division load group is required to perform nuclear safety functions. The loss of any si ngle ESF division does not prevent the performance of the minimum safety functions required for a safe shutdown.

Independent A-C Sources Each division a-c load group has a feed from ~ least one system auxiliary transformer (offsite) and one diesel genera tor (onsite), as shown in Figures 8.1-2 and 8.1-3. Divisions 1 and 2 are each prov ided with two off site power connections and Division 3 (HPCS) is provided with one off site power connection.

The diesel-generator breaker can be closed automatically only if the other source breakers to that load group are open, as shown in Figure 8.3-3.

Independence of D-C Sources Each division d-c load group has a feed from one battery charger and one battery as shown in Figures 8.3-9, through 8.3-12.

The redundant d-c load groups cannot be connected to each other. The d-c batteries and battery chargers cannot be connected to each other. Each d-c load group of one unit can be connected to the corresponding nonredundant d-c load group of the second unit and satisfy the temporary design load.

Independence of Standby Sources The diesel-generator breaker can be cl osed to its associ ated load group automatically only if the other source br eakers to that load group are open, as shown in Figure 8.3-2.

B.0-7 REV. 14, APRIL 2002 When the diesel-generator breaker is closed, no other source breaker can be closed automatically, as shown in Figures 8.3-2 and 8.3-3. No other means exist for connecting redundant load groups with each other.

Each of the redundant load groups is fed from only one diesel generator, as shown in Figure 8.1-2. No means are provided for transferring loads between the diesel generators.

Sufficient interlocks are provided to prevent paralleling the diesel generators manually by operator error during loss of o ff site power. Figures 8.3-2 and 8.3-3, and Subsection 8.3.1.1.2 describe the inte rlocks and interlock defeat mechanisms that prevent paralleling sources, including diesel generators.

Prime Mover Divisions 1, 2, and 3 diesel generators ar e provided with only one prime mover for each generator.

The HPCS standby power source and distribu tion system is redundant to the other two standby power sources and associated distribution systems in the plant.

No provisions exist for transferring loads from the HPCS power source to other power sources or transferring HPCS load groups either automatically or manually.

The provisions of this guide are incorporated into the LSCS design. Full compliance exists. REGULATORY GUIDE 1.7 - REV. 0 CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT Regulatory Guide 1.7 requires the cap ability be provided for (1) measuring hydrogen concentration, (2) mixing the at mosphere in the containment following a loss-of-coolant accident, (3) purging of the containment atmosphere through appropriate fission product removal systems, and (4) controlling combustible gas concentrations without reliance on purging.

Information on the control of containment gases for the post-LOCA situation is included in Subsection 6.2.5.

The LSCS design incorporates the guidance set forth in this regulatory guide; we believe that we comply with this guide.

B.0-8 REV. 14, APRIL 2002 REGULATORY GUIDE 1.8 - REV. 1R PERSONNEL SELECTION AND TRAINING Regulatory Guide 1.8 identifies acceptable criteria for the selection and training of nuclear power plant personnel.

LaSalle County Station's training and personnel selection is described in Section 13.2. We comply with the objectives set forth in the referenced revision of this regulatory guide, as indicated in the textual material mentioned above.

REGULATORY GUIDE 1.10 - REV. 1 MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES Regulatory Guide 1.10 provides acceptable procedures for splicing, crew qualification, visual inspection, tensile testing, tensile test frequency and procedure for handling substandard tensile test results for mechanical (cadweld) splices in reinforcing bars of Category I concrete structures.

Information on cadweld splices is provided in Appendix E.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.11 - REV, 0 INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT Regulatory Guide 1.11 describes an accept able basis on which to demonstrate the acceptability of instrument lines which penetrate or form a part of the reactor primary containment.

Instrument sensing lines that penetrate or connect to the primary reactor containment are designed to strict penetration requirements as discussed in Subsection 6.2.4 and shown in Table 6.2-21.

The provisions of this regulatory guide are incorporated into the design of the instrument sensing lines that penetrate or connect the primary containment at LSCS. We believe that we comply with the guidance set forth in this guide.

B.0-9 REV. 14, APRIL 2002 REGULATORY GUIDE 1.12 - REV. 1 INSTRUMENTATION FOR EARTHQUAKES Regulatory Guide 1.12 describes acceptable seismic instrumentation to determine promptly the seismic response of nuclear power plant features important to safety to permit comparison of such response with that used as the design basis.

The seismic instrumentation for LSCS utilizes three types of sens or-recorders with a playback capability available in the contro l room area. The location and function of these seismic devices were selected to provide adequately for the determination of seismic event loads into the structures via computerized analysis programs. Subsection 3.7.4 provides descriptive details on the seismic instrumentation.

The provisions of this guide were incorporated into the design, location, and function of the seismic instrumentation utilized at the LSCS.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.13 - REV. 1 SPENT FUEL STORAGE FACILITY DESIGN BASIS Regulatory Guide 1.13 requires that the fu el in pool storage be safeguarded from damage by missiles or the dropping of he avy loads from overhead cranes, that the fuel pool makeup system and filter system be designed to assure adequate water to cover the fuel, that low pool water level and high local radiation level be alarmed locally and in a continuously manned locati on, and that the pool be housed in a controlled leakage building having adequate ventilation and filtration to limit the potential release of radioactive materials.

The spent fuel pool, located inside the secondary containment, is designed to withstand the safe shutdown earthquake (SSE). (See Subsection 3.8.4 and Section 9.1.) The spent fuel storage racks are also designed to withstand the SSE.

A controlled leakage building is provided which encloses the fuel pool. The building is not designed to withstand extremel y high winds, but leakage is suitably controlled during refueling operations. The building is equipped with a ventilation and filtration system which is designed to limit the potential consequences of the release of that activity specified in Regulatory Guide 1.25 to those guidelines set forth in 10 CFR 100.

B.0-10 REV. 14, APRIL 2002 Detection of high radioactivity in the reactor building ventilation system initiates the standby gas treatment system. Analysis of a potential refueling accident is presented in Chapter 15.0.

The reactor building below the refueling floor is designed to be tornado proof as discussed in Sections 3.3 and 3.5.

The spent fuel shipping cask storage area is separated from the fuel pool by a spent fuel transfer canal. Under no circumstances is it necessary to transport the shipping cask over the fuel pool. A reactor building crane path control system and explicit crane operating procedures are employed to ensure that the shipping cask is not transported over the fuel pool. Administrative procedures also prohibit transport of heavy objects over the fuel pool. See Subsection 9.1.4 for further details. The refueling platform is designed to prevent it from toppling into the pools during an SSE. The design of the spent fuel pooling cooling and cleanup system is such that the failure of any component in the system would not drain the fuel pool or uncover the fuel. There are no drains in the fu el pool. (Refer to Section 9.1.)

Normal fuel pool makeup is provided from the cycle condensate storage system through a connection to the fuel pool sk immer surge tanks. In the event that a source of emergency makeup is required, a Seismic Category I portion of the CSCS equipment cooling water system (Subsection 9.2.1) is provided. This system is capable of supplying enough water to prevent the uncovering of the fuel. There is no permanent connection between the emergency makeup, fuel pool cooling, and filter/demineralizing systems.

A fuel pool water level alarm is provided in the control room. High radiation in the fuel pool area will start the standby gas tr eatment system and alarms in the control room. We believe that we comply with the guidan ce set forth in this regulatory guide.

B.0-11 REV. 14, APRIL 2002 REGULATORY GUIDE 1.14 - REV. 1 REACTOR COOLANT PUMP FLYWHEEL INTEGRITY Regulatory Guide 1.14 describes a method to minimize the potential for failure of flywheels of reactor coolant pump motors.

Flywheels are not used in the LSCS reactor coolant pumps or pump motors.

REGULATORY GUIDE 1.15 - REV. 1 TESTING OF REINFORCING BARS FOR CATEGORY I CONCRETE STRUCTURES Regulatory Guide 1.15 describes the tests and inspections of reinforcing bars for Category I concrete structures to assure that bars are tested to quality standards commensurate with the importance of th e safety function to be performed.

The testing of reinforcing bars for Seismic Category I structures is described in Appendix E.

We believe that we comply with the guidan ce set forth in this regulatory guide.

Regulatory Guide 1.16, Rev. 4 Regulatory Guide 1.16, Revision 4 as re vised by Generic Letter 97-02, "Revised Contents of the Monthly Operating Repo rt" presents an acceptable format for meeting the reporting requirements of the LaSalle Technical Specifications.

REGULATORY GUIDE 1.17 - REV. 1 PROTECTION OF NUCLEAR POSIER PLANTS AGAINST INDUSTRIAL SABOTAGE Regulatory Guide 1.17 describes physical security criteria that are generally acceptable for the protection of nuclear po wer plants against acts of industrial sabotage which could lead to a threat to the health and safety of the public.

The LaSalle County Station revised master security plan sets forth the principles, policies, and general requirements for industrial security at LSCS. Because of its sensitive contents it was separately submitted with the FSAR docket materials. It is B.0-12 REV. 14, APRIL 2002 intended to meet the requirements of the proposed 10 CFR 73.55. Implementing security procedures for onsite execution of the LSCS master security plan are a part of the station operating procedures. A reference is provided in Section 13.7.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.18 - REV. 1 STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS Regulatory Guide 1.18 describes an acceptable method of demonstrating the capability of a concrete primary containment to withstand postulated pressure loads.

The LSCS structural acceptance test is des cribed in Subsection 3.8.1.7. The concrete primary containment structure was designed as a Seismic Category I structure as described in Sections 3.7 and 3.8.

We believe that we comply with the struct ural acceptance test requirements as described in this regulatory guide.

REGULATORY GUIDE 1.19 - REV. 1 NONDESTRUCTIVE EXAMI NATI ON OF PRIMARY CONTA INMENT LINER WELDS Regulatory Guide 1.19 describes accept able procedures for demonstrating the leaktight integrity of the liner and penetrations of primary reactor containments of concrete construction by the use of nondestructive methods to examine the welds in the liner and penetrations.

Where the leak-chase-system channels are inst alled over liner welds, the channel- to liner-plate welds are tested for leaktightn ess by pressurizing the channels to the containment design pressure.

Further information concerning the examin ation of the primary containment liner welds are given in Appendix E.

We believe that we comply with the guidan ce set forth in this regulatory guide.

B.0-13 REV. 14, APRIL 2002 REGULATORY GUIDE 1.20 - REV. 2 VIBRATION MEASUREMENTS ON REACTOR INTERNALS Regulatory Guide 1.20 describes a comprehensive vibration assessment program for reactor internals during preoperational and initial startup testing. The vibration assessment program meets the requirements of Criterion 1, "Quality Standards and Records," of Appendix A to 10 CFR 50 and Section 50.34, "Contents of Applications; Technical Information," of 10 CFR 50.

An inspection test program has been defined for the confirmatory testing of this plant during preoperational flow tests based on the results of Browns Ferry and other vibration measurement programs. Core support structure and other reactor internals will be inspected at this plant for evidence of vibration following preoperational testing at specified recirculation flow conditions.

The Unit 1 pump adapter at its support plate, is a unique design, therefore it will be instrumented with vibration sinsois, as was done in the prototype plant, to qualify it to the same criteria as the remainder of the Unit 1 design.

Further information with respect to reacto r internals vibration inspection programs is given in Subsections 3.9.

2.3, 3.9.2.4, and 3.9.2.6.

We believe that we comply with the guidan ce set forth in this regulatory guide through incorporation of the alternate approach cited above.

REGULATORY GUIDE 1.21 - REV. 1 MEASURING, EVALUATING, AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT-WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.21 describes programs for measuring, reporting, and evaluating releases of radioactive materials in liqui d and gaseous effluents and guidelines for classifying and reporting the categories and curie content of solid wastes.

The process and effluent radiological monitoring and sampling system is designed to provide the monitoring and sampling capability required to make the measurements, evaluations, and reports recommended b this gu ide. Refer to Sections 7.3, 7.6, 7.7, and 11.5.

We believe that we comply with the guidan ce set forth in this regulatory guide.

B.0-14 REV. 14, APRIL 2002 REGULATORY GUIDE 1.22 - REV. 0 PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS Regulatory Guide 1.22 delineates acceptable design approaches that facilitate the periodic testing, during reactor oper ation, of actuation devices/equipment incorporated into the reactor protection system design.

The reactor protection system is designed so that it may be tested during plant operation from the sensor device to final actuator device. The test must be performed in overlapping portions so th at an actual reactor scram will not occur as a result of the testing. Refer to sections listed below for detailed information for the frequency of testing and minimum instrumentation availability.

Compliance for each system is described in the text as follows:

a. Reactor Protection System, Subsection 7.2.1.1;

b. Emergency Core Cooling System, Subsection 7.3.1.1;

c. PC RVICS, Subsection 7.3.2.2;

d. RCIC, Subsection 7.4.1.1;

e. Reactor Shutdown Cooling, Subsection 7 4.1.3;

f. Leak Detection System, Subsection 7.6.2.1;

g. HPCS Standby Power, Subsection 7.3.2.1;

h. RHR Containment Spray Cooling, Subsection 7.3.9.1;

i. Standby Liquid Control, Subsection 7.4.2.1;

j. Suppression Pool Cooling, Subsection 6.2.2.3; and

k. Process Radiation Monito r, Subsection 7.6.1.2.

We believe that we comply with the guidance set forth in this regulatory guide via NSSS analysis, design and equipment which is incorporated in the alternate approach cited above.

B.0-15 REV. 14, APRIL 2002 REGULATORY GUIDE 1.23 - REV. 0 ONSITE METEOROLOGICAL PROGRAMS Regulatory Guide 1.23 describes a suitable on site meteorological program to provide meteorological data needed to estimate potential radiation doses to the public as a result of the routine or accidental release of radioactive materials to the atmosphere and to assess other environmental effects.

The onsite meteorological program is discussed in Subsection 2.3.3. The accumulation of onsite meteorological data is extending into the second year. Correlations with the Dresden Nuclear Power Station and Peoria, Illinois weather station have established that LSCS meteorology is consistent with these stations. The original FSAR submittal utilized Dresden meteorology to indicate the type of diffusion and atmospheric dispersion attributable to the LSCS area. These data have now been superseded by a revised submi ttal in early 1978 which categorizes LSCS onsite meteorology into the typical joint frequency distributions by receptor sector around LSCS. Chi- over-Que's for the classical receptors were calculated for model and a realistic assessment of receptor dose via the NRC's models in Regulatory Guide 1.111. Of particular importance is the wind velocity and stability parameters applicable for the elevated single release point (station vent stack). The wind instrumentation was moved from the 33-foot level to the 200-foot level to provide a backup capability for the wind instruments at 375 fe et. This also minimizes th e building work problem for the north sector winds. Differential temp erature data for the atmospheric stability parameter now references33-375 feet; wind data references 375 feet (with backup at 200 feet). These data characterize the me teorology conditions for both normal and accidental releases. Adequate instrumentat ion is provided and sufficient operating experience will be accrued by the fuel loading date to assure the availability of reliable meteorological data for the analys is of routine and postulated accidental releases of gaseous radioactivity. Refine ments in the LSCS meteorological base will be made as the cumulative data is acquired.

We believe that we comply with the objectiv es set forth in this regulatory guide.

REGULATORY GUIDE 1.24 - REV. 0 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PRESSURIZED WATER REACTOR RADIOACTIVE GAS S'IORAGE TANK FAILURE Since LSCS uses a boiling water reactor, this guide is deemed not applicable.

B.0-16 REV. 14, APRIL 2002 REGULATORY GUIDE 1.25 - REV. 0 ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AND STORAGE FACILITY FOR BOILING AND PRESSURIZED WATER REACTORS Regulatory Guide 1.25 gives assumptions acceptable to the NRC that may be used in evaluating the radiological consequences of a postulated fuel handling accident resulting in damage to the fuel cladding and the subsequent release of radioactive materials.

A consonnant analysis of a fuel handling accident is provided in Subsection 15.7.4.

The key assumptions used in the analyses are as follows:

a. Site Boundary as near as 509m.

b. LPZ as near as 6400m.

c. SGTS Filter Efficiency 90% for all iodine forms.

d. All activity released to the environment is via the SGTS, yes.

e. 125 Fuel Rods damaged, yes.

f. The dose estimates at the site boundary and LPZ are dominated by the radiological consequences of Regulatory Guide 1.3 LOCA Analysis.

Parts of the models and conditions that ar e prescribed in Regulatory Guide 1.25 are demonstrably inconsistent with physical phenomena and for this reason additional realistic analyses are provided in Su bsection 15.7.4 of the LSCS UFSAR to demonstrate the conservative bias of the regulatory requirements.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.26 - REV. 3 QUALITY GROUP CLASSIFICATIONS AND STANDARDS FOR WATER-, STEAM-, AND RADIOACTIVE-WASTE-CONTAINING COMPONENTS OF NUCLEAR POWER PLANTS Regulatory Guide 1.26 describes a quality classification system for determining acceptable quality standards for safety-relat ed components containing water, steam, B.0-17 REV. 14, APRIL 2002 or radioactive material other than those components addressed in Section 50.55a of 10 CFR 50.

The definition of quality group classifications for LSCS was made initially and recorded in the PSAR in accordance with ASME Boiler and Pressure Vessel Code, Sections III and VIII. Quality group classi fications have been maintained during design and construction and are actively maintained during plant operations and modifications commensurate with the safety functions performed by the safety-related components. See Sections 3.9 and 5.2. This regulatory guide is applicable to Quality Groups B through D for pressure parts including piping, pumps, valves and vessels. Section 3.2 shows the Quality Group classification s of these parts.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.27 - REV. 1 ULTIMATE HEAT SINK FO R NUCLEAR POWER PLANTS Regulatory Guide 1.27 requires the ultima te heat sink (UHS) to be capable of providing sufficient cooling for at least 30 da ys based on the worst period of recorded weather conditions and to be capable of withstanding 1) the most severe natural phenomena taken individually, 2) the site-related events that historically have occurred or may occur, 3) reasonably pro bable combinations of less severe natural phenomena and/or site-related events, and 4) a single failure of man-made structural features.

The ultimate heat sink consists of an impoundment which holds 460 acre-feet of water with a surface area of 83 acres. In the unlikely event that the main dike is breached, the analysis in Subsection 9.2.6 supports the availability of a 30-day supply of water to ensure that design temper atures are not exceeded. It is based on the record period of worst weather condit ions in accord with the guidelines. Additional information is provided in Sectio ns 2.4 and 2.5, respec ting natural events and their effects on the UHS.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.28 - REVISION 3 QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION) Regulatory Guide 1.28 states that the general requirements and guidelines for establishing and executing a quality assurance program during the design and B.0-18 REV. 15, APRIL 2004 construction phases of nuclear power plan ts, which are included in ANSI N45.2 are acceptable and provide an adequate basis for compliance with Appendix B to 10 CFR

50. Chapter 17.0 cites a reference document NO-AA-10 which is the updated version of Exelon Generation Company's (EGC) Quality Assurance Program. This document was submitted as a topical report and wa s accepted on December 24, 2002 by the NRC as a docketed record adequate for cross-referencing in licensing matters.

The General Electric BWR Quality Assurance Program has been developed over the years such that at any point in time it has been in compliance with mandatory regulatory requirements such as 10 CFR 50, Appendix B, and the ASME Code. Implementation of the applicable ANSI-N45.2 series standards and the associated NRC Regulatory Guides (or NRC-approved GE alternate positions) has been an evolutionary process and although partial implementation has always been effected before the date of issue of the Regulatory Guide or "AEC Guidance on Quality Assurance" which recognized applicable ANSI standards, full implementation was not necessarily in place until the GE commitment date.

We comply with the objectives set forth in the referenced revision of this regulatory guide through the implementation of the id entified GE-BWR QA Program; S&L's QA Program; and the referenced Commonwealt h Edison QA Program under which all LSCS activity pertaining to nuclear safety has been conducted.

In a letter to Station Managers from C. Reed dated December 4, 1986, CECo committed to Regulatory Guide 1.28 Rev.

3 which endorses ANSI/ASME NQA-1-1983 and the ANSI/ASME NQA-1a-1983 Addenda.

NRC SER dated 12/24/02, approved Revision 70 of the Exelon Nuclear Quality Assurance Topical Report which utilizes ANSI/ASME NQA-1-1994 as the basis for Exelon Nuclear quality assurance program.

REGULATORY GUIDE 1.29 - REV. 2 SEISMIC DESIGN CLASSIFICATIONS Regulatory Guide 1.29 describes an acceptable method of identifying and classifying those features of light-water-cooled nuclear power plants that should be designed to withstand the effects of the SSE.

Information concerning the seismic design classification of structures, systems, and components of the LaSalle County Station, intended to withstand the effects of the SSE, is presented in Sections 3.

2, 3.7, 3.8, 3.9, and 3.10.

For the NSSS equipment this regulatory guid e was used as a basis for defining the systems and components which must meet Seismic Class I requirements. For the purpose of defining equipment that sh ould withstand the SSE, NSSS equipment B.0-19 REV. 14, APRIL 2002 conforms to the guide. Also the regulatory guide includes positions applicable to parts of the NSSS equipment as follows:

C.1(b) This guide is applicable to those reactor vessel internals which utilize engineered safety features, such as core spray piping, core spray spargers and hardware, etc.

The steam dryer, steam separator and feedwater spargers are not required to be Seismic Category I qualified components.

However, these components are designated Reliability Related at LaSalle and as such come under the control of the Quality Assurance Program for future modifications and repairs during refueling outages.

C.l(h) The component cooling water portions of the reactor recir culation pumps are not required to be Seismic Category I since the pumps do not perform a safety function.

C.1(i) The seismic classification of the off-gas charcoal decay system is based upon Paragraph C.l.i, Regulatory Guide 1.29, which requires that such systems be Seismic Category I unless simultaneous failure would not result in conservatively calculated potential offsite exposure comparable to th e guideline exposures of 10 CFR 100. GE Licensing Topical Report NEDO-l0734 is based upon calculations of release rates for off-gas charcoal delay system assuming seismically induced failure of the equipment. Offsite doses from component failure given in Table B-3 of that report show that the doses are well below the 10 CFR 100 limits.

Subsection 15.7.3 conservatively analyzes a postulated simultaneous failure of all the radwaste tanks in the radwaste building. The analysis assumes that 1% of the iodine is released to the atmosphere and at the time of failure, all tanks are filled to capacity (this condition is not expected). The analysis evaluates the possible control room site boundary and LPZ exposures to the whole body and thyroid. The results of the analysis indicate, in light of the requirements of Regulatory Guide 1.29, that the radwaste system is properly classified and changes are not required.

For the BOP equipment and design the follo wing clarification and exceptions are noted. C.l(d) The normal cooling system for the spent fuel storage pool is not a Seismic Class I system because the service-water side of the heat exchanger is not a Class I system. This system will be routinely used for coo ling the spent fuel storage pool, as the term normal implies. The emergency cooling system for the spent fuel storage pool is a B.0-20 REV. 14, APRIL 2002 Seismic Class I system when it is spooled into the loop for this service. Adequate post-SSE time is available to spool in this emergency system.

Alternately, a calculation of the radioiodine release from a boiling fuel pool has been made: thyroid doses are 4.5 x 10

-4 rem at the EAB and 6.0 x l0

-4 rem at the LPZ for 0-4 days inhalation. This shows that acceptable fuel storage exists even during boiling of the fuel storage pool without hazard to offsite personnel.

C.l (e) LaSalle has a special D+ category of piping (superpipe) between the outermost containment isolation valves and the turbine stop valves. It is structurally equivalent to Seismic Class B piping. It has an augmented inspection requirement. The turbine stop valves are not Seismic Class I valves.

The D+ category for LaSalle piping was defined and accepted via exemption at the PSAR review stage. This plant was desi gned, equipment was ordered and built, and installation of the non-safety-related turbine stop valve fixed approximately 3 years before this NRC position was formulated. Future maintenance of the superpipe will be under QA cognizance, but maintenance on the turbine stop valve will not be under QA coverage.

C.2 Those portions of structures, systems, or components whose continued function is not required but whose failure could reduce the functioning of any plant feature required for safe shutdown to an unacceptable safety level are designed and constructed so that the SSE would not cause such failure.

Plant features required for safe shutdown are described in Section H.4 "Safe Shutdown Analysis" of Appendix H.

We believe that we comply with the guidance set forth in the regulatory guide. The equipment and design utilized at LSCS as cited above, is in compliance with the intent of this guide via incorporation of the alternate approach specifically identified.

REGULATORY GUIDE 1.30 - REV. 0 QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING OF INSTRUMENTATION AND ELECTRIC EQUIPMENT Regulatory Guide 1.30 describes an acceptable method of complying with regulations regarding the quality assurance requirements for the installation, inspection, and testing of nuclear power plant instrumentation and electric equipment.

B.0-21 REV. 14, APRIL 2002 The testing, inspection, and proper installation of electrical and instrumentation equipment are described for various subsystems in Chapter 7.0.

For operation and maintenance, detailed station procedures cover inspection and testing of safety-related instrumentation and electrical equipment. They are reviewed and approved by the station superv isory and management staff prior to use.

The station activities affecting quality are accomplished in consonance with CE-lA by use of appropriate equipment, approval procedures, and station management controls.

For the NSSS, see compliance assessment made under Regulatory Guide 1.28 in this Appendix B.

We comply with the objectives set forth in the referenced revision of this regulatory guide because quality assurance for the installation, inspection, and testing of instrumentation and electric equipment ar e in accordance with ANSI N45.2.4 (1972) and IEEE 336-1971, as indicated in the textual material referenced above.

REGULATORY GUIDE 1.31 - REV. 1 CONTROL OF STAINLESS STEEL WELDING Regulatory Guide 1.31 describes an acceptable method of implementing requirements with regard to the control of welding when fabricating and joining austenitic stainless steel components and systems.

All austenitic stainless steel weld filler materials were supplied with a minimum of 5% delta ferrite. This amount of ferri te is considered adequate to prevent microfissuring in austenitic stainless steel welds.

An extensive test program performed by General Electric Company, with the concurrence of the Regulatory Staff, has demonstrated that controlling weld filler metal ferrite at 5% minimum produces production welds which meet the requirements of this regulatory guide.

A total of approximately 400 production welds in five BWR plants were measured and all welds met the requirements of the Interim Regulatory Position.

We believe that we comply with the intent of the guidance set forth in this regulatory guide via an assessed capability in the design and the related test program which demonstrated adequate welds.

B.0-22 REV. 15, APRIL 2004 REGULATORY GUIDE 1.32 - REV. 0 USE OF IEEE STANDARD 308-1971, "

CRITERIA FOR CLA SS 1E ELECTRIC SYSTEMS FOR NUCLEAR POWER GENERATING STATIONS

" Regulatory Guide 1.32 states that IEEE 3 08-1971 provides criteria that may be used in establishing some of the bases for the design of electric power systems, except that conflicts with General Design Criterion 17 should be resolved by:

a. provision of two immediate access circuits from the transmission network; and

b. the capacity of the battery char ger supply should be based on: the largest combined demands of the various steady-state loads and enough charging capacity to restore the battery from the minimum design charge state to a fully charged state, irrespective of the states of the plant. Two physically independent 3L15-kV transmission circuits (four transmission lines) occupying separate rights-of-way are available to each of the two LSCS reactor units so that each unit has an immedi ate access to a second of f site power source. Off site power is available to each unit from the systems auxiliary transformer of the opposite

unit via non-redundant bus ties between emer gency 4-ky buses. Switchyard power is available to both units as long as one of the four 3q5-kV transmission lines is available (Section 8.2).

Each Division 1 and 2 battery charger as well as each HPCS (Division

3) battery charger has sufficient capacity to restor e its battery to full charge under the maximum steady-state load (Section 8.3).

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.33 - REV. 2 QUALITY ASSURANCE PROGRAM REQUIREMENTS (OPERATION) Regulatory Guide 1.33 describes an acceptable method of complying with regulations regarding overall quality assurance program requirements for the operation of nuclear power plant structures, systems, and components.

The quality assurance program for nuclear generating stations is described in Exelon Nuclear Quality Assurance Program Topical Report NO-AA-10, referenced in Chapter 17.0. Under the alternate methods allowed by Regulatory Guide 1.33, the LSCS operating procedures were origina lly compiled per ANSI N18.7-l972 and ANSI B.0-23 REV. 15, APRIL 2004 N45.2-l97l; subsequently, they were upda ted, where applicable, to ANSI Nl8.7-1976 standards during station review and appr oval. Fulfillment of the recommendations of this regulatory guide indicates conformance to the requirements of the ANSI guidelines applicable to the BWR standardized operating procedures used at LaSalle.

For standardization purposes among Edison BWR plants, additional procedures were added and revisions were made to incorporate experience at similar operating stations such that uniform operating practi ces are utilized throughout Edison's BWR stations.

The Exelon Nuclear QA Program utilized for LSCS complies with the objectives set forth in the referenced revision of this regulatory guide, as indicated in the textual material mentioned above.

The NRC has approved the use of ANS 3.2, ANSI N18.7-1988 sections pertaining to biennial review of procedures, as defined in the Exelon Nuclear Quality Assurance Topical Report.

REGULATORY GUIDE 1.34 - REV. 0 CONTROL OF ELECTROS LAG WELD PROPERTIES

Regulatory Guide 1.34 describes an acceptable method of implementing requirements regarding control of weld properties when fabricating electroslag welds for nuclear components made of ferritic or austenitic materials.

Procurement specifications for ASME Code Section III Class 1 and 2 components require adherence to the Section IX, Wel ding Qualifications. Product quality certifications are required of vendors for applicable hardware. Records of qualifying tests and inspections are maintained by the vendor and are audited by the procuring contractor. Electroslag welding is not used on NSSS components; therefore this regulatory guide is not applicable thereto.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.35 - REVISION 3 INSERVICE INSPECTION OF UNGROUTED TENDONS IN PRESTRESSED CONCRETE CONTAINMENT STRUCTURES

Regulatory Guide 1.35 describes a surveillance program for "typical" containments which have vertical, hoop, and dome tendon

s. The LSCS containments do not have B.0-24 REV. 14, APRIL 2002 dome tendons. Therefore, the provisions of Regulatory Guide 1.35 are not directly applicable.

The LSCS tendon surveillance program is described in Sub section_3.8.1.7.

Regulatory Guide 1.70, Rev. 2 (09/75)

The previous commitment to this Regulatory Guide is superseded by the following.

As part of the ongoing effort to improve the quality of the UFSAR, the guidelines provided in NEI 98-03 Revision 1, 06/99 and endorsed by the NRC in Regulatory Guide 1.181, Revision 0, 09/99 will be used to further improve the content of the UFSAR. While the UFSAR will continue to follow the general organizational recommendations specified in this Regulatory Guide, the reformatting options described in NEI 98-03, Revision 1, will be used to simplify information contained in the UFSAR to improve their focus, clarit y and maintainability. Examples of the types of improvements that are anticipated are the identification of historical information and the removal or replacement of excessively detailed drawings as described in Appendix A to NEI 98-03 Revision1.

REGULATORY GUIDE 1.36 - REVISION 0 NONMETHLLIC THERMAL ISULATION FOR AUSTENITIC STAINLESS STEEL Regulatory Guide 1.36 requires that levels of leachable contaminants in nonmetallic insulation materials that come in contact with austenitic stainless steels of the AISI 3XX series used in fluid systems important to safety be carefully controlled so that stress-corrosion cracking is not promoted.

Reflective metallic insulation is used around all stainless steel Class I process fluid transporting equipment. Control of halogens is precisely specified in all materials which contact the austenitic stainless steels during fabrication, storage, erection, testing and service life. Use of nonmetallic thermal insulation is restricted to non-stainless steel parts of the plant.

We believe that we comply with the guidan ce set forth in this regulatory guide.

B.0-25 REV. 14, APRIL 2002 REGULATORY GUIDE 1.37 - REV. 0 QUALITY ASSURANCE REQUIREMENTS FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.37 states that requirements and recommendations for onsite cleaning of materials and components, clea nness control, and preoperational cleaning and lay-up of water-cooled nuclear power plant fluid systems included in ANSI N45.2.1-1973 provide an adeq uate basis for complying with pertinent quality assurance requirements of Appendix B to 10 CFR 50, subject to the following:

a. applicability and acceptability of references addressed elsewhere, b. should be used during operatio n phase of a nuclear power plant if applicable, c. water quality for final flushes to be at least equivalent to quality of operating system water, d. chemical compounds that could cont ribute to intergranular cracking or stress- corrosion cracking should not be used on austenitic stainless steels or nickel-based alloys, e. tools which contain materials that could contribute to intergranular cracking or stress-corrosion cracking or which may have become contaminated with such materials should not be used on the surfaces of corrosion-resistant alloys, and

f. ASTM A 393-63 is no longer a valid test.

Onsite cleaning of materials and components, cleanness control, and preoperational cleaning and lay-up of fluid systems for LSCS are based on ANSI N145.2.1 and acknowledge the specific guidelines enumerat ed. Cleaning and flushing instructions are incorporated in the criterion tests in addition to procedures for cleaning the plant's fluid systems.

Therefore we comply with the objectives set forth in the referenced revision of this regulatory guide.

B.0-26 REV. 14, APRIL 2002 REGULATORY GUIDE 1.38 - REV. 2 QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.38 describes a method acceptable to the NRC for complying with regulations regarding the quality assurance requirements for the packaging, shipping, receiving, storage, and handling of items for water-cooled nuclear power plants. Packaging, shipping, receiving, storage, and handling of BWR power plant equipment are covered by quality specificat ions based on ANSI N45.2.2-1972. These practices are audited for compliance in accordance with the Commonwealth Edison Quality Assurance Program previously refere nced. Compliance with these guidelines extends beyond the construction phase into the operational phase and are followed where appropriate.

Therefore we comply with the objectives set forth in the referenced revision of this regulatory guide.

REGULATORY GUIDE 1.39 - REV. 1 HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.39 states that the requirements and guidelines for the control of housekeeping activities, conditions and environments at water-cooled nuclear power plant sites which are included in ANSI N45.2.3-1973 are acceptable and provide an adequate basis for complying with pertin ent quality assurance requirements of Appendix B to 10 CFR 50, subject to the following:

a. Applicability and acceptability of ANSI references as addressed elsewhere in commission regulations.

b. Requirements should be considered applicable to the operation phase of a nuclear power plant.

Control of facility cleanness, care of material and equipment, fire prevention and protection, disposal of debris, protection of material and control of access are all standard practices of the station construction department during construction. After transfer of the plant to the operating sta ff, these functions are the responsibility of the production department. Normal mana gement attention and periodic audits B.0-27 REV. 14, APRIL 2002 under the CEC0 QA program will provide the desired result at LSCS. Independent audits by NRC Region III personnel also co ntribute to the effectiveness of good housekeeping practices.

We comply with the objectives set forth in the referenced revision of this regulatory guide. REGULATORY GUIDE 1.40 - REV. 0 QUALIFICATION TESTS OF CONTINUOUS-DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.40 states that the proced ures for conducting qualification tests of continuous-duty motors installed inside the containment of water-cooled nuclear power plants which are specified by IEEE 334-1971 are acceptable, subject to the following:

a. Auxiliary equipment that will be part of the installed motor assembly should also be qualified hereunder.

b. Tests should simulate as closely as practicable all design basis events which affect operation of the motor's auxiliary equipment.

c. Applicability and acceptability of references will be addressed elsewhere.

The recirculation pump is part of the RCPB and is designed to ASME III Class 1 requirements, however, the motor is not Class 1E. Performance of the safety function (i.e., coastdown) of the pump and motor combination is assured by specifying the following additional requirements:

a. A LOCA shall not degrade puirp and motor coastdown performance in the unbroken loop to the extent that the core is deprived of adequate cooling. b. The pump and motor bearing shall have sufficient dynamic load capability at rated operating conditio ns to withstand the safe shutdown earthquake.

There are no NSSS Class 1E safety-related continuous duty motors within the containment of this plant.

B.0-28 REV. 14, APRIL 2002 REGULATORY GUIDE 1.41 - REV. 0 PREOPERATIONAL TESTING OF REDUNDANT ONSITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS Regulatory Guide 1.41 describes an acceptable testing program to verify the existence of independence among redundant onsite power sources and their load groups. Each standby power system is designed to be tested independently of any other redundant load group in conformance with the provisions of this regulatory guide.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.42 - REV. 0 INTERIUM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGH T WATER-COOLED NUCLEAR POWER REACTORS This guide is no longer applicable.

REGULATORY GUIDE 1.43 - REV. 0 CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW-ALLOY STEEL COMPONENTS Regulatory Guide 1.43 describes acceptable methods for the selection and control of welding processes used for cladding ferritic steel components with austenitic stainless steel to restrict practices that could result in underclad cracking.

Procurement specification for ASME Code Section III structures and components using stainless steel cladding on low alloy steel require adherence to ASME Section IX, Welding Qualifications. Product certif ications are required of vendors for applicable hardware. Records of fabricatio n and qualification tests and inspections are maintained by the vendor. These and the process controls and performance test results are audited by the procuring contractor (such as GE for the NSSS)

B.0-29 REV. 14, APRIL 2002 The stainless steel clad on the blend radii of the principal RPV nozzles is being removed by a special cutting machine. Stainless steel safe-ends were replaced with carbon steel safe-ends.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.44 - REV. 0 CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL Regulatory Guide 1.44 describes acceptable methods for the control of the application and processing of stainless steel to avoid severe sensitization that could lead to stress corrosion cracking.

The purpose of this guide is to address 10 CFR 50 Appendix A, GDC's 1 and 4, and Appendix B requirements to control "the application and processing of stainless steel to avoid severe sensitization that could lead to stress corrosion cracking." The guide proposes that this should be done by limiting sensitization due to welding as measured by ASTM A262 Practice A or E, or another method that can be demonstrated to show nonsensitization in austenitic stainless steels.

Tests by General Electric indicate that th e test specified by A262 A or E (Detecting Susceptibility to Intergranular Attack in Stainless Steel) detects sensitization in a gross way, and that the tests do not provide a precise method of predicting susceptibility to stress corrosion cracking in the BWR environment.

All austenitic stainless steel for LSCS Units 1 and 2 was purchased in the solution heat treated condition in accordance with applicable ASME and ASTM specifications.

Carbon content was limited to 0.08% maximum, and cooling rates from solution heat treating temperatures were required to be rapid enough to prevent sensitization.

Welding heat input was restricted to 110,000 joules per inch maximum, and interpass temperature to 350° F. High heat welding processes such as block welding and electroslag welding were not permitted. All weld filler metal and castings were required by specification to have a minimu m of 5% ferrite. As deposited austenitic welds were controlled to have at least 3% fe rrite content. See Subsection 5.2.3.14 for specific reference to control of the use of sensitized stainless steel. Additionally, grinding of field erection welds was prohibited on the reactor fluid side of Class 1 stainless steel pressure boundary pipe.

Whenever any wrought austenitic stainless steel was heated to temperatures over 800° F by means other than welding or ther mal cutting, the mate rial was re-solution heat treated.

B.0-30 REV. 14, APRIL 2002 These controls were used to avoid severe sensitization and to comply with the intent of Regulatory Guide 1.44, "Control of the Use of Sensitized Stainless Steel."

We believe that we comply with the intent

~f this guide via incorporation of the alternate approach cited above.

REGULATORY GUIDE 1.45 - REV. 0 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS Regulatory Guide 1.45 describes methods acceptable to the NRC to assure that leakage detection and collection systems prov ide maximum practical identification of leaks within the reactor coolant pressure boundary.

The leak detection equipment consists of temperature, pressure, fission product monitoring and flow sensors with associ ated instrumentation and alarms. This equipment detects, annunciates, and isolates (in certain cases) leakages in the following systems:

a. Main steamlines

b. Coolant systems within the drywell

c. Reactor water cleanup (RWCU) system

d. Residual heat removal (RHR) system

e. Reactor core isolation cooling (RCIC) system

f. Feedwater system

g. High-Pressure Core Spray and Low-Pressure Core Spray.

Leakage is separated into identified and unidentified categories. The leakage to primary containment from identified sources is collected and isolated so that flow rates are separately monitored. For ex ample, the leakage from the MSIV's is separately identified and measured by localized sensors. Leakage to primary containment from unidentified sources is also collected and this accumulation rate is monitored. Small unidentified leaks (5 gpm and less) inside the drywell are detected by temperature changes, pressure changes, drain pump activities, fission product monitoring, and drywell cooler condensate flow monitoring.

B.0-31 REV. 14, APRIL 2002 Large leaks are also detected by changes in reactor water level and changes in flow rates in process lines.

Detection methods used to identify or quantify unidentified leakage include:

a. sump level and flow rate monitoring, b. airborne particulate radioactivity monitoring, and

c. airborne gaseous radioactivity monitoring.

These fission product detection methods are backed up with sump flow rate monitoring and drywell cooler condensate monitoring, thus providing diversity of methods. The leak detection equipment is capable of monitoring leakage of 1 gpm in less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The airborne activity monitors are used as trend devices only.

Isolation and/or alarm of affected systems and the detection methods used are summarized in Table 5.2-9 of the LSCS UFSAR.

Monitoring of coolant for radiation in the RHR and Reactor Water Heat Exchanger satisfies Position C.4 of the regulator gu ide. (For system details see Process Radiation Monitoring UFSAR Section 7.6.)

Leakage detection indicators and alarms are provided in the main control room. Operators use standard procedures to convert various indications to leakage equivalents.

The leakage detection systems are equipped with provisions to permit testing for operability and calibration during operation by the following methods:

a. Continuous monitoring of sump level compared to flow rates into sump.

b. Operability checked by compar ing one method to 'another.

c. Simulation of signals into trip monitors.

d. Channel "A" against Channel "B" of the same method.

Limiting conditions for identified and unidentified leakage are established in the Technical Specifications.

At least one particulate radioactivity monitoring channel is expected to remain functional following an SSE. Administrative procedures can be used to verify operability of this type of monitor following an SSE.

B.0-32 REV. 14, APRIL 2002 Additional information is presented in Sections 5.2, 7.1, 7.6, and 11.5.

The RCS leakage detection systems are co nsistent with the recommendations of Regulatory Guide 1.45.

REGULATORY GUIDE 1.46 - REV. 0 PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT Regulatory Guide 1.46 describes an acceptable basis for selecting the design locations and orientations of postulated breaks in fluid system piping within the reactor containment and for determining the measures that should be taken for restraint against pipe whipping that may result from such breaks.

The design of the containment structure, component arrangement, Class 1 pipe runs, pipe restraints and compartmentalization was done in consonance with the acknowledgement of protection against dynamic effects associated with postulated rupture of piping. Analytically sized and po sitioned pipe restraints were engineered to preclude damage based on the pipe break evaluation. See Appendix C for additional information.

This regulatory guide is applicable to recirculation pipe lines.

The design of the containment structure, component arrangement, Class 1 pipe runs, pipe whip restraints and compartmentaliz ation was done in consonance with the acknowledgment of protection against dynamic effects associated with postulated rupture of piping. Analytically sized an d positioned pipe whip restraints were engineered to preclude damage based on the pipe break evaluation.

Pipe whip requirements for fluid system piping within the primary containment that, under normal operation, has service temp eratures > 200°F or pressures > 275 psig, complied with ANS N176, "Design Basis for Protection Against Pipe Whip," and Regulatory Guide 1.46 except as delineated in the following criteria f or no breaks in Class 1 piping:

a. If Equation 10 of NB-3653-1, ASME Code III results in S < 2.4 S m for ferritic or austenitic steels, no othe r requirements need be met. Stress range should be calculated between any two load sets (including zero load set) according to NB-3600 for upset and an OBE event transient.

b. If Equation 10 results in 2.4 < S < 3.0 S m for ferritic or austenitic steels, the cumulative usage factor, U, calcul ated on the basis of Equation 14 of NB-3653.6, must be < 0.1.

B.0-33 REV. 14, APRIL 2002

c. If Equation 10 results in S > 3.0 Sm for ferritic or austenitic steels, then the stress value in Equations 12 and 13 of NB-3653..6 must not be > 2.4 S m. We believe that the recirculation piping is in compliance with the intent of this regulatory guide through incorporation of the alternate approach cited above. The remainder of the piping inside the containment has been evaluated as described in Subsections 3.6.1 and 3.6.2 and meets the intent of this Regulatory Guide.

REGULATORY GUIDE 1.47 - REV. 0 BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR POWER PLANT SAFETY SYSTEMS Regulatory Guide 1.47 requires that automatic indication at the system level be provided of the bypass or deliberately indu ced inoperability of the protection system, of systems actuated by the protection system, and of auxiliary or suppporting systems that effectively bypass or render inoperable the protection system.

Regulatory Guide 1.47 was issued after the Advisory Committee on Reactor Safeguards (ACRS) review of LSCS. The LaSalle County Station PSAR described a bypass and inoperable status indicator which was manually actuated by switches in the control room. This procedure supplements the normal administrative procedures followed by the Commonwealth Edison Company for rendering equipment inoperable.

Any bypass or deliberately induced inoperab ility of the protection system and those systems actuated or controlled by the protection system will be indicated at those system levels. Likewise, bypassed or deliberately induced inoperability of auxiliary or supporting systems are indicated at the sa me system level. By manually initiating this status indicator as part of the administrative procedure, or by automatically indicating a bypass or inoperative conditio n at the system level through the remote digital surveillance system described in Section 7.8 of the UFSAR the control room operators are made aware of potential technical specification violations prior to performing any improper bypass function.

Refer to Subsections 7.2.2, 7.3.1, and 7.

6.1 for additional information concerning bypass status indications which are annunciated.

We believe that we comply with the inte nt of the guidance set forth for LSCS.

B.0-34 REV. 14, APRIL 2002 REGULATORY GUIDE 1.48 - REV. 0 DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS Regulatory Guide 1.48 delineates acceptable design limits and appropriate combinations of loadings associated with no rmal operation, postulated accidents, and specified seismic events the design of Seismic Category I fluid system components.

The design limits specified for Class 1 components are in compliance with those specified in Article NB of Section III of the ASME B&PV Code, based on the following load combinations.

Normal Normal plant operation conditions. Upset a) Normal

+ 50% SSE b) SRV loads

+ 50% SSE Emergency Normal + 100% SSE Faulted Normal + 100% SSE + pipe rupture loads (if applicable)

The design basis was representative of good industry practices at the time of design, procurement and manufacture and is shown to be in general agreement with requirements of Regulatory Guide 1.48, with the following clarifications: (a) The probability of an OBE of the magnitude post ulated for LSCS is consistent with its classification as an emergency event. However, for design conservatism, loads due to the OBE vibratory motion have been includ ed under upset conditions, loads due to the OBE vibratory motion plus associated transients, such as a turbine trip, have been considered in the equipment design und er emergency conditions consistent with the probability of the OBE occurrence, and (I D) the use of increased stress level for Class 2 components is consistent with industry practice as specified in ASME Code Section III.

We believe that the design limits and load combinations for Seismic Category I fluid system components are in compliance with the intent of this guide through the incorporation of the alternate approach cited above.

REGULATORY GUIDE 1.49 - REV. 1 POWER LEVELS OF NUCLEAR POWER PLANTS Regulatory Guide 1.49 requires that the pr oposed licensed power level be restricted to a reactor core power level of 3800 megawa tts thermal or less and that analyses B.0-35 REV. 14, APRIL 2002 and evaluations in support of the applic ation should be made at 1.02 times the proposed licensed power level.

The original rated thermal power for LS CS reactors W25 3323 thermal megawatts.

The safety analyses and evaluations were made for an LSCS power level of 3454 thermal megawatts.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.50 - REV. 0 CONTROL OF PREHEAT TEMPERATURE FOR WELDING CF LOW-ALLOY STEEL Regulatory Guide 1.50 requires that weld fabrication for low-alloy steel components should comply with fabrication requirements of Section III and Section IX of the ASME B&PV Code supplemented by the following:

a. procedures qualification should require that a minimum preheat and maximum interpass temperature be specified and that the welding procedure be qualified at the minimum preheat temperature, b. preheat temperature be maintained until post-weld heat treatment has been performed, and

c. welding should be monitored to verify that limits on preheat and interpass temperatures are maintained.

For the NSSS equipment, the use of low-alloy steel is restricted to the reactor pressure vessel. Other ferritic components in the reactor coolant pressure boundary are fabricated from carbon steel materials.

Preheat temperatures employed for welding of low-alloy steel meet or exceed the recommendations of ASME Section III, Subsection NA. Components were either held for an extended time at preheat temperature to assure removal of hydrogen, or preheat was maintained until postweld heat treatment. The minimum preheat and maximum interpass temperatures were specified and monitored.

All welds were nondestructively examined by radiographic methods. In addition, a supplemental ultrasonic examination was performed.

By meeting and/or exceeding the recommendations of the ASME Code, the intent of this guide is satisfied even though the design was developed prior to its issuance.

B.0-36 REV. 14, APRIL 2002 REGULATORY GUIDE 1.51 - REV. 0 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR POWER PLANT COMPONENTS This regulatory guide was withdrawn by Mr. Robert B. Minogue's letter, dated July 15, 1975. The withdrawal letter states that current licensing practice is based on reference to Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. The LaSalle County Station, Units 1 and 2, is designed to comply with ASME Section XI.

REGULATORY GUIDE 1.52 - REV. 2 DESIGN, TESTING, AND MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ADSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.52 presents acceptabl e methods for implementing regulations with regard to the design, testing, and maintenance of air filtration and adsorption units of engineered-safety-feature atmospheric cleanup systems designed to mitigate consequences of postulated accidents.

A detailed description of the design and op eration of the gaseous effluent treatment equipment including filters, charcoal, etc.

is presented in Subsections 6.2.3, 9.4.1, and 11.3.1.

The ESF atmospheric cleanup units at LaSalle were designed and constructed to comply with the intent of the revision of Regulatory Guide 1.52, which was current at the time of their procurement.

The ESF atmospheric cleanup units at LaSalle that require in-place testing will be tested to comply with the intent of the now current Revision 2 of Regulatory Guide 1.52. REGULATORY GUIDE 1.53 - REV. 0 APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS

B.0-37 REV. 14, APRIL 2002 Regulatory Guide 1.53 requires that protec tion systems meet the requirements of Section 4.2 of IEEE 279-1971, which is also ANSI-N 42.7-1972 in that any single failure within the protection systems shall not prevent proper protective action at the system level when required. This guide provides guidance on an acceptable method of complying with this requirement.

Compliance is achieved by specifying, designing, and constructing the engineered safeguards systems to meet the single failure criterion, Section 4.2 of IEEE 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations," and IEEE 379- 1972, "IEEE Trial-Use Guide for th e Application of the single- Failure Criterion to Nuclear Power Generating Station Protection Systems." The reactor protection system has separate and red undant instrument channels and control logic to ensure that a failure in a sensing element or the decision logic or an actuator does not prevent or initiate protective action. Separated channels are employed, so that a fault affecting one ch annel does not prevent the other channels from operating properly.

The reactor protection system is normally energized, with two motor-generator sets for power, one for each separate trip system. Therefore, a single power failure produces a trip on only one channel. Comp lete loss of power trips the reactor since its protection system is designed for fail-safe operation.

The ECCS is divided into ADS, HPCS, LPCS, and RHR (LPCI) which meets the Single Failure criterion on a network basis.

Testing logic is provided so that the equipment can be operated in various test modes to confirm that it will operate properly when called upon. Testing incorporates all elements of the system under various test modes including sensors, logic, actuators, and actuated equipment.

Tests are performed at intervals so that there is an extremely low probability of fail ure in the intervening periods. During tests sufficient channels and systems are available to not degrade the protective function.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.54 - REV. 0 QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS Regulatory Guide 1.54 states that ANSI N101.4-1972 in conjunction with ANSI N45.2-1971 provides an ad equate basis for complying with quality assurance B.0-38 REV. 14, APRIL 2002 requirements for protective coatings applied to ferritic steels, aluminum, stainless steel, galvanized steel, concrete, or masonry subject to the following:

a. applicability and acceptability of references to be determined elsewhere; and

b. coatings and cleaning materials used with stainless steel shall not contain elements that could contribute to corrosion, intergranular cracking, or stress corrosion cracking.

Protective coatings are not applied to austenitic stainless steels at the LSCS. The provisions of ANSI N101.4-1972 apply to protective coatings applied to other materials within the power plant at LSCS.

We believe that we comply with the intent of guidance set forth in this regulatory guide. REGULATORY GUIDE 1.55 - REV. 0 CONCRETE PLACEMENT IN CATEGORY I STRUCI'URES Regulatory Guide 1.55 describes an acceptable method for implementing design control, material control, special processes control, and inspection and test control requirements with regard to the placement of concrete in Category I structures.

The applicant complies with the intent of the guidelines set forth in Regulatory Guide 1.55 and its Appendix. Refer to Appendix E for concrete placement in Category I structures.

Although the applicant complies with the regulatory position, the applicant does not comply with all the codes in the appendix of Regulatory Guide 1.55 in their entirety. Refer to Appendix E for exceptions taken to concrete codes with which the applicant complies.

REGULATORY GUIDE 1.56 - REV. 0 MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS Regulatory Guide 1.56 describes an acceptable method of implementing GDC 13, 14, 15, and 31 of 10 CFR 50 Appendix A with regard to minimizing the probability of B.0-39 REV. 14, APRIL 2002 corrosion-induced failure of the RCPB in BWR's by maintaining acceptable purity levels in the reactor coolant, and acceptable instrumentation to determine the condition of the reactor coolant.

Materials in the primary system are prim arily Type 304 stainless steel and Zircaloy cladding. The reactor water chemistry limi ts have been established to provide an environment favorable to these materials. Technical Specification limits are placed on conductivity and chloride concentratio ns. Operationally, the conductivity is limited because it can be continuously and reliably measured and gives an indication of abnormal conditions and th e presence of unusual materials in the coolant. Chloride limits are specified to prevent stress corrosion cracking of stainless steel.

The water quality requirements are further supported by General Electric Topical Report NEDO-10899.

Conductivity is continuously monitored on the primary coolant with instruments connected to redundant sources, the reactor water recirculation loop and the reactor water cleanup system inlet. The effluent from the reactor water cleanup system is also monitored for conductivity on a continuous basis. These measurements provide reasonable assurance for adequate surveillance of the reactor coolant.

Sufficient instrumentation is provided so that the conductivity of the condensate is known and so that the purity of the de mineralizer effluent is known. Other conductivity monitors provide redundant measurements of reactor water purity.

Refer to Subsection 5.2.3 in the UFSAR.

The condensate demineralizer system and reactor water cleanup system provide the recorded conductivity measurements and alarms of influents and effluents of the demineralizers and records of the flow rate through each demineralizer. Refer to Subsections 5.4.8 and 10.4.6.

Based upon the assessed capability of the design, we believe we comply with the intent of this regulatory guide through incorporation of the alternate approach cited above. REGULATORY GUIDE 1.57 - REV. 0 DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CO NTAINMENT SYSTEM COMPONENTS Regulatory Guide 1.57 delineates acceptable design limits and appropriate combinations of loadings associated with normal operation, postulated accidents, B.0-40 REV. 14, APRIL 2002 and specified seismic events for the design of components of metal primary reactor containment systems.

The applicant complies with the Regulatory Position to the extent described in.

Items a and b, below.

a. The design of ASME Code Cla ss MC components of primary metal containment systems, which are completely enclosed within Seismic Category I structures, is in accordance with Subsection NE of Section III and meets the intent of Regulatory Guide 1.57.

b. The design of piping, pump, and valve components of primary metal containment systems and of piping penetration head fittings, which are directly subjected to process pipe internal pressure, is in accordance with Subsection NB or NC (as applicable) of Section III and meets the intent of Regulatory Guide 1.48.

REGULATORY GUIDE 1.58 - REV. 1 QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL Regulatory Guide 1.58 describes an acceptable method of complying with regulations with regard to the qualification of nuclear power plant inspection, testing, and examination personnel during the construction, preoperational testing, startup testing, and operational phases.

Station Operational Phase For the operational phase, onsite inspection, examination and testing personnel are qualified in accordance with ANSI N 18-1-1971 as described in UFSAR Section 13.1.3.

Offsite inspection, examination and testing personnel are qualified in accordance with Regulatory Guide 1.58, Revision 1, dated September 1980.

Commonwealth Edison complies with the objectives set forth in the referenced revision of this regulatory guide, as indicated in the textural material described above.

B.0-41 REV. 14, APRIL 2002 REGULATORY GUIDE 1.59 - REV. 0 DESIGN-BASIS FLOODS FOR NUCLEAR POWER PLANTS Regulatory Guide 1.59 describes an acce ptable method of determining for sites along streams or rivers the design basis floods that nuclear po wer plants must be designed to withstand without loss of safety-related functions.

Although LSCS is approximately 255 feet above the Illinois River an analysis of design basis floods was ma de for the Illinois River and the onsite lake and watershed (drainage) areas at the plant for the defined PMF. Design and engineering of the lake structures, and plant structures was based on PMF with maximum wind wave runup. The methodol ogy was consistent with the extensive references provided as guidance. See Subsection 2.4.3.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.60 - REV. 1 DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS Regulatory Guide 1.60 describes a procedure acceptable to the NRC for defining the response spectra for the design of Seismic Category I structures, components, and systems. The requirements of current Regulatory Guide 1.60 (Revision 1) are not applicable to this facility as this revision of the re gulatory guide was not issued at the time construction permits were issued for LS CS. However, a comparison between the LSCS design spectra and Regulatory Guide 1.60, Revision 1 spectra, corresponding to 0.2g and 0.lg maximum ground accelerations and for appropriate values of dampings is presented in Figures B-1 thro ugh B-4. Regulatory guide spectra for OBE 4% damping and SSE 7% damping have been compared with the corresponding 2% and 5% damping LSCS design spectra. This is because conservative values of 2% and 5% dampings were used in actual design whereas 11% and 7% dampings are allowed in Regulatory Guide 1.61.

With acknowledgment of the vintage of th is plant vis-a-vis that of the current guidance, we believe we have complied with the intent of this regulatory guide via accepted industry practice at the time of the design. The response spectra used in the design were based on practical earthquake acceleration time-history records for this tectonic region.

B.0-42 REV. 14, APRIL 2002 REGULATORY GUIDE 1.61 - REV. 0 DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS Regulatory Guide 1.61 delineates damping va lues acceptable for use in the elastic modal dynamic seismic analysis of Seismic Category I structures, systems, and components.

With acknowledgement of the vintage of th is plant vis-a-vis that of the current guidance, we believe we have complied with the intent of this regulatory guide via alternate but locally conservative damping values used in the seismic analysis of major structures.

REGULATORY GUIDE 1.62 - REV. 0 MANUAL INITIATION OF PROTECTIVE ACTIONS Regulatory Guide 1.62 requires that manual initiation of each protective action at the system level be provided, that such initiation accomplishes all actions performed by automatic initiation, and that protective action at the system level go to completion once manually initiated. In addition, manual initiation should be by switches readily accessible in the control room, and a minimum of equipment should be used in common with automatically initiated protective action.

Means are provided for manual initiation of reactor isolation, emergency core cooling, and reactor manual scram at the system level through the use of armed pushbuttons, as described below:

Switch location Action initiated Number of Switches (Control Room Panel)

Reactor isolation four H13-P601 ADS four, two in Div. 1 and two in Div. 2 H13-P601 HPCS one switch in Div. 3 H13-P601 RHRA/LPCS one switch in Div. 1 H13-P601 RHRB/RHRC one switch in Div. 2 H13-P601 B.0-43 REV. 14, APRIL 2002 Reactor manual scram four H13-P601 Operation of these switches accomplishes th e initiation of all actions performed by the automatic initiation circuitry.

The amount of equipment common to both manual and automatic initiation of the above functions is kept to a minimum thro ugh implementation of manual activation at the final devices (relays, scram contractor) of the protection system. No failure in the manual, automatic, or common portions of the protection system will prevent initiation of a given function by manual or automatic means.

To prevent manual initiation of vessel depressurization when low- pressure core cooling capability is absent, the ADS manual initiation has an in terlock to assure proper conditions for depressurizatiOn (AC interlock). One interlock is provided for Division 1 ADS and a second independent interlock is provided for Division 2 ADS.

Manual initiation of any of the above functions, once initiated, goes to completion as required by IEEE 279-1971 Section 4.16.

For electric power systems, manual controls are provided to permit the operator to select the most suitable distribution path from the power supply to the load. An automatic start signal will override the test mode. Provision is made for control of the system from the control room as well as from an external location.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.63 - REV. 0 ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER-COOLED NUCLEAR POWER PLANTS This regulatory guide describes a method acceptable to the NRC for compliance with GDC 50 and Appendix B to 10 CFR 50 requirements of the design, construction, and installation of all electrical penetrations through the primary containment barrier.

The LaSalle County Station electrical penetrations are designed and applied to withstand, without loss of mechanical inte grity, the maximum possible fault current versus time conditions (which could occur because of single random failures of circuit overload protection devices) with in the two leads of any one single-phase circuit or the 3 leads of any one three-phase circuit.

B.0-44 REV. 14, APRIL 2002 The LSCS design meets the objective of Position 1 of Regulatory Guide 1.63 by having two breakers for interruption of fault current for the medium voltage penetrations and by having one breaker and self-fusing circuit characteristics to interrupt fault current for the low voltage penetrations. However, on the low voltage penetrations the self-fusing circu it characteristics are external to the penetrations. The LSCS low voltage penetrations are designed with oversized conductors through the penetration seal s such that they can withstand any conceivable fault current versus time condition not interrupted by the primary protective device. Should the backup protective device also fail to clear the fault, the ratio of conductor size within the seal s to the conductor si ze of the external circuit is such that failure of the electrical circuit external to the penetration seals is calculated to occur prior to failure of the penetration conductors. This capability is confirmed by testing (see Subsection 3.8.1. This oversizing of the penetration assembly, so that fault current will not cause overheating of the electrical penetration before fault current is interru pted, meets the requirement of Section 4 of IEEE 317-1972.

However, Commonwealth Edison will, by the end of the first refueling outage, install redundant (backup) circuit breakers at LaSalle on all 480-Vac circuits which penetrate primary containment and which may be energized during plant operation.

These circuit breakers will be directly in series with the primary protection circuit breakers. Also, by the end of the first refueling outage redundant (backup) circuit protection devices will be installed on all control circuits penetrating primary containment which do not already have backup circuit protection devices and which may be energized during plant operation. Note the control circuits to valve operators include their limit switches because their power comes from a dedicated transformer which is fed by the controller's power source hence limit switches will be protected with the same redundant ci rcuit protection devices as the parent controller. Control power is protected with fusing as the primary protection device.

No additional protection is planned for any instrumentation circuit which penetrates primary containment becauss all of the circuits are of such low energy that potential short circuit energy levels would have no adverse effect on the penetration seals.

Regulatory Guide 1.63 was issued after the LSCS construction permit. However, the LSCS design meets the objective of position 1 of Regulatory Guide 1.63 and meets Positions 2 and 4. Position 3 makes reference to other regulatory guides.

B.0-45 REV. 15, APRIL 2004 REGULATORY GUIDE 1.64 - REV. 2 QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS Regulatory Guide 1.64 describes an acceptable method for complying with the NRC's quality assurance requirements for the design of nuclear power plants.

Chapter 17.0 cities Exelon Nuclear's reference QA Topical Report NO-AA-10 which is the updated version of EGC-1-A describing the Exelon Nuclear Quality Assurance Program for design and control for its nucle ar generating stations. This document was accepted by the NRC as a docketed record adequate for cross-referencing in licensing matters.

We comply with the objectives set forth in the referenced revision of this regulatory guide. REGULATORY GUIDE 1.64 - REV. 1 QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS Regulatory Guide 1.64 describes an acceptable method for complying with the NRC's quality assurance requirements for the design of nuclear power plants.

Chapter 17.0 cites a reference document NO-AA-10 describing the Exelon Nuclear Quality Assurance Program. This document was submitted as a topical report and was accepted by the NRC as a docketed record adequate for cross-referencing in licensing matters.

We believe we comply with the objectives set forth in this regulatory guide.

REGULATORY GUIDE 1.65 - REV. 0 MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS Regulatory Guide 1.65 gives recommendations for the selection and application of the reactor vessel closure bolting materials, with the intent of preventing failures such as could occur from embrittlement, corrosion, and stress corrosion cracking.

B.0-46 REV. 14, APRIL 2002 The bolting materials used are SA-540, Grades 1323 and 1324, as specified in the regulatory guide.

The maximum reported tensile strength and minimum Charpy impact test properties are as tabulated below, from the certified materials test reports:

Tensile Strength Lateral Cv Energy Sn (psi) Expan. (mils) (ft-lb) LSCS 1: Studs 174,000 23 43 Nuts and washers 174,500 22 36 LSCS 2: Studs 163,500 25 40 Nuts and washers 156,000 25 43 As shown, the maximum tensile strength va lues of the LSCS 1 bolting exceed the 170-ksi maximum given in the Guide. Also , the impact test properties are lower than specified by Paragraph IV.A.4 of 10 CFR 50, Appendix G. It is noted, however, that the impact data shown are the lowest values measured from tests at 10°F), instead of the lowest preload or servic e temperature (70°F) specified for the Appendix G requirements. It is probable that the properties would compare more favorably with the Appendix G minimums of 45 ft-lb and 25 mils lateral expansion, if tested at the higher temperature.

The extent of compliance with the Item C.2 inspection requirements of the Guide is described in the following.

The bolting materials were ultrasonically examined after final heat treatment and prior to threading, as specified. The specific requirement for examination according to SA-388 was not complied wi th. However, the procedures approved for use in practice are judged to ensure compar able material quality and moreover are considered adequate on the basis of compliance with the applicable requirements of ASME Code Paragraph N-322. Addition ally, straight beam examination was performed on 100% of cylindrical surfaces and from both ends of each stud. In addition to the code required notch, the reference standard for the radial scan contained a 1/2-inch-diameter flat bottom hole with a depth of 10% of the thickness, and the end scan standard contained a 1/4-inch-diameter flat bottom hole 1/2 inch deep. Also, angle beam examination was performed on the outer cylindrical surfaces in both axial and circumferential surfaces in both axial and circumferential directions. Surface examinations were performed on the studs and nuts after final heat treatment and threading, as specifie d in the guide, in accordance with N-626 or N-627 of the applicable ASME Code.

As specified in Item C.1.b (3) of the guide, there are no metal platings applied to the closure bolting. A phosphate coating is applied to threaded areas to studs and nuts and bearing areas of nuts and washers to ac t as a rust inhibitor and to assist in retaining lubricant on these surfaces.

B.0-47 REV. 14, APRIL 2002 Item C.3 of the guide requires adequate corrosion protection of studs and flange holes during venting and filling of the vessel and while the head is removed.

General Electric practice allows exposure of these surfaces to high-purity fill water.

Nuts and washers are stored dry during refueling.

Volumetric examination of the studs is to be in accordance with ASME Code Section XI. REGULATORY GUIDE 1.66 - REV. 0 NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS Regulatory Guide 1.66 describes an acceptable method for implementing the requirements for nondestructive examination of tubular products used for components of the reactor coolant pressure boundary and other safety-related systems. Wrought tubular products were supplied in accordance with applicable ASTM/ASME material specifications. These specifications require a hydrostatic test on each length of tubing. Additionally, the specification for the tubular product used for CRD housing specified ultrasonic examination to Paragraph N132550 of the ASME Code,Section III.

These components met the requirements of ASME Codes existing at the time of placement of the order, which pr edated Regulatory Guide 1.66.

This regulatory guide was withdrawn on September 28, 1977 by the NRC because the additional requirements imposed were satisfied by the ASME Code. See Subsections 4.5.2 and 5.2.3.

REGULATORY GUIDE 1.67 - REV. 0 INSTALLATION OF OVERPRESSURE PROTECTION DEVICES Regulatory Guide 1.67 describes a method acceptable to the NRC for implementing GDC 1 of 10 CFR 50, Appendix A with regard to the design of pipi ng for safety valve and relief valve stations which have open discharge systems with limited discharge pipes and which have inlet piping that neit her contains a water seal nor is subject to slug flow of water upon discharge of the valves.

B.0-48 REV. 14, APRIL 2002 The MSL safety/relief valves of the LSCS plant relieve to closed discharge systems. The guidelines of this guide are therefore not applicable.

REGULATORY GUIDE 1.68 Initial Issue: Revision 0, November 1973 Current Issue: Revision 0, November 1973 LaSalle C.P. Issue: September 10, 1973 PREOPERATIONAL AND INITIAL STARTUP TEST PROGRAMS FOR WATER-COOLED POWER REACTORS Regulatory Guide 1.68 describes the NRC's requirements for the in itial startup test programs. It is applicable to such activities as precritical tests and flow power tests.

A test program has been established to ensure that all structures, systems, and components will satisfactorily perform their safety-related functions. This test program provides additional assurance that the plant has been properly designed and constructed and is ready to operate in a manner that will not endanger the health and safety of the public, that the procedures for operating the plant safely have been evaluated and have been demonstrated, and that the operating organization is knowledgeable about the plant and procedures and are fully prepared to operate the facility in a safe manner.

The test program includes simulation of equipment failures and control system malfunctions that could reasonably be expected to occur during the plant lifetime.

The test program also includes testing for interactions such as the performance of interlock circuits in the reactor protection system. It also determines that proper permissive and prohibit functions are performed and that circuits normally active and supposedly unaffected by the position of the mode switch perform their function in each mode. Care is taken to ensure that redundant channels of the equipment are tested independently.

We believe that the preoperational and initial startup test programs at LSCS comply with the intent of this regulatory guide REGULATORY GUIDE 1.69 - REV. 0 CONCRETE RADIATION SHIELDS FOR NUCLEAR POWER PLANTS

B.0-49 REV. 14, APRIL 2002 ANSI 101.6-1972 basically deals with field practices. It al so stresses quality assurance (QA) especially in the mixing and pouring of concrete. The parts of this document that coven shield design are more concerned with the protection and construction of shields than with the radiation protection aspects. The radiation protection is included in the station design criteria.

The shielding for LSCS has the following design:

a. the design density of the concrete used in shielding walls is 140 pounds per cubic foot;

b. the surrounding atmosphere does not exceed 1500 F;

c. all source areas have on e on all of the following: 1. leaktight liner,

2. sealed shields, 3. exhaust systems;

d. all shield penetrations and voids are considered;

e. removable walls and plugs are used where appropriate;

f. structural, thermal, missile, shie lding load, accident, and operational analyses are made where required;

g. shield walls are designed to maintain a design dose rate; and

h. ordinary concrete is being used in the reactor shield.

Precautions are to be taken to ensure that the pouring of the reactor shield is done correctly. Further discussion of the shield ing requirements is found in Section 12.1, and further discussion of the structural requirements is found in Section 3.8.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.70 - REV 2 (09/75)

The previous commitment to the Regulatory Guide is superseded by the following.

As part of the ongoing effort to improve the quality of the UFSAR, the guidelines provided in NEI 98-03 Revision 1, 06/99 and endorsed by the NRC in Regulatory Guide 1.181, Revision 0, 09/99 will be used to further improve the content of the UFSAR. While the UFSAR will continue to follow the general organizational recommendations specified in this Regulatory Guide, the reformatting options described in NEI 98-03, Revision 1, will be used to simplify information contained in the UFSAR to improve their focus, clarit y and maintainability. Examples of the B.0-49a REV. 15, APRIL 2004 types of improvements that are anticipated are the identification of historical information and the removal or replacement of excessively detailed drawings as described in Appendix A to NEI 98-03 Revision 1.

REGULATORY GUIDE 1.71 - REV. 0 WELDER QUALIFICATION_FOR AREAS OF LIMITED ACCESSIBILITY

Regulatory Guide 1.71 requires that weld fabrication and repair for wrought low-alloy and high-alloy steels or other materials such as static and centrifugal castings and bimetallic joints should comply with fabrication requirements of Section III and Section IX supplemented by the following:

B.0-50 REV. 14, APRIL 2002

a. Performance qualification should require testing of welders under simulated access conditions when welder's access to weld is restricted to 30 to 35 cm. in any direction from the joint.

b. Regulation is required if:

1. sufficiently different restricted accessibility occurs, or

2. any of essential welding variables listed in Section IX are changed. c. Production 'welding should be monitored and adherence to qualification requirements should be certified.

All ASME Section III welds were fabricated in accordance with the requirements of Sections III and IX of the ASME Boiler and Pressure Vessel Code.

Only a few restrictive welds are involved in the fabrication of LSCS pressure boundary components. Welder qualification for welds with the most restrictive access was accomplished via mockup welding. Mockups were examined with radiography or sectioning.

All reactor pressure boundary welding was performed in, accordance with ASME Section IX. Reactor internal components welding was performed in accordance with ASME Section IX or appropriate AWS requirements.

REGULATORY GUIDE 1.72 - REV. 0 SPRAY POND PLASTIC PIPING This regulatory guide is not applicable to LSCS.

REGULATORY GUIDE 1.73 - REV. 0 QUALIFICATION TESTS OF ELECTRIC VALVE OPERATORS INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS To the extent practicable, the electric and mechanical components integral to the electric valve operator mechanism and requ ired to operate and control valve action inside the containment have been tested in accordance with IEEE 382-1972.

B.0-51 REV. 14, APRIL 2002 The valve operators for Class lE electrically operated valves have been tested in accordance with the test sequence outlined in Section 4.5.2 of IEEE 382-1972. If the anticipated actual service operating sequence is expected to create a more severe operating condition, then this actual service sequence has been used in the qualifying tests.

The qualifying tests have been made under environmental conditions (temperature, pressure, humidity, radiation, dust) that ar e at least as severe as those that the valve operator will be exposed to during and following a design-basis accident (LOCA). The radiological source term is based upon the same source term described in the guide. We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.74 Initial Issue: Revision 0, February 1974 Applicable Issue: Revision 0, February 1974 LaSalle C.P. Issued: September 10, 1973 QUALITY ASSURANCE TERMS AND DEFINITIONS Regulatory Guide 1.74 identifies quality assurance terms and acceptable definitions that are important to the understanding of the quality assurance requirements for design, construction, and operation of nuclear power plant structures, systems, and components.

Chapter 17.0 cites Edison's QA Topical Report CE-lA document was accepted by the adequate for cross-referencing in licensing matters.

We comply with the objectives set forth in the referenced revision of this regulatory guide. REGULATORY GUIDE 1.75 - REV. 1 PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS

B.0-52 REV. 14, APRIL 2002 Regulatory Guide 1.75 describes a method acceptable to the NRC for assuring physical independence of the circuits and electric equipment comprising or associated with Class 1E power systems, protection systems, systems actuated or controlled by the protection system, arid auxiliary or supporting systems that must be operable for the protection systems to perform their safety-related functions.

Regulatory Guide 1.75 was issued after the LaSalle County Station construction permit. However, the LSCS design meets most of the regulatory positions, as described in Subsection 8.3.1. Aside from the two exceptions noted below (Positions 1 and 10) the LSCS design complies with th is guide. Position 2 is not applicable.

The LSCS design differs with Position 1 in that the basis for precluding the use of a fault-current-actuated interrupting device as an isolation device within the context of Regulatory Guide 1.75 is questionable. This conclusion seems to preclude the use of a circuit breaker as an isolating device and thus prevents achieving a practical design of an electrical auxiliary system.

A qualified (Class 1E) circuit breaker could otherwise fulfill the literal interpretation of Section 3.8 by preventing (by its successful interruption of fault current) a malfunction in one section of a circuit from causing unacceptable influences (i.e., failures) in other sections of the circuit (or other circuits). To preclude Class 1E circuit breakers from these applications casts doubt upon the safety-related portions of the auxiliary power systems in all nuclear plants. Such systems invariably utilize Class 1E metal-clad switchgear and molded case circuit breakers as circuit protection devices.

The LaSalle County Station (LSCS) design complies with Positions 3 through 9.

With respect to the separation of associated circuits, the LSCS design complies with Section 4.5(3) of IEEE 384-1974.

Although separation of associated circuits in some cases differs from that dictated in Sections 5.1.3 and 5.6.2 of IEEE-384, Clas s 1E circuits are not degraded below an acceptable level for the following reasons:

a. Cables associated with one safety-related division are never routed in cable trays or conduit containing cables of a redundant safety-related division. This is true for all gene ral plant areas. Lesser separation than that dictated by IEEE-384 occurs only within control panels located in the control room and auxiliary equipment room.

b. All cables used to interconnect associated circuits are the same high quality as that utilized in Class lE circuits, (i.e., all associated cables comply with the requirements of IEEE 383-1974). Therefore, this cable has been proven to be highly fire retardant by testing.

c. Cables with separation less than that dictated by Sections 5.1.3 and 5.6.2 of IEEE-384 are constrained to control and instrumentation B.0-53 REV. 14, APRIL 2002 circuits which, by their very nature, are low energy circuits. Control circuits are generally l20-Vac or 125-Vdc, whereas, the insulation rating of the cable utilized at LSCS is 600-V.

d. There are no power cables in contact with the control and instrumentation cables in the cable spreading area or in the control room and auxiliary equipment room. Also, there are no high energy sources located within control panels installed in these areas.

e. Fire stops are installed in the bo ttom entrances of all control panels.

With respect to the separation of non-Class 1E from Class 1E control and instrumentation circuits, LSCS complies with Section 4.6.2 of IEEE-384. Although the separation of non- Class 1E from Class 1E control and instrumentation circuits is in some cases, less than that requir ed by Sections 5.1.3 and 5.6.2 of IEEE-384, these circuits have been analyzed to show that Class 1E circuits are not degraded below an acceptable level because:

a. Non-Class lE cables are routed in separate cable trays from Class 1E and associated cables in general plant areas.

b. Non-Class lE cables which come in close proximity to Class 1E and associated cables at one end do not come in contact with redundant Class 1E or associated circuits at their other end. This has been confirmed by a study of in stalled cables at LSCS.

c. All cables used to interconnect non-Class 1E circuits are the same high quality as that. utilized in Class 1E circuits, (i.e., all associated cables comply with the requirements of IEEE 383-1974). Therefore, this cable has been proven to be highly fire retardant by testing.

d. Cable separation less than that required by Sections 5.1 and 5.6 of IEEE-384 is limited to control and instrumentation circuits which by their very nature are low energy circuits. Control circuits are generally 120-Vac or 125-Vdc, whereas, the insulation rating of the cable utilized at LSCS is 600-V.

e. There are no power cables in contact with the control and instrumentation cables in the cable spreading area or in the control room and auxiliary equipment room. Also, there are no high energy sources located within control panels installed in these areas.

f Fire stops are installed in the bo ttom entrances of all control panels.

Position 10 refers to Section 5.1.2 of IEEE 384-1974 concerning cable and raceway identification. The LSCS design utiliz es cable trays with permanent colored B.0-54 REV. 14, APRIL 2002 identification markers at each routing point which are assigned and alphanumeric code per Table 8.3-6 of the LSCS-FSAR.

Each cable is assigned a number and segregation code. This information is placed on a colored tag, of permanent design, which is affixed to each end of the cable. A similar tag is also affixed to the cable where it enters and exits a penetration.

The LSCS design complies with Positions 11, 12, 13, 14, 15, and 16.

IEEE 384-1974, Section 4.6, requires that Non-Class lE cable trays be separated from Class lE cable trays by the following minimum separation requirements:

a. 1 ft horizontally and 3 ft vertically in cable spreading areas.

b. 3 ft horizontally and 5 ft vertically in general plant areas.

The LaSalle County Station (LSCS) criteria specifies that the minimum distance between safety-related and non-safety-r elated cable trays shall be 3 inches horizontally and 1 foot vertically. Cable trays at LSCS were installed to this specific criterion prior to issuance of IEEE-384.

At LSCS, the routing of safety-related c ables and non-safety-related cables was controlled during design such that only a few such instances of close proximity exist in the auxiliary building.

Although LSCS criteria are less than that prescribed by IEEE-3 84, the LSCS cable tray installation is considered to be adequate for the following reasons:

a. There are no non-safety-related cable trays throughout the entire reactor building and primary containment.

b. There are no safety-related cable trays throughout the entire turbine building and outlying structures (service building, screen house, etc.).

c. Non-safety-related cable trays ar e installed in close proximity to safety-related cable trays only at a minimal number of locations in the auxiliary building.

d. In those few instances where non-safety-related cable trays come in close proximity to safety- related tr ays, only one division of safety-related trays are present in the room. The lone exception to this statement is the corridor between the diesel generator rooms and the switchgear room. However, trays in this area will be covered with a flame retardant material as a result of the Fire Hazards Analysis.

e. All cables used to interconnect non-safety-related circuits are the same high quality as that utilized in Cla ss 1E circuits; (i.e., all non-safety-B.0-55 REV. 14, APRIL 2002 related cables comply with th e requirements of IEEE 383-1974).

Therefore, this cable has been proven to be highly fire retardant by testing. f. Fire stops are installed in all cabl e trays, including non-safety-related cable trays, where they penetrate walls and floors.

g. All cable trays installed at LSCS are, for the most part, solid bottom cable trays. Therefore, where vertical stacking occurs, an additional barrier is provided between each layer of cable trays beyond that recognized by Section 5.1.1.3 of IEEE 384-1974.

Regulatory Guide 1.75 Regulatory Position C6 Analysis performed in accordance with Sections 4.5(3), 4.6.2, and 5.1.1.2 should be submitted as part of the Safety Analysis Report and should identify those circuits installed in accordance with these sections.

Applicants Position The referenced analysis, when performed to justify deviation from specific requirements of standard IEEE 384-1974, shall be prepared on a case-by-case basis, shall be documented and be on permanent file, available for NRC review, but will not be an integral part of the safety analysis report.

Justification of Applicants Position The applicant's position is consistent with that taken for other plant design records; e.

g., routine design calculations, design document revisions, etc.

REGULATORY GUIDE 1.76 - REV. 0 DESIGN BASIS TORNADO FOR NUCLEAR POWER PLANTS The tornado design parameters are in accordance with the PSAR. Regulatory Guide 1.76 was issued after the construction permits were granted.

A comparison of the design parameters for the maximum wind velocity (360 mph) and the maximum pressure drop (3 psi) with the regulatory guide values indicates that these parameters are in accordance with the aforementioned guide. The design radius (227 feet) of maximum wind velocity is larger than that of the regulatory guide valued (150 feet) and therefore, the design tornado engulfs a larger portion of the structure.

The rate of pressure drop of the design tornado (1 psi/sec) is smaller than the value given in the regulatory guide (2 psi/sec). However, this difference in the rate of pressure drop has no effect on the design, since the natural periods of the building B.0-56 REV. 14, APRIL 2002 structures in comparison to either rate of pressure drop allow this load to be treated as a static load.

Thus the tornado design parameters are either equivalent or conservative compared to the parameters given in Regulatory Guide 1.76.

Refer to Section 3.3 for additional discussion.

We believe that we comply with the guidan ce set forth in this regulatory guide.

REGULATORY GUIDE 1.77 - REV. 0 ASSUMPTIONS USED FOR EVALUATING A CONTROL ROD EJECTION ACCIDENT FOR PRESSURIZED WATER REAC'IORS This regulatory guide is not applicable to LSCS.

REGULATORY GUIDE 1.78 - REV. 0 ASSUMPTIONS FOR EVALUATING THE HABITABILITY OF A NUCLEAR POWER PLANT CONTROL ROOM DURING A POSTUATED HAZARDOUS CHEMICAL RELEASE Regulatory Guide 1.78 describes assumptions acceptable for use in assessing the habitability of the control room during an d after a postulated external release of hazardous chemicals and describes criteria that are generally acceptable for the protection of the control room operators.

The requirements of this regulatory guide (for those plants that have received a construction permit prior to the issuance of this regulatory guide) are that the applicant should be in conformance to the criteria that were in effect at the time of the granting of the construction permit. The LaSalle County Station, Units 1 and 2, conform to all the criteria in this area that were in effect on the date that the construction permits were granted.

REGULATORY GUIDE 1.79 - REV. 1 PREOPERATIONAL TESTING OF EM ERGENCY CORE COOLING SYSTEMS FOR PRESSURIZED WATER REACTORS This regulatory guide is not applicable to LSCS.

B.0-57 REV. 14, APRIL 2002 REGULATORY GUIDE 1.80 PREOPERATIONAL TESTING OF INSTRUMENT AIR SYSTEMS The preoperational testing of instrument air systems for the LaSalle County Station, Units 1 and 2, is in conformanc e with Regulatory Guide 1.80 in that all components of the instrument air system are tested to demonstrate that they will perform satisfactorily in service. The tests are designed to verify design response of safety-related equipment to a loss of instrument air accident and to verify that the failure of nonsafety-related components will not jeopardize safety-related components. The tests will also demonstrat e the adequacy of operating procedures.

REGULATORY GUIDE 1.81 - REV. 1 SHARED EMERGENCY AND SHUTDOWN ELECTRIC SYSTEMS FOR MULTI-UNIT NUCLEAR POWER PLANTS Regulatory Guide 1.81 describes an acceptable method for complying with GDC 5 of 10 CFR 50 Appendix A with respect to the sharing of onsite emergency and shutdown electric systems for mu ltiunit nuclear power stations.

The criteria utilized in designing the two unit LaSalle County Station is that each unit shall operate independently of the other and any malfunction of equipment or operator error in one unit will not initiate a malfunction or error in the other unit nor affect the continued operation of the other unit.

The separate HPCS diesel-generators for LSCS Units 1 and 2 do not share other loads within each unit nor with other external units, therefore this guide is fulfilled. The HPCS is in electrical Division 3 and is isolated from electrical Divisions 1 and

2. No safety-related systems, structures, or components are shared unless such sharing has been evaluated to ensure th at there will be no significant adverse impact on safety functions.

Regulatory Guide 1.81 was issued after the LaSalle County Station construction permit. However, the design generally meets the intent of the positions of Regulatory Guide 1.81.

REGULATORY GUIDE 1.82 - REV. 0 SUMPS FOR EMERGENCY CORE COOL ING AND CONTAINMENT SPRAY SYSTEMS B.0-58 REV. 14, APRIL 2002 Regulatory Guide 1.82 applies to pressurized water reactors; therefore, it is not applicable to the LaSalle County Station.

REGULATORY GUIDE 1.83 - REV. 1 INSERVICE INSPECTION OF PRESSURIZED WATER REACTOR STEAM GENERATOR 'IUBES This regulatory guide is not applicable to LSCS.

REGULATORY GUIDE 1.84 - REV. 6 CODE CASE ACCEPTABILITY - ASME SECTION III DESIGN AND FABRICATION This guide provides a list of ASME Design and Fabrication Code Cases that have been generically approved by the Regulatory Staff as of June 1975. Code Cases on this list may, for design purposes, be used until appropriately annulled. Annulled cases are considered "active" for equipm ent that has been contractually committed to fabrication prior to the annulment.

The guide, Revision 6 and later, revisions, requires specific approval by the NRC on all ASME Section III, Class 1, 2, and 3 Code Cases not listed. NRC regulation 10 CFR 50.55a requires NRC approval only fo r Code Cases for Class 1 Components.. GE's procedure for meeting the regulatory requirements is to obtain NRC approval for Code Cases applicable to Class 1 components only. NRC approval of Class 2 and 3 Code Cases 'was not required at the time of the design of LSCS and is not required by 10 CFR 50.55a, however, all Class 2 and 3 equipment has been designed to ASME Code or ASME approved Code Cases.

Conformance for Class 1 components is acco mplished by the use of only those ASME Code Cases which are listed as being a cceptable in this Guide. Design and construction of the reactor vessel complies with Regulatory Guide 1.84 to the extent described as follows: Code Cases whic h were used were 1332-5, 1359-1, 1401, 1420, 1441-1 and 1459-1. For Code Case compliance, 1332-4 was authorized under provisions of Regulatory Guide 1.85 to allow use of a Code Case for which a subsequent revision (1332-6) has been appr oved in the Guide. Similarly, 1401 is allowed under the provision of Regulatory Guide 1.85 for use of a previously listed Code Case which has been subsequently annulled. Code Cases 1359-1, 1420, 1441-1, and 1459-1 were incorporated in the AMSE Co de prior to issuance of the regulatory guides, so their application would also be consistent with the intent of the above provisions of the guides.

B.0-59 REV. 14, APRIL 2002 In addition to the approved Code Cases listed above, for Class 2 and 3 components LSCS is utilizing Code Cases N-237, N-240, and N-241 for which NRC approval has been requested.

Based upon the assessed capability of the Class 1 equipment utilized at LSCS, we believe that this facility is in compliance with the intent of this guide through incorporation of the alternate approach cited above.

Regulatory Guide 1.84 The NRC has approved the use of ASME Code Case N-411 for class 1,2, and 3 piping at LaSalle in a letter from E.G.

Adensam (NRC) to D.

L. Farrar (CECo) dated April 1, 1986.

REGULATORY GUIDE 1.85 - REV. 6 CODE CASE ACCEPTABILITY - ASME SECTION III MATERIALS This guide provides a list of ASME Materials Code Cases that have been generically approved by the Regulatory Staff as of June 1974.

Code Cases on this list may, for design purposes, be used until appropriately annulled. Annulled cases are considered "active" for equipment that has been contractually committed to fabrication prior to the annulment.

The guide, Revision 6 and later revisions, require specific approval by the NRC on all ASME Section III, Class 1, 2, and 3 Code Cases not listed. NRC regulation 10 CFR 50.55a requires NRC approval only for Code Cases for Class 1 components.

GE's procedure is to obtain NRC approval of Code Cases on Class 1 components only. NRC approval of Class 2 and 3 Code Cases was not required at the time of the design of LSCS and is not required by 10 CFR 50.55a; however all Class 2 and 3 equipment has been designed to ASME code or ASME approved Code Cases.

Conformance is accomplished by the use of only those ASME Code Cases which are listed as being acceptable in this guide. Design and construction of the reactor vessel complies with Regulatory Guide 1.85 to the extent described as follows. Code Cases which were used were 1332-5, 1359-1, 1401, 1420, 1441-1 and 1459-1. Of these, 1332-4 was authorized under provisions of Regulatory Guide 1.85 which allow use of a Code Case for which a subsequent revision (1332-6) has been approved in the Guide. Similarly, 1401 is allowed under the provision to regulatory guide 1.85 for use of a previously listed Code Case which has been subsequently annulled. Code Cases 1359-1, 1420, 14141-1 , and 1459-1 were incorporated in the ASNE Code prior to issuance of the regu latory guides, so their application would also be consistent with the intent of the above provisions of the guides.

B.0-60 REV. 14, APRIL 2002 Based upon the assessed capability of Class 1 materials utilized in the RCPB at LSCS, we believe that this facility is in compliance with the intent of this guide via the alternate approach cited above.

Regulatory Guide 1.85

Code cases 1713 (endorsed in Revision 5) and N-242-1 (endorsed in Revision 18) of Regulatory Guide 1.85 has been adopted for use at LaSalle in a letter from H. L. Massin (BWR Engineering) to G. J. Diederich dated February 5, 1987.

REGULATORY GUIDE 1.86 - REV. 0 TERMINATION OF OPERATING LI CENSES FOR NUC LEAR REACTORS Regulatory Guide 1.86 describes acceptable methods and procedures for the termination of operating licenses for nuclear reactors.

The LaSalle County Station, Units 1 and 2, will conform to the legal requirements of the appropriate NRC regulations when the operating license is terminated.

REGULATORY GUIDE 1.87 - REV. 0 GUIDANCE FOR CONSTRUCTION OF CL ASS 1 COMPONENTS IN ELEVATED TEMPERATURE REACTORS Regulatory Guide 1.87 applies to high-tem perature gas-cooled reactors (HTGRS), liquid metal fast-breeder reactors (LMFBR's) and gas-cooled fast-breeder reactors (GCFBR's); therefore, it does not apply to the La Salle Ccunty Station.

REGULATORY GUIDE 1.88 - REV. 2 COLLECTION, STORAGE, AN D MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS Regulatory Guide 1.88 describes a method acceptable to the NRC for complying with the requirements to collect, store, and maintain quality assurance records for nuclear power plants.

B.0-61 REV. 15, APRIL 2004 Per U.S. NRC letter from E.S. Beckjord dated June 17, 1991 (

Subject:

Withdrawal of Regulatory Guides), Revision 2 of Re gulatory Guide 1.88 was withdrawn due to obsolescence.

The ANSI standards endorsed by the associated guides were incorporated into ANSI/ASME NQA-1-1983, "Quality Assurance Program Requirements for Nuclear Facilities." ANSI/ASME NQA-1-1983 was endorsed by Revision 3 of Regulatory Guide 1.28, "Quality Assurance Program Re quirements (Design and Construction)", which was issued in August 1985.

Reference to Regulatory Guide 1.88 Revision 2 is being maintained for archival purposes.

UFSAR Chapter 17.0 cities Exelon Nuclear's QA Topical Report NO-AA-10 which is the updated version of EGC-1-A that described the QA program. That document outlines the Exelon document practices. The station fire protection afforded the archives storage area conforms to N FPA 232 criteria with the exception of documented deviations (See NFPA Code Report). This document was accepted by the NRC as a docketed record adequate for cross-referencing in licensing matters. The station central files is an area for the retention and reference of semicurrent records pending their ultimate dispositio

n. It does not conform to NFPA 232 criteria. Fire protection is afforded by a fire detection system which automatically initiate preaction fire extinguishing system that has a local alarm, and an alarm in the control room. Additional protection is afforded by personnel occupancy during normal working hours and periodic patrols on nights, weekends and holidays.

The following information is also applicable to the station central file:

Completed Quality Assurance records will reside in this area for a nominal period of 1 year. Completed Quality Assurance records are stored in fully enclosed metal cabinets.

Adequate access and aisle ways are maintained at all times throughout the facility.

Smoking is prohibited.

Although a copying machine and two machines for making hard copy prints of aperture cards are located within the area , they are segregated from the records area. Another copying machine is located in a separate enclosed room. The machines for drawing reproduction are located in the front room and are separated from the records area by a central file service counter. The area is used for microfilming.

Access to the records area is restricted to authorized personnel.

The station central file is located on the fourth floor of the So uth Service Building.

B.0-62 REV. 15, APRIL 2004 Additional storage of records is provided for on the 2nd floor of the South (New) Service Building. Historical hard copies of safety related design calculations are maintained in this area for reference purp oses. Filmed copies of these records are maintained on microfiche for quality records retention purposes in a QA vault in building 30.

The copies of more recent design calculation records are filed in this area while the records await filming to microfiche. These calculations are maintained for reference

following filming.

The original hard copies of older non-safety related design calculation records are also maintained in this area.

The fire protection is a wetpipe system, which automatically initiates, and alarms locally and in the control room.

Exelon Nuclear complies with the objectives set forth in the referenced revision of this regulatory guide, with the exception noted above.

Selection, Design, Qualification and Test ing of Diesel-Generator Units used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants

Regulatory Guide (RG) 1.9 Revision 3 endorses IEEE Standard 387-1984, "IEEE Standard Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations". In addition to this standard RG 1.9 Rev. 3 provides supplemental regulatory position

s. The applicant complies with these supplemental regulatory positions in Revision 3 with the following clarifications regarding:

1. Paragraph C.2.2. Exception is taken to the last sentence in this paragraph, "Jumpers and other nonstandard configurations or arrangements should not be used subsequent to initial equipment start up conditions". In order to successfully accomplish certain diesel tests it is necessary to use jumpers to si mulate particular engine signals. The use of jumpers is a normal practi ce in diesel engine testing, and the safe use of the jumpers are ensured with detailed procedures.

2. Paragraph C.2.2.1 "Start Test".

Each EDG undergoes a start up test on a monthly basis from "standby co nditions" Once every six months this test is supplemented by verifying proper start up from "normal standby conditions". This test is covered in paragraph C.2.2.3, and is further discussed in Item 3 below.

B.0-63 REV. 17, APRIL 2008 3. Paragraphs C.2.2.1, C.

2.2.4, C.2.2.5, and C.2.2.

6. Exception has been taken against use of the term 'stand by conditions' to denote 'normal standby conditions'.

The term normal 'standby condition' is interpreted as any conditional state of the EDG in which the EDG is considered operable. More specifically, standby conditions for an EDG refer to a condition whereby the diesel engine lube oil is being continuously circulated and

engine jacket water and lube oil temperatures are consistent with manufacturer's recommended operating range (low lube oil and jacket water temperature alarm settings to the high lube oil and jacket water temperature alarm settings).

The term 'ambient standby condition' defines a conditional state of the EDG in which lube oil and jacket water temperatures are within the prescribed temperature bands of th ese subsystems when the EDG has been at rest for an extended period of time with the pre-lube oil and jacket water circulating systems operat ional. It should be noted that the semi-annual fast start test de scribed in paragraph C.2.2.3 is performed from 'normal standby conditions'

4. Paragraph C.2.2.6, "Combined SIAS and LOOP Tests". Exception is taken to the statement that the EDG be tested for proper response to a loss-of-offsite power (LOOP) in conjunction with a safety injection actuation signal (SIAS) in whatever sequence they might occur. The EDGs are tested for response to a LOOP, to a SIAS and to a LOOP and SIAS when they occur concurrently. Performing a LOOP/SIAS EDG test in whatever sequence they might occur is beyond the Stations' original licensing basis, provides no additional value, and was not included in Regulatory Guide 1.9, Revision 2 or Regulatory Guide 1.108. 5. Paragraph C.2.2.7 and C.

2.2.8, Single Load Rejection Test and "Full-Load Rejection Test". Exception is taken to the specified power factor requirements of this paragraph. Single and full load rejection tests have been successfully performed at rated power factor. During these tests the generator transient voltage approached the manufacturers recommended voltage limit. Such an event is considered to border on destructive testing and can cause premature generator aging. As a result, the station will continue to perform single load and full-load rejection tests; however, not at rated power factor.

B.0-64 REV. 17, APRIL 2008 Regulatory Guide 1.108 has been superseded by Regulatory Guide 1.9 Revision 3 A majority of the guidance provided in RG 1.108 has been transferred to RG 1.9 Rev. 3. The following discussion contains the applicable responses to these sections of RG 1.9 Rev. 3.

The preoperational test program (FSAR Table 14.2-36) was designed to show functional compliance to the operational requirements set forth in this guide.

A separate demonstration test program was exercised on a one-time basis to show compliance with the availability requirements of this guide. The demonstration test of the LaSalle HPCS diesel generator (all LaSalle diesel generators are design equivalents) was in response to the NRC' s request letter of December 17, 1976, O. D. Parr to GE's Dr. Sherwood. A proposed test program was accepted with minor comments and recommendations by the NRC staff by letter of March 31, 1978 from O. D. Parr to Dr. Sherwood. GE completed the tests with the staff recommendations and reported the results in NEDO-10905 (Amendment 3) which was submitted to O. D. Parr from J. F. Quirk (GE) on December 20, 1979. These test results and conclusions demonstrate the acceptability of this HPCS diesel generator configuration for most BWR-5 pl ants and all BWR-6 plants. Specific reference was made to the LaSalle HPCS diesel generator because the demonstration test was performed at LaSalle with LaSalle equipment.

The following objectives were demonstrated successfully:

1. Generator acceleration and bulk load acceptance within acceptable time and performance parameters.

2. Confirmation of analytic prediction models used for design and factory test of this diesel engine and load combination.

3. Established a 0.99 reliability for DG set starting through load acceptance (for the LaSalle diesel generator per se, and for identical units elsewhere).

4. Margins of 20% excess rated engine torque; attainment of max rated bhp in 19 seconds with an 8 second margin on the requirement; and a 10% margin on minimum load carrying capability without exceeding manufacturer's design limits. The long term continuous load rating was demonstrated to be 30%. This is the power margin for operation during accident conditions when the reactor pressure is high. Full test

B.0-65 REV. 14, APRIL 2002 results are included in NEDO-10905-3 which is part of the LaSalle docket. Sixty-nine consecutive start-ups through load acceptance form both cold and hot engine conditions demonstrated the 0.99 performance reliability number for the LaSalle units.

From these two test programs the following test objectives are covered: initial start; loss of a-c power; rated load; load shedding; load transfer; and, stability of long term operation.

The demonstration test program of Pa ragraphs C.2.a and C.2.b of RG 1.108 (Paragraphs C.2.2 and C.2.3 of RG 1.9 Rev.

3) are not repeated every 18 months. However, it should be noted that the demonstration test program of paragraph C.2.a(9) of RG 1.108 (paragra ph C.2.3.1 of RG 1.9 Rev.

3) regarding 69/n start test, is only a requirement during the plant preoperational test program and not required every 18 months was complied with only during preoperational testing.

This clarification was made by a NRC errata sheet dated September 1977 to RG 1.108. By letter of April 7, 1980 from O. D. Parr to GE's Dr. Sherwood, the NRC concluded that these test resu lts qualified the DG units fo r LaSalle. As a result of this qualification testing and the comprehensive pre-operational tests it is understood that Paragraphs C.2.a and C.

2.b of RG 1.108 (Paragraphs C.2.2 and C.2.3 of RG 1.9 Rev. 3) are not a part of the normal operational surveillance requirement. The Edison alternate to these recent requirements promulgated

under RG 1.108 (RG 1.9 Rev. 3) has been previously submitted with LaSalle Technical Specification.

Regulatory Guide 1.9 Rev. 2 has been superseded by Regulatory Guide 1.9 Revision 3. A majority of the guidance provided in RG 1.9 Rev. 2 has been transferred to RG 1.9 Rev.3. The following discussion contains the applicable responses to these sections of RG 1.9 Rev. 3.

UFSAR Chapter 8.0 shows that the continuous rating of the diesel generator is greater than the maximum coincidental steady-state loads requiring power at any time. Intermittent loads such as motor-op erated valves are not considered for long-term loads.

UFSAR Table 8.3-1 contains the maximum expected coincidental loads for each of the diesel-generator sets, for LOCA conditio ns, and for safe shutdown conditions The 2000-hour rating of each standby diesel generator is 2860 KW, and the 30-minute rating is 3040 KW (UFSAR Table 8.3-3).

UFSAR Chapter 8.0 illustrates that the 2000-hour rating of the HPCS diesel-generator, the 90% of 30-minute rating, and the maximum coincidental load, are in conformance with this guide.

During preoperational testing, the predicted standby diesel-generator loads for each ESF Division were verified, as well as the capability of the diesel generators to carry these loads.

B.0-66 REV. 14, APRIL 2002 Preoperational test results for the LaSalle Un it "0" and "1A" diesel generators were reported to the NRC on March 25, 1982. Initial test results indicated noncompliance with the underfrequency an d undervoltage guidelines of Position C.4. RG 1.9 Rev. 2 (Position C.1.4 of RG 1.9, Rev. 3). Result indicated an undervoltage of 63.7% (2650 volts) and an underfrequency of 94.3% (56.5 Hz) and a frequency recovery of the load sequence interval of 82.5% versus the C.4-RG 1.9 Rev. 2 (C.1.4-RG 1.9 Rev. 3) guide lines of 75%, 95% and 60%, respectively.

Justification of the LaSalle diesel generator design is as follows:

1. The preoperational test used actual emergency core cooling system loads and demonstrated that the diesel generators have the capability

to start and accelerate for all loads to rated speed within the required time period without failing.

2. A number of margin tests demonstrated that the diesel generators can accept a step load increase that is greater than the actual step load and can endure voltage drops of 60, 58.7, and 51.5 percent without experiencing instability resulting in generator voltage collapse or inability of the voltage to recover.

3. During load sequencing, no loads tripped due to undervoltage.

4. The voltage recovered to within 10% of the nominal voltage in less than 50% of the load sequence in terval (approximately 2 seconds).

5. The frequency recovered to within 2% of nominal before the next load was applied automatically.

In addition, motors connected to the Clas s 1E buses that will be subject to the undervoltage transients, are designed to start and accelerate their loads with terminal voltages at 80%. The time interv al that the bus voltage will be below 80% due to the starting transient is less than 1 second and will not appreciably reduce the overall reliability of the motors.

Based upon these justifications, the LaSalle County Station diesel generator performance demonstrates practical compliance with Position C.4 of RG 1.9 , Rev. 2 (Position C.1.4 of RG Rev. 3) A requirement was levied to monitor that the undervoltage trip signal is less than or equal to 3 seconds when subjected to a 40%

voltage drop.

Position C.2.a.(8) of RG 1.108 (Position C.2.2 of RG 1.9, Rev. 3) allows the Division 1 and Division 2 diesel generators to be ru nning in standby at 50 droop. Under this rare condition the Division 1 and 2 diesel generators do not meet the voltage requirements of RG 1.90 during the ECCS pump starts. A test of the Division 1 diesel generator (worst case) was done for two conditions. One test was a B.0-67 REV. 14, APRIL 2002 simultaneous ECCS initiation with bus underv oltage, while the dies el was in a test mode at 50 droop. The other test was an ECCS initiation signal with the diesel in a test mode at 50 droop, followed later by bus undervoltage. The tests verified the capability of the diesels to pick up the pump loads with no indication of any possible failures due to the voltage transient and recovery time.

The suitability of each standby diesel generator is confirmed by prototype qualification test data and by preoperational tests.

The HPCS diesel-generator unit is considered as a unique application, justifiable departure from the strict conformance to RG 1.9 regarding voltage and frequency limits during the initial loading transient.

The HPCS system consists of one large pump and motor combination which represents more than 90% of the total load; consequently, limiting the momentary vo ltage drop to 25% and the momentary frequency droop to 5% would not significantly enhance the reliability of HPCS operation. To meet the specific Regulatory Guide requirements, a diesel-generator unit approximately two to three times as large as that required to carry the continuous rated load, would be necessary. The specified diesel engine-electric generator-pump assembly was designed specifically for this integral operation. The frequency and voltage overshoot requirements of RG 1.9 are met. A factory testing program on a prototype unit has verified the following functions:

1. system fast-start capabilities, 2. load-carrying capability, 3. load shedding capability, 4. ability of the system to accept and carry the required loads, and

5. the mechanical integrity of the diesel-engine generator unit and all of the major system auxiliaries.

GE licensing Topical Report, "HPCS Power Supply, NEDO-10905", describes the theoretical analytical aspects of the uni que application including prototype and reliability test considerations.

The design of the HPCS diesel-generator conforms with the applicable sections of IEEE Criteria for Class I.E "Electrical Systems for Nuclear Power Generation Stations", IEEE Standard 308-1971.

The generator has the capability of providing power for starting the required loads with operationally acceptable voltage and frequency recovery characteristics. A partial or complete load rejection will not cause the diesel-engine to trip on overspeed.

B.0-68 REV. 14, APRIL 2002 A special prototype test was conducted at the LSCS facility to field (site) verify the hardware real load aspects of the HPCS power supply concept. This test was conducted in February 1979. This prototyp e test verified the acceptability of the HPCS power supply concept.

The HPCS diesel generators utilized in LSCS Units 1 and 2 are in compliance with the intent of this guide through the alternate approach cited above. The other standby-power diesel-generators and their loading schemes are in compliance with the guidance set forth in this guide.

REGULATORY GUIDE 1.89 - REV. 0 QUALIFICATION OF CLASS 1E EQUIPMENT FOR NUCLEAR POWER PLANTS Regulatory Guide 1.89 describes a method for complying with regulations regarding to design verification of Class 1E equipment for service in light-water-cooled and gas-cooled nuclear power plants.

All environmental and seismic qualification testing of Class 1E equipment within the NSSS scope was in compliance with IEEE 323-1971 and IEEE 344-1971 which were NRC accepted standards and whic h GE believes represent an adequate protection standard. Specific evaluations are given in Subsections 3.10, 3.11, 7.2.2, and 7.3.2.

With regard to environmental qualification of all balance-of-plant (non-NSSS) instrumentation, control, and electrical equipment important to safety, the applicant complies with the intent of IEEE 323-1971. Additional information is provided in Subsection 8.3.3.

Based upon the assessed capability of the LSCS design, we believe that the Class 1E equipment at LSCS is in compliance with the intent of this guide via the alternate approach cited above.

REGULATORY GUIDE 1.90- REV. 0 INSERVICE INSPECTION OF PRESTRESSED CONCRETE CONTAINMENT STRUCTURES WITH GROUTED TENDONS This regulatory guide is not applicable to the LaSalle County Station, Units 1 and 2, since this guide does not cover the ty pe of containment utilized on LSCS.

B.0-69 REV. 14, APRIL 2002 REGULATORY GUIDE 1.91- REV. 0 EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANT SITES Regulatory Guide 1.91 describes a method acceptable to the NRC for determining safe distances from a nuclear power plan t to a transportation route over which explosive material (not including gases) may be carried.

The applicant complies with Regulatory Guide 1.91 in the sense that the tornado design pressure is higher than the overpressures of the postulated explosions on the nearby transportation routes. For details, refer to Section 2.2.

We believe that we comply with the objectiv es set forth in this regulatory guide.

REGULATORY GUIDE 1.92- REV. 1 COMBINATION OF MODES AND SPATIAL COMPONENTS IN SEISMIC RESPONSE ANALYSES This regulatory guide describes methods acceptable to the NRC for combining the values of the response spectrum modal dynamic analyses, and combining maximum values of a time-history dynamic analyses or in combining maximum values of a spectrum dynamic analyses.

a. Combination of Modal Response

1. In response spectrum method of analysis, the probable ma>~imum response is obtained by using the double sum criterion as explained in Subsection 3.7.2.7 for BOP structur es, and in Subsection 3.7.3.7 for BOP and NSSS subsystems. The method takes into account the separated modes as well as the closely spaced modes.

2. All piping and equipment analyzed or supplied by General Electric are evaluated to the requirements of this regulatory guide as described in Subsection 3.7.3.7.

b. Combination of Spatial Components

1. Time-History Method

B.0-70 REV. 14, APRIL 2002 When the time-history dynamic analysis is used and the three spatial components are statistically independent, the response due to each spatial component is algebraically added at each time instant and the maximum is obtained from the combined response.

2. Response Spectrum Method For NSSS components and equipment, the calculated maximum responses due to one horizont al directional earthquake excitation are combined with the responses due to the vertical earthquake by the sum of the absolute values method. The maximum responses due to another perpendicular horizontal earthquake are also combined with the responses due to the vertical earthquake in the same manner. The larger of the two values is used for the design,. (i.e., the larger of [x+y] and [y+z]

where x and z are horizontal and y is vertical.

REGULATORY GUIDE 1.93 - REV. 0 AVAILABILITY OF ELECTRIC POWER SOURCES Regulatory Guide 1.93 describes acceptable operating procedures and restrictions which should be implemented if the available electric power sources are less than the limiting conditions of operation.

Regulatory Guide 1.93 was issued after the LaSalle County Station construction permit. The LSCS design generally meets the intent of the regulatory positions, as outlined in Chapter 8.0. A specific network stability analysis for LSCS is included in Section 8.2.

REGULATORY GUIDE 1.94 - REV. 1 QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PH ASE OF NUCLEAR POWER PLANTS The construction of LaSalle County Station was well underway at the time Regulatory Guide 1.94 was published. Howe ver, all quality assurance programs for LSCS are in line with ANSI N45.2.5-1974 ex cept that in some cases the type and frequency of these tests are not identical. For LSCS quality control programs on structural concrete and steel see Appendix E.

LaSalle County Station has cadwelded all positive connections for reinforcing bars; therefore the section on welding reinforcing bar splices is not applicable.

B.0-71 REV. 14, APRIL 2002 REGULATORY GUIDE 1.95 - REV. 0 PROTECTION OF NUCLEAR POWER PLANT CONTROL ROOM OPERATORS AGAINST AN ACCIDENTAL CHLORINE RELEASE The only potential source for accidental chlorine release are chlorine-carrying barges which occasionally pass about 5 miles north of the Site. The effects of this accident are discussed in Subsection 2.2.3.2.

The detailed description of the control room HVAC system covered under Section 6.4 and Subsection 9.4.1 of LSCS-UFSAR.

The detection systems installed at LSCS ar e consistent with the recommendation of this regulatory guide.

REGULATORY GUIDE 1.96 - REV. 1 DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR BOILING WATER REACTO R NUCLEAR POWER PLANTS LSCS Unit 1 MSIV-LCS has been abandone d in place, and Unit 2 MSIV-LCS has been removed.

REGULATORY GUIDE 1.97 - REV. 2 Regulatory Guide 1.97 Rev. 2, Instrumentation for Light-Watered-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions during and following an Accident, describes a method acceptable to the NRC for monitoring plant variables and systems during and following an accide nt. LaSalle is committed to compliance with Reg. Guide 1.97 Rev. 2 as documented in reference 1 with noted exceptions and deviations. Additional information and clarification was provided to the NRC in response to an interim report to CECo from the NRC. (References 2 and 3). A final report summary of Human Factors Review of Reg. Guide 1.97, Rev. 2 was later submitted to the NRC (Ref. 4). The NRC accepted the Commonwealth Co. position for LaSalle County Station Units 1 and 2 as complying with the recommendations of Reg. Guide 1.97 Rev. 2 with the exception that CECo commit to installing and having operational monitoring instrumentation for neutron flux which fully meets the recommendations of Reg. Guide 1.97, Rev. 2 (Reference 5). Commonwealth Edison responded in later correspon dence that LaSalle would follow the recommendations of the BWR Owners Group in regard to upgrade of the neutron monitoring system when available. (Reference 6). The NRC accepted ComEd's B.0-72 REV. 14, APRIL 2002 submittal of compliance to Reg. Guide 1.97 Rev. 2 for neutron flux instrumentation with justification for non-conformance from the NEDO 31558-A that is consistent with the BWROG position (References 7 and 8).

In response to an NRC inspection of LaSalle for compliance with Reg. Guide 1.97, additional clarification was provided in regard to the noted deviations of the instrumentation ranges of the Containmen t Hydrogen Concentration Monitors and the Reactor Vessel Water Level (References 9 and 10). In addition a concern for the acceptability of electrical isolation devices between Reg. Guide 1.97 instrument circuitry and the non-safety process comput er was addressed by CECo with analysis and testing (References 11 and 12)

References

1. 6/29/82 Letter from C.W. Schroed er CECO to A. Schwencer NRC on Commonwealth Edison's position of Reg Guide 1.97 for LaSalle Station with noted exceptions. 2. 12/13/84 NRC letter from A Schwencer to Dennis L. Farrar CECO, Interim Report on CECo response to Reg Guide 1.97 Rev 2. 3. 2/22/85 letter from G.L. Alexander CECO to H. R. Denton NRC on CECO response to NRC interim report on CECO compliance to Reg Guide 1.97 Rev 2. 4. 8/1/86 letter from C. M. Allen CECO to H. R. Denton NRC on LaSalle Human Factors Review of Reg Guide 1.97 Instrumentation. 5. LaSalle Reg Guide 1.97 SER transmitted to L. D. Butterfield on 8/20/87 from D. Muller on Emergency Response Cap ability, Conformance to Reg Guide 1.97 Rev 2 for LaSalle Unit 1 and 2. 6. 10/7/87 letter from C.M. Allen CECO to NRC Document Control Desk on Neutron Flux Monitoring. 7. 6/21/99 letter from Jeffery A. Benjamin to the NRC on Compliance with Reg. Guide 1.97 - BWR Neutron Flux Monitoring. 8. 9/7/99 NRC letter from Donna M. Skay NRC to O. Kingsley ComEd on Reg Guide 1.97 - BWR Neutron Flux Monitoring - LaSalle County Station Units 1 and 2 (TAC No. M77660). 9. 12/5/88 NRC letter from Hubert J.

Miller NRC to Commonwealth Edison, Notice of violation for items identified by RG 1.97 Inspection. Unresolved items. 10. 3/2/89 CECO Response to NRC from W. E. Morgan CECo to A Bert Davis NRC. 11. 11/9/90 letter from W. E. Morgan CECo to A Bert Davis, NRC on Response to Inspection Report 373-90022 and 374-90023. 12. 6/21/91 letter from Peter Piet CECo to NRC on Supplement to response on Inspection Report 373-90022 and 374-90023.

Regulatory Guide 1.99 Rev. 2 Regulatory Guide 1.99 Rev. 2, "Radiation Embrittlement of Reactor Vessel Material," date May 1988, required updating Technical Specification B.0-73 REV. 14, APRIL 2002 Pressure/Temperature Curves and have been approved by NRC in Amendments 71 and 55 dated for LaSalle Unit 1 and Unit 2 respectively.

Regulatory Guide 1.100, Rev. 1 Regulatory Guide 1.100, Rev. 1, descri bes a method acceptable to the NRC for seismic qualifications testing of electric equipment. LaSalle complies with the intent of the guidance provided in that:

a) Equipment qualified prior to the issuance of the Facility Operating License was either:

1) Qualified in accordance with Regu latory Guide 1.100, Rev. 1, (which endorses IEEE 344-1975, with a few limitations), or

2) Qualified in accordance with IEEE 344-1971 and evaluated and found to meet the intent of Regu latory Guide 1.100, Rev. 1.

b) New and replacement (except for "like-for-like" replacement) equipment is qualified in accordan ce with Regulatory Guide 1.100, Rev. 1.

c) The guidance is also applied to mechanical equipment and instrumentation that is required to be seismically qualified.

Regulatory Guide 1.106, Rev. 1 Regulatory Guide 1.106, Rev. 1, "Thermal Overload Protection for Electric Motors on Motor-Operated Valves," describes a method acceptable to the NRC for design, erection, and testing of thermal overload protection for electric motors on motor-operated valves important to safety.

Except for the Hydrogen Recombiner System motor-operated valves, both NSSS and Non-NSSS safety-related equipment meet the requirements of the Regulatory Position, C.1, using both alternate positions, as determined to be appropriate based on the design function of each motor-op erated valve. The Hydrogen Recombiner System motor-operated valves meet Regulato ry Position C.2. The methodology used to meet Regulatory Position C.2 is in accordance with IEEE-741-1990. More information concerning actual implementation of this Regulatory Guide is discussed in Subsection 6.3.2.2.13.

REGULATORY GUIDE 1.108 Initial Issue: Revision 0, August 1976 B.0-74 REV. 15, APRIL 2004 Current Issue: Revision 1, August 1977 LaSalle C.P. Issued: September 10, 1973 PERIODIC TESTING OF DIESEL GENERATORS USED AS ONSITE ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS Regulatory Guide 1.108 describes a method acceptable to the NRC for preoperational and periodic testing of diesel electric power units to ensure their availability requirements (0.99 reliability at 50% confidence).

This Regulatory Guide has been superseded by Regulatory Guide 1.9, Revision 3.

REGULATORY GUIDE 1.116 - REV. 0 QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF MECHANICAL EQUIPMENT AND SYSTEMS

This guide describes a method acceptable to the NRC for fulfilling QA requirements on the installation, inspection, and testing of mechanical equipment and systems for water- cooled nuclear power plants.

ANSI N45.2-1975 is used to delineate preoperational tests and other functional tests.

The Exelon Nuclear Topical Report which outlines the Exelon Nuclear QA program with respect to inspection, and testing of safety-related equipment applies to these mechanical systems.

We comply with the objectives of the referenced revision of this regulatory guide as indicated by the practices referenced above.

REGULATORY GUIDE 1.118 - REV. 1 PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS Regulatory Guide 1.118 describes a method acceptable to the NRC for periodic testing of the protection system and electrical power systems important to safety. This is the implementation of IEEE Standard 338-175 which is ancillary to IEEE 279-1971 and IEEE 380-1974.

The NSSS design complies with this guideline to enable response time testing based on relay contact internals, but not sensor responses to stimulating media.

B.0-75 REV. 15, APRIL 2004 Relative to the regulatory position given in paragraphs C.l through C.7, the LSCS design generally meets the intent of this guide with the exception of provisions for response time testing. Because of the issue date of this document, no specific design features were originally included in the non-NSSS systems to facilitate response time testing.

NRC Implementation: To be used in the evaluation of submittals with construction permit applications docketed after January 1, 1978.

REGULATORY GUIDE 1.123 - REV. 1 QUALITY ASSURANCE REQUIREMENTS FOR CONTROL OF PROCUREMENT OF ITEMS AND SERVICES FOR NUCLEAR POWER PLANTS

This regulatory guide describes a method acceptable to the NRC for fulfilling QA requirements relative to the procurement of items and services during design, construction and operations ph ases of nuclear power plants.

The LSCS FSAR was docketed with the OL application prior to issuance of this regulatory guide. Nevertheless, the Edison "QA Progran for Nuclear Generating Stations," CE-1-A app1ies to procurement of modification items and services for safety-related systems as out lined in that Topical Report.

Edison complies with the objectives of the referenced revision of this regulatory guide. Regulatory Guide 1.181, Rev. 0 (09/99)

As part of the ongoing effort to improve the quality of the UFSAR, the guidelines provided in NEI 98-03 Revision 1, 06/99 and endorsed by the NRC in Regulatory Guide 1.181, Revision 0, 09/99 will be used to further improve the content of the UFSAR. While the UFSAR will continue to follow the general organizational recommendations specified in Regulato ry Guide 1.170, Rev. 2 (09/75), the reformatting options described in NEI 98-03, Revision 1, will be used to simplify information contained in the UFSAR to improve their focus, clarity and

maintainability. Examples of the types of improvements that are anticipated are the identification of historical inform ation and the removal or replacement of excessively detailed drawings as described in Appendix A to NEI 98-03 Revision 1.

Regulatory Guide 1.190, dated March 2001 Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" provides state of the art calculation and measurement procedures that are acceptable to the NRC for determining Reactor Pressure Vessel (RPV) neutron fluence.

B.0-75a REV. 15, APRIL 2004 In Technical Specification Amendments 160 (Unit 1) and 146 (Unit 2), the NRC approved LaSalle's participation in the In tegrated Surveillance Program. Specific details are provided in UFSAR section 4.1.4.5. However, as a co ndition of granting the Amendments, LaSalle committed that all future evaluations of RPV fluence will be completed using a method in accordance with the recommendations of Regulatory Guide 1.190.

Regulatory Guide 1.32 Rev. 2 "Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants" Regulatory Guide 1.32 endorses IEEE Standard 450-1975, "IEEE Recommended Practice for Maintenance, Testing, an d Replacement of Large Lead Storage B.0-76 REV. 14, APRIL 2002 Batteries for Generating Stations and Substations." The regulatory guide specifies what type of battery tests and the test ing interval based on IEEE Standard 450-1975. The applicant complies with these regu latory positions in Revision 2 with the following clarification.

1. Paragraph C.1.c. Exception is taken to the requirements of IEEE Standard 450-1975 for battery testing. IEEE Standa rd 450-1995 will be utilized to perform battery testing. A battery service test will be performed to verify battery capacity is adequate to supply, and ma intain in OPERABLE status, the required emergency loads for the design duty cycle. A battery performance test discharge test or a modified performance discharge test will be performed to verify the battery capacity is >80% of the manufacture's rating. The modified performance test may be performed in lieu of the service test provided the modified performance test completely envelops the service test. The battery testing intervals are in accordance with the Technical Specification requirements.

LSCS-UFSAR C-i REV. 15, APRIL 2004 APPENDIX C ANALYSIS OF EFFECTS OF HIGH-ENERGY LINE BREAKS OUTSIDE PRIMARY CONTAINMENT TABLE OF CONTENTS PAGE C.1 ABSTRACT C-1 C.2 INTRODUCTION C-1 C.3 DEFINITIONS C-2 C.4 METHOD AND CRITERIA USED TO DETERMINE THE NECESSITY OF A COMPLETE ANALYSIS C-2 C.5 PIPE WHIP C-3 C.6 METHOD FOR CALCULATING THE JET IMPINGMENT FORCE C-3 C.7 CONCLUSIONS C-4 ATTACHMENT C1 - DESCRIPTION OF "PWUR" COMPUTER PROGRAM C1-1 LSCS-UFSAR C-ii REV. 15, APRIL 2004 APPENDIX C ANALYSIS OF EFFECTS OF HIGH-ENERGY LINE BREAKS OUTSIDE PRIMARY CONTAINMENT LIST OF TABLES NUMBER TITLE C-1 Postulated Rupture Orientation and Number Of Design Basis Breaks Used in Rupture Analysis Outside Containment C-2 Results of Dynamic Analysis for Postulated Pipe Rupture Outside Containment

LSCS-UFSAR C-iii REV. 15, APRIL 2004 APPENDIX C ANALYSIS OF EFFECTS OF HIGH-ENERGY LINE BREAKS OUTSIDE PRIMARY CONTAINMENT LIST OF FIGURES NUMBER TITLE C-1 Whip Restraints on the Steam Tunnel Area C-2 Feedwater System - Postulated Breaks and Restraint Locations C-3 Main Steam - Postulated Breaks and Restraint Locations C-4 Condensate Booster - Postulated Breaks and Restraint Locations C-5 RCIC - Postulated Breaks and Restraint Locations C-6 Jet Model - Flashing/Expanding Fluid C-7 Jet Model - Non-Flashing/Non-Expanding Fluid

LSCS-UFSAR C.0-1 REV. 15, APRIL 2004 APPENDIX C ANALYSIS OF HIGH-ENERGY LINE BREAKS OUTSIDE PRIMARY CONTAINMENT The analytical methods and results of th e high-energy line break analysis for the LaSalle plant is documented in this Appendix.

LSCS-UFSAR C-1 REV. 15, APRIL 2004 APPENDIX C ANALYSIS OF EFFECTS OF HIGH-ENERGY LINE BREAKS OUTSIDE PRIMARY CONTAINMENT C.1 ABSTRACT The purpose of this appendix is to record in detail the methods used in the pipe whip analysis for the LaSalle County Station (LSCS). It also contains the results of the pipe whip and jet impingement analysis of systems outside the containment vessel at LSCS, as well as an exposition of the method of analysis and the criteria for postulating breaks. The conclusions obtained for LSCS Unit 1 are also appropriate for LSCS Unit 2.

The appendix concludes with the action taken in the specific areas or components needing protection and in the design of restraints required for this purpose.

C.2 INTRODUCTION Piping systems which exceed temperatures above 200

°F and/or pressures above 275 psi during "normal plant conditions" are cla ssified as high-energy lines according to NRC guidelines. These high-energy lines ar e considered as possible sources of pipe whip. Some systems, such as the RHR system , exceed these criteria less than 2% of the time that the systems operate as mode rate energy systems, and therefore, do not qualify as high energy systems.

See MEB 3-1, 8/5/77, paragraph B.2.c.

The postulation of circumferential and longitudinal breaks on high-energy lines of 4-inch diameter or more is based on th e A. Giambusso letter of December 15, 1972 which recommends that the breaks be cons idered at points where stresses exceed 0.8 (S A + S h). However, per Branch Technical Position MED 3-1 (Revision 0), the postulation of Class 2 and 3 piping (ASME Code,Section III) pipe ruptures in areas other than containment penetration is postulated at the following locations in those portions of each piping and branch run:

a. At terminal ends. b. At intermediate locations selected by the following criteria: At each location where th e stresses exceed 0.8 (1.2 S h + S A) but at not less than two separated locations chosen on the basis of highest stress. Where the piping consists of a straight run without fittings, welded attachment, or valves, and all stresses are below 0.8 (1.2 S h + S A), a minimum of one location chosen on the basis of highest stress.

LSCS-UFSAR C-2 REV. 15, APRIL 2004 Normally break locations are assumed at locations that would maximize EQ (Environmental Qualification Values) impact not on pipe stress. The above break exclusion criteria is applied when the postulated break analysis produced consequences which are too conservative.

The following systems were analyzed:

a. feedwater heater drain lines, b. reactor core isolation cooling lines, c. reactor water cleanup lines, d. condensate booster lines, e. extraction steam lines, f. control rod drive hydraulic lines, g. main steamlines, and h. feedwater lines.

The effects of pipe rupture in all of the above systems on essential components and structures were determined for the case of both pipe whip and jet impingement.

The method of analysis for pipe wh ip is obtained in Attachment C1.

There are no safe shutdown SSCs in the i mmediate vicinity of breaks postulated for the extraction steam system and the feedwater heater drain lines. Therefore, no further evaluation is necessary of these systems.

Double-ended guillotine breaks are postulated in Rooms 20A, 21A, 22A, 23A, 26, 27, 28A, 29A, 30A, 30B, 30C, 36, 45A, 48, 49, and 50 throughout the reactor water cleanup (RT) piping.

C.3 DEFINITIONS See Attachment C1.

C.4 METHOD AND CRITERIA USED TO DETERMINE THE NECESSITY OF A COMPLETE ANALYSIS Breaks are postulated by marking the hi gh-energy lines on a set of composite drawings at every pipe fitting and identifying the breaks in numerical sequence. (The original composite drawings are historical and have not been maintained.) If it is determined that there is no compon ent or structure affected by a particular break, no further analysis is required. If a component or structure is affected, further analysis is only required when:

LSCS-UFSAR C-3 REV. 15, APRIL 2004

a. The component is essential to ensure safe shutdown of the plant or is required to mitigate the consequences of the accident. b. The ruptured pipe collides with or otherwise affects another pipe of equal or smaller size. c. The component is a cable pan ca rrying electrical safety-related hardware. d. The structure is Seismic Category I or safety-related Class I. e. Examination of the composite drawings may or may not provide sufficient information to enable the determination of whether or not any of the conditions (above items a through d) are applicable. In case sufficient information is not available, it is assumed that conditions of items a through d pertain.

C.5 PIPE WHIP Based on the information contained in th e "PWUR" computer output, certain areas in the steam tunnel are provided with pi pe whip protection. Restraints RMS-1 through 8 and RFW-1 through 9 were located at selected points for this purpose. In order to meet the valve vendors allowabl e loading, it was necessary to include additional restraints RFW-10 and RFW-11 and design restraints RMS-1, RMS-2, RMS-3, RMS-4, RFW-1, and RWF-2 to allow deflections of 0.3 inch or less at that location. These restraints are located as shown in Figure C-1.

In the above-grade areas of the Turbine Driven Reactor Feed Pumps (TDRFP) rooms, additional restraints were prov ided to protect structural components.

The postulated pipe rupture orientation an d number of design basis breaks used in the analysis of high energy systems outside containment are shown in Table C-1.

The individual break locations within each system and the whip restraints are provided in Table C-2. Figures C-2 through C-5 schematically illustrate the information provided in Table C-2.

It is normally recommended that a pipe whip restraint be connected to a pipe at a distance from the elbow less than the minimum distance to plastic hinge formation. This ensures that a hinge cannot form ahead of the restraint or at the restraint itself. C.6 METHOD FOR CALCULATING THE JET IMPINGMENT FORCE The models used for calculating jet impingement forces are shown in Figures C-6 and C-7.

LSCS-UFSAR C-4 REV. 15, APRIL 2004 The impingement force on a target is found by first calculating the average pressure of the jet at the location of the target according to the equation.

jet A F jet P= Where Pjet = average jet pressure at the required location F = blowdown force Ajet = area of the jet at the required location, The impingement force is then F imp = impingement force on target Where F imp = impingement force on target K o = shape factor Pjet = average jet pressure as defined above A target = area of the target.

Affected cable pans that cannot withstand these forces will be rerouted or shielded from the jet.

C.7 CONCLUSIONS Analysis of the effects of high-energy line breaks is performed to determine where pipe whip restraints are necessary. Based on the information provided by the analysis, pipe whip restraints are prov ided where needed to protect essential components or structures. Likewise, analysis of jet impingement due to high-energy line breaks is performed to determine if shields are necessary to protect essential components or structures.

Subsequently to the original analysis presented here, the Class 1 piping stress reports were completed. At that time the piping stress and fatigue factors were LSCS-UFSAR C-5 REV. 15, APRIL 2004 used to identify potential HELB's in accordance with the appropriate guidelines (USNRC Branch Technical Position MEB 3-1). The use of the pipe stress results reduced the number of required whip rest raints and also eliminated 19 of the original 22 "jet impingement hits" on cable pans. With this reduced number of pipe breaks, combined with the Safe Shutdown Analysis (UFSAR Appendix H) and the EQ System Safety Evaluation, it is concluded that neither jet impingement shields nor equipment shields are required. This conclusion is based on the physical independence or alternate safety systems designed for safe reactor shutdown and removal of decay heat from the shutdown reactor system.

LSCS-UFSAR C1-i REV. 15, APRIL 2004 ATTACHMENT C1 DESCRIPTION OF "PWUR" COMPUTER PROGRAM -

ENERGIES AND LOADS FOR UNRESTRAINED PIPE WHIP TABLE OF CONTENTS PAGE C1.0 Introduction C1-1 C2.0 Definitions C1-2 C2.1 Pipe Breaks C1-2 C2.1.1 Circumferential Break C1-2 C2.1.2 Longitudinal Break C1-2 C2.2 Blowdown Force C1-2 C2.3 Plastic Hinge C1-3 C2.4 Plastic Hinge Distance C1-3 C2.5 Minimum Distance to Plastic Hinge Formation C1-3 C2.6 Terminal Ends C1-4 C2.7 Restraint C1-4 C2.8 Pipe Run C1-4 C2.9 Distance to Affected Structure C1-4 C3.0 Procedure C1-4 C4.0 Interpretation of Results C1-5 C5.0 References C1-6

LSCS-UFSAR C1-ii REV. 15, APRIL 2004 ATTACHMENT C1 ENERGIES AND LOADS FOR UNRESTRAINED PIPE WHIP LIST OF FIGURES NUMBER TITLE C1-1 Physical Layout for Circumferential Break C1-2 Physical Layout for Longitudinal Break C1-3 Possible Hinge Formulations C1-4 Model for Problem Formulation C1-5 Circumferential Break Model - Original Formulation C1-6 Circumferential Break Model - Reformulation C1-7 Circumferential Break Model - Formulate L-Break C1-8 Longitudinal Break Model - L rl < L s C1-9 Longitudinal Break Model - Reformulation C1-10 Longitudinal Break Model - Formulate C-Break

LSCS-UFSAR C1-1 REV. 15, APRIL 2004 C

1.0 INTRODUCTION

According to NRC regulatory guides, essential (safety-related) systems and components required to safely shutdown the plant need to be protected from the effects of high energy pipe breaks, including pipe whip.

Depending on the nature of the break, four different kinds of protection can be provided against pipe whip:

a. reinforcement, such as in the case of a structure; b. dissipation of energy by a restraint; c. shielding plates such as in the case of jet impinging upon a cable pan; and d. separation criteria for systems which operate in different areas of the containment.

The study of the effect of pipe whip considering an individual component leads to the following conclusions:

a. The component is capable of assi milating the impact energy generated by pipe whip. b. The component would be able to assimilate the impact energy if reinforcement is provided. c. The component cannot be reinforced and is not suited to assimilate the energy of impact.

The object of the PWUR computer program is to provide information and necessary documentation in aspects such as:

a. source and nature of break; b. break number and type; c. affected component, whether it is a wall, another pipe, valve, electrical or mechanical equipment; d. location of the break and affected component; e. energies and jet impingement loads resulting from circumferential or longitudinal breaks; f. plastic hinge location; and g. minimum distance to plastic hinge formations.

LSCS-UFSAR C1-2 REV. 15, APRIL 2004 C2.0 DEFINITIONS C2.1 Pipe Breaks C2.1.1 Circumferential Break Circumferential breaks are perpendicular to the pipe axis. Dynamic forces resulting from such breaks are assumed to separate the pipe axially and cause whipping in the plane determined by the pipe axis and the next pipe run.

C2.1.2 Longitudinal Break Longitudinal breaks are parallel to the pipe axis. Dynamic forces resulting from such breaks act perpendicular to the pipe axis. The nature of this break is two-fold:

a. The pipe deforms plastically and impacts surrounding components. This situation is referred to as "longitudinal impact." b. A jet is created in the opposite direction of the force deforming the pipe and impinges upon other component(s). This situation is referred to as "jet impingement."

Longitudinal and circumferential breaks do not occur simultaneously at the same location on the pipe, although for analysis purposes, they may be postulated as such when essential components can be affected by either one.

C2.2 Blowdown Force The sudden change of pressure when a break occurs has a two-fold effect:

a. A thrust force of magnitude equal to the initial internal pressure times the break area is developed. b. The flow velocity increases from zero to a limiting terminal velocity, that is, if it is assumed that the pressure P o inside the pipe remains constant for as long as it takes the flow to reach V T. Therefore, the initial stage of the flow where only the static pressure P o exerts a total force of P o times the blowdown area changes to a stage where the total blowdown force is the result of the static pressure plus dynamic pressure times the blowdown area.

F. J. Moody (Reference 1) has shown that for an ideal gas under isen tropic conditions this total force is LSCS-UFSAR C1-3 REV. 15, APRIL 2004 F = 1.26 x P o x A P o = internal operating pressure A = break area.

The blowdown thrust for subcooled water was derived using 2P o A as the steady-state value for frictionless non-flashing water. The magnitude of the steady-state thrust coefficient varies from 1.26 to 2.0 for frictionless fluid depending on the degree of subcooling.

However, the steady-state blowdown forces are reduced by taking frictional effects into consideration.

Therefore, the blowdown thrust values used for calculating pipe motion take into account frictional effects of the piping system.

C2.3 Plastic Hinge A plastic hinge is the point on the pipe where the pipe will deform plastically under stress caused by bending moments.

C2.4 Plastic Hinge Distance When a force is considered acting at a point of a pipe with infinite length, the distance from the force action point to the plastic hinge is defined as the plastic hinge distance.

C2.5 Minimum Distance to Plastic Hinge Formation If the pipe has a terminal end or a restraint located along its axis, then the pipe will become stiffer than a pipe of infinite length. However, if th e constraint is not applied within the plastic hinge distance, it will behave as a pipe of infinite length. When the constraint is within this distance, the plastic hinge will form exactly at the constraint point. As the constraint is moved toward the force application point, the pipe becomes stiffer until it reaches a point when the couple induced by the constraint and the force is equal to the plastic moment of the pipe. The distance from the point of application of the force to this point is labeled as "The Minimum Distance to Plastic Hinge Formation" and is given by B o s F M L=

LSCS-UFSAR C1-4 REV. 15, APRIL 2004 where: M o = plastic moment of the pipe.

C2.6 Terminal Ends Terminal ends of pipe runs originate at any point of maximum constraint such as components rigidly anchored to struct ures, anchor, penetrations, elbows, or restraints.

C2.7 Restraint A restraint is a structural component located around the pipe for the purpose of constraining the pipe. In some cases the restraint is designed to yield under the load caused by the break, but the energy assimilated by the restraint in the deformation process decreases the effect of the pipe impact on the component.

C2.8 Pipe Run Usually taken as the distance between two consecutive elbows or between the elbow adjacent to the break and the next possible place where a hinge could form, such as a terminal end. See Figures C1-1 and C1-2.

C2.9 Distance to Affected Structure For a circumferential break, the distance to the affected structure is taken as the space separating the elbow adjacent to the break and the structure or component located along the line of action of the blowdown force. See Figure C1-1.

For a longitudinal break, this distance is the space separating the point where the break occurs and the structure located along the line of action of the thrusting force.

See Figure C1-2.

C3.0 PROCEDURE The simple model shown in Figure C1-4 is used (Reference 2) to calculate energies imparted by two objects colliding plastically. The mass of the pipe shown in the longitudinal position is lumped into an equivalent mass concentrated at the tip of the pipe along with the existing tip weight, which is the weight of the elbow plus the portion of pipe where the circumferential break occurs.

The velocity at which the pipe moves is assumed to be linear, which implies constant acceleration, which in turn implies that the limiting blowdown force is constant.

LSCS-UFSAR C1-5 REV. 15, APRIL 2004 C4.0 INTERPRETATION OF RESULTS Depending on the nature of the break and the local configuration of the pipe, the following cases are possible:

Under the circumferential break conditions (see Figure C1-5):

a. If the pipe run provided is too small for a plastic hinge to form there will be a message generated "No plastic hinge formed along pipe run." This condition exists when the pi pe run length is less than the minimum distance for pl astic hinge formation. b. If the pipe run provided is large and exceeds the value at which the pipe will hinge under unrestrained conditions. In this case, a message "Hinge formed at a distance 'S' from break elbow" will appear. c. If the pipe run provided is large and exceeds the minimum distance for hinge formation, but not the 'S' valu

e. The message "Hinge formed at the end of the run" will depict this condition.

Under longitudinal break conditions:

d. For a given pipe length there is a force F that will deform the pipe plastically. If this condition is not met, the message "Force is not large enough to deform pipe plastically" will appear. e. The longitudinal break model considers the force applied at the midsection of the pipe. If this is not the case, i.e., the force is applied closer to one end, the smaller distan ce to the elbow or possible hinge point is compared with the minimum distance for hinge information. In the event that the force application point clears this last distance, then the force is considered as if it were applied at the midsection. But if the point of application does not clear the minimum distance for hinge formation, then the message "L break too close to elbow and must be considered separately, see documentation" will appear.

Since the program is not designed to stor e the entire configuration of the pipe and/or the surrounding components, pipes structures, etc., changes cannot be made when the aforementioned conditions of items a or e arise. These cases are reformulated after the results are available according to the following considerations:

For case a:

Reformulate the c-break by considering the pipe run, input previously, as part of the portion of pipe where the break occurs such as indicated in Figure C1-6. Thus, this "new" portion will be composed of the "old

" portion of pipe where the break occurs; LSCS-UFSAR C1-6 REV. 15, APRIL 2004 the "old" pipe run plus the additional pipe portion ranging between the end of the "old" run and the nearest elbow. Close observation of this pipe configuration indicates that a longitudinal break in impact may need to be formulated as well, if there are components at a radial distance from pipe. See Figure C1-7.

For case e:

Two possible considerations are applied in this case:

a. The portion of pipe labeled "D", shown in Figure C1-8, is less than the minimum distance for plastic hinge formation. Therefore, the two elbows adjacent to "D" are disreg arded as possible hinge formations and the break is reformulated as shown in Figure C1-9. b. The portion of pipe labeled "D" is greater than L s, as shown in Figure C1-10; therefore, no longitudinal break in impact can exist at this location, and no formulation is required.

C

5.0 REFERENCES

1. F. J. Moody, "Prediction of Blowdown Thrust and Jet Forces," ASME Trans., 69-HT-31, 1969.

2. H. G. Hopkins, "On the Behavior of Infinitely Long Rigid Beams Under Transverse Concentrated Load," J Mech. Phys. Solids, Vol. 4, 1955.

Table C-1 REV. 15, APRIL 2004 TABLE C-1 POSTULATED RUPTURE ORIENTATION AND NUMBER OF DESIGN BASIS BREAKS USED IN RUPTURE ANALYSIS OUTSIDE CONTAINMENT NUMBER OF BREAKS CIRCUMFERENTIALLONGITUDINAL SYSTEM UNIT 1 UNIT 2 UNIT 1 UNIT 2 Main Steam 8 8 -- -- Feedwater 10 10 3 3 Condensate Booster 7 4 -- -- RCIC 1 1 1 1 Table C-2 REV. 15, APRIL 2004 TABLE C-2 (Sheet 1 of 2)

RESULTS OF DYNAMIC ANALYSIS FOR POSTULATED PIPE RUPTURE OUTSIDE CONTAINMENT SUB- SYSTEM LINE NO. RESTRAINT NO. BREAK LOCATION BLOWDOWNLOAD (KIPS)

BREAK TYPE FW-03 1FW02HA-24 2FW02HA-24 RFW-2 C-444 254.4 C FW-03 1FW02BA-24 2FW02BA-24 RFW-10 C-430 240.3 C FW-03 1FW02BA-24 2FW02BA-24 RFW-11 C-429 167.4 C FW-04 1FW02HB-24 2FW02HB-24 RFW-1 C-435 254.4 C FW-04 1FW02BB-24 2FW02BB-24 RFW-3 C-448 240.3 C FW-05 1FW02AB-30 2FW02AB-30 RFW-5 C-396 562.6/480 C/L FW-05 1FW02AA-30 2FW02AA-30 RFW-6 C-407 587.2/480 C/L FW-05 1FW-34.5 2FW-34.5 RFW-7 C-412 667.6 C FW-05 1FW02A-30 2FW02A-30 RFW-8 C-417 415.6 C FW-05 1FW02BB-24 2FW02BB-24 RFW-9 C-421 216/385 C/L MS-05 1MS01BA-26 2MS01BD-26 RMS-3 C-17 334.5 C MS-05 1MS01BA-26 2MS01BD-26 RMS-7 C-19 334.5 C MS-06 1MS01BB-26 2MS01BD-26 RMS-4 C-25 334.5 C MS-06 1MS01BB-26 2MS01BC-26 RMS-8 C-27 334.5 C

Table C-2 REV. 15, APRIL 2004 TABLE C-2 (Sheet 2 of 2)

RESULTS OF DYNAMIC ANALYSIS FOR POSTULATED PIPE RUPTURE OUTSIDE CONTAINMENT SUB- SYSTEM LINE NO. RESTRAINT NO. BREAK LOCATION BLOWDOWN LOAD (KIPS)

BREAK TYPE MS-07 1MS01BC-26 2MS01BB-26 RMS-1 C-3 334.5 C MS-07 1MS01BC-26 2MS01BB-26 RMS-5 C-5 334.3 C MS-08 1MS01BD-26 2MS01BA-26 RMS-2 C-9 334.34 C MS-08 1MS01BD-26 2MS01BA-26 RMS-6 C-11 334.34 C CB-01 1CB06CA-30 R-134 C-134 119.89 C CB-01 1CB06CB-30 2CB06CA-30 R-135 C-135 119.89 129.1 C CB-01 1CB06CB-30 R-136 C-136 337.89 C CB-01 1CB06CB-30 R-137 C-137 84.77 C CB-01 1CB06CB-30 2CB06CB-30 R-138 C-138 334.5 286.7 C CB-01 1CB06CA-30 R-139 C-139 353.61 C CB-03 1CB06B-36 2CB06B-36 RCB-2 C-658 545 C CB-03 2CB06D-24 RCB-140 C-696 262 C RI-01 1RI01C-4 2RI01C-4 R-141 C-77B 9.748/9.748 C/L

LSCS-UFSAR D.0-1 REV. 14, APRIL 2002 APPENDIX D PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS Detailed analyses of the "worstcase" sequ ence of the "Nuclear Safety Operational Analysis" (NSOA) for the LaSalle plant is documented in Chapter 15 of the UFSAR.

LSCS-UFSAR E.0-i REV. 13 APPENDIX E-CONSTRUCTION MATERIAL STANDARDS AND QUALITY CONTROL PROCEDURES TABLE OF CONTENTS PAGE

E.1 CONCRETE STANDARDS E.1-1 E.1.1 General E.1-1 E.1-2 Strength Requirements E.1-1 E.1-3 Material Requirements and Quality Control E.1-1 E.1.3.1 Cement E.1-1 E.1.3.2 Aggregates E.1-1 E.1.3.3 Fly Ash E.1-2 E.1.3.4 Admixtures E.1-2 E.1.3.5 Water E.1-2 E.1.3.6 Concrete Testing E.1-2 E.1.4 Concrete Mix Design E.1-3 E.1.5 Batching, Mixing, Delivery, and Placement E.1-4 E.1.6 Witness and Inspections E.1-4 E.1.7 Exceptions Taken to Concrete Codes E.1-5 E.2 REINFORCING STEEL STANDARDS E.2-1 E.2.1 Material Requirements E.2-1 E.2.2 Quality Control E.2-1 E.2.2.1 Testing E.2-1 E.2.2.2 Reinforcing Bar Fabrication E.2-1 E.2.3 Cadwelding E.2-1 E.2.3.1 Qualification of Operators E.2-2 E.2.3.2 Procedure Specifications E.2-2 E.2.3.3 Joint Testing E.2-3 E.2.3.4 Joint Acceptance Standards E.2-3 E.2.3.5 Repairs E.2-3

E.3 POSTTENSIONING TENDONS E.3-1 E.3.1 General E.3-1 E.3.2 Materials E.3-1 E.3.2.1 Tendon Material E.3-1 E.3.2.2 Buttonheads E.3-1 E.3.2.3 Tendon Sheathing E.3-1 E.3.2.4 Grease E.3-1 E.3.2.5 Anchor Heads E.3-2 E.3.2.6 Bearing Plate E.3-2

LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE E.0-ii REV. 13 E.3-3 Quality Control E.3-2 E.3.3.1 Testing E.3-2 E.3.3.1.1 Tendon Tests E.3-2 E.3.3.1.2 Tests on Wires and Buttonheads E.3-2 E.3.3.1.3 Grease Tests E.3-2 E.3.3.1.4 Anchor Hardware Tests E.3-2 E.4 STRUCTURAL STEEL E.4-1

E.4.1 Structural Steel Materials E.4-1 E.4-2 Structural Steel Connections and Connection Material E.4-1 E.4-2.1 Bolted Connections E.4-1 E.4.2.2 Welded Connections E.4-1 E.4.3. Quality Control E.4-2 E.4.3.1 General E.4-2 E.4.3.2 Testing and Inspection of Weldments E.4-2 E.4.3.3 Fabrication E.4-2 E.5 CONTAINMENT LINER, TH E REACTOR SUPPORT PEDESTAL LINER, AND THE DRYWELL FLOOR LINER WITHIN THE CONTAINMENT BACK ED BY CONCRETE E.5-1 E.5.1 General E.5-1 E.5.2 Materials E.5-1 E.5.3 Quality Control E.5-1 E.5.3.1 Testing of Welds E.5-1 E.5.3.1.1 General E.5-1 E.5.3.1.2 Liner Plate Seam Welds E.5-2 E.5.3.1.2.1 Radiographic Examinations E.5-2 E.5.3.1.2.2 Ultrasonic Examinations E.5-2 E.5.3.1.2.3 Magnetic Particle Examination E.5-2 E.5.3.1.2.4 Liquid Penetrant Examination E.5-2 E.5.3.1.2.5 Vacuum Box Soap Bubble Test E.5-3 E.5.3.1.3 Leak Test Channels E.5-3 E.5.3.2 Fabrication and Erection E.5-3 E.5.3.2.1 General E.5-3 E.5.3.2.2 Qualification of Welders E.5-3

E.6 CONTAINMENT STEEL BOUNDARY NOT BACKED BY CONCRETE E.6-1 E.6.1 Materials E.6-1 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE E.0-iii REV. 13 E.6.2 Quality Control E.6-1 E.6.2.1 Testing E.6-1 E.6.2.1.1 General E.6-1 E.6.2.1.2 Testing of Welds E.6-1 E.6.2.2 Fabrication and Erection E.6-2 E.6.2.2.1 General E.6-2 E.6.2.2.2 Qualification of Welders E.6-2 E.7 STAINLESS STEEL POOL LINERS E.7-1 E.7.1 Materials E.7-1 E.7.2 Welding E.7-1 E.8 OTHER ELEMENTS E.8-1 E.8.1 Stainless Steel Requirements E.8-1

E.9 EXCEPTIONS TO ASME B&PV CODES E.9-1 E.9.1 Concrete E.9-1 E.9.1.1 Cement E.9-1 E.9.1.2 Fly Ash E.9-1 E.9.1.3 Aggregates E.9-1 E.9.1.4 Admixtures E.9-1 E.9.1.5 Water and Ice E.9-1 E.9.1.6 Concrete Testing E.9-1 E.9.2 Steel Reinforcement E.9-1 E.9.3 Prestressing System E.9-1 E.9.3.1 Materials for Prestessing E.9-1 E.9.3.2 Fabrication and Installation of Prestressing Systems E.9-2 E.9.3.3 Examination of Prestressing System E.9-2 E.9.4 Containment Liner, Reactor Support Liner and Drywell Floor Liner E.9-2 LSCS-UFSAR E.1-1 REV. 13 APPENDIX E CONSTRUCTION MATERIAL STANDARDS AND QUALITY CONTROL PROCEDURES

E. 1 CONCRETE STANDARDS E.1.1 General All concrete work done is in accordance with ACI 318-71, "Building Code Requirements for Reinforced Concrete," and ACI 301-72, "Specifications for Structural Concrete for Buildings," with the exceptions indicated in Subsection E.1.7. The concrete work also complies wi th all the sections of ACI 349 and ACI 359 which are based on ACI 318-71.

E.1.2 Strength Requirements The specified minimum compressive strengths are:

Prestressed concrete and the drywell floor f 'c

= 4,500 psi at 91 days Seismic Category I f 'c

= 4,000 psi at 91 days structures All others f 'c = 3,500 psi at 91 days E.1.3 Material Requirements and Quality Control E.1.3.1 Cement Cement is Type I portland cement. The sa mpling and qualification tests are done in conformance with ASTM C150-72.

Control tests are made for every 1000 tons or once a month on car or truck samples taken at the batch plant. These test s are ASTM Standard Tests: C109-70T, C266-71, C191-71, and C151-71.

E.1.3.2 Aggregates

Aggregates consist of gravel or crushed st ones that are clean, hard, and durable.

The sampling and qualification tests confor m with ASTM C33-71a. Tests that are done from this specification are the fo llowing ASTM Standard Tests: D75-71, LSCS-UFSAR E.1-2 REV. 13 C136-71, C117-69, C40-66, C 87-69, C88-71a, C142-71, C123-69, C131-69, C535-69, C227-71, Cl28-68, and C29-71.

Control tests in accordance with ASTM C136-71 and C40-66 are done daily. ASTM C29-71 and C117-69 are done weekly on samples taken at the batch plant.

E.1.3.3 Fly Ash Fly ash is used in a portion of the construction. It conforms to ASTM C618-72, Pozzolan Class F, except that in the Table I chemical requirements the maximum percent of sulfur trioxide is 7%. Qualification and control tests are in accordance with the previously mentioned specificat ion. Control tests for approximately 200 tons or about once a month were made on batch plant samples. The control test set consists of the sulfur trioxide, sodium oxide, loss on ignition, and fineness tests of A8TM C618-72.

E.1.3.4 Admixtures Air entraining admixtures used in all construction conform to ASTM C260-69.

Retarding and water-reducing admixtures when used conform to ASTM C494-71.

E.1.3.5 Water An initial qualification test in accordan ce with AASHO T26-70 is made on samples taken from the water source to be used in the concrete mix. A monthly control test in accordance with ASTM D512-67 is taken to determine the chloride ion content in the mix.

E.1.3.6 Concrete Testing During construction, concrete is sampled and tested to ascertain conformance to ACI 318-71 and ACI 301-72. Using the previously specified ingredients the following physical properties are determined for each sample:

a. Temperature of concrete

b. Temperature of air

c. Slump of concrete (ASTM C143-71)

d. Air content (ASTM C231-72T)

The physical properties stated above are determined for each control test performed. Samples are secured in accordance with ASTM C172-71 at the forms when concrete is pumped or at the LSCS-UFSAR E.1-3 REV. 13 mixer truck when a conveyor or bucket is used. The samples are taken per each 50 yd 3 of concrete, or for each pour, or one out of five consecutive trucks.

e. Compressive strength of concrete Six cylinders are filled with fresh concrete and molded into test cylinders. The set of six test cylinders is taken with the following frequencies:

Total Cubic Yards of Concrete in Each Continuous Placement of Concrete Frequency 100 yd 3 or less each day's placement 100 to 500 yd 3 each 100 yd 3 500 to 2000 yd 3 each 200 yd 3 over 2000 yd 3 each 300 yd 3 The cylinders are molded and cured in accordance with ASTM C31-69. Two cylinders are tested at 7 days, two cylinders at 28 days, and two cylinders at 91 days in accordance with ASTM C39-71. Evaluation of compression test results and acceptance of concrete is done in accordance with ACI 214-65 and with Subsection 4.3.3 of ACI 318-71.

f. Tensile splitting strength For every ten samples taken of BA45 concrete, two additional cylinders are molded. These are tested at 91 days for the tensile splitting strength in accordance with ASTM C496-71.

E.1.4 Concrete Mix,Design

The concrete mix is designed in accordance with ACI 613-64 using the previously specified concrete materials. The trial mixes are tested in accordance with ASTM standards listed as follows:

LSCS-UFSAR E.1-4 REV. 13 Test

ASTM Designation

a. Making and curing of the test cylinders in the laboratory C192-69 b. Air content C231-72T or C173-71

c. Slump C143-71

d. Compressive strength C39-71

Six cylinders are cast from each design mix. Compressive strength tests are made at 7, 28, and 91 days,. Two cylinders are used for each test.

Test cylinders are cast from the mix prop ortions selected for the containment and the following concrete properties determined:

a. uniaxial creep,

b. modulus of elasticity and Poisson's ratio, c. plastic shrinkage (both autogenous and paste shrinkage), d. thermal diffusivity,

e. volume changes (expansion and contraction), and

f. coefficient of thermal expansion.

E.1.5 Batching, mixing, Delivery, and Placement Batching, mixing, and delivery equipment, including their operation, conforms to the requirements of' ASTM C94-72, Chapters 7, 8, and 9. The extent to which placement complies with the requirements of Regulatory Guide 1.55 is discussed in Appendix B.

E.1.6 Witness and Inspections A testing agency is retained to make in spections of the manufacturer's, suppliers, and contractor's plants or installations to investigate their testing facilities and procedures.

LSCS-UFSAR E.1-5 REV. 13 The testing agency inspects the batch plant and stationary and truck mixers or agitators prior to production operation to verify conformance with ASTM C94-72, Sections 7, 8, and 9 of ACI 304-73 (Chapt er 4 - Mixing), and the "Check List for Certification at Ready-Mixed Concrete Production Facilities" of the Material Ready-Mixed Concrete Association. After the co ncrete batch plant and mixers are placed in successful production operation, the testing agency inspects the production facilities and secures samples to determine that the concrete is being produced in accordance with Section 7.2 of ACI 301-72.

E.1.7 Exceptions Taken to Concrete Codes ACI 301-72 Specifications for St ructural Concrete for Buildings Chapter 5- Reinforcement Section 5.4 - Fabricating and Placing Tolerances: The following requirements apply in place of the requirements specified for items 5.4.2.1 and 5.4.2.4 of Paragraph 5.4.2, "Bars are placed to the following tolerances":

5.4.2.1 Clear distance to formed surfaces:

For #3 through #11 =

+/- 1/4 inch for straight bars.

+/- 1/2 inch for bent bars. For #14 and #18 =

+/- 1 inch with minimum cover 1-1/4 inches. +/- 1/2 inch with minimum cover 1-1/2 inches.

5.4.2.4 Spacing tolerance betw een bars parallel to the neutral axis for the following bar sizes is as follows, providing the

minimum spacing is not violated:

For #3 through #11, #14, and #18 bars +/- 2 inches.

Chapter 12 - Curing and Protection Section 12.2 - Preservation of Moisture: The following requirements apply in place of all requirements specified in Section 12.2.

Initial Curing:

Initial curing immediately follows the finishing operation. Concrete is kept continuously moist at least overnight in accordance with the following requirements:

a. Unformed Surfaces at Construction Joints:

These surfaces are protected and cured by one of the following methods:

LSCS-UFSAR E.1-6 REV. 13 1. absorptive mat or fabric kept continuously moist, and 2. any one of the sheet materials specified in ASTM C171. This protection and curing is placed over the entire surface immediately after completion of placing concrete and of any required surface treatment.

b. Unformed Surfaces for Rough or Finished Slabs, Floors, or Roofs:

These surfaces are protected and cured with membrane curing compound applied immediately after completion of required surface treatment.

c. Formed Surfaces: All forms are protected, on either the face in contact with concrete or on the opposite face, with a suitable sealer to prevent moisture escape from the forms.

Final Curing: Immediately following the initial curing and before the concrete has dried, additional curing is accomplished in accordance with the following requirements:

a. Unformed-Surfaces at Construction Joints:

Continue the same method used for initial curing until the next placement of concrete is made or until completion of the entire curing periods specified, whichever is the lesser time interval.

b. Unformed Surfaces for Ro ugh or Finished Slabs Floors or Roofs: The requirements for initial curing provide the necessary protection for final curing.

c. Formed Surfaces:

1. If forms are left in place for the entire curing periods herein stated under "Exceptions" for Se ction 14.5 of ACI 301, then no further curing is required.

2. If forms are stripped before completion of the specified curing periods, membrane curing compound is applied to these stripped surfaces immediately after completion of specified surface treatment, or these stripped su rfaces are kept continuously moist with a continuous water spray for the balance of the specified curing periods.

LSCS-UFSAR E.1-7 REV. 13 Duration of curing is stated under "Exceptions" for Section 14.5 of ACI 301.

Chapter 14 - Massive Concrete Section 14.4 - Placing: The following requirements apply in place of Subsections 14.4.1 and 14.4.2:

14.4.1 Slump: Slump is not less than 2 inches, nor more than 4 inches, in place of a 2 inch maximum.

14.4.2 Placing Temperature: Maximum temperature of concrete when deposited is 75° F, in place of 70

° F. 14.4.3 Layers: Restriction to 18 inch heig hts does not apply. Section 14.5 - Curing and Protection: The following requirements apply in place of all requirements of Section 14.5:

14.5.1 Minimum Curing Period: In place of a minimum curing period of 2 weeks, the following curing periods apply:

Curing Periods for Temperatures Above Freezing:

The following curing periods apply when the temperature of the surrounding air is above 40° F and is expected to remain above this temperature, for the entire curing period indicated:

a. Concrete floors and stair treads with monolithic or steel trowel finish 10 days b. Roof slabs of job-placed concrete 7 days c. All other concrete work 5 days Curing Periods During Freezing weather:

a. When the temperature of the surrounding air is 40

° F or below, or is expected to drop to 40

° F or below in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , provide adequate means for maintaining the following temperatures of surrounding air for the curing periods indicated:

1. Concrete floors and stair treads with monalitic or steel trowel finish 10 days at 50

° F or 8 days at 70

° F 2. Roof slabs of job-placed concrete. 7 days at 50

°F or LSCS-UFSAR E.1-8 REV. 13 5 days at 70

° F 3. All other concrete work 5 days at 50

° F or 3 days at 70

°F The design mix follows the standards of ACI 301-66.

ACI 306 Recommended Practice for Cold Weather Concreting:

The following addition applies:

Requirements for winter concreting conform to ACI 306-66, with the added requirement that curing periods are as here in stated under "Exceptions" for Section 14.5 of ACI 301-72.

ACI 605 Recommended Practice for Hot Weather Concreting:

The following exceptions apply:

Temperature of Concrete as Placed: The maximum temperature is 75

° F when deposited, in place of 90

° F maximum.

Curing: Is as herein stated under "Exceptions Taken to Concrete Codes", for Subsection 14.5.1 of ACI 301-72.

LSCS-UFSAR E.2-1 REV. 13 E.2 REINFORCING STEEL STANDARDS E.2.1 Material Requirements Reinforcing bars for- concrete consist of full size deformed bars meeting the requirements of ASTM A615-72.

Grade classifications are as follows:

Sizes Grade #3 to #11 60 #14 to #18 50

Grade 50 reinforcing bars are specially ordered reinforcement modified to obtain a guaranteed minimum yield strength of 50,000 psi and a minimum ultimate strength of 75,000 psi but maintaining a ductility of 12% elongation in an 8-inch gauge length.

E.2.2 Quality Control

E.2.2.1 Testing Tests done on reinforcing bars measure the deformations, tensile and bending properties, and the chemical composition of the reinforcing bars in accordance with ASTM A615-72 and A370-72.

At least one full-diameter specimen from each bar size is tested for each 50 tons or fraction thereof from each heat. However, only one chemical analysis is done per heat. E.2.2.2 Reinforcirq Bar Fabrication Reinforcing bar fabrication conforms wi th the requirements of ACI 318-71, Chapter 7.

E.2.3 Cadwelding Splices in reinforcing bar sizes #11 and sm aller are lapped in accordance with ACI 318-71 or by cadwelding. Bar sizes #14 and #18 are spliced by cadwelding. The splice is designed to develop the specified minimum ultimate strength of the reinforcing bar.

LSCS-UFSAR E.2-2 REV. 13 E.2.3.1 Qualification of Operators Prior to production splicing, each cadweld crew member prepares two joints for each position used in his work. These are tested and meet the joint acceptance standards of this subsection for workmanship, visu al quality, and minimum tensile strength.

E.2.3.2 Procedure Specifications All joints are made in accordance with the manufacturer's instruction sheets "Rebar Instructions for vertical Column Joints," plus the following addi tional requirements:

A manufacturer's representative, experienced in cadweld splicing of reinforcing bars, is present at the jobsite at the outset of the work to demonstrate the equipment and techniques used for making quality splices. He is present for the first 25 production splices to observe and verify that the equipment is being used correctly and that quality splices are being obtained.

The splice sleeves, exothermic powder, and gr aphite molds are stored in a clean dry area with adequate protection from the el ements to prevent absorption of moisture.

Each splice sleeve is visually examined immediately prior to use to ensure the absence of rust and other foreign materi al on the inside diameter surface.

The graphite molds are preheated with an oxyacetylene or propane torch to drive off moisture at the beginning of each shift when the molds are cold or when a new mold is used.

Bar ends to be spliced are powerbrushed to remove all loose mill scale, rust, concrete, and other foreign material. Prior to power-brushing, all water, grease, and paint are removed by heating the bar ends with an oxyacetylene or propane torch. A permanent line is marked 12 inches back from the end of each bar for a reference point to confirm that the bar ends are properly centered in the splice sleeve.

Immediately before the splice sleeve is placed into final position, the previously cleaned bar ends are preheated with an oxyacetylene or propane torch to ensure complete absence of moisture.

Special attention is given to maintaining the alignment of sleeve and guide tube to ensure a proper fill.

When the temperature is below freezing or the relative humidity is above 65%, the splice sleeve is externally preheated with an oxyacetylene or propane torch after all materials and equipment are in position.

LSCS-UFSAR E.2-3 REV. 13 E.2.3.3 Joint Testing To ensure that the strength of each cadweld joint is equal to or greater than the specified minimum tensile strength fo r the particular bar size and ASTM specification, tensile tests are made at the following intervals:

a. one production splice out of the first ten production splices is tested, b. one production and three sister sp lices for the next ninety production splices is tested, and

c. three splices, either production or-sister splices for the next subsequent 100 splices are tested. At least 1/4 of the total number of splices tested are production splices.

E.2.3.4 Joint Acceptance Standards Sound, nonporous filler metal is visible at both ends of the splice sleeve and at the tap hole in the center of the splice sleeve. Filler metal is usually recessed 1/4 inch from the end of the sleeve due to the packing material and is not considered a poor fill. Splices which contain slag or porous metal in the riser tap hole or at the ends of the sleeves (generally porosity) are rejected. A single shrinkage bubble present below the riser is not detrimental and is distinguished from general porosity as described above.

There is evidence of filler material between the sleeve and bar for the full 360

°, however, the splice sleeves need not be exactly concentric or axially aligned with the bars. The tensile strength of each sample tested is equal or exceeds 125% of the minimum yield strength specified in the ASTM standard appropriate for the grade of reinforcing bar using loading rates set forth in ASTM A370-72.

The average tensile strength of each group of 15 consecutive samples is equal to or exceeds the guaranteed ultimate tensile strength specified for the reinforcing bar.

E.2.3.5 Repairs Splices that fail to pass visual inspection are discarded and replaced and are not used as tensile test samples.

LSCS-UFSAR E.2-4 REV. 13 If any production or sister splice tested fails to meet the tensile test specification and the observed rate of splices that fail the tensile test at that time does not exceed 1 for each 15 consecutive test samples, the sampling procedure is started anew.

If any production or sister splice used for testing fails to meet the tensile test specification and the observed rate of splices that fail the tensile test exceeds 1 for each 15 consecutive test samples, mechanical splicing is stopped. In addition, the adjacent production splices on each side of the last failed splice and 4 other splices distributed uniformly throughout the balan ce of the 100 production splices under investigation is tested, and an independent laboratory analysis is made to identify the cause of all failures.

If 2 or more splices from an y of these 6 additional splice samples fail to meet the tensile test specification, the balance of the 100 production splices under investigation is rejected and replaced.

When mechanical splicing is resumed, the sampling procedure is started anew.

If the average tensile strength of the 15 consecutive samples fails to meet the provisions of Paragraph 6.5, the acceptability of the reduced average tensile strength is evaluated with respect to the required strength at the location from which the samples were taken.

LSCS-UFSAR E.3-1 REV. 13 E.3 POSTTENSIONING TENDONS E.3.1 General A BBRV posttensioning system is used. Tendons consist of 90 1/4-inch-diameter parallel lay wires. Positive anchorage at ends is provided by button-heading. The materials, erection and fabrication proc edures, and testing requirements are in general compliance with the technical provisions of Sections CC-2400, CC-4400, and CC-5400 of the ASME B&PV Code,Section III, Division 2, July 1, 1977 edition with exceptions noted in Section E.9.

E.3.2 Materials

E.3. 2. 1 Tendon Materia1 A 1/4-inch-diameter wire conforms to cold-drawn ASTM A421-65, Type BA, stress-relieved, having a guaranteed minimum ultimate tensile strength, (f PU), of 240,000 psi and a minimum yield strength not less than 0.80 f PU as measured by the 1.0%

PU extension under load method.

E.3.2.2 Buttonheads The positive anchorage of tendons to anchor heads is provided by button-heading of the wires.

All buttonheads are cold-formed after threading wires through wire holes of anchor heads. Buttonheads are formed symmetrically about the axis of wires and are free from harmful seams, fractures, and flaws.

E.3.2.3 Tendon Sheathing Tendon sheathing through the foundation consists of black seamless steel pipe, ASTM A106-72a, Grade A and/or ASTM A53.

Wall sheathing is a black interlocked steel strip conduit, 24-gauge minimum wall th ickness, fabricated to be watertight.

The outside diameter of the sheathing is approximately 4 inches. All splices are sealed to prevent intrusion of cement paste. The tendon sheath splice is made using a snugly fitting coupling approximately 1 foot long. The joints between the sheath and the coupling are taped. The minimum radius of curvature used is 20 feet.

E.3.2.4 Grease

A corrosion-preventing grease is used as a tendon casing filler.

LSCS-UFSAR E.3-2 REV. 13 E.3.2.5 Anchor Heads The anchor heads conform to AISI C1141 or HR 1144/11L44.

E.3.2.6 Bearing Plate

The bearing plate material conforms to ASTM A36-70a.

E.3.3 Quality Control E.3.3.1 Testing E.3.3.1.1 Tendon Tests Tensile tests are done on the full 90-wire tendon. Tensile tests are done in accordance with applicable ASTM standards. These tests are done on 1 out of every 100 tendons fabricated.

E.3.3.1.2 Tests on Wires and Buttonheads

Sample tests are done on the wires in accordance with ASTM A421-65. All buttonheads are inspected visually. In addition, 10% are tested with a go and no-go gauge. E.3.3.1.3 Grease Tests Tests are made on each shipment of grease. These tests are ASTM D512-67, ASTM D992-71, and APHA tests on the sulphide content.

E.3.3.1.4 Anchor Hardware Test A minimum of 10% of the anchor heads are subjected to the Rockwell Hardness Test.

LSCS-UFSAR E.4-1 REV. 13 E.4. STRUCTURAL STEEL E.4.1 Structural Steel Materials Structural support steel is ASTM A36-70a, ASTM A572-72, Grade 50, and ASTM 588-71 high strength, low alloy corrosion-resis tant steel. Round tubular bracing is ASTM A53-72a Type E or S Steel.

E.4.2 Structural Steel Connections and Connection Materials E.4.2.1 Bolted Connections Structural steel bolted connections us e ASTM A325-71 Type 1 and ASTM A490-71 friction-type high-strength bolts, except at sliding connections, where high strength bolts in bearing type are used. High-s trength bolted connections conform to "Specification for Structural Joints Using ASTM A325 or A490 Bolts" issued by the Research Council on Riveted and Bolted Structural Joints of the Engineering Foundation and endorsed by the AISC, and to Table I or 11 of AISC Manual-69.

ASTM A36 nuts are used with all ASTM A36 threaded rods and all ASTM A307-68 headed bolts. A194-71, Grade 2 (or bette r) ANSI series nuts are used with all ASTM A193-71, Grade B7 (or better), threaded rods.

Structural steel galleries have been provided at seven elevations in the drywell area of the containment and in various areas of the plant where platforms are needed for access to the valves and other equipment. Sliding connections are provided to these gallery framings, to accommodate movement due to thermal expansion or movement due to containment wall growth under pipe break condition. The gallery framing inside containment consists of radial beams spanning between the sacrificial shield and containment wall and tangential beams spanning between the radial beams. Both radial and tangenti al beams have a sliding connection at one end to allow for expansion and a fixed connection at the other end. Bearing type connections have been used to transfer the load at the sliding end of beams.

Friction type connections have been provid ed at the fixed end. The bolts at the sliding connection are provided with a jam nut in addition to the regular high strength hexagonal nut to ensure against loosening or loss of the high strength nut.

To bring connection angles and connection plates to a snug position, the sliding connection bolts are tightened to a pretension load of 5.0k to 10.0k.

E.4.2.2 Welded Connections Standard welded beam connections confor m to Table III or IV of AISC Manual-69.

Welding procedures for shop and field welding are in accordance with the AWS specifications listed in Table 3.8-2. Selection of electrodes and recommended minimum preheat and interpass temper ature are in accordance with AWS LSCS-UFSAR E.4-2 REV. 13 requirements. All welders and welding operators are certified by an approved testing laboratory and are qualified under AWS procedure set forth in AWS specifications.

E.4.3 Quality Control

E.4.3.1 General Quality assurance requirements apply to the fabrication and testing of structures and components. Certified material test reports are furnished stating the actual results of all chemical analyses and mechanical tests required by ASTM specifications. Identifying heat numbers are furnished on all structural steel to trace the steel to the specific heat in which the steel was made.

E.4.3.2 Testing and Inspection of Weldments The following test methods are used for the inspection of welds:

a. radiographic inspection, b. ultrasonic inspection, c. magnetic particle inspection, and

d. liquid penetrant inspection.

The above nondestructive test methods are in compliance with the following ASTM specifications: E94-68, E142-72, E114-63 , E164-65, E109-63, E 138-63, and E165-65.

Visual weld inspection is in accordance with guidelines prepared by the Nuclear Construction Issues Group, NCIG-01, Rev. 2, titled "Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants".

E.4.3.3 Fabricaticn The fabrication of structural steel conforms to AISC specifications.

LSCS-UFSAR E.5-1 REV. 13 E.5 CONTAINMENT LINER, TH E REACTOR SUPPORT PEDESTAL LINER, AND THE DRYWELL FLOOR LINER WITHIN THE CONTAINMENT BACK ED BY CONCRETE E.5.1 General

The materials, erection and fabrication procedures, and testing requirements are in general compliance with the technical provisions of Sections CC-2500, CC-4500, and CC-5500 of the ASME B&PV Code,Section III, Division 2, July 1, 1977 edition with exceptions as noted in Section E.9.4.

E.5.2 Materials

The containment liner materials performing only a leaktight function (excluding leak test channels), the reactor support pedestal liner and the drywell floor liner within the containment backed by concre te meet the requirements of the 1971 ASME Boiler and Pressure Vessel Code (ASME B&PV Code), Summer 1972 Addenda,Section III, Paragraph NE-2300, and comply with the following specifications:

APPLICATICN SPECIFICATION Drywell liner SA 516-72 GRADE 60 Suppression chamber, reactor support pedestal, and drywell floor liners SA 240-72 TYPE 304

Containment liner anchors A36-70a Reactor support pedestal and drywell floor anchors ASTM A108-69 E.5.3 Quality Control E.5.3.1 Testinq of welds E.5.3.1.1 General All nondestructive examination procedures are in accordance with Appendix X of section III and Section V of the 1971 ASME B&PV Code. Acceptance criteria for

nondestructive examinations are in a ccordance with the 1971 ASME B&PV` Code,Section III, NE-5300.

LSCS-UFSAR E.5-2 REV. 13 E.5.3.1.2 Liner Plate Seam welds E.5.3.1.2.1 Radiograohic Examinations The first 10 feet of weld for each welder and welding position is 100% radiographed. Thereafter one spot radiographing of not less than 12 inches in length is taken for each welder and welding position in each additional 50-foot increment of weld. In any case, a minimum of 2% of linear seam welds is examined by radiography. All radiographic examinations are performed as soon as possible after the weld is placed. The spots selected for radiography are randomly picked. No two spots chosen for radiographic examination are closer than 10 feet apart. If a weld fails to meet the acceptance standards specified in NE-5120,Section III of the ASME B&PV Code, two additional spots are radiographically examined. These two additional spots are at least 1 foot from the spot of initial examination. If either of these two additional welds fails to meet the acceptance standards, then the entire weld test unit is considered unacceptable. Either the entire unacceptable weld is removed and the joint rewelded, or the entire weld unit is completely radiographed and the defective welding repaired. The repaired areas are spot radiographed. Records of all radiographs are stored in accordance with provisions of the 1971 ASME B&PV Code,Section III.

E.5.3.1.2.2 Ultrasonic Examinations Ultrasonic examinations are done on 100% of the liner seam welds where radiographic examinations of liner seam welds are either not feasible or the welds are not accessible for radiographic examination. If a weld fails to meet the acceptance standards specified in NE-5120 of Section III of the-ASME B&PV Code, the weld is repaired and reexamined according to the above code using an ultrasonic examination.

E.5.3.1.2.3 Magnetic Particle Examination The magnetic particle examination is pe rformed on 100% of liner seam welds for ferritic material. If a weld fails to meet the acceptance standards specified in NE-5120 of Section III of the ASME B&PV code, the weld is repaired and reexamined according to the above code using the magnetic particle examination.

E.5.3.1.2.4 Liquid Penetrant Fxamination The liquid penetrant examination is perfor med on 100% of liner seam welds for austenitic material. If a weld fails to meet the acceptance standards specified in NE-5120 of Section III of the ASME B&PV code, the weld is repaired and reexamined according to the above code using the liquid penetrant method of examination.

LSCS-UFSAR E.5-3 REV. 13 E.5.3.1.2.5 Vacuum Box Soap Bubble Test The vacuum box soap bubble test is performed on 100% of liner seam welds for leaktightness. If leakage is detected, th e test is done again after the weld is repaired.

E.5.3.1.3 Leak Test Channels The leak-chase-system channels are installe d over the liner welds. The channel-to-liner plates are tested for leaktightness by pressurizing the channels to the containment design pressure and doing a pneumatic test on 100%

of the welds.

A 2-psi change in pressure over a 2-hour hold period is allowed due to the variation in temperature during the holding period.

E.5.3.2 Fabrication and Erection E.5.3.2.1 General The fabrication and erection of the contai nment steel boundaries backed by concrete are in accordance with the 1971 ASME B&PV Code,Section III, Division I, Subsection NE-4000.

E.5.3.2.2 Qualification of welders The qualifications of welders and welding procedures are in accordance with Sections III and IX of the 1971 ASME B&PV Code.

LSCS-UFSAR E.6-1 REV. 13 E.6 CONTAINMENT STEEL BOUNDARY NOT BACKED BY CONCRETE E. 6.1 Materials The materials comply with the requirements of the 1971 ASME B&PV Code, Summer 1972 Addenda,Section III, Division I, Paragraph NE-2300, and also to the following specifications:

APPLICATION SPECIFICATION Drywell head SA516-72 Grade 70 Personnel access airlock and equipment hatch SA516-72 Grade 70 All pipe penetrations SA312-72 Seamless Grade TP-304 SA376-72 Seamless Grade TP-304 SA403-71 Forged Grade NP-304 E.6.2 Quality Control E.6.2.1 Testing E.6.2.1.1 General The testing of the containment leak boundaries not backed by concrete is in accordance with the 1971 ASME B&PV Code ,Section III Division I, Subsection NE-5000. E.6.2.1.2 Testing of welds One hundred percent of all we lds between penetration and flued fitting and flued fittings and pipelines is examined by radiographic examinations. One hundred percent of all welds in the drywell head, equipment hatch, personnel airlock, and penetration sleeves is inspected also by radi ographic examination. If a weld-fails to meet the acceptance standards specified in NE-5120,Section III of the ASME B&PV Code, two additional spots are radiographically examined. These two additional spots are at least 1 foot from the spot of initial examination. If either of these two additional welds fails to meet the acceptance standards, then the entire weld test unit is considered unacceptable. Either the entire unacceptable weld is removed and the joint rewelded, or the entire weld unit is completely radiographed and the defective welding repaired. The repaired areas are spot radiographed. Records of all radiographs are stored in accordance with provisions of Subsection NE-5000 of the 1971 ASME B&PV Code,Section III, Division I.

LSCS-UFSAR E.6-2 REV. 13 E.6.2.2 Fabrication and Erection E.6.2.2.1 General The fabrication and erection of the containment steel boundaries not backed by concrete are in accordance with the 1971 ASME B&PV Code,Section III, Division I, Subsection NE-4000.

E.6.2.2.2 Qualification of Welders The qualif ications of welders and welding procedures are in accordance with Section III and IX of the 1971 ASME B&PV Code.

LSCS-UFSAR E.7-1 REV. 13 E.7 STAINLESS STEEI, ,POOL LINERS E.7.1 Materials Stainless steel pool liners are fabric ated from SA-240-72, Type 304 material.

E.7.2 Welding Welding and fabrication of the liners is in accordance with the applicable parts of the 1971 ASME B&PV Code, Unfired Pressure Vessels, Subsection B, Part UW. All liner seam welds have complete penetration welds. Acceptance criteria f or nondestructive examinations are in a ccordance with the 1971 ASME B&PV Code,Section III, NE-5210, NE-5220, and NE-5230.

The liner seam welds are examined and tested as follows:

a. radiographic examination on 2% of liner seam welds, b. ultrasonic examination on 100% of liner seam welds wherever the joint detail does not permit radiographic examination to be performed in

accordance with ASME Code, c. liquid penetrant examination on 100% of liner seam welds for the austenitic material, and

d. vacuum box soap bubble test on 100% of liner seam welds.

These welds are tested in the same manner as the welds discussed in Subsection E.5.3.

LSCS-UFSAR E.8-1 REV. 13 E.8 OTHER ELEMENTS E.8.1 Stainless Steel Requirements Stainless steel plate bars and flats, angles , and pipe are hot rolled, annealed, and pickled. The finish is comparable to No. 2B sheet finish in accordance with ASTM A480-71.

For further discussion on austenitic stainless steel, refer to Subsection 5.2.3.4.

LSCS-UFSAR E.9-1 REV. 13 E.9 EXCEPTIONS TO ASME B&PV CODES E.9.1 Concrete E.9.1.1 Cement

For ASTM testing methods C114, C185, C204, C115, C451, and C186, one initial test is made by each test supplier.

E.9.1.2 Fly Ash For ASTM testing methods C618 and C311, one initial test is made by each supplier. Sulfur trioxide, sodium oxide, loss on ignition, and fineness tests are done on every 200 tons.

E.9.1.3 Aggregates ASTM tests C142, C117, C123, C128, C88, C131, C535, C87, and C227 are done once by each supplier. ASTM tests C 117 and C29 are done weekly. ASTM tests C235, C125, C39, C78, C666, C289, C295, C342, C586, and CRDC119 are done on

request. E.9.1.4 Admixtures ASTM tests C260 and C949 are done on request. ASTM test C512 is done monthly.

E.9.1.5 Water and Ice Tests C109, C191, C151, and D1888 are not done. ASHO Designation T-26-70 is done initially by each supplier.

E.9.1.6 Concrete Testing Test C94 is not done.

E.9.2 Steel Reinforcement ASTM test A615 for deformations is done on every 50 tons or one time per heat.

E.9.3. Prestressing System

E.9.3.1 Materials for Prestressing CC2423, CC2424, CC2433.

2.2, CC2442.3.2

LSCS-UFSAR E.9-2 REV. 13 E.9.3.2 Fabrication and Installation of Prestressing Svstems CC4431.1, CC4432.5, CC4451, CC4464.1 E.9.3.3 Examination of Prestressing System

CC5421, CC5423.1, CC5423.2 E.9.4. Containment Liner, Reactor Support Liner and Drywell Floor Liner CC2511, CC2532, CC4534, CC4542.1, CC4543.5, CC4544, CC4552, CC4560, CC5521, CC5524, CC5525, CC5531, CC5535, CC5536, CC5546.1.1, and CC5547.

F.0-i REV. 15, APRIL 2004 APPENDIX F - COMPUTER PROGRAMS TABLE OF CONTENTS PAGE F.0 COMPUTER PROGRAMS FOR STRUCTURAL ANALYSIS AND DESIGN F-1 F.1 CONCRETE F-2 F.2 CSEF-III F-3 F.3 DAPS - II F-6 F.4 DYNAS F-7 F.5 DYNAX F-9 F.6 EASE F-11 F.7 EMBANK F-12 F.8 INDIA F-14 F.9 SETTLE F-15 F.10 KALSHEL F-17 F.11 LAFD F-18 F.12 MESHG F-20 F.13 PLFEM - II F-21 F.14 QUAD4 F-23 F.15 RGS F-25 F.16 SEPOL F-26 F.17 SHAKE F-27 F.18 SLOPE F-29 F.0-ii REV. 15, APRIL 2004 TABLE OF CONTENTS (Cont'd)

F.19 SLSAP F-30 F.20 SOR - III F-32 F.21 SSANA F-33 F.22 STAND F-34 F.23 STRESS F-36 F.24 STRUDL - II F-37 F.25 TEMCO F-38 F-26 WARLOC F-40 F.27 PIPSYS F-42a F.28 PENAN F-42b F.29 FAST F-42c F.30 REFERENCES F-43

F.0-iii REV. 15, APRIL 2004 APPENDIX F - COMPUTER PROGRAMS LIST OF TABLES NUMBER TITLE PAGE F.1-1 Results for 28-Day Strength F-47 F.2-1 Natural Periods for the Eight Lowest Flexural Modes F-48 F.3-1 Structural Frequencies F-49 F.4-1 Probable Maximum Story Displacements F-50 F.5-1 Absolute Maximum Story Shears F-51 F.6-1 Probable Maximum Story Shears F-52 F.7-1 Natural Periods for the Eight Lowest Flexural Modes F-53 F.8-1 Interaction Diagram Axial Load Versus Bending Moment Referred to the Plastic Centroid of the Section F-54 F.9-1 Computer Output Displacements F-59 F.10-1 Computer Output Anchor Forces F-60 F.11-1 Stiffness F-61 F.12-1 Weight Moment of Inertia About X-Axis (Ip) F-62 F.13-1 Rolled Beam Design Problem F-63 F.14-1 Composite Beam Design Problem F-64 F.15-1 Column Design Problem F-65 F.16-1 Plate Girder Design Problem F-66 F.17-1 Section and Material Properties for Problems F-67 F.18-1 Results of Problems F-68

F.0-iv REV. 15, APRIL 2004 APPENDIX F - COMPUTER PROGRAMS LIST OF FIGURES NUMBER TITLE F.1-1 Deflection of a Circular Plate Due to Uniform Pressure and Axisymmetric Edge Moment F.2-1 Radial Moment Due to Unif orm Pressure and Axisymmetric Edge Moment F.3-1 Radial Moment - Simply Supported Circular Plate, Linearly Varying Pressure Load F.4-1 Deflection - Simply Supported Circular Plate, Linearly Varying Pressure Load F.5-1 Response History Analysis of Cantilever Beam for DAPS-II F.6-1 Cantilever Response F.7-1 Three-Story Shear Building F.8-1 Response History Analysis of Cantilever Beam for DYNAS Program F.9-1 Cantilever Response F.10-1 Shallow Spherical Shell F.11-1 Axial Displacement Shallow Spherical Shell F.12-1 Meridional Moment Shallow Spherical Shell F.13-1 Finite Element Idealization of Thick-Walled Cylinder F.14-1 Stresses and Displacements in Thick-Walled Cylinders F.15-1 Cylinder Under Harmonic Loads F.16-1 Meridional Moments and Deflections of Cylinder - N=0 and N=2 F.17-1 Meridional Moments and Deflections of Cylinder - N=5 and N=20 F.18-1 Suddenly Applied Ring Line Load F.19-1 Radial Displacement Vs. Time F.20-1 Bending Moment Vs. Time - Suddenly Applied Ring (Line) Load F.21-1 Spherical Cap F.22-1 Axial Displacement of Spherical Cap Under Dynamic Load F.23-1 Meridional Tension of Spherical Cap Under Dynamic Load F.24-1 Hyperbolic Cooling Tower F.25-1 Spectrum of Design Earthquake F.26-1 Cooling Tower Meridional Force F.27-1 Comparison of Measured Displacements in Otter Brook Dam With Displacements Calculated Using EMBANK F.28-1 Mesh for Sample Problem (EMBANK) F.29-1 Comparison of Results From Original Version of EMBANK and Sargent & Lundy Version F.30-1 Settlement Profile Co mputed by SETTLE and SEPOL F.31-1 KALSHEL Validation Example - Eccentric Line Load on Conical/Cylindrical Shell F.0-v REV. 15, APRIL 2004 LIST OF FIGURES (Cont'd)

NUMBER TITLE F.32-1 Meridional Moment and Meridional Force Along the 0

° Plane F.33-1 Radial Shear Along the 0

° Plane F.34-1 Hoop Moment and Hoop Force F.35-1 Meridional Force, Meridional Moment, and Radial Shear Force F.36-1 Rectangular Tank Filled With Water F.37-1 Moment of M y at Horizontal Cross Section of Walls F.38-1 Moment M y at Top of Wall F.39-1 Moment M x Along Cross Section of Long Wall F.40-1 Plate With Circular Hole Under Uniform Tension F.41-1 Stresses in Plate With Ci rcular Hole Under Uniform Tension F.42-1 Square Plate With Rectangular Hole Subjected to Temperature Variation F.43-1 Moments in Plate Due to Temperature Variation F.44-1 Soil Profile and Finite Element Representation Used for Sample Problem F.45-1 Strain-Compatible Damping and Modulus Values Used in Analysis F.46-1 Distribution of Maximum Shear Stresses and Accelerations F.47-1 Acceleration Spectra for Computed Surface Motions F.48-1 Validation for a One-Degree-of-Freedom Damped System (RSG) F.49-1 Soil Profile and Layered Representation Used for Sample Problem F.50-1 Comparison of Shear Stresses and Accelerations F.51-1 Comparison of Spectral Values for Surface Motions F.52-1 Bending Moments in a Cantilever Beam F.53-1 Bending Moments in a Simply Supported Plate F.54-1 Circular Plate for SOR-III Example F.55-1 Moment Comparison of SABOR-III and SOR-III F.56-1 Radial Shear Comparison for SABOR-III and SOR-III F.57-1 Building Complex F.58-1 Shear and Moment Diagrams

LSCS-UFSAR F.1 REV. 15, APRIL 2004 APPENDIX F COMPUTER PROGRAMS The computer programs referred to in Sectio ns 2.5, 3.7, and 3.8 by their acronyms are described herein. All programs are verified, within the stated assumptions and limitations, for correctness of utilized theory and validity of obtained results for a variety of typical problems. Results are checked against known solutions, solutions obtained from other programs, or hand calculations. Examples of validation problems are included with the program descriptions. Whenever applicable, internal checks such as equilibrium and orthogonality checks are included as an aid in checking the validity of the results obtained from the computer program for each problem analyzed. Additionally this appendix describes some computer programs that are not referenced in the UFSAR, but th at have historical significance. This is not an all-inclusive list of computer programs, but highlights those of historical significance.

LSCS-UFSAR F.2 REV. 15, APRIL 2004 F.1 CONCRETE CONCRETE is a computer program used for statistical evaluation of concrete strength. It sorts and analyzes the field data collected on concrete samples and presents it in a convenient form for interpretation.

CONCRETE was developed by Sargent & Lundy.

The compressive strength test results of concrete cylinders are statistically analyzed to obtain the mean, standard deviation, co efficient of variation, moving averages, and other statistical parameters required in the quality appraisal of concrete according to ACI 214-65. The strength resu lts are also compared with the quality control limits fixed according to the ACI 318-71 Code and the ASTM Manual on Quality Control of Materials, 15-C. Any violations or inadequacies are clearly pointed out in the output.

To demonstrate the validity of the program, a sample problem from "Notes on ACI 381-71 Building Code Requirements with De sign Applications" (Reference 1) was processed on CONCRETE. The problem de termines the average 28-day strength and standard deviation for 46 test cylinders. CONCRETE's results, shown in Table F.1-1, are identical to the hand calculation solutions in the ACI notes.

LSCS-UFSAR F.3 REV. 15, APRIL 2004 F.2 CSEF-III CSEF-III (Circular Slab on an Elastic Foundation), analyzes any circular slab for arbitrary load conditions. The analysis is based on the solution to the basic differential equation for a plate on an elastic foundation, modified to include a Fourier series representation for the circumferential variation. For each Fourier harmonic, a matrix of linear equations is formulated using a finite difference representation of the plate equation an d boundary equations. The system of equations is then solved by means of a Gauss elimination method.

Arbitrary normal pressure loading, ed ge moment and/or shear loads, and axisymmetric thermal gradient loads may be considered in the analysis. At any radius, the entire circumference may be fixed against rotation or displacement, allowing for a variety of support conditions or boundary conditions.

The program output includes deflection, soil pressure, radial moment, hoop moment, twisting moment, radial shear, tangential shear, and Kirchhoff shear for each Fourier harmonic. Results from individual harmonics are superimposed and output along the radii at specified an gles. Moments may be converted to a Cartesian coordinate system for comparison with other programs.

Version III of CSEF was developed by Sargent & Lundy in 1971 Two plate analyses are presented as examples of validation.

The first example is a concrete circular plate resting on an elastic foundation. The plate parameters are as follows:

Radius a = 50 ft, Thickness h = 5 ft, Elastic Modulus E = 576,000 ksf, Poisson's Ratio

µ = 0.17, and Foundation Modulus K = 518.4 kip/ft

3. The plate is analyzed for a uniform pressure load:

P = 50 kip/ft 2 And a uniform edge load:

M = 1000 kip/ft/ft.

LSCS-UFSAR F.4 REV. 15, APRIL 2004 The results of the CSEF-III analysis are compared with a hand-calculated solution.

The hand calculations are based on equa tions presented in Reference 2. The displacements for the distributed load and edge moments are found independently and superimposed. For the uniform pressure load the displacement is:

w = p/k and for the edge moment:

w = D 1 Z 1 (r) + D 2 Z 2 (r) where: = D k/4 Z 1 (x) = R e J o (x i) Z 2 (x) = I m J o (x i) J o = Bessel function of the first kind and D 1 = K M 2 a)(a)(a-1 a)(a)()()()(µ2 2 2 1 2 1 2 1 1 Z Z Z Z a Z a Z a Z++ D 2 = K M 2 a)(a)(a-1 a)(a)()()()(µ2 2 2 1 2 1 2 1 2 Z Z Z Z a Z a Z a Z++ .)(dx x dZ Z and Z=2 1 The results obtained by evaluating th ese equations for displacement and the appropriate expressions for radial moment are presented along with those obtained from CSEF-III in Figures F.1-1 and F.2-1, respectively. As illustrated in these figures, the independent solutions compare favorably.

LSCS-UFSAR F.5 REV. 15, APRIL 2004 For a second example, a simply supported circular plate under a linearly varying pressure load is presented. The solution is obtained for a plate having the following properties and loading:

Radius a = 100 in., Thickness h = 2 in., Elastic Modulus E = 3.0 x 10 7 psi, Poisson's Ratio = 0.3, and Loading Intensity q =

a pr , cos where: p = 100 psi.

Results from CSEF-III are compared with those obtained for Kalnin's Shell of Revolution (Reference 3). The comparison for radial moment and displacement are shown in Figures F.3-1 and F.4-1. Solution s obtained from both programs favorably compare with a solution presented by Timoshenko and Woinowski-Krieger in Reference 4.

F.3 DAPS-II DAPS (Dynamic Analysis of Plane Strain or Plane Stress Systems) analyzes the dynamic response of two-dimensional plane stress or plane strain finite element systems to an earthquake. The equati ons of motion are solved by a stable step-by-step integration procedure or modal analysis. Quadrilateral, triangular, truss, beam, shear spring, and rocking spring elements are included. For step-by-step integration, the program output includes relative nodal displacements, absolute accelerations and stresses at selected time intervals, maximum relative displacements, absolute accelerations and stresses, response spectra frequencies, and modal vectors.

For modal analysis, the output includes maximum displacements, velocities, accelerations, stresses, response spectra, modal frequencies and vectors, participation factors, and an orthogonality check.

DAPS was originally developed by E. L.

Wilson of the University of California, Berkeley (Reference 5). It wa s acquired by Sargent & Lundy.

LSCS-UFSAR F.6 REV. 15, APRIL 2004 One of the problems used to validate the program is presented. The ground acceleration is applied at the fixed end of a cantilever beam (Figure F.5-1). Table F.2-1 and Figure F.6-1 compare the results obtained by Wilson et al using the SAP-IV program (Reference 6) with the results obtained from DAPS-II. As illustrated in these figures, the results show good agreement.

F.4 DYNAS DYNAS (Dynamic Analysis of Structures) is designed to perform dynamic analysis of structures which can be idealized as three-dimensional space frames and/or rigid slabs connected together by translational or torsional springs. The program considers the combined effects of translational, torsional, and rocking motions on the structure. The program uses either th e response spectrum or the time-history method of analysis, depending on the type of forcing function available. Each method uses a normal mode approach. In the case of time-history analysis, the decoupled differential equations of motion are numerically integrated using Newmark's -method (Reference 7).

The DYNAS program is capable of analyzing structures with parts having different associated dampings. An option is also available to analyze a large structural system using a modal synthesis technique. In this option, the system is divided into subsystems whose modal characteristics are computed separately and then synthesized to obtain the response of the complete system. The input base motion can be applied simultaneously in two orthogonal directions. A response spectrum can be generated at specified slabs or joints.

The program output includes modal responses, probable maximum responses, a time-history of structural response, and a response spectrum at specified joints.

The DYNAS program was originally developed by Sargent & Lundy in 1970.

The solutions of two of the problems us ed for validating DYNAS are presented.

In the first problem, a three-story shear building is analyzed and compared to a solution obtained by Biggs (Reference 8). The structure in conjunction with the applicable masses and stiffness values is re presented by the closed-coupled system shown in Figure F.7-1. For this analysis the following response spectrum was used:

Frequency Displacement 1.00 cps 3.30 in. 2.18 cps 1.40 in. 3.18 cps 0.66 in.

LSCS-UFSAR F.7 REV. 15, APRIL 2004 The results obtained by Biggs and from DYNAS are compared in Tables F.3-1 through F.6-1.

In the second example, results of DYNAS ar e compared to those obtained by Wilson et al. (Reference 5) using the SAP-IV program.

An acceleration is applied at the fixed end of a cantilever beam (Figure F.8-1). The natural periods calculated by both SAP-IV and DYNAS are shown in Table F.7-1. A comparison of the bending moment at the fixed end of the cantilever beam is shown in Figure F.9-1. As demonstrated in both examples, DYNAS performs an accurate analysis.

F.5 DYNAX DYNAX (Dynamic Analysis of Axisymmetric Structures) is a finite element program capable of performing both static and dyna mic analyses of axisymmetric structures. Its formulation is based on the theory of small displacement.

Three types of finite elements are available: quadrilateral, triangular, and shell. The geometry of the structure can be general as long as it is axisymmetric. Both isotropic and orthotropic elastic material pr operties can be modeled. Discrete and distributed springs can be used to model elastic foundations, etc.

For static analysis, input loads can be structural weight, nodal forces, nodal displacements, distributed loads, or thermal loads. Loads can be axisymmetric or nonaxisymmetric. For solids of revolution, the program outputs nodal displacements and element stresses in the global system (radial, circumferential, and axial) and element and nodal stress resultants in a shell coordinate system. For shells of revolution, the output consists of nodal displacements as well as element and nodal stress resultants in a shell coordinate system (meridional, circumferential, and normal).

For dynamic analysis, three methods are available: direct integration method, modal superposition method, and response spectrum method. Dynamic analysis by direct integration or modal superposition method uses a forcing function input via either 1) nodal force components versus time for any number of nodes, or 2) vertical or horizontal ground acceleration versus time. For nonaxisymmetric loads, the equivalent Fourier expansion is used. Dy namic analysis by the response spectrum method uses a spectral velocity versus natural frequency input with up to four damping constants. The output of dynamic analysis provides nodal displacements, element stresses, and resultant forces and moments at specified time steps. When the modal superposition method is used for earthquake response analysis, the prescribed number of frequencies and mode shapes are computed and printed along with the cumulative response of all specified modes by the root sum square (RSS) method and the absolute sum method.

LSCS-UFSAR F.8 REV. 15, APRIL 2004 DYNAX was originally developed under the acronym ASHAD by S. Ghosh and E. L.

Wilson of the University of California, Berkeley, in 1969 (Reference 9). It was acquired by Sargent & Lundy.

Validation of the major analytical capabilities of DYNAX is demonstrated by a comparison of the results from six documented problems with DYNAX results.

The first problem is taken from S. Timoshenko's book Theory of Plates and Shells (Reference 10). A clamped shallow spherical shell, shown in Figure F.10-1, is analyzed for displacements and stresses produced by a uniform pressure applied on its outside surface. DYNAX and Timoshenko's solutions are compared in Figures F.11-1 and F.12-1.

The second problem, taken from Theory of Elasticity by Timoshenko and Goodier (Reference 11), is a plane st rain analysis of a thick-walled cylinder subjected to external pressure. The finite element idealization and the loading system used for this case are shown in Figure F.13-1. Re sults of the DYNAX analysis are compared with the exact solution in Figure F.14-1. The agreement for both stresses and displacements is excellent.

The third problem is taken from an article by Budiansky and Ra dkowski (Reference 12). The structure, illustrated in Figure F.15-1, is a short, wide cylinder with a moderate thickness to radius ratio. The applied loads and the output stresses are pure uncoupled harmonics. For this finite element analysis, the cylinder is divided into 50 elements of equal size. This problem solves for harmonic deflections, element stresses, and forces. Figures F.16-1 and F.17-1 compare DYNAX results with the results given in the article.

The fourth problem is taken from an article by Reismann and Padlog (Reference 13). A ring (line) load of magnitude P (500 pounds) is suddenly applied to the center of a freely supported cylindrical shell. The dimensions of the shell and the time-history of the load are shown in Figure F.18-1. Because of symmetry, only one-half of the cylinder is modeled using 80 elements of equal size. The time-history of radial deflection and meridional moments from DYNAX and from Reismann and Padlog are compared and ar e shown in Figures F.19-1 and F.20-1, respectively.

For the fifth problem, the method of mode superposition is used to solve a shallow spherical cap with clamped support under the action of a suddenly applied uniformly distributed load. The dimensions of the shell and the load time-history are shown in Figure F.21-1. The first 12 modes were considered to formulate the uncoupled equations of motion. Each of these equations was solved by the step-by-step integration method using a time step of 0.1 x 10

-4 seconds. The results LSCS-UFSAR F.9 REV. 15, APRIL 2004 are compared graphically with those obtained by S. Klein (Reference 14) in Figures F.22-1 and F.23-1.

The sixth problem is a hyperbolic cooling tower, as shown in Figure F.24-1. The tower is analyzed for horizontal earthquake motion. A response spectrum for 2%

damping, as shown in Figure F.25-1, was used for this analysis. The RMS values of the meridional force are compared with those obtained by Abel et al. (Reference 15) in Figure F.26-1.

As shown in these six examples, DYNAX is capable of producing accurate results for both static and dynamic analyses of shells.

F.6 EASE EASE (Elastic Analysis for Structural Engin eering) performs static analysis of two- and three-dimensional trusses and frames, plane elastic bodies, and plate and shell structures. The finite element approach is used with standa rd linear or beam elements a plane stress triangular element, or a triangular plate bending element. The EASE program accepts thermal loads as well as pressure, gravity, or concentrated loads.

The program output includes joint displacements, beam forces, and triangular element stresses and moments.

EASE was developed by the Engineering An alysis Corporation, Redondo Beach, California, in 1969 and is in the public do main. It was used by Sargent & Lundy.

F.7 EMBANK EMBANK (Embankment Analysis) is a fi nite element program used for the evaluation of embankment stresses and deformations. Because construction sequence plays an important part in the deformations developed in earth embankments, an incremental version of the finite element method is employed to account for this effect. In addition, this method has the capability of employing nonlinear, stress-dependent soil stress-strain behavior. The soil constitutive relations are expressed in terms of bulk modulus and deformation modulus, and it is assumed that only the deformation modulus varies with strain.

To take account of incremental construction, the finite element idealization is arranged in horizontal layers corresponding to construction lifts. The analysis then involves the evaluation of stresses and deflections in a succession of structures, corresponding to various stages of proposed construction. In the first analysis, for example, only the lowest layers of elements are considered. The program evaluates the stiffness of these elements and the dead weight forces which they produce and LSCS-UFSAR F.10 REV. 15, APRIL 2004 then calculates the stresses and displacements developed in the lowest layer due to its own weight.

After sorting the stresses and displacements resulting from the first lift analysis, the program next considers the structure co nsisting of the lowe st two lifts. The stiffness of all elements in these two lifts is evaluated, but only the dead weight forces of the upper lift are considered. The program calculates stresses and displacements developed throughout the structure as a result of this new increment of load and adds these stress and displacement results to those obtained for the preceding increment. By repeating this procedure for each successive construction lift, the complete history of the development of stresses and displacements can be obtained.

The program output includes the displace ments of each node and the resultant direct stresses, shear stresses, and principal stresses with their associated directions for each element.

EMBANK is the Sargent & Lundy version of the LSBUILM program. It was originally developed by F. H. Kulhawy, J. M. Duncan, and H.

B. Seed at the University of California, Berkeley, in 1969.

The validity of the original version was checked by Kulhawy, Duncan, and Seed (Reference 17) by comparing the measured movements of the Otter Brook Dam.

Otter Brook Dam, a rolled earth dam cons tructed in New Hampshire, is about 130 feet high and has 2.5:1 slopes upstre am and downstream. The calculated displacements from the program were compared to the measured values and are shown in Figure F.27-1. The variations of the horizontal displacement with height above the base of the dam are shown on th e left, and the measured and calculated variations of the displacement of a bridge pier located midway on the upstream slope are shown on the right. It can be seen that in both cases the agreement is satisfactory, indicating that the analysis used in the program provides a rational basis for estimating embankment deformations.

To validate Sargent & Lundy's UNIVAC 1106 version of the program, example problems provided by the original authors were rerun. One of these examples is a 30-foot-high embankment (Figure F.28-1) built in three layers. The results for the final shear stresses at the center of various elements of a layer and the x-displacements at a layer junction are shown in Figure F.29-1.

The results obtained from the Sargent & Lundy version were identical with those from the original version of the program, thus indicating a successful implementation on the UNIVAC 1106.

LSCS-UFSAR F.11 REV. 15, APRIL 2004 F.8 INDIA INDIA (Load-Moment Interaction Diagram) is a program used to compute the coordinates and plot the bending moment-axial load interaction diagram for a rectangular reinforced concrete section. The program will plot interaction curves for ultimate strength, yield strength, and working stress methods. Both compression and tension axial loads are consid ered as well as positive and negative moments for appropriate cross sections.

The procedures used for the working stress and yield stress methods are taken from ACI 318 Code. Equations used for the ultimate stress method are taken from a University of Illinois Civil En gineering Study (Reference 18).

INDIA was originally developed at Sargent & Lundy on the IBM 1130 in 1971.

To demonstrate the validity of the program, a sample problem, shown in Table F.8-1, was executed. Calculations were ma de by hand, and all results were found to be consistent with the theoretical approach.

F.9 SETTLE SETTLE is a computer program developed to predict the magnitude of settlements of shallow foundation caused by the fo undation load. Janbu's tangent modulus method (Reference 19) is employed to account for the nonlinear stress-strain behavior of soil.

The distribution of contact pressure co nsidering the effect of the foundation structure is taken into account by assu ming that the foundation is rigid and, therefore, the settlement distribution is approximately linear. The variations in contact pressure can then be determined from the conditions of force equilibrium and compatibility of the foundation and soil settlements.

The foundation settlement calculation is based on the following fundamental assumptions:

a. The soil profile can be divided into homogeneous horizontal layers with uniform thickness. b. The stress increment caused by th e applied loads can be approximated by the Boussinesq formula. c. At any point, the stress increment contributed by each of the loading areas can be superimposed to calc ulate the total stress increment at that point.

LSCS-UFSAR F.12 REV. 15, APRIL 2004 Settlements at a point are computed by summing the individual settlement of each soil sublayer of a predetermined thickness. The following calculations are performed for each settlement point:

a. The stress increment caused by ea ch loading area is computed, and the total influence at the center of ea ch soil sublayer caused by all the loading areas is accumulated. b. After the stress increments have been accumulated, the settlement they produced is computed and accumulated. Settlements are computed by Janbu's tangent modulus concept. c. The settlement beneath a point is considered as the total of the individual settlement of each soil sublayer. d. After the settlement at one point has been obtained, the SETTLE computer program proceeds to calcul ate settlement for the next point.

An iterative procedure is used so as to make the settlement patterns of the foundation and the subsoil compatible. If settlement patterns are not compatible, the distribution of contact pressure underneath the foundation is recalculated to satisfy the deformation prescribed by the subsoil. The new contact pressure distribution is then used to compute the new settlement pattern of the subsoil. The iteration continues until the predetermined convergence criteria are satisfied.

SETTLE was developed by Sargent &

Lundy following the SETTLE tangent modulus concept. Results of settlement computation have been validated with those of the ICES-SEPOL program (Reference 20) combined with hand calculations.

A typical example consists of a soil layer which is 31 feet thick and has two different soil types. This layer was loaded by three rectangular loading areas with varying intensities. Figure F.30-1 shows the cl ose agreement of the settlement profiles computed by SETTLE and ICES-SEPOL. A 20- x 30-foot rectangular mat foundation loaded by four different loading areas was used to validate the stress distribution calculation. The mat foundation was subdivided into six elements, and a spring was assumed under the center of each element. The results of the SETTLE program and the hand calculation are exactly the same. These satisfactory results indicate a successful development of this program.

F.10 KALSHEL KALSHEL (Kalnins' Shell of Revolution) is a computer program used to analyze thin axisymmetric shells of revolution for arbitrary load conditions. The solution is obtained by transforming the H. Reissner-Neisser equations to eight first-order ordinary differential equations (Reference 21). An Adams method of numerical integration is used as the basis for the solu tion of transformed equations. Since the program is based on classical shell theory, it has the same limitations.

LSCS-UFSAR F.13 REV. 15, APRIL 2004 The shell wall may vary in thickness along th e meridian and consists of up to four layers of different isotropic or orthotropic materials. Branch shells may be connected to the main shell. Surface load s and line loads in the radial, tangential, and/or meridional directions and meridional moments as well as temperature distributions, which are assumed to vary linearly across the thickness, may be considered in the analysis.

All loads may be asymmetric.

The program output includes shell displa cements in the radial, tangential, and meridional directions, meridional rotations, meridional moment, hoop moment, meridional force, hoop force, transverse shear force, and twist shear force. In addition, outer fiber stresses calculated from the stress resultants may be obtained.

The program was originally developed by A. Kalnins of Lehigh University (Reference 3). It was acquired by Sa rgent & Lundy in 1969. This version was modified by Sargent & Lundy to sum disp lacements and stress resultants of the individual Fourier harmonics along meridians at specified angles.

A number of test cases were run to ch eck the program options and validity of solution. One practical example included th e analysis of a conical shell subjected to eccentric line load. The shell is made of two parts, cylindrical and conical, and both are of reinforced concrete with different thicknesses, as shown in Figure F.31-1.

This problem was analyzed by this program and also by the public domain program SABOR-III (Reference 22).

Results from the two programs are compar ed in Figures F.32-1 through F.35-1.

Figures F.32-1 and F.33-1 show a comparison of shell forces along a meridian at 0

° (symmetric with respect to the load).

Figures F.34-1 and F.35-1 show a comparison of shell forces around the circumference at an elevation where the load is applied. As shown in these figures, the results indicate comparable values.

F.11 LAFD LAFD (Analysis of Linear Anchor Forces and Displacements) calculates the maximum force and displacement of anchors resulting from local buckling of thin plate liners anchored to concrete walls. The solution method used in LAFD is described in Reference 23.

First, anchor displacements are found fo r an assumed postbuckling load by a relaxation technique. Then, using this maximum displacement, the anchor force and the strain in the buckled plate are calculated. The stress-strain relation given in a paper by Young and Tate (Reference 24) is reestablished in the program. Using the calculated strain, first stress is found and then a new load. The new load is then used to find a new set of displacements. The procedure is repeated to find a LSCS-UFSAR F.14 REV. 15, APRIL 2004 second new load. This load is then compared to the load used in the previous cycle.

The procedure is repeated until the loads obtained in the last two cycles are approximately zero.

The program is capable of analyzing four type s of anchors: Nelson studs of 1/2-, 5/8-and 3/4-inch diameter, and 3- x 3-1/4-inch angle continuous rib anchors. The force-deformation relations of these anchors are obtained from the manufacturer's publication (Reference 25).

The program output includes the maximum anchor force, the maximum anchor deformation, and the postbuckling load of the buckled plate.

LAFD was developed by Sargent & Lundy in 1971.

To validate the program, significant calculations were verified with hand calculations. As an example of this validation, a comparison of these calculations is presented for a strip of liner ha ving the following properties:

Strip span a = 17.5 in., Plate thickness t = 0.375 in., Strip width w = 9 in., Modulus of Elasticity E = 30 x 10 3 ksi, and Yield Stress o = 36 ksi.

5/8-inch-diameter Nelson studs are used as anchors.

The anchor displacements, U n, the force in the anchor adjacent to the buckled panel, f 1 , and the postbuckling load P as calculated by the program are shown in Table F-9. Substituting these displacements into the appropriate force-deformation relationship for a 5/8-inch-diameter Nelson stud yields the anchor forces contained in Table F-10.

LSCS-UFSAR F.15 REV. 15, APRIL 2004 The validity of the solution is checked using the displacements and anchor forces given in Tables F.9-1 and F.10-1 to verify the equality of the original equations:

Fo-P =

a EA (U 1 -U 2) + f 1 (1) O = a EA (2U n - U n-1 - U n+1) + f n (2) N = 1, 2, 3 -. N The postbuckling load, P, as determined by Equation 1, is equal to 21.864K as compared to 21.978K obtained from the pr ogram. Substitution into Equation 2 satisfies the equation; equilibrium having been verified, the results obtained from the program are valid.

F.12 MESHG MESHG (Mesh Generator) checks input data for finite element programs. Using the member incidences and node point coor dinates as prepared for a finite element program such as SLSAP, PLFEM, DYNAX, and QUAD4, the program produces a Calcomp plot of the mesh.

Several isometric views of 3-D data may be obtained, axes may be rotated for 2-D data, and scaling may be specified. Element numbers are plotted proportional in size to element areas for ease in detecting errors in element connectivity or nodal coordinates.

The program was developed by Sargent &

Lundy in 1970. Validity of the program is verified repeatedly by inspection of each plotted mesh.

F.13 PLFEM-II PLFEM (Plate Finite Element Method) analyzes plane elastic bodies, plates, and shell structures by the stiffness matrix method. The program uses two finite elements, a rectangular element and a triangular element.

Elastic spring supports and/or an elasti c foundation may be considered in the analysis. Orthotropic materials may also be considered in conjunction with the rectangular element. Pressure loads, concentrated forces, nodal displacements, and thermal loads may be considered in the analysis. All loading cases may be factored and/or combined in any manner.

LSCS-UFSAR F.16 REV. 15, APRIL 2004 The program output includes deflections and rotations of all joints and membrane stresses (normal, shearing, and principal) at the center of each element, the resultant moments (X, Y, twisting principa l), and shears and reaction forces. An equilibrium check is made to determine the accuracy of the results.

PLFEM was developed by Sargent & Lundy.

Three sample problems are presented to demonstrate the validity of PLFEM. Plots of the computer results obtained are compared with theoretical results by other methods. The first problem is an analysis of a rectangular tank filled with water which was presented by Y. K. Cheung and J. D. Davies (Reference 26). The finite element used was presented by Zienkiewicz and Cheung in August 1964 (Reference 27).

Experimental results agreed exactly with the finite element results except at a few isolated points where very small differences were noted. The PLFEM grid and loading for the tank problem are shown in Figure F.36-1. The grid used is the same size as the one used by Cheung and Davies. Moments in three regions of the tank are plotted along with the PLFEM results in Figures F.38-1 and F.39-1.

As a second example, a rectangular plate subjected to a uniform plane stress and having a circular hole in its center is analyzed. The grid used in the PLFEM analysis is shown in Figure F.40-1. Because of double symmetry, only one quarter of the plate is analyzed. Results obtained from the PLFEM analysis are plotted in Figure F.41-1 against the exact values as gi ven by S. Timoshenko and J. Goodier in Reference 11.

As a final example, a square plate having a rectangular hole in its center is analyzed for the effect of a thermal gradient through the plate. The grid used in the PLFEM analysis is shown in Figure F.42-1. Only one quarter of the plate is analyzed because of the double symmetry.

Moment values obtained by PLFEM are plotted for two regions of the plate in Figure F.43-1. For comparison, values of the moments obtained by an analysis based on the Hrennekoff framework analogy are also shown.

F.14 QUAD4 QUAD4 is a finite element program which evaluates the seismic response of soil structures using a different damping ratio for each individual element. The base motion can be applied simultaneously in tw o orthogonal directions. In addition, the procedure allows incorporation of strain-d ependent stiffness and damping values for each element.

LSCS-UFSAR F.17 REV. 15, APRIL 2004 The program has been written for elements in plane-strain; triangular and quadrilateral elements can be used in re presenting the continuum. The solution proceeds by assigning modulus and damping va lues to each element. Because these values are strain-dependent, an iteration procedure is adopted. Thus values of shear moduli and damping are estimated at the outset, and the analysis is performed. Using the computed values of average strain developed in each element, new values of modulus and damping are dete rmined from appropriate data relating these values to strain. Proceeding in this way, a solution is obtained incorporating modulus and damping values for each element which are compatible with the average strain developed.

The program output includes the strain compatible soil properties, response spectrum at specified nodes, probable ma ximum nodal accelerations, and probable maximum direct stresses, shear stresses, and shear strains for all elements. The shear stress time-history can also be obtained for specific elements.

QUAD4 was originally developed by I. M. Idriss, J. Lysmer, R. Hwang, and H. B.

Seed of the University of California, Berkeley (Reference 28). It was acquired by Sargent & Lundy.

To validate Sargent & Lundy's version of QUAD4, a sample problem was taken from the original program documentation (Reference 28).

A 100-foot layer of dense sand as shown in Figure F.44-1 has been analyzed.

The properties of the sand were considered to be as follows:

Total unit weight = 125 pcf, (K 2) max = 65, and K o = 0.5. The parameter (K

2) max relates the maximum shear modulus, G max , and effective mean pressure at any depth, y, below the surface as follows:

G max = 1000 (K

2) max m 1/2 Where: 'm = 3 2 1)(o K+ ' and Ko = coefficient of lateral pressure at rest and ' = effective vertical pressure at depth y.

LSCS-UFSAR F.18 REV. 15, APRIL 2004 Damping values and variations of modulus values with strain were based on data given in Reference 28.

The response of the sand layer was evaluated using the time-history of accelerations recorded at Taft during the 1952 Kern Count y earthquake as base excitation. The ordinates of this time-history were adjusted to provide a maximum acceleration of 0.15g. The sand layer has been represented by the finite element mesh shown in Figure F-44, which consists of 20 elements and 42 nodal points. To simulate a semi-infinite system, nodal points 1 through 40 have been fixed in the vertical direction and are therefore permitted to move only in the horizontal direction. Nodal points 41 and 42 are fixed to the base. A comparison of the results obtained from this validation run and the published results is shown in Figures F.45-1 an d F.47-1. The values of damping and modulus compatible with the strain level computed in each element are presented in Figure F.45-1. The variations of maximum shear stresses and the maximum accelerations with depth are shown in Figure F.46-1. The acceleration spectrum for the computed surface moti ons is shown in Figure F.47-1. As illustrated by these figures, th e comparisons are favorable.

F.15 RSG RSG (Response Spectrum Generator) generates dynamic response spectra (displacement, velocity, and acceleration) for single-degree-of-freedom elastic systems with various dampings, subjected to a prescribed time-dependent acceleration. The differential equation of motion is solved using Newmark's -method of numerical integration (Reference 7).

The program may also be used to obtain a response-spectrum-consistent time-history in which the response spectrum of the generated time-history closely envelopes the given spectrum.

The program has the capability of plotting the input time, acceleration function, and the response spectra output on the tripartite and/or acceleration versus period frequency grids.

Depending on the option, the program outp ut includes the response spectra of a given time-history or the response-spectrum-consistent time-history.

RSG was developed by Sargent & Lundy in 1969.

One of the comparisons used for validation is presented.

LSCS-UFSAR F.19 REV. 15, APRIL 2004 The response spectrum for a one-degree-of-freedom damped system as presented by Biggs (Reference 8) was determined using RSG. The system was subjected to the sinusoidal ground acceleration shown in Figu re F.48-1. A damping factor of 0.2 was used for this example. The response sp ectra obtained by Biggs and from RSG are also shown in Figure F.48-1. As demons trated by this comparison, RSG generates an accurate response spectrum.

F.16 SEPOL SEPOL (Settlement Problem-Oriented Language) is used to aid in the calculation of predicted magnitudes and progress of settleme nt of shallow foundations. It has four functions: INSITU establishes the soil profile and calculates required soil properties; STRESS DISTRIBUTION is used to calculate stresses due to foundation loading; SETTLEMENT calculates th e magnitude of settlement; and RATE calculates the progress of settlement. These computations may be performed individually, or they may be combined in to a single integrated computer run.

In STRESS DISTRIBUTION, the stresses and strains in three orthogonal directions in a loaded soil mass are calculated using Boussinesq's linear theory of elasticity, assuming the mass is homogeneous, isotropic, and semi-infinite. The input surface load may be built up by superimposing available basic load segments: rectangular, circular, or strip loading configurations (uniform, ramp, inverse ramp, or any). The required input soil properties are Young's modulus and Poisson's ratio.

In SETTLEMENT, the initial settlement is ca lculated first, then the consolidation settlement is calculated, and, finally, the initial and consolidation settlements are summed to obtain the total settlement at gi ven surface locations. The calculations are made on a layer-by-layer basis and accumulated only for specified layers or finite depths. The consolidation settlement can be obtained by summing experimentally observed strains (that is, by inputting Young's modulus and Poisson's ratio over the compressible layers); by using the volume compressibility, m v, the principal stresses, and the pore pressure parameter A; or by using the one-dimensional compressibility (inputting either a v , C c , or m c). In RATE, generation and dissipation of excess pore pressure using Terzaghi's one-dimensional consolidation theory and an arbitrary load history are calculated. The required input soil properties are the coefficients c v and m v. Both single and double drained layers are considered by SE POL. The output consists of excess pore pressure, total pore pressure, and the degree of consolidation at given points and time, as well as time-settlement values a given surface points.

SEPOL (Reference 29) was developed at the Massachusetts Institute of Technology and is a subsystem of ICES (Integrated Civil Engineering Systems). It has been in the public domain since 1967. It was used by Sargent & Lundy.

LSCS-UFSAR F.20 REV. 15, APRIL 2004 F.17 SHAKE SHAKE (Soil Layer Properties and Resp onse/Earthquake) is a program which computes response in a horizontally layered semi-infinite system subjected to vertically traveling shear waves. Strain-compatible soil properties are computed within the program. Earthquake motion can be specified at any level of the soil profile, and a resulting motion can be computed anywhere else in the profile. The method is based on the continuous solution of the shear wave equation. For soil liquefaction studies, plots of stress time-histories at various levels in a soil profile can also be obtained.

The input for the program includes property data for the soil profile, curves of strain versus shear moduli and damping ratios, and the input earthquake motion.

The output includes the strain-compatible so il properties, response spectra of object and computed motions, printer and CALCOMP plots of time-histories, Fourier spectra, and response spectra. Stress time-history plots are also included.

SHAKE originally was developed by John Lysmer and P. B. Schnabel of the University of California, Berkeley (Reference 31). It was modified by Sargent &

Lundy. To verify Sargent & Lundy's version of SHAKE, results from the program were compared with results from a problem in a paper by Idri ss and Seed (Reference 32).

The 100-foot layer of dense sand shown in Figure F.49-1 was analyzed. The properties of the sand were considered to be as follows:

Total unit weight = 125 pcf, (K 2)max = 65, and K o = 0.5.

The parameter (K 2)max relates the maximum shear modulus, G max, and effective mean pressure at any depth, y, below the surface as follows:

Gmax = 1000 (K 2)max'm 1/2 where: 'm = 3 2 1)(o K+ ' K o = coefficient of lateral pressure at rest, and LSCS-UFSAR F.21 REV. 15, APRIL 2004

' = effective vertical pressure at depth y.

Damping values and the variation of modulus values with strain were based on published data for sa nds (Reference 33).

The response of the sand layer was evaluated using the time-history of accelerations recorded at Taft during the 1952 Kern Count y earthquake as base excitation. The ordinates of this time-history were adjusted to provide a maximum acceleration of 0.15g. The results obtained from SHAKE and the published results are compared in Figures F.50-1 and F.51-1. The maximum shear stresses and accelerations from both solutions are compared in Figure F.

50-1; the response spectra of the surface motions are compared in Figure F.51-1. As illustrated in these figures, the two solutions compare favorably.

F.18 SLOPE SLOPE (Slope Stability Analysis) utilizes the theory of equilibrium of forces to determine the factor of safety against s liding of any embankment or slope. It contains the Bishop, Fellenius, and Morgenstern-Price methods of two-dimensional stability analysis. In the Bishop and Fellenius methods, the factor of safety against failure is estimated along a circular surface of failure, whereas any arbitrary failure surface may be chosen for the Morgenstern-Price method.

The input includes the slope geometry, soil profile, soil properties (density, cohesion, and the friction angle) and the piezometric surface(s). The program also has the capability to introduce an earthquake loading assumed as a horizontal gravitational force. Once the problem is input, several options can be used to determine the factor of safety by the various methods. In addition, different stages such as end-of-construction, full-lake, and sudden-drawdown, can be considered in a single run. The output includes factors of safety for each trial surface and a plot of the slope cross section having slope profile, soil profile, water table conditions, and failure surface for the minimu m factor of safety.

SLOPE was developed and put under ICES (Integrated Civil Engineering Systems) by William A. Bailey at the Massachusetts In stitute of Technology. It has been in the public domain since 1967. It was used by Sargent & Lundy. (Reference 34).

LSCS-UFSAR F.22 REV. 15, APRIL 2004 F.19 SLSAP SLSAP (Sargent & Lundy Structural Analysis Program) performs static and dynamic analyses of linear structural systems. The program uses the stiffness matrix method to analyze two- and three-dimensional frames, trusses, and grids; three-dimensional elastic solids; axially sy mmetrical solids, plates, and shells; and straight and curved pipes for arbitrary static loads. Dy namic analyses for frequencies and mode shapes, spectral anal ysis, and numerical integration analyses are also possible.

The program allows materials with arbitrary elastic constants, combined loadings, rigid members, elastic supports, and a combination of different element types.

The program output includes displacement and rotations of all joints or nodes, forces or stresses in members or elemen ts, frequencies and mode shapes, and dynamic response in terms of displacements and forces.

SAP was developed by E. L. Wilson of the University of California at Berkeley. The original version of the program dates back to 1968.

Sargent & Lundy used the SAP IV version (Reference 6). It is used primarily for static analysis.

Results from the program have been compared with several other static and dynamic computer programs and classical solutions. Two examples of these validation problems are presented.

The first example is a cantilever beam under both uniform and concentrated load. (The beam was modeled for SAP using ten equal length beam elements.) It has a cross-sectional area of 1 x 2 inches, a le ngth of 10 inches, and Young's modulus 30 x 10 3 ksi. A uniform load (q = 2 kips/inc h) and a concentrated load (10 kip) are applied at one end of the beam.

The results from the program are compared to analytical results obtained by Timoshenko and Gere (Reference 35). Fi gure F.52-1 shows excellent agreement for the bending moment obtained in both solutions.

In the second example a simply supported square plate under uniform loading is analyzed. A 10-inch square by 1-inch thick square plate with Poisson's ratio = 0.3 and Young's modulus = 30 x 10 3 ksi was loaded with 1 ksi pressure.

The results obtained were compared to those presented by S. Timoshenko and S. Woinowsky-Krieger (Reference 36). The bending moments M xx and M yy for both the x and y symmetry lines obtained in the two solutions are shown in Figure F.53-1.

LSCS-UFSAR F.23 REV. 15, APRIL 2004 The maximum bending moment which occurs at the center of the plate differs by only 1.05%.

F.20 SOR-III SOR-III (Shell of Revolution) is a computer program used to analyze thin shells of revolution subjected to axisymmetric loading by employing a generalized Adams-Moulton method to integrate numerically the governing differential equations.

Arbitrary distribution of normal, tangenti al, and moment surface loadings as well as edge forces and deflections may be analyzed in the axisymmetric loadings. Input of boundary conditions allows for the consideration of elastic support conditions. Temperature variations along the meridian or across the thickness may also be considered.

The program output includes shell displacements, outer fiber stresses, and strains and stress resultants.

SOR-III was developed at Knolls Atomic Power Laboratory for the United States Atomic Energy Commission (Reference 37). Version III was acquired by Sargent &

Lundy in 1969.

Results from this program have been frequently compared with other available solutions and other computer programs to ch eck the validity of th e program. One of these comparisons is the analysis of a circular flat reinforced concrete plate. The details of the problem and the boundary co nditions are shown in Figure F.54-1. Results of the SOR-III analysis were compared with the finite element program, SABOR-III (Reference 22). Figure F.

55-1 shows the bending moment in the meridional and hoop directions, respective ly. Figure F.56-1 shows the comparison of radial shear. As shown in these figures, results compare favorably.

F.21 SSANA SSANA (Spring-Slab Analysis Program) was written to facilitate the reduction of the reinforced concrete shear building and the equipment in the building to a system of rigid slabs inter-connected by weightless linear springs. The program calculates the centroid, total weight, and the weight moment of inertia about the vertical and two horizontal centroidal axis of each slab. The program also calculates the spring stiffness of concrete walls and its distance from the mass centroid.

LSCS-UFSAR F.24 REV. 15, APRIL 2004 Spring constants of shear walls are computed based on the following equation:

K = D T 1 where: DT = DF + DS and DF = Flexural deflection/unit load DS = Shear deflection/unit load.

SSANA was written by Sargent & Lundy.

Hand calculations were used to validate the program. As an example of this validation, stiffness and rotary mass were calculated for elements of the structure shown in Figure F.57-1. The comparis on of results from SSANA and hand calculations shown in Tables F.11-1 and F.11-2 demonstrates the accuracy of the program. F.22 STAND STAND (Structural Analysis and Design) is an integrated structural code which is programmed to perform analysis and design of structural steel members according to the 1969 AISC Specification. It co nsists of the following subsystems:

a. beam edit, b. rolled beam design, c. composite beam design, d. plate girder design, e. column edit, f. column design, and g. column base plate design.

The program input consists of member geom etry and basic loadings. The design is performed for specified combin ations of basic loadings and overstress factors. For floor framing systems, the program is capable of automatic transfer of reactions from tributary beams to supporting members. There are many design control parameters available, such as minimum and maximum depth limitations, shape of LSCS-UFSAR F.25 REV. 15, APRIL 2004 the rolled section, location of the lateral support of the compression flange, material grade or yield stress, deflection limitations, flange cutoff criterion, and location of stiffeners.

For columns, the program is capable of accounting for axial loading as well as uniaxial or biaxial bending.

For column base plate design, only axial load and column combinations are considered.

The program output includes the complete final design and provides the designer with sufficient intermediate information to enable him to evaluate the results. For rolled and composite beam designs, complete details of shopwelded and field-bolted end connections are contained in the ou tput. Supplementary information for economic evaluation of the design is also provided.

STAND was developed by Sargent & Lundy. Some of the principal applications include the design of steel floor framing us ing various types of horizontal structural elements and the design of columns or beam columns.

To validate STAND, results from the program were compared with results from example design problems in the Manual of Steel Construction (Reference 38). Four problems are given.

The first is a rolled beam design problem (E xample 1, pp. 2-5).

A beam of 36-ksi steel is designed for a 125-kip/ft bending moment, assuming its compression flange is braced at 6-foot intervals. The result s, listed in Table F.13-1, show that STAND selects a more efficient section.

The second is a composite beam design problem (Example 1, pp. 2-143 and 2-144). A noncoverplated composite interior floor beam is designed. Limits of 1 1/2 inches for dead load deflection and 1 2/10 inches for live load deflection are imposed. The results, shown in Table F.14-1 are nearly identical.

The third is a column design problem with three examples, (Examples 1, 2 and 5, pp. 3-4, 5, and 9).

The first of these examples is the design of a W12 column of 36-ksi steel that will support a concentric load of 670 kips. The effective length with respect to its minor axis is 16 feet and to its major axis, 31 feet.

The second example is the design of an 11-foot-long W12 interior bay column of 36-ksi steel that will support a concentric load of 540 kips. The column, rigidly framed at the top by 30-foot-long W30 x 116 girders connected to each flange, is braced normal to its we b at the top and the base.

LSCS-UFSAR F.26 REV. 15, APRIL 2004 The third example is the design of a W14 column of 36-ksi steel for a tier building of 18-foot story height that will support a 600-kip gravity load and a 190-kip/ft maximum wind moment, assuming K = 1 relative to both axes and bending is about the major axis.

The results from all three checks are identical to those in the AISC Manual, and are shown in Table F.15-1.

The fourth problem is a plate girder design problem (Example 1, p. 2-108). A welded plate girder is designed to support a uniform load of 3 kips/ft and two concentrated loads of 70 kips as shown in Figure F.58-1. The compression flange of the girder is supported laterally only at points of concentrated load. The close results are shown in Table F.16-1.

F.23 STRESS STRESS analyzes plane and space trusses and frames and plane grid by the stiffness matrix method subjected to static loads only.

The structure can be analyzed for arbitrar y joint loads, member loads, temperature changes, and joint displacements. Plotting features are available with the program.

Output includes joint displacements, equilibrium checks, and reactions and member forces. STRESS was originally developed at the Massachusetts Institute of Technology and is in the public domain (References 39 and 40). The version used by Sargent &

Lundy was adapted by the Chi Corporatio n, Cleveland, Ohio, (Reference 41).

F.24 STRUDL-II STRUDL-II (Structural Design Language) is used primarily for static analysis of frame and truss structures. The program is, however, capable of performing linear static or dynamic analyses for finite element representatives of structures using stiffness matrix methods. Nonlinear st atic problems and stability problems may also be treated.

The program is capable of analyzing plane trusses and frames , grids and elastic bodies, space trusses and frames, or thre e-dimensional elastic solids subjected to arbitrary loads, temperature changes, or specified displacements. Either earthquake accelerations or time-history force may be used for dynamic analysis. Anisotropic materials may also be used. In addition to analysis, the program is capable of performing structural steel desi gn according to AISC Code and reinforced or prestressed concrete design according to ACI Code.

LSCS-UFSAR F.27 REV. 15, APRIL 2004 The program output depends upon the type of finite element used and the analysis that was performed. Included in the output are displacements and member forces and moments or element stresses and moments. Eigenvalues, eigenvectors, and time-history response or nodal response may be obtained for dynamic analyses. Member sizes may be obtained if the design portion is used.

STRUDL-II was developed as part of the Integrated Civil Engineering System at the Massachusetts Institute of Technology (Reference 42).

The program has been in the public domain since 1968. Two vers ions were used by Sargent & Lundy, one maintained by the McDonnell Douglas Automation Company (Reference 43) and one maintained by UNIVAC (Reference 44).

F.25 TEMCO TEMCO (Reinforced Concrete Sections Under Eccentric Loads and Thermal Gradients) analyzes reinforced concrete sections subject to separate or combined action of eccentric loads and thermal grad ients. The effect of temperature is induced in the section by reactions created by the curvature restraint.

The analysis may be done assuming either a cracked or an uncracked section. Material properties can be assumed to be either linear or nonlinear. The program is capable of handling rectangular as well as nonrectangular sections.

The program input consists of section dimensions, areas and location of each layer of reinforcing steel, loads, load co mbinations, and material properties.

The curvature and axial strain corresponding to the given eccentric loads (axial load and bending moment) are determined by an iterative procedure. Thermal gradient is applied on the section by inducing reactions created by the curvature restraint, i.e., there is no curvature change due to a thermal gradient on the section. The axial expansion is assumed to be free after thermal gradient is applied. An iterative procedure is employed again for finding the final strain distribution such that equilibrium of internal and external loads is satisfied.

The program output consists of the echo of input, combined loads, final location of neutral axis, final stresses in steel and concrete, and final internal forces. Similar intermediate results (before thermal gradient is applied) can also be output if desired. The program has applications to a wide variety of reinforced concrete beams and columns, slabs, and containment structures subject to various combinations of external loads and thermal gradients.

LSCS-UFSAR F.28 REV. 15, APRIL 2004 This program was developed by Sargent & Lundy.

To demonstrate the validity of TEMCO, program results are compared with hand calculated results. Three example problems are considered. The section and material properties for each problem are given in Table F.17-1, along with the applied external forces and thermal gradients.

The first problem considered involves a section with two layers of steel under the action of a compressive force applied at the centerline of the section, a bending moment, and a thermal gradient. A cracked analysis of the section is required assuming nonlinear material properties.

The second problem considered involves a section with two layers of steel under the action of a tensile force applied at the centerline of the section, a bending moment, and a thermal gradient. A cracked analysis of the section is required assuming nonlinear material properties.

The third problem considered involves a se ction with two layers of steel under the action of a tensile force applied at the centerline of the section, a bending moment, and a thermal gradient. A cracked analysis of the section is required assuming linear material properties.

The hand-calculated solution was obtained according to the following procedure:

a. Assume the location of neutral axis and the stress distribution to be the same as those given by the program under the given mechanical loading. b. Compute the strain distribution under the given mechanical loading. c. Compute the stress resultants by integration and using the proper stress-strain relationships. d. Check for equilibrium with external mechanical loads. e. If equilibrium is satisfied, compute the curvature imposed on the section by the given thermal gradient. f. Compute the final curvature by subtracting the thermal curvature from the mechanical curvature. g. Compute the new axial strain su ch that equilibrium is satisfied keeping the curvature constant. h. Compute the final stress resultan ts by integration and using the proper stress-strain relationships. i. Compute the thermal moment. j. Check for equilibrium and compare program results with hand calculated results.

LSCS-UFSAR F.29 REV. 15, APRIL 2004 Results obtained using this procedure together with those computed by TEMCO are presented in Table F.18-1.

It is concluded that results given by the program agree very well with results obtained by hand calculations and that equilibrium between internal and external forces is satisfied for all three problems.

F.26 WARLOC The basic FLASH-4 program, which uses a generalized multicell, multiflowpath thermal hydraulic transient calculational method, was improved by S&L to include an expanded network with air as a flowing fluid component in addition to a water/steam component and a number of ot her new features. The S&L modified version of the FLASH-4 program, designated WARLOC, thus has the major physical modeling capability to describe system transients in both boiling water and pressurized water reactors (BWR and PWR), with the associated balance-of-plant compartments. As an option, the blowdown fluid generation model (reactor vessel simulation) can be coupled with or decoupled from the compartment transient analytical model. When this subcompartme nt transient analysis is performed, the decoupled option is utilized (i.e., the reactor vessel simulation model is not used), and the blowdown data is read in or input into the program.

The basis for the computer program is a network of fluid control volumes (fluid nodes) and flow paths (interconnecting co ntrol volumes), for which the conservation equations of mass, momentum, and energy fo r air-steam-water mixtures are solved in space and time. Superimposed on the network are computer subroutines which permit physical modeling of the reactor system, the containment, plant subcompartments, safeguard fluid systems, and the pipe rupture flow. Flow types ranging from homogeneous to stratified can be specified for each node, with transition from one flow type to another.

Nonthermal equilibrium treatment of the nodes enables the nodal atmosphere to ha ve low initial relative humidities and allows steam to coexist with subcooled liquid in the same node, if desired.

Leak flows from ruptured lines are calcul ated using the Moody two-phase critical flow model with a multiplier of 1.0 or orific e flow model. The vent flow process at a subcritical flow condition is described by the one-dimensional momentum equation for a homogeneous two-phase, two-compon ent flow (air-steam-water mixture) through a flow path connecting two pressure nodes. Time-dependent blowdown mass flow rates with their a ssociated enthalpies can be read in as input to model pipe breaks if this information is known from other blowdown codes.

For the normal flow path, the choked mass flow rate can be calculated either by using Moody's correlation with a multiplier or by using the isentropic expansion LSCS-UFSAR F.30 REV. 15, APRIL 2004 method. The actual mass flow rate can be determined by taking the minimum flow rate predicted from any one of the three following comparisons:

a. the inertia flow rate and choked flow rate from Moody's correlation; b. the inertia flow rate and the chok ed flow rate from the isentropic expansion method; or c. the inertia flow rate, the choked flow rate from Moody's correlation, and the choked flow rate from the isentropic expansion method.

The choked mass flow rate using Moody's correlation with a multiplier of .6 is usually more conservative than that obtained using the isentropic expansion method. In the calculation, allowance is made for the presence of liquid droplets in the vent flow. For the air-steam-water mixture system with large pressure differences, isentropic flow of a homogen eous compressible fluid through an orifice or nozzle can be used as one of the choked flow calculations. First the exponent of isentropic expansion, is calculated for the mixture of air and water.

= (1-y air) w + air y air where: air = isentropic exponent for air = 1.4, w = isentropic exponent, = 1.035 + 0.1 X for wet steam, = 1.135 for saturated steam and 1.3 for superheated steam, x = steam quality, and y air = air mole fraction in the flowing fluid.

Using this exponent, the critical pressure ratio is evaluated from

/ (-1) source sink P P = +1 2 where: P sink = pressure of downstream node, and P source = pressure of upstream node.

LSCS-UFSAR F.31 REV. 15, APRIL 2004 The actual pressure ratio is compared to this value. If the actual pressure ratio is supercritical, the choked mass flow rate follows

(+1) / (2-2) wchoked = 144 C D A +1 2 (g c P sink P source)1/2 where: C D = discharge coefficient (normally set to 1.0 for pipes), A = cross-sectional area of path, and Psource = density of fluid in upstream node.

For determining the choked flow rate from Moody's correlation for an air-steam-water system, the total nodal pressure and the enthalpy of the mixture corresponding to the mixture density from the steam table were used with an input Moody multiplier.

The program has the option of using the instantaneous phase separation assumption or the homogeneous fine mist assumption for the air-steam-water system. F.27 PIPSYS PIPSYS (Piping Analysis Information Syst em) is the primary pr ogram used in the analysis of stresses on piping systems. It is a composite of several lesser programs. It combines thermal, weight, and seismic analysis into one program. It may also be broken down into a static analysis portion, a dynamic analysis po rtion, and a stress combination portion, which is in accordan ce with ASME Boiler and Pressure Vessel Code,Section III, Articles NB-3000 and NC-3000.

NUPIPE subroutine is the nuclear code part which combines and computes the stresses in piping components according to ASME Boiler and Pressure Vessel Code,Section III, Articles NB-3000 and NC-3000. It was validated by intercomparison with the results of a Teledyne Corporation program for the same sample problem. The final results were essentially identical.

Like NUPIPE, TEE is also an independen t program as well as a subroutine of PIPSYS. It also was validated by hand calculation for a problem identical to that processed by the computer.

LSCS-UFSAR F.32 REV. 15, APRIL 2004 TEE determines the blowdown force at the discharge pipe due to the sudden opening of a safety/relief valve. It ca lculates the stress levels at the branch intersection points and valve outlets according to Power Piping Code ANSI B31.10.

Finally, it checks the proper discharge pipe sizing at the valve outlet in order to sustain the dynamic force due to the valve blowout.

DYNAPIPE analyzes the dynamic model and upper bound response of a structure (modeled as a space frame) subjected to prescribed dynamic forces at discrete points or subjected to earthquake type motion prescribed as acceleration response spectra. Elbows and tees are treated per piping code.

F.28 PENAN PENAN handles the analysis of axially sy mmetric solids of revolution, which are composed of orthotropic materials (with temperature-dependent properties), and which are subjected to asymmetric and time-dependent heating and loading.

Mainly, the structure of this program is a combination of two, suitably modified, axisymmetric finite element programs. The two programs are:

a. NOHEAT (Nonlinear Heat Transfer Analysis Program), by I. Farhoomand and E. Wilson (Structural Engineering Laboratory, University of California, Berkeley, California. Report Number UC SESEM 71-6, April 1971), and b. ASAL (Finite Element Analysis-A xisymmetric Solids with Arbitrary Loads), by R. S. Dunham and R. E. Nickell (Structural Engineering Laboratory, University of California, Berkeley, California, Report Number 67-6, June 1967).

Outlined below is a brief description of the program's most significant features.

a. PENAN is designed to handle automatic finite element mesh generation and plotting for various penetration assembly configurations. b. It has a built-in materi al property bank covering temperature-dependent mechanical and thermal properties (including fatigue design curves) for all Section III materials. c. It forms optimal loads and load-range combinations for the various Code-specified loading categories and generates Fourier series coefficients for all asymmetrically applied loads. d. Through repeated use of the same stiffness matrix and unit-load stresses, the program carries out all required multiple stress evaluations.

LSCS-UFSAR F.33 REV. 15, APRIL 2004

e. PENAN calculates the allowable stresses for all stress and loading categories, makes the necessary stress comparisons, and generates the entire Penetration Assembly Stress Analysis Report.

F.29 FAST FAST is used primarily for the dynamic analysis of linear axisymmetric structures using transfer functions (modeled as a multi-degree-of-freedom system). The structural model is given as input either in the form of its eigenvalues, eigenvectors, participation factors and modal damping ratios, or in the form of its responses for a typical dynamic load. This program is used to obtain the time history, transfer function and response spectra for various response components of the model. The results are given in print and plot forms.

FAST was developed at Sargent & Lundy in 1975.

F.30 REFERENCES

1. Portland Cement Association, "Notes on ACI 381-71 Building Code Requirements with Design Applications," pp. 1-14, July 1972.

2. M. Hetenyi, Beams on Elastic Foundation, The University of Michigan Press, Ann Arbor, Michigan, pp. 100-106, 1946.

3. A. Kalnins, "Static, Free Vibration and Stability Analysis of Thin, Elastic Shells of Revolution, " Technical Report AFFDL-TR-68-144 , March 1969.

4. S. Timoshenko and S. Woinowski-Krieger, Theory of Plates and Shells , McGraw-Hill, New York, pp. 285-287, 1959.

5. E. L. Wilson, "A Computer Program for the Dynamic Stress Analysis of Underground Structures," Report to U. S. Army Engineering Waterways Experiment Station, January 1965.

6. Bathe, et al. "SAP-IV A Structural Analysis Program for Static and Dynamic Response of Linear Syst ems," Report No. EERC-73-11, University of California, Berkeley, June 1973.

7. Newmark, Nathan and Rosenblueth, Emilio, Fundamentals of Earthquake Engineering , Prentice-Hall, Englewood Cliffs, N. J., p. 15, 1971.

8. J. Biggs, Introduction to Structural Dynamics, McGraw-Hill, New York, p.

266, 1964.

LSCS-UFSAR F.34 REV. 15, APRIL 2004

9. S. Ghosh and E. Wilson, "Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading," Report No. EERC 69-10 , University of California, Berkeley, September 1969.

10. S. Timoshenko and S. Woinowsky-Krieger, Theory of Plates and Shells , McGraw-Hill, New York, pp. 553-554, 1959.

11. S. Timoshenko and J. Goodier, Theory of Elasticity , McGraw-Hill, New York, 1970. 12. B. Budiansky and P. Radkowski, "Numerical Analysis of Unsymmetric Bending of Shell of Revolution," Journal of the American Institute of Aeronautics and Astronautics, August 1963.

13. H. Reismann and J. Padlog, "Forced, Axisymmetric Motions of Cylindrical Shells," Journal of the Franklin Institute , Vol. 284, No. 5, pp. 308-319, November 1967. 14. S. Klein, "A study of the Matrix Displacement Method as Applied to Shells of Revolution," Proceedings, Conference on Matrix Methods in Structural Mechanics, Wright-Patterson Air Force Base, Ohio, 1965.

15. J. F. Abel, P. P. Cole, D. P. B illington, "Maximum Seismic Response of Cooling Towers," Report No. 73-SM-1 , Department of Civil and Geological Engineering Research, Princeton University, March 1, 1973.

16. "EAC/EASE - Elastic Analysis for Structural Engineering," Control Data Corporation, Cybernet Service, June 1972.

17. F. H. Kulhawy, J. M. Duncan, and H. B. Seed, "F inite Element Analyses of Stresses and Movements in Embankments During Construction," Report No.

TE-69-4, College of Engineering Office of Research Services, University of California, Berkeley, California.

18. E. O. Pfrang and C. T. Siess, "Ana lytical Study of the Behavior of Long Restrained Reinforcement Columns," Struct ural Research Series #214, University of Illinois, June 1971.

19. N. Janbu, "Settlement Calculation Based on the Tangent Modulus Concept,"

Three Guest Lectures at Moscow State University, Bulletin Number 2 of Soil Mechanics and Foundation Engineering of the Technical University of Norway, Troudheim, 1967.

LSCS-UFSAR F.35 REV. 15, APRIL 2004

20. ICES-SEPOL - Soil Profile Sett lement Analysis, McDonnell Douglas Automation Company, 1974.

21. A. Kalnins, "Analysis of Shells of Revolution Subjected to Symmetrical and Nonsymmetrical Loads," Journal of Applied Mechanics , ASME Vol. 31, pp. 467-476, 1969. 22. J. H. Percy, D. R. Navaratna, S. Klein, "Stress and Bending Analysis for Shells of Revolution III," ASRL TR 121-7 , MIT Aeroelastic and Structures Laboratory.

23. J. M. Doyle and S. L. Chu, "Liner Plate Buckling and Behavior of Stud and Rip Type Anchors," Proceedings, First International Conference on Structural Mechanics in Reactor Technology, Berlin, Germany, Volume 4, Part H, September 1972. 24. A. G. Young and L. A. Tate, "Design of Liners for Reactor Vessels,"

Proceedings, Conference on Prestressed Concrete Pressure Vessels, Institute of Civil Engineers, London, 1967, Paper J57.

25. "Nelson Study Welding Applications in Power Generating Plants," Nelson Study Welding Company, Lorain, Ohio.

26. Y. K. Cheung and J. D. Davies , "Analysis of Rectangular Tanks,"

CONCRETE, London, England, Vol. 1, pp. 169-174, May 1967.

27. O. Zienkiwicz and Y. Cheung, "The Finite Element Method for Analysis of Elastic Isotropic and Orthotropic Slabs," Proceedings of the Institute of Civil Engineering, London, England, pp. 471-487, August 1964.

28. I. M. Idriss et al., "QUAD4-A Computer Program for Evaluating the Seismic Response of Soil Structures by Variable Damping Finite Element Procedures," Report No. EERC 73-16, Earthquake Engineering Research Center, University of California, Berkeley, California, July 1973.

29. R. L. Schiffman et al., "ICES SE POL-1, A Settlement Problem Oriented Language," R67-73 , Department of Civil Engineering, M.I.T., December 1967.

30. "SEPOL - Soil Profile Settlement Analysis System," McDonnell Douglas Automation Company, May 1968.

LSCS-UFSAR F.36 REV. 15, APRIL 2004

31. P. B. Schnabel et al., "SHAKE:

A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites," Report No. EERC 72-12 , Earthquake Engineering Research Center, University of California, Berkeley, December 1972.

32. I. M. Idriss and H. B. Seed, "Seismic Response by Variable Damping Finite Elements," Journal of the Geotechnical Engineering Division, ASCE, Vol. 100, No.

GT1, Proc. Paper 10284, pp. 1-13, January 1974.

33. H. B. Seed and I. M. Idriss, "Soil Moduli and Damping Factors for Dynamic Response Analyses," Report No. EERC 70-10 , Earthquake Engineering Research Center, University of California, Berkeley, California, December 1970.

34. "ICES SLOPE - Slope Stability Analysis System," McDonnell Douglas Automation Company, 1974.

35. S. Timoshenko and J. Gere, Mechanics of Materials , Van Nostrand Reinhold Company, New York, pp.99-100, 1972.

36. S. Timoshenko and S. Woinowsky-Krieger, Theory of Plates and Shells , McGraw-Hill, New York, p. 118, 1959.

37. "A Program to Perform Stress Analys is of Shells of Revolution," Knolls Atomic Power Laboratory, Schene ctady, New York, September 1963.

38. Manual of Steel Construction , Seventh Edition, American Institute of Steel Construction, New York, N. Y.

39. S. J. Fenves et al., "STRESS:

A User's Manual," MIT Press, 1964.

40. S. J. Fenves et al., "STRESS:

A Reference Manual," MIT Press, 1965.

41. "Chi Stress," Chi-1008, Chi Corporation, Cleveland, Ohio, February 1972.

42. R. D. Logcher et al., "ICES STRUDL II, The Structural Design Language Engineering User's Manual," Department of Civil Engineering, M.I.T., 1st ed., November 1968.

43. "ICES STRUDL Improvements," McDonnell Douglas Automation Company, February 1973.

44. "ICES Application Brief," UNIVAC Marketing Support, Sperry Rand Corporation, 1972.

LSCS-UFSAR TABLE F.1-1 REV 15, APRIL 2004 TABLE F.1-1 (Sheet 1 of 1)

RESULTS FOR 28-DAY STRENGTH RESULTS FOR 28-DAY STRENGTH Number of Samples Collected - - - 46 Mean Observed Strength - - - - - - 3456.1 Allow. Design Str. - - - - - - - 2955.1 Spec. Design Str. - - 2500.0 Observed Standard Deviation - - - 373.0 C.O.V. % - - - - - 10.8 Expected C.O.V. % - 15.0 Within Test Std. Deviation - - - - - .0 C.O.V. % - - - - - .0 Expected C.O.V. % - 5.0 Mean Observed Range - - - - - - - - .0 Total Number of Bad Samples - Average Strength 0 Number of Times Inefficient Testing Observed 0 CONTROL ACCORDING TO ACI MANUAL Number of Samples Falling Below FC - - - - - - - - - - - - 0 Percent of Samples Collected - - - - .00 Number of Samples Falling Below FC-500 - - - - - - - - 0 Percent of Samples Collected - - - - .00 Number of Times Moving Avg Fell Below FC - - - - - - 0 Percent of Samples Collected - - - - .00

LSCS-UFSAR TABLE F.2-1 REV 15, APRIL 2004 TABLE F.2-1 NATURAL PERIODS FOR THE EIGHT LOWEST FLEXURAL MODES PERIODS (seconds)

MODE NUMBER SAP-IV DAPS-II 1 525.79 526.54 2 85.368 86.213 3 30.965 31.658 4 16.059 16.676 5 9.9006 10.439 6 6.8276 7.2634 7 5.1865 5.4680 8 4.3777 4.4676 LSCS-UFSAR TABLE F.3-1 REV 15, APRIL 2004 TABLE F.3-1 STRUCTURAL FREQUENCIES STRUCTURAL FREQUENCY (cps)

MODE NUMBER BIGGS DYNAS 1 1.00 1.00 2 2.18 2.18 3 3.18 3.18 LSCS-UFSAR TABLE F.4-1 REV 15, APRIL 2004 TABLE F.4-1 PROBABLE MAXIMUM STORY DISPLACEMENTS PROBABLE MAXIMUM STORY DISPLACEMENT (inches)

MODE NUMBER BIGGS DYNAS 1 1.50 1.51 2 3.22 3.20 3 4.86 4.68 LSCS-UFSAR TABLE F.5-1 REV 15, APRIL 2004 TABLE F.5-1 ABSOLUTE MAXIMUM STORY SHEARS ABSOLUTE MAXIMUM STORY SHEAR (kips)

MODE NUMBER BIGGS DYNAS 1 3020 3010 2 2080 2068 3 1345 1353 LSCS-UFSAR TABLE F.6-1 REV 15, APRIL 2004 TABLE F.6-1 PROBABLE MAXIMUM STORY SHEARS PROBABLE MAXIMUM STORY SHEAR (kips)

MODE NUMBER BIGGS DYNAS 1 2250 2262 2 1740 1757 3 895 902 LSCS-UFSAR TABLE F.7-1 REV 15, APRIL 2004 TABLE F.7-1 NATURAL PERIODS FOR THE EIGHT LOWEST FLEXURAL MODES PERIODS (seconds)

MODE NUMBER SAPIV DYNAS 1 525.79 525.69 2 85.368 85.369 3 30.965 30.964 4 16.059 16.060 5 9.9006 9.9010 6 6.8276 6.8279 7 5.1865 5.1866 8 4.3777 4.3778 LSCS-UFSAR TABLE F.8-1 REV 15, APRIL 2004 TABLE F.8-1 (Sheet 1 of 5)

INTERACTION DIAGRAM AXIAL LOAD VS. BENDING MOMENT REFERRED TO THE PLASTIC CENTROID OF THE SECTION.

LIST OF SYMBOLS B = WIDTH OF SECTION, IN. T = HEIGHT OF SECTION, IN. D = DEPTH OF TENSILE STEEL, IN. AS = AREA OF TENSILE STEEL, SQ IN. DC = DEPTH OF COMPRESSION STEEL, IN. ASC = AREA OF COMPRESSION STEEL, SQ IN. ES = MODULUS OF ELASTICITY OF REINFORCING STEEL, KSI SSY = YIELD STRESS OF REINFORCING STEEL, KSI. SSU = ULTIMATE STRESS OF REINFORCING STEEL, KSI. PRESTR= INITIAL PRESTRAIN OF REINFORCING STEEL ULTSTR= ULTIMATE STRAIN OF REINFORCING STEEL CUS = 28 DAY STRENGTH OF CONCRETE CYLINDER, KSI. EPSZ = CONCRETE STRAIN FOR MAXIMUM STRESS EPSU = CONCRETE STRAIN AT CRUSHING NSIESS = NUMBER OF POINTS IN INTERACTION DIAGRAM EPSL = MAXIMUM TOP STRAIN FOR WHICH ID. IS COMPUTED INPUT DATA B= 12.000 T= 48.000 D= 45.000 AS= 2.7500 DC= 10.000 ASC= 1.2500 ES= 29000.0000 SSY= 80.0000 SSU= 90.0000 ULTSTR= .020000 PRESTR= .000000 CUS= 4.5000 EPSZ= .002000 EPSU= .004000 NSIESS= 20 EPSL= .003000 RESULTS GIVEN IN THE FOLLOWING ORDER COUNTER CURVATURE TOP STRAIN AXIAL LOAD BENDING MOMENT AXIAL LOAD BENDING MOMENT C.R.FACTOR REDUCED AXIAL REDUCED BENDING (KIPS) (FT.-KIP) DIMENSIONLESS DIMENSIONLESS PHI LOAD (KIPS) MOMENT (FT.-KIP) INITIAL POINT UNDER UNIFORM COMPRESSION STRAIN - EPSZ 1 .00000000 .00200000 2419.8999 6.0875 1.0984 .0007 .8981 2173.3915 5.4674 2 .00001562 .00225000 2349.4036 96.0163 1.0654 .0109 .8982 2110.2042 86.2406 3 .00003125 .00250000 2216.4314 258.0601 1.0060 .0293 .8983 1990.9977 231.8128 4 .00004687 .00275000 2023.3391 495.1479 .9184 .0562 .8984 1817.8463 444.8601 5 .00006250 .00300000 1770.1265 807.2801 .8034 .0916 .8986 1590.6962 725.4495 PRECEDING POINT HAD BOTTOM FIBER STRAIN = ZERO 6 .00007316 .00300000 1377.5877 1191.7730 .6253 .1352 .8989 1238.3646 1071.3289 7 .00008966 .00300000 1163.2063 1325.1266 .5280 .1504 .8991 1045.8417 1191.4246 8 .00010115 .00300000 985.3649 1400.5915 .4472 .1589 .8992 886.0792 1259.4674 9 .00011264 .00300000 833.6598 1447.7928 .3784 .1643 .8994 749.7575 1302.0822 BALANCED POINT. TENSILE STEEL STRAIN = - EPSY 10 .00016245 .00300000 518.9094 1288.7116 .2355 .1462 .8996 466.8107 1159.3245 11 .00021226 .00300000 339.3652 1137.9047 .1540 .1291 .8997 305.3398 1023.8163 12 .00026207 .00300000 218.2740 1021.0622 .0991 .1159 .8998 196.4099 918.7840 13 .00031188 .00300000 127.3296 930.6383 .0573 .1056 .8999 114.5842 837.4830 14 .00036169 .00300000 52.4308 857.1947 .0238 .0973 .9000 47.1856 771.4405 15 .00041149 .00300000 -11.2036 797.4642 -.0051 .0905 .9000 -10.0834 717.7247 16 .00046130 .00300000 -67.2224 747.8149 -.0305 .0849 .9001 -60.5036 673.0722 PRESENT POINT HAD TENSILE STEEL STRAIN = - ULTSTR 17 .00044047 .00200001 -174.2335 571.9126 -.0791 .0649 .9001 -156.8336 514.7982 18 .00041964 .00100000 -269.5452 392.8982 -.1223 .0446 .9002 -242.6467 353.6901 19 .00039880 .00000000 -317.0636 299.7994 -.1439 1.6329 .9002 -285.4348 269.8928 20 .00000000 -.02000000 -360.0000 273.9367 -.1634 .0311 .9003 -380.0000 273.9367 DISTANCE OF PLASTIC CENTROID TO BOTTOM FIBER.IN = 24.9313 LSCS-UFSAR TABLE F.8-1 REV 15, APRIL 2004 TABLE F.8-1 (Sheet 2 of 5)

AS AND ASC CHANGED VALUES, READ MOMENT WITH OPPOSITE SIGN INPUT DATA B= 12.000 T= 48.000 D= 45.0000 AS= 1.2500 DC= 10.0000 ASC= 2.7500 ES= 29000.0000 SSY= 60.0000 SSU= 90.0000 ULTSTR= .020000 PRESTR= .000000 CUS= 4.5000 EPSZ= .002000 EPSU= .004000 NSIESS= 20 EPSL= .003000 RESULTS GIVEN IN THE FOLLOWING ORDER COUNTER CURVATURE TOP STRAIN AXIAL LOAD BENDING MOMENT AXIAL LOAD BENDING MOMENT C.R.FACTOR REDUCED AXIAL REDUCED BENDING (KIPS) (FT.-KIP) DIMENSIONLESS DIMENSIONLESS PHI LOAD (KIPS) MOMENT (FT.-KIP)

INITIAL POINT UNDER UNIFORM COMPRESSION STRAIN + EPSZ 1 .00000000 .00200000 2419.8999 -1.8527 1.0984 -.0002 .8981 2173.3915 -1.6640 2 .00001562 .00225000 2371.9226 84.3211 1.0766 .0073 .8982 2130.3893 57.7713 3 .00003125 .00250000 2258.0534 206.7475 1.0249 .0235 .8983 2029.3136 185.7126 4 .00004687 .00275000 2083.4750 431.3483 .9457 .0489 .8984 1871.7781 387.5200 5 .00006250 .00300000 1843.1875 738.1239 .8389 .0838 .8986 1660.7331 663.2583 PRECEDING POINT HAD BOTTOM FIBER STRAIN = ZERO 6 .00007816 .00300000 1484.8191 1109.2716 .6739 .1259 .8989 1334.6301 997.0735 7 .00008966 .00300000 1292.5999 1224.0333 .5887 .1389 .8990 1162.0507 1100.4092 8 .00010115 .00300000 1133.6276 1273.4436 .5145 .1445 .8991 1019.2732 1144.9854 9 .00011264 .00300000 999.4452 1290.4057 .4536 .1464 .8992 898.7299 1160.3700 BALANCED POINT. TENSILE STEEL STRAIN = - EPSY 10 .00016245 .00300000 669.1896 1129.0560 .3037 .1281 .8995 601.9351 1015.5674 11 .00021226 .00300000 474.8520 963.1077 .2155 .1093 .8996 427.1928 866.4440 12 .00026207 .00300000 339.6791 826.0203 .1542 .0937 .8997 305.6222 743.2018 13 .00031188 .00300000 234.6630 712.3114 .1065 .0808 .8998 211.1542 640.9513 14 .00036169 .00300000 143.7225 611.7068 .0652 .0694 .8999 129.3743 550.4683 15 .00041149 .00300000 64.0464 523.6752 .0291 .0594 .9000 57.8386 471.2818 16 .00046130 .00300000 -8.0140 444.9537 -.0036 .0505 .9000 -7.2127 400.4611 PRECEDING POINT HAD TENSILE STEEL STRAIN = - ULTSTR 17 .00044047 .00200001 -135.5442 255.9720 -.0615 .0280 .9001 -122.0039 230.4015 18 .00041964 .00100000 -232.6858 84.0563 -.1056 .0095 .9002 -209.4590 75.6658 19 .00039980 .00000000 -282.0341 -6.7816 -.1280 -.0369 .9002 -253.8921 -6.1049 20 .00000000 -.02000000 -360.0000 -83.3721 -.1634 .0095 .9003 -360.0000 -83.3721 DISTANCE OF PLASTIC CENTROID TO BOTTOM FIBER.IN = 23.7166 LSCS-UFSAR TABLE F.8-1 REV 15, APRIL 2004 TABLE F.8-1 (Sheet 3 of 5)

INTEPACTION DIAGRAM --- P VS. M ABOUT C. G. OF UNCRACKED TRNSFORMED SECTION YIELD-STRENGTH THEORY WIDTH OF SECTION (IN.) = 12.000 AREA OF TENSILE STEEL (IN.) = 2.750 HEIGHT OF SECTION (IN.) = 48.000 DEPTH OF TENSILE STEEL (IN.) = 45.000 ELASTIC MODULUS, STEEL (KSI) = 29000. AREA OF COMPRESSIVE STEEL (IN.) = 1.250 ELASSTIC MODULUS, CONCRETE (KSI) = 3865. DEPTH OF COMPRESSIVE STEEL (IN.) = 10.000 28-DAY STRENGTH OF CONCRETE CYLINDER (KSI) = 4.500 YIELD STRESS FOR REINFORCING STEEL (KSI) = 54.000 DEPTH OF C.G. OF UNCRACKED TRANSFORMED SECTION (IN.) = 24.435 MODULAR RATIO = 8.000 MAXIMUM STRESS OF CONCRETE (KSI) = 3.825 MAXIMUM STRESS OF REINFORCING STEEL (KSI) = 48.600 UNIT WEIGHT OF CONCRETE (LS./CU. FT.) = 145.000 PHI1 IS THE CAPACITY REDUCTION FACTOR POSITION NO. AXIAL LOAD (KIPS) BENDING MOMENT (FT.-KIP)

PHI1 REDUCED AXIAL LOAD (KIPS) REDUCED BENDING MOMENT (FT.-KIP) 1 2302.69 .00 .8992 2068.33 .00 2 1130.43 796.74 .8991 1016.45 716.37 3 1056.93 841.20 .8992 950.38 756.39 4 262.26 846.88 .8998 235.98 752.02 5 -147.15 212.81 .9001 -132.45 191.55 6 -194.40 155.97 .9001 -174.99 140.40 7 -71.30 -54.99 .9001 -64.18 -49.50 8 -51.38 -90.70 .9000 -46.24 -81.63 9 316.89 -674.33 .8998 285.12 -606.73 10 940.50 -909.37 .8993 845.77 -817.77 11 1172.20 -796.61 .8991 1053.92 -716.23 12 2302.68 .00 .8982 2068.33 .00 LSCS-UFSAR TABLE F.8-1 REV 15, APRIL 2004 TABLE F.8-1 (Sheet 4 of 5)

INTERACTION DIAGRAM --- P VS. M ABOUT C.

G. OF UNCRACKED TRANSFORMED SECTION WORKING-STRESS DESIGN METHOD WIDTH OF SECTION (IN.) = 12.000 AREA OF TENSILE STEEL (IN.) = 2.750 HEIGHT OF SECTION (IN.) = 48.000 DEPTH OF TENSILE STEEL (IN.) = 45.000 ELASTIC MODULUS STEEL (SKI) = 29000. AREA OF COMPRESSIVE STEEL (IN.) = 1.250 ELASTIC MODULUS CONCRETE (KSI) = 3865. DEPTH OF COMPRESSIVE STEEL (IN.) = 10.000 ALLOWABLE STRESS OF CONCRETE IN BENDING (KSI) = 2.700 ALLOWABLE STRESS IN REINFORCING STEEL (KSI) = 20.000 POSITION NO. AXIAL LOAD (KIPS) BENDING MOMENT (FT.-KIP) 1 1717.20 .00 2 824.64 618.32 3 768.37 653.56 4 340.10 659.09 5 -60.56 94.98 6 -80.00 60.75 7 -29.34 -23.89 8 -21.14 -38.23 9 389.08 -627.77 10 718.18 -704.17 11 892.56 -618.00 12 1717.20 .00 LSCS-UFSAR TABLE F.8-1 REV 15, APRIL 2004 TABLE F.8-1 (Sheet 5 of 5)

AXIAL LOAD 1 IN. = 650. KIP.

INTERACTION DIAGRAM - AXIAL LOAD VS. BENDING MOMENT. TEST 4 48 X 12 STEEL T=2.75/40. C=1.25/10. FCP=4.5 FP=60. -1600 -800 -400 800 400 012001600MOMENT 1 IN. = 400. FT.-KIP.

6501300195026002000 USD T S D WSD LSCS-UFSAR TABLE F.9-1 REV 15, APRIL 2004 TABLE F.9-1 COMPUTER OUTPUT DISPLACEMENTS LOCATION VALUE (inches)

U 1 .059492 U 2 .045083 U 3 .033292 U 4 .023913 U 5 .016642 U 6 .011246 U 7 .007491 U 8 .004830 U 9 .002874 U 10 .001338 f 1 16.293 kips Post-Buckling Load P = 21.978 kips

LSCS-UFSAR TABLE F.10-1 REV 15, APRIL 2004 TABLE F.10-1 COMPUTER OUTPUT ANCHOR FORCES LOCATION VALUE (kips) f 1 16.270 f 2 15.430 f 3 14.149 f 4 12.338 f 5 10.935 f 6 9.531 f 7 6.348 f 8 4.093 f 9 2.436 f 10 1.134 LSCS-UFSAR TABLE F.11-1 REV 15, APRIL 2004 TABLE F.11-1 STIFFNESS ELEMENT STIFFNESS PROGRAM HAND CALCULATIONS (kip/ft.)

1 398821 398880 2 398821 398880 3 398821 398880 4 398821 398880 LSCS-UFSAR TABLE F.12-1 REV 15, APRIL 2004 TABLE F.12-1 WEIGHT MOMENT OF INERTIA ABOUT X-AXIS (Ip)

ELEMENT WEIGHT INERTIA PROGRAM HAND CALCULATIONS (ip) 1 2005 2005 2 531 531.25 3 531 531.25 4 32 31.5 LSCS-UFSAR TABLE F.13-1 REV 15, APRIL 2004 TABLE F.13-1 ROLLED BEAM DESIGN PROBLEM MAXIMUM MOMENTS (kip-ft.)

SECTION SELECTED SECTION MODULUS (in 3) AISC 125 W16x40 64.6 STAND 125.58 W18x40 68.4 LSCS-UFSAR TABLE F.14-1 REV 15, APRIL 2004 TABLE F.14-1 COMPOSITE BEAM DESIGN PROBLEM BENDING MOMENTS (kip-ft.)

CONSTRUCTION LOAD DESIGN LOAD MAXIMUM SHEAR (kips) STEEL SECTION NUMBER OF SHEAR CONNECTORS AISC. 71.3 237.2 26.4 W21x44 42 STAND 71.3 236.5 26.3 W21x44 42 LSCS-UFSAR TABLE F.15-1 REV 15, APRIL 2004 TABLE F.15-1 COLUMN DESIGN PROBLEM ITEMS AISC EXAMPLE 1 AISC EXAMPLE 2 AISC EXAMPLE 5 Column Design Parameters 670 k 540 k

600 kips 100 k-ft.

AISC W12x161 W12x99 W14x142 SOLUTION STAND W12x161 W12x99 W14x142 SOLUTION 670 k 540 k 190 k-ft.600 kips LSCS-UFSAR TABLE F.16-1 REV 15, APRIL 2004 TABLE F.16-1 PLATE GIRDER DESIGN PROBLEM RESULTS AISC STAND Maximum Bending Moment (kip-ft.)

2054 2045 Maximum Vertical Shear (kips) 142 141.3 Web Section 1 plate, 70x5/16 1 plate, 70x5/16 Flange Section 2 plates, 18x3/4 2 plates, 18x3/4 Stiffener End Spacing (ft.)

3.5 3.56 Stiffener Intermediate Spacing (ft.)

6.75 6.72 Area of

Stiffeners Furnished (in

2) 2.0 1.88 *required area is 1.78 in 2

LSCS-UFSAR TABLE F.17-1 REV 15, APRIL 2004 TABLE F.17-1 SECTION AND MATERIAL PROPERTIES FOR PROBLEMS SECTION AND MATERIAL PROPERTIES 1 PROBLEM NUMBER 2 3 Thickenss (in.)

42.0 30.0 42.0 Width (in.)

12.0 12.0 12.0 Area of 1st steel layer (in

2) 6.25 2.25 3.12 Distance of 1st steel layer (in.)

3.0 3.0 3.0 Area of 2 nd steel layer (in

2) 6.25 4.0 3.12 Distance of 2 nd steel layer (in.)

37.0 25.0 37.0 Concrete unit weight (lb/ft

3) 150.0 150.0 150.0 Concrete compressive strength (lb/in

2) 4000.0 4000.0 4000.0 Concrete coef. of thermal expansion (in/

°F) 5.56 x 10-6 5.56 x 10-6 5.56 x 10-6 Steel yield strength (kips/in

2) 45.0 45.0 45.0 Steel modulus of elasticity (kips/in

2) 29000.0 29000.0 29000.0 Material properties Nonlinear Nonlinear Linear Applied axial force (kips)

-38.25 76.53 34.65 Applied bending moment (ft-kips) 129.75 -9.49 206.25 Inside temperature (°F) 82.50 67.50 247.50 Outside temperature (°F) 52.50 0.0 115.50 LSCS-UFSAR TABLE F.18-1 REV 15, APRIL 2004 TABLE F.18-1 RESULTS OF PROBLEMS RESULTS 1 PROBLEM NUMBER 2 3 Equilibrating axial force given by TEMCO program (kips)

-38.25 76.53 34.65 Equilibrating axial force computed by hand (kips)

-38.253 76.53 34.65 Equilibrating bending moment given by TEMCO program (ft-kips) 129.75 -9.49 206.25 Equilibrating bending moment computed by hand (ft-kips) 129.752 -9.493 206.25 Thermal moment given by TEMCO program (ft-kips)

-54.58 -21.07 -137.75 Thermal moment computed by hand (ft-kips)

-54.585

-21.071

-137.757

LSCS-UFSAR G.0-i REV. 15, APRIL 2004 APPENDIX G REACTOR RECIRCULATION SYSTEM TABLE OF CONTENTS

PAGE G.0 REACTOR RECIRCULATION SYSTEM G.1.1-1 G.1 INTRODUCTION G.1.1-1

G.1.1 Purpose G.1.1-1 G.1.2 Discussion of Design Changes G.1.2-1 G.1.2.1 Jet Pumps G.1.2-1 G.1.2.2 Flow Control G.1.2-1 G.1.2.3 Pump Motor G.1.2-2 G.1.2.4 Low-Frequency Motor-Generator Set G.1.2-2 G.1.2.5 Cavitation Protection G.1.2-2 G.1.2.6 Piping G.1.2-3 G.1.2.7 Recirculation Pump Flow Measurement G.1.2-3 G.1.2.8 Recirculation Pump Trip (RPT) G.1.2-4 G.1.2.9 Core Flow Measurement G.1.2-4 G.1.2.10 Recirculation System Operation G.1.2-4

G.2 RECIRCULATION SYSTEM DESIGN DESCRIPTION G.2.1-1 G.2.1 Plant Configuration Data G.2.1-1 G.2.1.1 Recirculation System Configuration G.2.1-1 G.2.1.1.1 Description G.2.1-1 G.2.1.2 Thermal, Hydraulic, and Physical Data G.2.1-4 G.2.1.2.1 Recirculation Loop Piping G.2.1-4 G.2.1.2.1.1 Piping Description G.2.1-4 G.2.1.2.1.2 Operating Conditions G.2.1-4 G.2.1.2.2 Valves G.2.1-5 G.2.1.2.2.1 Flow Control Valves G.2.1-5 G.2.1.2.2.1.1 Valve Description G.2.1-5 G.2.1.2.2.1.2 Valve Operating Conditions G.2.1-5 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-ii REV. 15, APRIL 2004 G.2.1.2.2.1.3 Definitions G.2.1-5 G.2.1.2.2.2 Isolation Valves G.2.1-6 G.2.1.2.2.2.1 Suction Line Isolation Valve G.2.1-6 G.2.1.2.2.2.2 Discharge Line Isolation Valve G.2.1-6 G.2.1.2.3 Recirculation Pump G.2.1-6 G.2.1.2.3.1 Pump Description G.2.1-6 G.2.1.2.3.2 Operating Conditions G.2.1-7 G.2.1.2.3.3 Performance at Normal Speed G.2.1-7 G.2.1.2.3.4 Performance at Low Speed G.2.1-7 G.2.1.2.4 Recirculation Pump Motor G.2.1-8 G.2.1.2.4.1 Motor Description G.2.1-8 G.2.1.2.4.2 Operating Conditions G.2.1-8 G.2.1.2.4.3 Required Performance G.2.1-8 G.2.1.2.5 Jet Pumps G.2.1-8 G.2.1.2.5.1 Description G.2.1-8 G.2.1.2.5.2 Design Operating Conditions G.2.1-9 G.2.1.2.5.3 Performance G.2.1-9 G.2.1.2.6 Equipment Interlocks G.2.1-9 G.2.1.2.6.1 Pump Cavitation Protection G.2.1-9 G.2.1.2.6.2 Flow Control Valve Cavitation Protection G.2.1-9 G.2.1.2.7 Recirculation Pump Flow Measurement G.2.1-10 G.2.1.2.7.1 Description G.2.1-10 G.2.1.2.7.2 Performance G.2.1-10 G.2.2 Operating Restriction G.2.2-1 G.2.2.1 Cavitation G.2.2-1 G.2.2.1.1 Background G.2.2-1 G.2.2.1.1.1 Cavitation G.2.2-1 G.2.2.1.1.2 Net Positive Suction Head (NPSH) G.2.2-1 G.2.2.1.1.3 Cavitation Damage G.2.2-1 G.2.2.1.2 Recirculation System Cavitation G.2.2-2 G.2.2.1.2.1 Recirculation Pump G.2.2-2 G.2.2.1.2.2 Flow Control Valve G.2.2-2 G.2.2.1.2.3 Jet Pump Nozzle G.2.2-2 G.2.2.1.2.4 Jet Pump Suction G.2.2-2 G.2.2.1.2.5 Cavitation Coefficients G.2.2-2 G.2.2.1.2.6 NPSH G.2.2-2 G.2.2.1.3 Operating Restrictions G.2.2-3 G.2.2.1.3.1 Automatic Interlocks G.2.2-3 G.2.2.1.3.2 Operating Procedures G.2.2-3 G.2.2.1.3.2.1 One Recirculation Pump Operation G.2.2-3 G.2.2.1.3.2.2 Residual Heat Removal System Operation G.2.2-3 G.2.2.1.3.2.3 Vessel Hydrostatic Testing G.2.2-3 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-iii REV. 15, APRIL 2004 G.2.2.2 Loop Flow Balance G.2.2-4 G.2.2.3 Thermal Shock G.2.2-4 G.2.2.4 Restart of One Recirculation Pump G.2.2-4 G.2.2.5 25% to 100% Speed Transfer G.2.2-4 G.2.3 Operation and Performance G.2.3-1 G.2.3.1 General Description G.2.3-1 G.2.3.2 Normal Operation G.2.3-2 G.2.3.2.1 Start G.2.3-2 G.2.3.2.2 Heatup and Pressurization G.2.3-3 G.2.3.2.3 Low Thermal Power G.2.3-3 G.2.3.2.4 High Thermal Power G.2.3-3 G.2.3.2.5 Natural Circulation G.2.3-3 G.2.3.2.6 Thermal Power - Core Flow Operating Map G.2.3-4 G.2.3.2.7 Mechanical and Control Features G.2.3-4 G.2.3.3 Abnormal Operation G.2.3-6 G.2.3.3.1 Safety G.2.3-7 G.2.3.3.1.1 Moderate and Infrequent Events G.2.3-7 G.2.3.3.1.2 Accident Events G.2.3-7 G.2.3.3.2 Power Generation G.2.3-7 G.2.3.3.2.1 Cavitation Interlocks G.2.3-7 G.2.3.3.2.2 Recirculation FCV Runback G.2.3-7 G.2.3.3.2.3 Flow Control Valve Minimum Position Interlock G.2.3-7 G.2.3.3.2.4 LFMG Output Breaker Control G.2.3-7 G.2.3.3.2.5 High Loop Flow Mismatch G.2.3-8 G.2.3.3.2.6 Loop Suction and Discharge Isolation Valve Position G.2.3-8 G.2.3.3.2.7 Trip to 25% Speed G.2.3-8 G.2.4 DELETED G.2.4-1

G.2.4.1 DELETED G.2.4-1 G.2.4.2 DELETED G.2.4-1 G.2.4.3 DELETED G.2.4-1 G.2.4.4 DELETED G.2.4-2 G.2.4.4.1 DELETED G.2.4-2 G.2.4.4.2 DELETED G.2.4-3 G.2.4.4.3 DELETED G.2.4-3 G.2.4.4.4 DELETED G.2.4-3 G.2.4.4.5 DELETED G.2.4-3

LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-iv REV. 15, APRIL 2004 G.2.5 Thermal and Hydraulic Characteristics Summary: Tables and Figures G.2.5-1 G.2.5.1 Table G.2.5-1 System Design Data G.2.5-1 G.2.5.2 Figure G.2.5-1: Two Pump Thermal Power-Core Flow

Operating Map G.2.5-1 G.2.5.3 Figure G.2.5-2: One Pump Thermal Power-Core Flow Operating Map G.2.5-1 G.2.5.4 Table G.2.5-2 Pressures and Temperatures at Selected Conditions G.2.5-1 G.2.5.5 Table G.2.5-3: System Trip and Start Functions G.2.5-1 G.2.5.6 Figure G.2.5-3: Flow Control Valve Characteristics Curve G.2.5-1 G.2.5.7 Table G.2.5-4: Reacto r Recirculation System Design Characteristics G.2.5-1 G.2.5.8 Table G.2.5-5: Plant Configuration Data G.2.5-1 G.3 SYSTEM DESIGN EVALUATIONS G.3.1-1

G.3.1 Thermal/Hydraulic Conc eptual Design Studies for BWR/5 and /6 G.3.1-1 G.3.1.1 Transients G.3.1-1 G.3.1.1.1 Flow Increase Transient G.3.1-1 G.3.1.1.2 Turbine Trip G.3.1-2 G.3.1.1.3 Trip of One Feedwater Pump G.3.1-2 G.3.1.1.4 Other Transients G.3.1-2 G.3.1.2 Accidents G.3.1-2 G.3.1.2.1 Design-Basis Loss-of-Coolant Accident (DB-LOCA) G.3.1-2 G.3.1.2.2 Anticipated Transients Without Scram (ATWS) G.3.1-3 G.3.2 Design Bases G.3.2-1

G.3.2.1 General Design Bases G.3.2-1 G.3.2.1.1 Safety Design Bases G.3.2-1 G.3.2.1.2 Power Generation Design Bases G.3.2-1 G.3.2.2 Safety Design Bases G.3.2-1 G.3.2.2.1 Normal Operation G.3.2-1 G.3.2.2.2 Transient Operation G.3.2-2 G.3.2.2.3 Accident Operation G.3.2-2 G.3.2.3 Power Generation Design Bases G.3.2-3 G.3.2.3.1 General Requirements G.3.2-3 G.3.2.3.2 Normal Operation G.3.2-3 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-v REV. 15, APRIL 2004 G.3.2.3.3 Transient Operation G.3.2-3 G.3.3 Safety and Power Generator Evaluations G.3.3-1 G.3.3.1 Component Classification G.3.3-1 G.3.3.2 System Design G.3.3-1 G.3.3.2.1 Transient Performance G.3.3-1 G.3.3.3 System Component Design G.3.3-1 G.3.3.3.1 General G.3.3-1

G.3.3.3.1.1 Pressure Integr ity (External Loop Piping, Recirculation Pump Bowl, Block Valve Body, Flow Control Valve Body) G.3.3-1 G.3.3.3.1.2 Inspection and Testing G.3.3-1 G.3.3.3.2 Jet Pumps G.3.3-2 G.3.3.3.3 Recirculation Pump and Motor G.3.3-2 G.3.3.3.3.1 Analytical Methods for Evaluation of Pump Speed and Bearing Integrity G.3.3-2 G.3.3.3.3.1.1 Pump Shaft Critical Speed G.3.3-2 G.3.3.3.3.1.2 Pump Bearing Integrity G.3.3-2 G.3.3.3.4 Recirculation System Piping G.3.3-3 G.3.3.3.4.1 Pipe Rupture G.3.3-3 G.3.3.3.5 Suction and Discharge Block Valves G.3.3-3 G.3.3.3.5.1 Safety Requirement G.3.3-3 G.3.3.3.5.2 Design Description G.3.3-3 G.3.3.3.5.3 Safety Evaluation G.3.3-3 G.3.3.3.6 Flow Control Valv e, Actuator and Hydraulic Power Unit G.3.3-3 G.3.3.3.6.1 Safety Requirements G.3.3-3 G.3.3.3.6.2 Design Description G.3.3-3 G.3.3.3.6.2.1 Flow Control Valves G.3.3-3 G.3.3.3.6.2.2 Flow Control Valve Actuator G.3.3-4 G.3.3.3.6.2.3 Hydraulic Power Unit G.3.3-4 G.3.3.3.6.2.4 Hydraulic Line Design G.3.3-4 G.3.3.3.6.2.4.1 Design Pressure Rating of Interconnecting Piping G.3.3-4 G.3.3.3.6.2.5 Hydraulic Line Isolation Valves G.3.3-5 G.3.3.3.6.2.6 Hydraulic Fluid G.3.3-5 G.3.3.3.6.3 Safety Evaluation G.3.3-5 G.3.3.3.6.3.1 Flow Control Valve G.3.3-5 G.3.3.3.6.3.2 Flow Control Va lve Seizure: Pressure Boundary Integrity G.3.3-5 G.3.3.3.6.3.3 Hydraulic Actuator and Hydraulic Power Unit G.3.3-6 G.3.3.3.6.3.3.1 Actuator Velocity Limits G.3.3-6 G.3.3.3.6.3.3.2 Summary of Failure Analysis G.3.3-7 G.3.3.3.7 Flow Control System G.3.3-7 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-vi REV. 14, APRIL 2002 G.3.3.3.7.1 Safety Requirements G.3.3-7 G.3.3.3.7.2 Design Description G.3.3-8 G.3.3.3.7.3 Safety Evaluation G.3.3-8 G.3.3.3.7.3.1 Flow Control Valve Stroke Rate G.3.3-8 G.3.3.3.7.3.2 Loss-of-Coolant Accident G.3.3-8 G.3.3.3.7.3.3 DELETED G.3.3-9 G.3.3.3.8 Recirculation Pump Trip (RPT) System G.3.3-9 G.3.3.3.8.1 Safety Requirements G.3.3-9 G.3.3.3.8.2 RPT Design G.3.3-9 G.3.3.3.8.3 Failure Mode and Effects Analysis (FMEA) G.3.3-10 G.3.3.3.9 Low-Frequency Motor-Generator (LFMG) Set G.3.3-10 G.3.3.3.9.1 Special Design Requirements G.3.3-10 G.3.3.3.9.1.1 Safety G.3.3-10 G.3.3.3.9.1.2 Power Generation G.3.3-10 G.3.3.3.9.2 Design Parameters G.3.3-11 G.3.3.3.9.2.1 General G.3.3-11 G.3.3.3.9.2.2 Interlocks G.3.3-11 G.3.3.3.9.3 Failure Mode and Effects Analysis (FMEA) G.3.3-12 G.3.3.3.10 Control Interlocks G.3.3-12 G.3.3.3.10.1 Power-Generation Design Requirements G.3.3-12 G.3.3.3.10.2 Design Description G.3.3-12 G.3.3.3.10.2.1 Reactor Lo w Power Level - LFMG Set Auto Sequence G.3.3-12 G.3.3.3.10.2.2 Reactor Normal Power Level Low Speed to High Speed Transfer G.3.3-12 G.3.3.3.10.2.3 High Speed to Low Speed Auto Transfer Sequence G.3.3-13 G.3.3.3.10.3 Failure Mode and Effects Analysis (FMEA) G.3.3-13 G.3.3.3.11 Power Supplies G.3.3-13 G.3.3.3.11.1 Design Requirements G.3.3-13 G.3.3.3.11.2 Power Supply Arrangement G.3.3-13 G.3.3.3.11.3 Failure Mode and Effects Analysis (FMEA) G.3.3-13 G.4 TESTING G.4.1-1 G.4.1 Development Tests G.4.1-1 G.4.1.1 Five-Nozzle Jet Pumps G.4.1-1 G.4.1.1.1 Development of Jet Pumps G.4.1-1 G.4.1.1.1.1 Improvement and Evaluation of Performance G.4.1-1 G.4.1.1.1.2 First-Generation Jet Pumps G.4.1-2 G.4.1.1.1.3 Second-Generation Jet Pumps G.4.1-2 G.4.1.1.2 Cavitation Tests G.4.1-4 G.4.1.1.2.1 Effects of Cavitation G.4.1-4 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-vii REV. 14, APRIL 2002 G.4.1.1.2.2 Cavitation Tests on the Laguna Verde Production Jet Pumps G.4.1-4 G.4.1.1.3 Production Tests G.4.1-5 G.4.1.2 Flow Control Valve Testing G.4.1-6 G.4.1.3 Low Frequency Motor-Generator (LFMG) Set Testing G.4.1-6 G.4.1.4 Hydraulic Power Unit and Actuator Testing G.4.1-7 G.4.2 Preoperational Tests G.4.2-1 G.4.2.1 Recirculation System G.4.2-1 G.4.2.1.1 Test Conditions G.4.2-1 G.4.2.1.2 Test Requirements G.4.2-1 G.4.2.1.2.1 Functional Tests G.4.2-1 G.4.2.1.2.2 System Performance Tests G.4.2-1 G.4.2.2 Vessel Internals Vibration Tests G.4.2-1 G.4.2.2.1 Vibration Test Conditions G.4.2-1 G.4.2.2.2 Inspection of Internals G.4.2-2

G.4.3 Startup Tests G.4.3-1 G.4.3.1 Recirculation System G.4.3-1 G.4.3.1.1 General Description G.4.3-1 G.4.3.1.2 Specific Tests G.4.3-1 G.4.3.1.2.1 Cavitation Tests G.4.3-1 G.4.3.1.2.2 Stratification Tests G.4.3-1 G.4.3.1.2.3 Flow Increase Operation Test G.4.3-1 G.4.3.1.2.4 Steady-State Performance Test G.4.3-2 G.4.3.1.2.5 Jet Pump and Recirculation Pump Flow Measurement G.4.3-2 G.4.3.1.2.6 Recirculation Control System Test G.4.3-2 G.4.3.1.2.7 Recirculation Pump Trip Test G.4.3-2

G.5 RECIRCULATION PUMP TRIP AND FLOW CONTROL FEATURES G.5.1-1 G.5.1 Recirculation Pump Trip (RPT) System G.5.1-1 G.5.1.1 Identification G.5.1-1 G.5.1.2 Power Supplies G.5.1-1 G.5.1.3 Equipment Design G.5.1-1 G.5.1.3.1 Circuit Description G.5.1-1 G.5.1.3.2 Bypasses and Interlocks G.5.1-2 G.5.1.3.3 Redundancy G.5.1-3 G.5.1.3.4 Testability G.5.1-3 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-viii REV. 15, APRIL 2004 G.5.1.4 Environmental Considerations G.5.1-3 G.5.1.5 Operational Considerations G.5.1-3 G.5.1.6 Analysis G.5.1-4

G.5.1.6.1 Conformance to Functional IEEE Requirements G.5.1-4 G.5.1.6.2 Conformance to Regulatory Guides G.5.1-8 G.5.1.6.3 General Design Criteria Conformance

10 CFR 50 Appendix A G.5.1-8 G.5.2 Recirculation Flow Control G.5.2-1 G.5.2.1 Identification G.5.2-1 G.5.2.1.1 Function G.5.2-1 G.5.2.1.2 Classification G.5.2-1 G.5.2.2 Power Sources G.5.2-1 G.5.2.2.1 Control System G.5.2-1

G.5.2.2.2 Pump Drive Motor and Controls G.5.2-1 G.5.2.3 Equipment Design G.5.2-2 G.5.2.3.1 General G.5.2-2 G.5.2.3.2 LFMG Set G.5.2-3 G.5.2.3.3 Block Valves (Suction/Discharge) G.5.2-3 G.5.2.3.4 Flow Control Valve Position Control Components G.5.2-4 G.5.2.3.4.1 DELETED G.5.2-4 G.5.2.3.4.2 DELETED G.5.2-5 G.5.2.3.4.3 DELETED G.5.2-5 G.5.2.3.4.4 DELETED G.5.2-5 G.5.2.3.4.5 DELETED G.5.2-5 G.5.2.3.4.6 Manual/Automatic Transfer Stations G.5.2-6 G.5.2.3.4.7 Flow Controller G.5.2-6 G.5.2.3.4.8 Limiter G.5.2-6 G.5.2.3.4.9 Hydraulic Power Unit G.5.2-6 G.5.2.3.4.9.1 General Description G.5.2-6 G.5.2.3.4.9.2 Operation G.5.2-7 G.5.2.3.4.9.3 Protection Interlocks G.5.2-8

G.5.2.3.4.9.4 Hydraulic Fluid Description G.5.2-9 G.5.2.3.4.10 Hydraulic Line Design G.5.2-9 G.5.2.3.4.11 Actuator Design G.5.2-9 G.5.2.3.5 Inspection and Testing G.5.2-9 G.5.2.4 Environmental Considerations G.5.2-10

G.5.2.5 Operational Considerations G.5.2-10 G.5.2.5.1 General Information G.5.2-10 G.5.2.5.2 Reactor Operator Information G.5.2-10 G.5.2.6 General Functional and Specific Regulatory Requirements Conformance G.5.2-11 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

G.0-ix REV. 15, APRIL 2004 G.6 TRANSIENT ANALYSIS G.6.1-1 G.7 REFERENCES G.7.0-1 ATTACHMENT G.A: REACTOR RECIRCULATION SYSTEM DESIGN

DRAWINGS ATTACHMENT G.B: FAILURE MODES AND EFFECTS

ANALYSES G.B-i

LSCS-UFSAR G.0-x REV. 14, APRIL 2002 APPENDIX G REACTOR RECIRCULATION SYSTEM LIST OF TABLES G.2.5-1 System Design Data at Rated Thermal Power and Core Flow G.2.5-2 Pressures and Temperatures at Selected Conditions G.2.5-3 Recirculation System Trip and Start Functions

G.2.5-4 Reactor Recirculation System Design Characteristics G.2.5-5 Plant Configuration Data G.3.3-1 Recirculation System Equipment Classification

G.5.2-1 Reactor Recirculation Pump Trip Functions

LSCS-UFSAR APPENDIX G REACTOR RECIRCULATION SYSTEM LIST OF FIGURES G.0-xi REV. 15, APRIL 2004 NUMBER TITLE G.2.1-1 Schematic of Recirculation System G.2.1-2 Elevation A-A: Recirc ulation System External Loop Piping Layout

G.2.1-3 Plan 1: Recirculation System External Loop Piping Layout G.2.1-4 Plan 2: Recirculation System External Loop Piping Layout G.2.1-5 Plan 3: Recirculation System External Loop Piping Layout G.2.1-6 Recirculation Pump characteristics G.2.3-1 LFMG Set: Flow Control Valve Recirculation System G.2.3-2 Typical Reactor Power Flow Operating Map

G.2.4-1 DELETED

G.2.4-2 DELETED

G.2.4-3 DELETED G.2.5-1 Unit 2 Two-Pump Thermal Power - Core Flow Operating Map G.2.5-1a Unit 1 Two-Pump Thermal Power - Core Flow Operating Map G.2.5-2 One-Pump Thermal Power - Core Flow Operating Map G.2.5-3 Flow Control Valve Characteristic Curve

G.3.1-1 Transient Analysis Figures - 30%/sec Valve Stroking Rate LSCS-UFSAR APPENDIX G REACTOR RECIRCULATION SYSTEM LIST OF FIGURES (cont'd)

G.0-xii REV. 15, APRIL 2004 NUMBER TITLE G.3.3-1 Highest Single-Failure Stroke Rate G.3.3-2 RPT Logic Diagram G.3.3-3 Breaker Designation

G.3.3-4 Power Supply Development "A" Recirculation Loop G.3.3-5 Layout of Auxiliary Power System for Unit 1 and Unit 2 Reactor Recirculation System

G.4.1-1 Jet Pump Terminology

G.4.1-2 Operating Principle of Jet Pump G.5.1-1 Recirculation Pump Trip System A G.5.1-2 Turbine Control Valve Fast Closure Sensors

G.5.1-3 Turbine Stop Valve Sensors G.5.2-1 Configuration of the RRFC System G.5.2-2 Valve Actuator Hydraulic Power Unit LSCS-UFSAR G.0-xiii REV. 14, APRIL 2002 DRAWINGS CITED IN THIS APPENDIX*

DRAWING* SUBJECT M-93 Nuclear Boiler & Reactor Recirculating System, Unit 1 M-139 Nuclear Boiler & Reactor Recirculating System, Unit 2 M-2093 Reactor Recirculating System "RR", Unit 1 M-2139 Reactor Recirculating System "RR", Unit 2 1E-1(2)-4205 Schematic Diagrams for Re actor Recirculation System "RR"

The listed drawings are included as "General References" only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.

LSCS-UFSAR G.1.1-1 REV. 13 APPENDIX G - REACTOR RECIRCULATION SYSTEM G.1 INTRODUCTION G.1.1 Purpose

The purpose of this appendix is to describe the design of the La Salle County Station (LSCS) recirculation system in sufficient detail to show compliance with NRC nuclear reactor design requirements.

LSCS-UFSAR G.1.2-1 REV. 13 G.1.2 Discussion of Design Changes LSCS has an improved recirculation system design that retains many of the features of the BWR/3 and BWR/4 product lines. The LSCS design is typical of the BWR/5 and /6 generation of plants. In order to put this design in perspective, the design changes made with the introduction of the BWR/5 are compared to the BWR/3 and /4 design in the following.

G.1.2.1 Jet Pumps Jet pumps are common to BWR/3, /4, /5, and /6. The design of the jet pump was changed with introduction of the BWR/5 to implement a more efficient design which allowed a decrease in pump motor size and horsepower. The major difference in the jet pump design is the five-nozzle jet configuration, where five jets induce driving flow through each jet pump throat rather than the single jet typical of the BWR/3 and /4 generation.

G.1.2.2 Flow Control The flow control design was changed with the BWR/5 to improve the response and

load-following capabilities of the previous design. The control concept changed from recirculation pump speed control to direct recirculation flow control using a valve in each external recirculation loop. The sa lient design features are as follows:

BWR/3 and /4 Variable speed pump and motor

Drive motor - fluid coupler - generator - pump motor:

The M-G varies the line frequency supplied to the pump motor. The fluid coupler provides slip between the moto r and generator, thereby governing the output frequency of the generator. The control design incorporates both manual and automatic features, using feedback loops of generator speed and turbine load demand.

BWR/5 and /6 Constant speed pump and motor Flow control valve (FCV):

Varies the flow directly by increasing or decreasing the pressure drop in the recirculation loop.

LSCS-UFSAR G.1.2-2 REV. 15, APRIL 2004 Hydraulic actuator - power unit:

The FCV position is varied by a hydr aulic cylinder mounted on the valve.

Hydraulic lines lead from the cylinder to a power unit located outside the primary containment. The power unit controls the line pressure and the fluid flow to or from either side of the actuator piston, thereby changing flow control valve position.

Electrical Control:

The remainder of the system incorporates manual features using feedback loops of FCV position and velocity.

Automatic load following capabilities have been removed.

G.1.2.3 Pump Motor The coastdown characteristic of the recircu lation system after the power supply is tripped is important to mitigate some tr ansient and accident events. The factor that governs this coastdown for the BWR/5 design is the inertia of the pump impeller and motor rotor whereas the BWR/3 and /4 design was governed by the pump-motor inertia plus the inertia of the motor-fluid coupler-generator set. Since the MG set was deleted from the BWR/5 design, the inertia of the pump motor rotor was increased by approximately 20% in order to obtain a coastdown characteristic similar to a BWR/3 and /4.

G.1.2.4 Low-Frequency Motor-Generator Set

Since feedwater flow is low during plant heatup and low thermal power operation, there is little net positive suction head (NPSH) available to the recirculation pumps, flow control valves, and jet pumps.

The NPSH available downstream of the recirculation pump is a function of the NPSH available to the pump and the heat addition to the fluid due to pump inefficiency. The LFMG set operates the pump at 25% of normal speed to minimize pump heat input, thereby avoiding cavitation in the loop during plant heatup and low-power operation. At high thermal power

levels, there is enough feedwater subcoolin g to allow normal speed operation of the pump. G.1.2.5 Cavitation Protection The introduction of the flow control valve and five-nozzle jet pumps made it necessary to provide different cavitation protection interlocks, since these components produced different cavitation characteristics than the BWR/3 and /4 design. The design features are as follows:

LSCS-UFSAR G.1.2-3 REV. 15, APRIL 2004 BWR/3 and 4 Cavitation is prevented by a feedwater flow interlock set at 20% to 30% of normal flow, which reduces pump speed to below 30%

of rated. The setpoint is determined by the thermal power level where cavitation would occur when the recirculation pump was operated at rated flow. This is conservative for all lower flows, since NPSH requirements decrease at lower flows, while available NPSH remains essentially the same for any given thermal power level.

BWR/5 and /6 The cavitation characteristics (thermal power level at which cavitation will occur for a given flow) of the jet pumps and recirculation pump are similar to a BWR/3 and /4 except that jet pump cavitation occurs at higher thermal power levels. Protection is provided by measuring the available subcooling and tripping the pumps to 25% speed when there is inadequate subcooling. A temperature element in the loop suction line and a pressure transmitter measuring vessel pressure in conjunction with a function generator measure the amount of subcooling available to the system.

When the recirculation pump is operating at 100% speed, the flow control valve will cavitate at low flows due to the large pressure drop across the valve. Cavitation protection is provided by a feedwater flow interlock that trips the pumps to 25%

speed when there is not enough feedwater flow subcooling to allow operation at low flow control valve positions.

G.1.2.6 Piping The recirculation system discharge piping configuration was changed to allow room for the flow control valve between the recirculation pump and discharge isolation valve. Pipe diameter was also reduced to minimize drywell space requirements.

G.1.2.7 Recirculation Pump Flow Measurement

The recirculation loop flow element is used to provide indirect core flow indication for the flow bias scram system and to indicate pump performance and jet pump drive flow. In general, these functions do not require high measurement accuracy, although repeatability is required. The flow bias scram system requires a signal that is proportional to pump flow. The signal is used as an indicator of core flow.

The proportionality constant (calibration coefficient - pressure drop versus flow) is unimportant as long as that constant does not change, that is, the element is repeatable. Flow is controlled through a Manual/Auto (M/A) station. Individual FCV position demand is used as feedback for the ganged position control setpoint. For ganged position control, the common position setpoint is fed to each loop where it is compared to actual position to develop the FCV position demand signal.

LSCS-UFSAR G.1.2-4 REV. 14, APRIL 2002 These functional requirements are satisf ied by an elbow flow element where the pressure from the inside to the outside of the elbow is proportional to flow. Consequently, the flow nozzle in the BWR/4 design was replaced by an elbow pressure tap flow element in the recirculation pump suction line for the BWR/5 and

/6 design.

G.1.2.8 Recirculation Pump Trip (RPT)

The recirculation pumps are tripped for ma ny reasons, among which are low NPSH, anticipated transient without scram (ATWS) mitigation, and electrical faults such as short circuits. Only one trip function is currently required to be safety grade, and that function is given the name RPT. The purpose of RPT is to mitigate the thermal consequences of the turbine trip and generator trip transients by tripping the recirculation pumps early in the even t, producing rapid pump flow coastdown and additional core voiding, which results in a core reactivity reduction. This system is linked to the reactor protection system (RPS) such th at both a scram and a pump trip occur when the turbine stop valves start to close and when turbine control valve fast closure occurs. Both scram and RPT are bypassed at low thermal power levels.

Since only one power source is available to a BWR/4 pump motor, RPT trips the pumps completely off. The BWR/5 activates the 25% speed source (the low-frequency M-G set) when the pump has coasted down to that speed.

G.1.2.9 Core Flow Measurement The core flow measurement system is unchanged from the BWR/4 design. For BWR/5 and /6, as an operating convenience, individual jet pump pressure drop signals are fed to the process computer to calibrate the system and obtain the jet pump integrity surveillance data requir ed by the Technical Specifications.

G.1.2.10 Recirculation System Operation Due to the changes described in Subsecti ons G.1.2.2 and G.1.2.5, the startup and operation of the BWR/5 recirculation system is significantly changed from previous systems. As a result, new control interlocks were necessary to prevent significant transients, equipment damage, or unnecessar y scrams. Electrical interlocks were installed between the LFMG set and the normal power supply to prevent damage to the LFMG set and on the flow control valv e to prevent cavitation damage. These interlocks also protect against flow-increase transients when starting the system or transferring to the normal power supply.

LSCS-UFSAR G.2.1-1 REV. 14, APRIL 2002 G.2 RECIRCULATION SYSTEM DESIGN DESCRIPTION G.2.1 Plant Configuration Data G.2.1.1 Recirculation System Configuration

G.2.1.1.1 Description The function of the recirculation system is to circulate reactor coolant through the core and to provide a means of controlling reactor power output by varying the rate of flow of the coolant. In this system, water is circulated through the core by the use of jet pumps within the reactor pressure vessel. The arrangement of equipment and the flow paths are shown schematically in Figures G.2.1-1 to G.2.1-5.

Driving flow for the jet pumps is drawn from the bottom of the annular region surrounding the reactor core into two symmetrically oriented piping loops. Each loop contains a single-stage, vertical, centrifugal pump installed at an elevation several feet below the reactor vessel to provide the required net positive suction head (NPSH). Gate valves are located in both suction and discharge piping. The pump discharge flow return line connects to a manifold formed into a near semicircle to fit close to the reactor vessel. Smaller, vertical pipes from the manifold (one for every two jet pumps) return the high-pressure flow to the reactor at nozzles distributed circumferentially on the vessel wall. Inside the reactor vessel, the circuit is continued through internal risers which pass upward in the annular region (downcomer) between the core shroud and the vessel wall. Each internal riser is terminated with a special fitting which distributes the flow to the nozzle-supply passages of twin jet pumps.

Reactor recirculation system equipment that is part of the reactor pressure boundary is designed as Seismic Category I equipment. As such, it is designed to resist sufficiently the response motion at the installed location within the supporting structure for the safe shutdown earthquake. The pump is assumed to be filled with water for the analysis. Vibrat ion snubbers located at the top of the motor and at the bottom of the pump casing are designed to resist the horizontal reactions.

The recirculation piping, valves, and pump s are supported by hangers to avoid the use of piping expansion loops that would be required if the pumps were anchored.

In addition, the recirculation loops are provided with a system of restraints designed so that reaction forces associated with a split or circumferential pipe break do not jeopardize drywell integrity. This restraint system provides adequate clearance for normal thermal expansion movement of the loop.

The recirculation system piping, valves, and pump casings are covered with thermal insulation having a maximum average heat transfer rate of 65 Btu/hr-ft 2 of LSCS-UFSAR G.2.1-2 REV. 16, APRIL 2006 insulation extension surface with the system at rated operating conditions. This heat loss includes losses through joints, laps, and other openings that may occur in normal application.

The insulation is all-metal reflective type. It is prefabricated into components for field installation. Removable insulation is provided at various locations to permit periodic inservice inspection of the equipment.

The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high-pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the driven flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated by the driving flow. The flows, both driving and driven, are mixed in the jet pump throat section and result in partial pressure recovery. The balance of recovery is obtained in the jet pump diffusing section.

During startup and low-power operation, the recirculation pumps are run at 25% of rated speed to provide forced circulation. Two 15-Hz motor-generator sets provide the power to the pump motors during these periods. The pump is first started with the flow control valve in its minimum position (with the flow control valve at its minimum position, approximately 34% of rated pump flow will pass with the pump operating at high speed). The main (60-Hz) power source brings the pump up to full speed and is then tripped. During the following coastdown, the low-frequency motor generator (LFMG) set starts and energizes th e pump at 25% speed. The flow control valve is then opened to prov ide at least 20% pump flow.

The allowable heatup rate for the recirculati on pump casing is the same as for the reactor vessel. If one loop is shut down, th e idle loop can be kept hot by leaving the discharge and suction gate valve open; this permits the active jet pump head to cause reverse flow in the idle loop.

Because the removal of the reactor recirculation gate valve internals would require unloading the core, or the use of alternate isolation devices such as the RR Suction plug or Jet Pump plugs, the objective of valv e trim design is to minimize the need for maintenance of the valve internals. The valves are provided with high-quality backseats that permit renewal of stem packing while the system is full of water.

When the pump is operating at 25% speed (15 Hz), the head provided by the elevation of the reactor water level above the recirculation pump and the jet pumps is sufficient to provide the required NPSH. When the pump is operating at 100%

speed (60 Hz), most of the NPSH is supplied by the subcooling provided by LSCS-UFSAR G.2.1-3 REV. 14, APRIL 2002 feedwater flow. An accurate temperature element is provided in the recirculation line to measure recirculation loop temperature. A pressure transmitter measuring vessel pressure in conjunction with a function generator is provided to measure vessel water saturation temperature. The difference between these two readings is a direct measurement of the subcooling. If the subcooling falls below 10.1° F, the 60-Hz power supply is tripped to the 15-Hz supply to prevent cavitation of the recirculation pump and jet pump.

The transition of 100% speed is made after thermal power is sufficient to provide enough feedwater subcooling to prevent flow control valve cavitation. At this point the flow control valves are moved to the minimum positions and the pumps brought to 100% speed. The 25% to 100% speed tr ansition is prevented by interlocks requiring sufficient thermal power and the control valve in minimum position.

During preparation for hydrostatic tests, the nuclear system temperature must be raised above the vessel nil ductility transition temperature limit. The vessel is heated by operating the residual heat removal pumps and by core decay heat.

When NPSH is adequate, th e recirculation pumps may be run at 100% speed to provide additional heat.

Each recirculation pump is equipped with mechanical shaft seal assemblies. Each of these assemblies consists of two seals built into a cartridge that can be readily replaced without removing the motor from the pump. Each individual seal in the cartridge is designed for pump design pressures so that any one seal can adequately limit leakage in the event that the other seal should fail. The pump shaft passes through a breakdown bushing to reduce leakage in the event of a gross failure of both shaft seals. The cavity temperature and pressure drop across each individual seal is monitored.

Each recirculation pump motor is a vertical solid-shaft, totally enclosed, air/water-cooled induction motor. The combined rotating inertias of the recirculation pump and motor determine the coastdown for both the loss-of-coolant accident and the turbine-generator trip transients. This inertia requirement is met without a flywheel.

The pump discharge flow control valve throttles the discharge flow of the pump. The recirculation loop flow rate can be changed rapidly within the expected flow range in response to rapid changes in system demand. The maximum required valve actuator stroking rate is 10% to 11% of full stroke per second.

A design objective for the recirculation sy stem equipment is to provide units that will not require removal from the system for rework or ov erhaul. Pump casings and valve bodies are designed for a 40-year life and are welded to the pipe. The pump drive motor, impeller, wear rings, and flow control valve internals are designed for as long a life as is practical. Pump mechanical seal parts and the valve LSCS-UFSAR G.2.1-4 REV. 14, APRIL 2002 packing are expected to have a life expectancy which affords convenient replacement during the refueling outages.

The recirculation system piping is of all-welded construction and is designed and constructed to meet the requirements of either the ASME Section III Code or the ANSI B31.7 Code.

G.2.1.2 Thermal, Hydraulic, and Physical Data

G.2.1.2.1 Recirculation Loop Piping G.2.1.2.1.1 Piping Description

a. Number of loops - 2

b. General arrangement - See Figures G.2.1-1 through G.2.1-5

c. Pump suction line nominal size - 24 in.

d. Pump discharge line nominal size - 24 in.

e. Ring header nominal size - 16 in.

f. Number of risers - 10

g. External riser nominal size 12 in.

h. Internal riser nominal size - 10 in.

G.2.1.2.1.2 Operating Conditions

a. Nominal loop velocity (at pump rated flow) - 41.5 ft/sec

LSCS-UFSAR G.2.1-5 REV. 14, APRIL 2002 G.2.1.2.2 Valves G.2.1.2.2.1 Flow Control Valves G.2.1.2.2.1.1 Valve Description

a. Nominal size - 24 in.

b. Type - Ball

c. Actuator - Hydraulic G.2.1.2.2.1.2 Valve Operating Conditions

a. Valve wide-open flow - 45,600 gpm

b. Valve wide-open pressure drop - 22.7 psi

c. Fluid specific volume - 0.0212 ft 3/lb d. Valve C v - 7770 e. Stroke rate (% of stroke) - 10

+/- 1% per second G.2.1.2.2.1.3 Definitions Closed Position - the fully closed position of the valve ball. This is a reference position for identifying valve ball angular position. Physically, the valve cannot reach this position with the actuator connected.

Minimum Actuator or Zero Position:

The actuator stop. This is the point where the valve pressure drop is largest. At this position, a flow path exists between the valve body and valve seat. This is defined as 0% position.

Minimum Position:

The position of the flow control valve to which the valve must be closed in order for the reactor recirculation pump start/upshift to be permitted to occur. The nominal value of the minimum position is 20%, relati ve to the zero and 100% valve positions.

Wide Open Position - the actuator stop. Th is is the point where valve pressure drop is smallest, and is defi ned as 100% position.

LSCS-UFSAR G.2.1-6 REV. 14, APRIL 2002 Stroke - the movement between zero and wide open positions of the actuator.

G.2.1.2.2.2 Isolation Valves G.2.1.2.2.2.1 Suction Line Isolation Valve

a. Nominal size - 24 in.

b. Body type - Gate

c. Actuator - Motor operator

d. Valve wide-open loss coefficient (max) (f L/D) - 0.25

e. Stroke time - 2 min

+/- 10% f. Closure pressure differential - 50 psid G.2.1.2.2.2.2 Discharg e Line Isolation Valve

a. Nominal size - 24 in.

b. Body type - Gate

c. Actuator - Motor operator

d. Valve wide-open loss coefficient (max) (f L/D) - 0.25

e. Stroke time - 12"/min +

10% f. Closure pressure differential - 400 psid G.2.1.2.3 Recirculation Pump G.2.1.2.3.1 Pump Description

a. Rated flow - 47,200 gpm; 17.85 x 10 6 lb/hr b. Total developed head at rated flow - 805 ft

c. Nominal speed - 1,780 rpm

d. Specific speed - 2,559

LSCS-UFSAR G.2.1-7 REV. 14, APRIL 2002 e. Nominal low speed - 427 % 18 rpm

f. Type - Vertical centrifugal single stage

g. Seals - Mechanical shaft

G.2.1.2.3.2 Operating Conditions

a. Water temperature - 534° F

b. Pump suction static pressure - 1,023 psia

c. Suction velocity at rated flow - 41.5 ft/sec

G.2.1.2.3.3 Performance at Normal Speed

a. Continuous flow capability (min) 115% of rated

b. Head-capacity characteristic type - See Figure G.2.1.-6

c. Minimum pump efficiency (rated flow) - 87%

d. Pump brake horsepower (rated flow) - 8,340 hp

e. Maximum pump brake horsepower - 8,900 hp

f. NPSH required at rated flow (see Figure G.2.1-6) - 115 ft

G.2.1.2.3.4 Performance at Low Speed

a. Low speed rated flow - 11,330 gpm

b. Head at rated flow - 46 ft

c. Minimum pump efficiency (at rated flow) - 67%

d. Maximum pump brake horsepower -hot - 170 hp

e. Maximum pump brake horsepower -cold - 226 hp

f. Available NPSH (minimum) - 69 ft

LSCS-UFSAR G.2.1-8 REV. 14, APRIL 2002 G.2.1.2.4 Recirculation Pump Motor The recirculation pump decoupler has been eliminated from the design. (Note: See the GE report "Analysis of Recirculation Pump Motor Accident Conditions," January 14, 1977.)

G.2.1.2.4.1 Motor Description

a. Type - Totally enclosed air/water cooled

b. Minimum nameplate rating - 8,900 hp

c. Normal speed - 1,780 rpm

d. Rotational inertia - 21,500 lb-ft 2 G.2.1.2.4.2 Operating Conditions

a. Output range at norm al speed - 5,310-8,189 hp

b. Output range at low speed - 101-161 hp

c. Starting power range, hot/cold - 96/135 hp G.2.1.2.4.3 Required Performance

a. Minimum efficiency at rated power - 94%

G.2.1.2.5 Jet Pumps G.2.1.2.5.1 Description

a. Number of jet pumps - 20

b. Type of pump - Five-jet nozzle

c. 180° bend internal diameter - 7.0 in.

d. Nozzle internal diameter at tip - 1.30 in.

e. Throat internal diameter - 6.40 in.

f. Discharge diameter - 19.0 in.

LSCS-UFSAR G.2.1-9 REV. 13 G.2.1.2.5.2 Design Operating Conditions

a. Driving flow per jet pump - 1.628 x 10 6 lb/hr; 4,307 gpm

b. M-ratio (suction flow/drive flow) - 2.33

c. Universal efficiency - 41.25%

d. Total developed head - 88.2 ft

e. Nozzle velocity - 208 ft/sec

f. Throat velocity - 143 ft/sec

g. Diffuser exit velocity - 16.2 ft/sec G.2.1.2.5.3 Performance

a. Peak efficiency (design) - 43.36%

G.2.1.2.6 Equipment Interlocks G.2.1.2.6.1 Pump Cavitation Protection

a. Function

To prevent cavitation damage to either the recirculation pumps or the jet pumps, the system is provided with an interlock to trip the pump motor to low speed if the temper ature difference between the vessel water saturation temperature and the recirculation pump inlet temperature is less than the setpoint value. A high accuracy RTD is provided to measure recirculation pump inlet temperature. A pressure transmitter measuring vessel pressure in conjunction with a function

generator is provided to measure ve ssel water saturation temperature.

b. Setpoint adjustment range: 0°-100° F

c. Setpoint: 10.1° F G.2.1.2.6.2 Flow Control Valve Cavitation Protection

a. Function

LSCS-UFSAR G.2.1-10 REV. 14, APRIL 2002 1. To prevent flow control valve cavitation, the system is provided with interlocks which prevent oper ation of the plant at conditions where cavitation will occur.

2. In addition to the pump cavitation interlock, an interlock is provided to trip the pump motors to low speed wh en the feedwater flow rate is less than the specified setpoint value.

3. An interlock is provided to prevent increasing the pump motor speed above its low speed unless th e flow control valve is in the minimum position and the feedwater flow rate is above the setpoint value.

b. Feedwater setpoint adjustment range: 1.4-5.0 x 10 6 lb/hr c. Feedwater setpoint: 2.83 x 10 6 lb/hr G.2.1.2.7 Recirculation Pump Flow Measurement G.2.1.2.7.1 Description

a. Flow element type - Elbow taps

b. Rated flow type - 47,200 gpm

c. Flow element location - Pump suction line

G.2.1.2.7.2 Performance

a. Range 115% rated pump flow

b. Flow element accuracy (% rated pressure drop) -

+/-5%

LSCS-UFSAR G.2.2-1 REV. 13 G.2.2 Operating Restrictions G.2.2.1 Cavitation G.2.2.1.1 Background

G.2.2.1.1.1 Cavitation Cavitation is the term used to describe the formation and subsequent collapse of vapor bubbles in a fluid stre am. Vapor bubbles can be formed in the stream when the local static pressure becomes less than the vapor pressure of the fluid. These bubbles then collapse when the local static pressure increases.

G.2.2.1.1.2 Net Positive Suction Head (NPSH)

NPSH is a term initially used by pump manufacturers to describe the minimum static to saturation vapor pressure differenc e that is needed at the pump suction to avoid cavitation in the pump. This term has been generalized to define the available pressure difference in a fluid stream. Consequently, NPSH is a measure of the pressure margin available before vapor bubbles start to form in the fluid.

G.2.2.1.1.3 Cavitation Damage Cavitation damage is caused by the collapse of the bubbles. It will occur only when the static pressure increases downstream of the point where these bubbles form. While cavitation is usually referenced to the component that creates the vapor bubbles this is generally not where the bubbles collapse and the consequent damage

occurs. General practice is to avoid cavitation in hydraulic systems. While the mechanism is not well understood, industri al data available indicate that cavitation can result in severe material damage if it continues over a prolonged period of time. However, little is known about the damage potential versus bubble velocity, rate of static pressure increase, surface material pr operties, and density of the fluid.

The largest body of knowledge comes from the centrifugal-axial pump impeller and ship propeller design areas where cavitation has caused material pitting and erosion, leading in severe cases to failure of the impeller or propeller. The damage is attributed to rapid collapse of the vapor bubbles on or very near the surface of the impeller. Such bubble collapse shocks the surface, since it must absorb the energy of the fluid that is accelerated to close the vapor bubble void. The potential for severe shocks decreases as the distance of the bubble from the surface increases. The distance the bubble must be from the surface in relation to its size in order to avoid these shocks is essentially unknown. Consequently, little is known about the need to prevent free-stream cavitation relatively far from solid boundaries.

LSCS-UFSAR G.2.2-2 REV. 13 In cases of severe cavitation, abnormal vibration of the associated component could occur. Whether this vibration is excessive depends on the cavitation severity and the rigidity of the component.

G.2.2.1.2 Recirculation System Cavitation

G.2.2.1.2.1 Recirculation Pump Low NPSH would produce vapor bubbles near the center of the impeller. These would collapse as they move toward the impeller exit. Consequently, cavitation affects the impeller surfaces and not the bowl or discharge region.

G.2.2.1.2.2 Flow Control Valve Low NPSH would produce vapor bubbles in the fluid stream before it reaches the valve ball due to the decrease in static pressure. These bubbles would collapse in the pressure recovery areas at the ball and valve exits affecting material in these areas. G.2.2.1.2.3 Jet Pump Nozzle Vapor bubbles could form in the nozzle as the pressure decreases and velocity increases toward the nozzle exit. These would collapse in the pressure recovery area in the jet throat, affecting material in the throat area.

G.2.2.1.2.4 Jet Pump Suction

Vapor bubbles could form in the suction flow as pressure decreases and velocity increases from the downcomer annulus to th e throat. These would also collapse in the throat pressure recovery area.

G.2.2.1.2.5 Cavitation Coefficients The recirculation pump, jet pump, and flow control valve are tested to determine their cavitation coefficients so that prolon ged operation in cavitating regimes can be avoided. G.2.2.1.2.6 NPSH System NPSH depends on reactor water level and water subcooling below vessel dome saturation temperature. During normal operation, feedwater flow provides the most subcooling by mixing with the steam separator return flow making the downcomer water temperature less than saturation temperature.

LSCS-UFSAR G.2.2-3 REV. 15, APRIL 2004 G.2.2.1.3 Operating Restrictions Operation in cavitating conditions is prevented in most cases by automatic interlocks and in others by operating procedures.

G.2.2.1.3.1 Automatic Interlocks As mentioned in Subsection G.2.1.2.6, cavi tation interlocks are based on feedwater flow and the water temperature difference between the suction line temperature and the vessel water saturation temperature. These interlocks activate 15 seconds after the low feedwater flow condition is indicated, or 15 minutes after a low temperature difference is indicated, in order to prevent spurious trips from temporary changes in temperature or feedwater flow. These interlocks trip the pumps from 100% to 25% speed. Water level interlocks sequentially trip the pumps from 100% to 25% speed and th en from 25% speed to zero if level drops to the point where a loss-of-coolant accident or anticipated transient without scram is indicated. (These interlocks are also discussed in Section G.5.)

A bypass function is provided on the Operator Station to bypass the cavitation interlocks. Use of these bypasses are administratively controlled.

G.2.2.1.3.2 Operating Procedures G.2.2.1.3.2.1 One Recirculation Pump Operation The automatic interlocks do not prevent jet pump cavitation in this operating mode. Reactor operation is limited to the areas above the jet pump nozzle cavitation line "N" of Figure G.2.5-2. The flow control valv e cavitation interlock remains functional in this operating mode.

G.2.2.1.3.2.2 Residual Heat Removal System Operation The RHR system can use the recirculation loop jet pumps to provide circulation through the reactor core. Operating rest rictions limit RHR operation to regions where jet pump cavitation does not occur. NPSH requirements for RHR operation were established during preoperational testing.

G.2.2.1.3.2.3 Vessel Hydrostatic Testing In preoperational and startup modes of oper ation, recirculation pump operation at 100% speed is needed to heat the vessel wa ter up to the temperature required to perform hydrostatic tests. A bypass function is provided to disable the cavitation interlocks. Use of this function and system operation with the cavitation interlock disabled are governed by operating proced ures. These procedures were determined during preoperational testing.

LSCS-UFSAR G.2.2-4 REV. 13 G.2.2.2 Loop Flow Balance The design-basis loss-of-coolant accident analyses assume that the total flow through one bank of jet pumps is close to the total flow of the other bank. If it is assumed that one bank is at a higher flow than the other and a design-basis pipe rupture occurs in the high-flow loop, calcul ated peak cladding temperatures higher than those calculated in the DBA analysis could result, since the low flow loop could not provide as much flow during the first few seconds of the transient.

Consequently, loop flow imbalance restrictions are placed on the system.

G.2.2.3 Thermal Shock When the recirculation system is inactive and thermal power is too low to support much natural circulation flow, stratification of water in the recirculation loops or vessel bottom head can occur. If core flow is increased in such a condition, rapid replacement of cold with hot water shocks the adjacent components increasing their usage. Although the reactivity transient on the core is not significant, the increased usage is not acceptable. Consequently, rest rictions are placed on allowable inactive loop suction temperature before startup and on core flow incr eases based on bottom head drain line to vessel saturation temperature.

G.2.2.4 Restart of One Recirculation Pump In order to maintain plant availability, it is necessary to follow specific procedures for restart of one pump to return to two-pump operation. Otherwise, a scram will result. Specific procedures to avoid scram were determined during startup testing and subsequent analyses. Th e procedures include reducing active pump flow to at least 50%, reducing inactive flow control valve position to restart permissive level, and verifying that delta temperature between coolant in the idle loop and in the reactor pressure vessel is less than or equal to 50

° F, before starting the inactive pump. G.2.2.5 25% to 100% Speed Transfer Transfer from low to high speed also causes an increase in reactivity. If this is done at high thermal power, a reactor scram will result. The specific thermal power range where transfer can be made without scram is determined based on analysis of startup test results and subsequent analyses and is included in plant operating procedures.

LSCS-UFSAR G.2.3-1 REV. 13 G.2.3 Operation and Performance G.2.3.1 General Description The reactor recirculation system pumps reactor coolant through the core. This is accomplished by two recirculation loops external to the reactor vessel but inside the

primary containment. Each external loop contains one high-capacity motor-driven recirculation pump, a flow control valve, and two motor-operated gate valves for suction shutoff and discharge shutoff purpos es. Each pump suction line contains a flow-measuring system.

The internal portion of the loop consists of the jet pumps, whic h contain no moving parts. The jet pumps provide a continuous internal circulation path for the major portion of the core coolant flow. The je t pumps are located in the annular region between the core shroud and the vessel inner walls. Any recirculation line break would still allow core flooding to approxim ately two-thirds of the core height, the level of the inlet of the jet pumps.

In the jet pump nozzle, the high static head developed by the centrifugal pumps is converted to a high-velocity jet at a low static pressure. The low pressure at the nozzle discharge draws the surrounding fluid into the jet pump throat, where it is mixed with the driving flow and then passes into the diffuser. In the diffuser, the velocity is reduced and the flow leaves th e jet pumps at a pressure somewhat higher than the pressure in the downcomer.

The main resistances against which the jet pumps work are the core pressure drop and the steam separator pressure drop.

There is a small, beneficial natural circulation driving head arising from the density difference between the two-phase mixture in the core and the subcooled liquid in the downcomer. Core flow is very nearly a linear function of the jet pump driving flow, except at low flow rates where natural circulation becomes significant.

Jet pump driving flow rate and thus core flow rate is varied by means of control valves located in each loop immediately downstream of the centrifugal pumps.

To avoid system cavitation during startu p, hot standby and low-power operation, the recirculation system is provided with the capability of operating the centrifugal pumps at 25% of rated speed. Low-speed operation is accomplished by supplying power at a fixed low frequency to the pump-motor from a motor-generator set provided for this purpose.

LSCS-UFSAR G.2.3-2 REV. 13 G.2.3.2 Normal Operation Normal plant operation presumes that no abnormal condition exists due to such events as scram, operator error, or turbine trip. In this context, an operator error is considered an action that is contrary to plant operating procedures.

Monitoring of recirculation and jet pump flow distributions is performed periodically to verify expected performance. Routine monitoring for potential jet pump nozzle plugging was implemented after Unit 2 Cycle 2 because a 2.7 inch diameter spacer was separated from the 2B discharge valve, 2B33-F067B. The potential operational and safety effects of the disk were evaluated and approved.

No measurable effect was produced.

G.2.3.2.1 Start The system is started after the core is loaded, the vessel internals installed, the head is in place and the water level is at least at the normal level. Residual heat removal operation through the recirculation system is terminated and the system valves are oriented to their startup positions as follows:

a. initial conditions (recirculation system):

1. discharge valve closed (preve nts RHR flow backwards through the pump); 2. suction valve open (allows cleanup suction from the recirculation loops); 3. flow control valve in minimum position; and

4. pump off.

b. startup procedures (Figure G.2.3-1):

1. terminate RHR operation through the recirculation loop discharge lines; 2. open discharge gate valve;

3. start pump (the automatic sequencing enables the 100% speed power supply to start the pumps, closes the LFMG feeder breaker to bring it up to speed, trip s the 100% speed supply when 90 to 100% speed is reached and closes the LFMG output breaker when the pump has coasted down to 25% speed); and

LSCS-UFSAR G.2.3-3 REV. 13 4. open FCV to the wide open position (100% open).

G.2.3.2.2 Heatup and Pressurization During vessel pressurization and heatup to rated conditions, the system is run at 25% speed with the FCV wide open (100% open).

G.2.3.2.3 Low Thermal Power Power ascension up to 30% to 40% power is accomplished by pulling control rods. The recirculation system operating mode is 25% speed with the flow control valve wide open.

G.2.3.2.4 High Thermal Power The system is transferred to 100% speed after the flow control valve cavitation interlock is cleared (approximately 25% th ermal power/20% feedwater flow). This allows further power ascension by core flow control as well as by control rod movement.

a. transfer procedure (power ascension):

1. Close the FCV to the minimum position.

2. Transfer to 100% speed - FCV minimum position (the automatic sequencing opens the LFMG generator breaker, opens the LFMG feeder breaker and closes the 100% speed breaker).

b. transfer sequence (normal shutdown):

1. Close FCV's to minimum position (100% speed).

2. Transfer to 25% speed - FCV minimum position (the automatic sequencing opens the 100% speed br eaker, closes the LFMG feeder breaker and closes the LFMG generator breaker when the pump has coasted down to 25% speed).

3. Open FCV's to wide open position.

G.2.3.2.5 Natural Circulation

Power ascension without forced circulation creates the possibility of shocks to vessel components due to vessel fluid stratification. Although the piping systems are designed to minimize this effect, plant heatup must be more gradual than with LSCS-UFSAR G.2.3-4 REV. 13 forced circulation due to the need to pr event buildup of stratified layers. Consequently, natural circulation is considered an abnormal startup procedure.

G.2.3.2.6 Thermal Power -

Core Flow Operating Map A boiling water reactor must operate with certain restrictions because of pump net positive suction head (NPSH), overall plant control characteristics, core thermal power limits, etc. A typical power-flow map for the power range of operation is shown in Figure G.2.3-2. The nuclear system equipment, nuclear instrumentation, and the reactor protection system, in conjunction with operating procedures, maintain operations within the area of this map for normal operating conditions.

The Core Operating Limits Report (COLR) co ntains cycle specific transient analyses that defines the operational boundaries of interest on the power to flow map. Points along the APRM rod block line and the increased core flow line are analyzed each cycle to determine appropriate thermal limit values. All other information on the power to flow map is considered reference only. A description of the power to flow map is as follows:

a. Natural Circulation Line (A) -

The Operating state of the reactor moves along this line for normal control rod withdrawal sequence in the absence of recirculation pump operation.

b. 100% FCL - This line passes through 100% power at 100% flow. The operation state for the reactor follows this line for recirculation flow changes with a fixed control rod pattern. The line is based on a constant xenon concentration at rated power and flow.

c. Cavitation Protection Lines - 20% FW Flow for Flow control valve cavitation (reference 4) and 10.1 F for jet pump nozzle cavitation protection (reference 5).

d. APRM Rod Block Line - defined in Technical Specifications.

e. 105% flow line - defines the in creased core flow (ICF) region.

f. REG A/B - Interim Corrective Action (ICA) regions for instability concerns. High power and low flow conditions are known areas where instabilities may occur.

G.2.3.2.7 Mechanical and Control Features

The following limits and design features ar e employed to maintain power conditions at the required values sh own in Figures G.2.3-2:

LSCS-UFSAR G.2.3-5 REV. 15, APRIL 2004 a. Minimum Power Limit at Intermediate and High Core Flows - To prevent cavitation in the recircu lation pump, jet pump, and flow control valves, the recirculation system is provided with an interlock to trip off the 60-Hz source and close in the 15-Hz power source if the difference between vessel water saturation temperature and recirculation pump inlet temperature is less than a preset value.

b. Minimum Power Limit at Low Core Flow - During lower power, low-loop flow operations, the temperat ure differential interlock does not provide cavitation protection to the flow control valves. Therefore, the system is also provided with an interlock to trip off the 60-Hz power source and close in the 15-Hz powe r source when the feedwater flow falls below a preset level.

Discussion When reactor power is greater than appr oximately 25% of rated (20% feedwater flow), the low feedwater flow interlock is cleared, and the main recirculation pumps can be switched to the 100% speed power sou rce. The flow control valve is closed to the minimum position before the speed change to prevent large increases in core power and a potential flux scram. The system is then brought to the desired power flow level within the normal operating area of the map by opening the flow control valves and withdrawing control rods.

Control rod withdrawal with constant flow control valve position will result in power-flow changes along lines of constant C v (constant position). Flow control valve movement with constant control rod position will result in power-flow changes along, or nearly parallel to, the rated flow control line.

Large negative operating reactivity coefficients, inherent in the boiling water reactor, provide the following important advantages:

a. good load-following with well-damped behavior and little undershoot or overshoot in the heat transfer response, b. load-following the recirculation flow control, and

c. strong damping of spatial power disturbances.

LSCS-UFSAR G.2.3-6 REV. 13 Because of the negative void coefficient, load-following is accomplished by varying the reactor recirculation flow. To increase reactor power, the recirculation flow rate is increased. The increased core flow sweep s some of the voids from the moderator and causes an increase in core reactivity. As reactor power increases, more steam is formed and the reactor stabilizes at a new power level, with the transient excess reactivity balanced by the new void formation. No control rods are moved to accomplish this change of power level. Conversely, when a power reduction is required, the recirculation flow rate is reduced. The resultant formation of more voids in the moderator automatically decreases the reactor power level to that commensurate with the new recirculation flow rate. Again, no control rods are moved to accomplish the power reduction.

Varying the recirculation flow rate (flow control) is more advantageous, relative to load-following, than using control rod positioning. Flow variations perturb the reactor uniformly in the horizontal planes ensure a flatter power distribution and reduced transient allowances. As flow is varied, the power and void distributions remain approximately constant at the steady-state end points for a wide range of flow variations. After adjusting the power distribution by positioning the control rods at a reduced power and flow, the operator can then bring the reactor to rated conditions by increasing flow, with the a ssurance that the power distribution will remain approximately constant.

G.2.3.3 Abnormal Operation Abnormal operation is defined as any operation other than normal routine and includes such events as operator error, plant operational transients such as MSIV closure and turbine trip, or an accident condition, such as a recirculation line double-ended pipe break.

The recirculation system is designed to achieve the following two major objectives for abnormal operation:

a. Plant Safety - those system functions required to mitigate an abnormal operational event, that is function s for which credit is taken in the event analysis; and

b. Power Generation - those functions necessary to maximize power generation capability and ensure longevity of equipment. These functions prevent steady-state operat ion of system equipment in modes where damage can occur and preven t unnecessary reactor protection system (RPS) activation. Equipment damage would make it necessary to inspect or replace equipment during a schedule outage thereby increasing plant unavailability. Unnecessary scrams also increase plant unavailability.

LSCS-UFSAR G.2.3-7 REV. 15, APRIL 2004 G.2.3.3.1 Safety G.2.3.3.1.1 Moderate and Infrequent Events Trip of the recirculation pumps is needed to mitigate the effects of turbine or generator trip events.

G.2.3.3.1.2 Accident Events Trip of the recirculation pumps is needed to mitigate the vessel overpressure transient and reduce core power level for the ATWS event.

G.2.3.3.2 Power Generation G.2.3.3.2.1 Cavitation Interlocks Cavitation interlocks are installed to provide protection against:

a. operator errors, and

b. rapid transients such as scram where NPSH conditions deteriorate rapidly and the operator is not expected to respond in time to prevent long-term operation in the cavitation region.

G.2.3.3.2.2 Recirculation FCV Runback

A recirculation FCV runback occurs if th e total steam flow exceeds the feedwater flow capacity of the feedwater pumps running and reactor water level is less than L4. A 5% conservative bias is added to total steam flow, to keep total steam flow well within the capacity of the feedwater system. The reduction in recirculation flow helps avoid a low level reactor scram by reducing the steaming rate. The recirculation FCVs will runback as long as the condition is valid.

The recirculation FCV runback limits the valve position demand and the flow control A/B trouble alarm is annunciated in the 1(2)H13-P602 panel. When the FCV runback occurs, an automatic transfer to manual mode of the FCV's is performed.

G.2.3.3.2.3 Flow Control Valve Minimum Position Interlock The interlock is installed to prevent syst em startup or transfer from 25% to 100% speed unless the valve is in the minimum position. The objective is to prevent LSCS-UFSAR G.2.3-7a REV. 15, APRIL 2004 scrams due to a rapid flow increase resulting from an operator failure to close the FCV prior to the start of speed transfer.

G.2.3.3.2.4 LFMG Output Breaker Control The LFMG set output breaker closing logic includes interlock which prevent breaker closure until the 100%

power supply is tripped and residual voltages in the motor have decayed. These interlocks prevent possible damage to the generator which could result if the generator wa s connected to the pump while the 100%-speed power supply was still active.

LSCS-UFSAR G.2.3-8 REV. 13 G.2.3.3.2.5 High Loop Flow Mismatch Mismatch of the flow in one loop to the other loop of greater than 50% is known to create abnormal conditions in the jet pumps having the lower flow. Such operation is normally precluded by operating procedures. Following a trip of one of two operating pumps, the tripped pump coasts down to zero speed rather than 25%

speed to avoid automatic generation of a 100% - 25% mismatch condition.

G.2.3.3.2.6 Loop Suction and Discharge Isolation Valve Position The pump is tripped at less than 90% open position to prevent pump damage from no flow if isolation valve closure is in advertently initiated while the pumps are running.

G.2.3.3.2.7 Trip to 25% Speed The LFMG set is activated in most 100% sp eed trip cases to avoid scram recovery delays due to vessel bottom head fluid stratification. The specific cases are discussed in Table G.2.5-3.

LSCS-UFSAR G.2.4-1 REV. 15, APRIL 2004 This page intentionally left blank.

LSCS-UFSAR G.2.4-2 REV. 15, APRIL 2004 This page intentionally left blank.

LSCS-UFSAR G.2.4-3 REV. 15, APRIL 2004 This page intentionally left blank LSCS-UFSAR G.2.5-1 REV. 14, APRIL 2002 G.2.5 Thermal and Hydraulic Characteri stics Summary: Tables and Figures G.2.5.1 Table G.2.5-1 Syst em Design Data at Rated Thermal Power and Core Flow G.2.5.2 Figure G.2.5-1: Two Pump Thermal Power-Core Flow Operating Map

G.2.5.3 Figure G.2.5-2: One Pump Thermal Power-Core Flow Operating Map The LSCS 1999 Power Uprate Project did not re-analyze the reactor operating domain for single loop operation (SLO).

G.2.5.4 Table G.2.5-2 Pressures and Temperatures at Selected Conditions

Key to Table:

POINT = location in the recirculation loop corresponding to numbers on Figure G.2.1-1 W = flow rate in percent of given "Normal Flow" PDOME+ = static pressure above the given "Normal Dome Pressure" (psi)

TDOME+ = temperature above the given "Normal Dome Temperature" (° F).

G.2.5.5 Table G.2.5-3: Syst em Trip and Start Functions G.2.5.6 Figure G.2.5-3: Flow Control Valve Characteristics Curve G.2.5.7 Table G.2.5-4: Reactor Recirculation System Design Characteristics G.2.5.8 Table G.2.5-5: Plant Configuration Data

LSCS-UFSAR G.3.1-1 REV. 13 G.3 SYSTEM DESIGN EVALUATIONS This section discusses the factors that we re considered in determining the system safety requirements (Subsection G.3.1), the system design bases for safety and power generation (Subsection G.3.2), and the system safety evaluation (Subsection G.3.3).

G.3.1 Thermal-Hydraulic Conceptual Design Studies for BWR/5 and BWR/6 Conceptual thermal-hydraulic design stud ies were performed on the BWR/5 and /6 preliminary system design to determine the system design parameters that would have an effect on the transient and accident event evaluations and to determine system functions that would be necessary to mitigate transients and accidents. Subsequently the individual plant design requirements for the parameters determined from these studies were determ ined (Subsection G.3.2). Reference 1 of Section G.7 describes the analytical methods.

G.3.1.1 Transients G.3.1.1.1 Flow Increase Transient The worst core effects generated by a recirculation system fault came from rapid core flow increases. The following table summarizes the nominal BWR/5 and /6 effects if one flow control va lve is stroked to the wide open position at a constant rate. 30%/sec Valve Stroking Rate

Initial power 56% NBR Initial core flow 35% NBR Reactor scram time 1.15 sec Peak neutron flux 281.9% NBR

Fuel surface heat flux 79.2% NBR Peak heat flux time 2.15 sec Fuel center temperature increase 349° F CPR 0.19 MCPR >1.06

Initial licensing analysis results LSCS-UFSAR G.3.1-2 REV. 15, APRIL 2004 Figure G.3.1-1 is a plot of various plant parameters as a function of time and core flow for the 30% per second valve stroking rate. Consequently, it was important to limit valve opening rate in the system design.

G.3.1.1.2 Turbine Trip

Events during which a turbine trip or generator load rejection occurs result in a rapid pressurization condition which could lead to overpower conditions in the core. Trip of the recirculation pumps (RPT) and rapid coastdown during the first few seconds of this event mitigate the neutron flux excursion by creating additional core voids. In order to act fast enough for this negative reactivity to benefit the event, it is also necessary to sense the event at the turbin e stop and control valves just as the reactor protection system does for the scram function and to minimize the recirculation pump motor and pump inertia such that a rapid coastdown will occur when the pump is tripped.

G.3.1.1.3 Trip of One Feedwater Pump

Loss of a feedwater pump capability, as di scussed in Section G.

2.3.3.2.2, does not produce a severe thermal transient but does cause the vessel level to decrease to the scram setpoints if thermal po wer level is not decreased to the level that corresponds to the remaining feedwater pump capacity.

Runback of the recirculation flow is needed to accomplish this function since a reduction in core flow reduces core power level and steam flow and thus re quires less feedwater makeup.

G.3.1.1.4 Other Transients During all other events, action by the recirc ulation system is not needed to mitigate the transients. The equipment protective interlocks are presumed to operate when the setpoints are reached, but this action is not critical to the events. Failure of these interlocks in conjunction with the event would be compared to an accident since multiple failures are required. Such events are normally bounded by the consequences of a loss-of-coolant accident.

G.3.1.2 Accidents G.3.1.2.1 Design-Basis Loss-of-Coolant Accident (DB-LOCA)

For reasons discussed earlier in this UFSAR, the design-basis accident for the

emergency core cooling system (ECCS) is an instantaneous double-ended pipe rupture in one of the recirculation loops. During the first few seconds of this postulated event, core flow is maintained by coastdown of the recirculation pump in the unbroken loop. This core flow is important to the subsequent core thermal LSCS-UFSAR G.3.1-3 REV. 15, APRIL 2004 response; consequently, the recirculation system must be designed so that flow in the unbroken loop is maintained to the extent assumed in the accident analyses. Specifically, high pump inertia is desired, and the unbroken loop recirculation valves must not close at a ra te that will affect the coast down flow, as discussed in subsequent sections.

The ECCS design-basis permits no credit to be taken for a-c power other than that provided by the onsite emergency power supplies. Consequently, the recirculation pump in the unbroken loop is assumed to trip and coast down during the postulated DBA. In addition, the DBA analysis assumptions limit recirculation pump flow as a function of quality at the pump suction, and core flow is assumed to cease at the time of jet pump suction uncovery. (These assumptions and their conservatism are fully discussed in General Electric's ECCS analytical models (Initial Cycle discussed in Reference 2 of Section G.7; SAFER/GESTR analysis discussed in Reference 3 of Section G.7. These conservatism and assu mptions are addressed in Reference 6 for FANP fuel. The EXEM BWR analysis is discussed in References 7 and 10 of Section G.7.) in conformance with 10 CFR 50 Appendix K. With postulated pipe breaks smaller than the DB-LOCA, the rate of increase in pump suction enthalpy is slower and the time of jet pump suction uncovery is extended; thus there is more credit for coast down flow. Consequently, the general requirement would be to design the recirculation system so that the coast down flow rate is unaffected.

As a measure of the sensitivity of the BWR5/6 DBA-LOCA analysis results to the flow in the unbroken loop, an analysis was conducted to bound the effect. In this analysis, the flow in the unbroken loop was assumed to decrease instantaneously from 100% to 25% concurrent with the pipe break. Instantaneous FCV closure was not physically realistic, since there would be some finite time for the valve to close; however, in order to determine how sensitive PCT was to the flow rate, the assumption served to indicate sensitivity.

The increase in calculated peak cladding temperature with this assumption was on the order of 300° F. Although such an incre ase is certainly not sufficient to cause a safety concern, it indicates the importance of ensuring through design that the valves would not affect the coast down flow rate.

If a-c power is not lost, the recirculation pumps are expected to trip when the cavitation interlocks are exceeded (low water level, or steamline to recirculation suction temperature) or recirculation pump trip (RPT) is actuated.

G.3.1.2.2 Anticipated Transients Without Scram (ATWS)

Trip of the recirculation pumps from 100%

to zero speed would be needed to limit the initial vessel pressure rise and to reduce the steam flow rate to the suppression pool by reducing thermal power. Both recirculation pumps are tripped on high LSCS-UFSAR G.3.1-4 REV. 13 reactor dome pressure or on low-low reactor water level using one-out-of-two taken twice logic.

Both recirculation pumps are tripped on hi gh reactor dome pressure or on low-low water level using one-out-of-two taken twic e logic. The sensed variables originate at the level and pressure transmitters of the Alternate Rod Insertion (ARI) system.

The trip relays of ARI are common to the ATWS-RPT logic such that initiation of ARI from either or both divisions occurs simultaneously with actuation of ATWS-RPT in the corresponding division(s).

Although the ARI and ATW-RPT logics share the trip relays utilized for automatic actuation, reset of the ATWS-RPT logic is completely independent of the reset of ARI. Manual initiation of ARI is also completely independent of the ATWS-RPT logic and will not actuate ATWS-RPT.

Initiation of an ATWS-RPT in either division annunciates an alarm window in the Main Control Room. The ATWS-RPT signal is sealed-in until reset by deliberate operator action in the Main Control Room.

A spurious level or pressure signal in any one ARI/ATWS-RPT instrument channel will not trip either one or both recirculation pumps. A failure in any one transmitter, trip unit, or trip relay will not result in inadvertent actuation of the ATWS-RPT.

The ATWS-RPT logic is designed to be test ed during normal plant operation up to and including the relays that energize th e trip coils of the recirculation pump breakers. Test switches are installed for each ATWS-RPT division on Auxiliary Electrical Equipment Room (AEER) panels to avoid inadvertent actuation. The test position of each switch annunciates an al arm window in the Main Control Room. When either switch is in the test position, a trip signal will illuminate an amber light located just above the switch.

Each division of the ATWS-RPT logic is powered by the corresponding 125 VDC divisional bus.

Although the ATWS-RPT equipment does not perform any safety function, NRC Generic Letter 85-06, "Quality Assurance Gu idance for ATWS Equipment that is not Safety-Related", indicates special quality requirements must be applied. In accordance with this Generic Letter, the test and reset switch es of the ATWS-RPT logic were procured safety-related with seismic qualification.

LSCS-UFSAR G.3.2-1 REV. 13 G.3.2 Design Bases G.3.2.1 General Design Bases G.3.2.1.1 Safety Design Bases

The reactor recirculation system is designed to meet the following safety design bases: a. Provide adequate fuel barrier thermal margin during postulated transients. That is, all transients due to normal operation and to single operator error or equipment malfunction result in a minimum critical power ratio (MCPR) greate r than the safety limit MCPR.

b. Failure of piping integrity d oes not compromise the ability of the reactor vessel internals to provide a refloodable volume.

c. System pressure integrity duri ng abnormal and accident events to minimize consequences of such events.

G.3.2.1.2 Power Generation Design Bases The reactor recirculation system is designed to meet the following bases in order to maximize plant availability and capacity factor:

a. Ability to achieve rated core flow throughout the design life of the plant. b. Flexibility to adjust core power output over the range of 65% to 100% thermal power.

c. Minimize maintenance situations that require significant outage time such as those requiring core disassembly and fuel removal.

G.3.2.2 Safety Design Bases G.3.2.2.1 Normal Operation The following safety design requirements are met for normal (unfaulted) operation of the system such as flow control, system startup and power ascension:

a. Pressure integrity of all primary system components.

LSCS-UFSAR G.3.2-2 REV. 13 b. Normal flow control valve stroke ra te in the open direction is less than effective rate of 30%/sec such that transient MCPR is greater than the MCPR safety limit.

c. One pump restart to return to two pump operation transient MCPR is greater than the MCPR safety limit.

G.3.2.2.2 Transient Operation The following design requirements are met for all abnormal events, that is, those due to an expected operating transient such as turbine trip, any single equipment failure, or an operator error:

a. Pressure integrity of all primary system components.

b. Flow control valve stroke rate in the open direction is less than the effective rate of 30%/sec such that transient MCPR is greater than the MCPR safety limit.

c. One-pump restart to return to two-pump operation transient MCPR is greater than the MCPR safety limit.

d. Turbine trip:

1. Trip of the recirculation pumps occurs in less than or equal to 190 msec after initiation of turb ine stop or control valve fast closure. 2. Recirculation Pump-Motor combined inertia is minimal to ensure fast pump coastdown.

G.3.2.2.3 Accident Operation The following design requirements are met for accident events:

a. Pressure integrity of primary system components such that a loss-of-coolant accident does not result from the event.

b. LOCA:

1. Unbroken loop valves do not close such as to affect the coastdown flow rate of the recirculation pump.

2. Integrity of the jet pumps such that a refloodable volume is maintained up to the jet pump throat elevation.

LSCS-UFSAR G.3.2-3 REV. 13

3. Recirculation pump-motor combined inertia is maximized to ensure slow pump coastdown.

c. ATWS: Recirculation pump trip on high vessel pressure or low water level. d. SSE: Establish containment by providing for isolation of the flow control valve hydraulic lines.

G.3.2.3 Power Generation Design Bases G.3.2.3.1 General Requirements Power generation deals with the ability of the plant to stay on line at high capacity. Two general concerns that address this objective are scram avoidance and equipment damage.

Scram Avoidance

Reactor protection system activation makes the plant unavailable. It is desirable to minimize the number of times that a scram is needed by designing with large margins to the RPS setpoints.

Equipment Damage Damage to system equipment can call for replacement, repair, or inspection that increases outage time and decreases plant availability. It is desirable to avoid situations where damage can occur by including automatic features to protect the equipment.

G.3.2.3.2 Normal Operation Scram Avoidance FCV open/close rate is less than 11%/sec.

G.3.2.3.3 Transient Operation Scram Avoidance Provisions

a. One-pump trip - pump inertia is greater than 21,500 lb m -ft 2 to allow coastdown without an RPS trip.

LSCS-UFSAR G.3.2-4 REV. 15, APRIL 2004 b. Flow increase transients - positi on switch on the FCV prohibits speed increases unless the FCV is in the minimum position.

c. FCV runback given the loss of capability of one feedwater pump and vessel low level.

Equipment Damage Provisions

a. Cavitation interlocks for the recirculation pump, flow control valve and jet pumps: since cavitation produces material damage after long-term operation, and the damage potential decreases with an increase in water temperature, short periods of cavitation during a transient or accident are not a concern. However, long-term operation that might

occur due to delayed operator action to trip the pumps is of sufficient concern to call for inspections during the next refueling outage. Consequently, to avoid the need for such inspections, automatic interlocks are installed.

b. Suction/discharge valve closure interlocks to protect the recirculation pump. c. Electrical interlocks to prevent connection of the LFMG set while the 100%-speed power supply is still active.

d. Loop flow mismatch interlocks to prevent automatic generation of a high flow loop - low flow loop cond ition after a one-pump trip: this condition is known to produce acceptable but nevertheless abnormally high stresses in the low flow jet pumps and shall not be automatically generated by activating the LFMG set during a one-pump trip. Hence, it would require an operator error to produce this condition.

Other Activate LFMG in most 100%-speed trip case s to avoid scram recovery delays due to vessel bottom head fluid stratification.

LSCS-UFSAR G.3.3-1 REV. 14, APRIL 2002 G.3.3 Safety and Power Generation Evaluations G.3.3.1 Component Classification The classification of the recirculation syst em equipment is shown in Table G.3.3-1.

Active components are those components th at are required to actuate mechanically or electrically during the transients or accidents.

Passive components are those components that are not required to actuate mechanically or electrically duri ng the transients or accidents.

The active/passive designation is not applicable to those components that are not inherently capable of action, such as piping.

G.3.3.2 System Design G.3.3.2.1 Transient Performance Reactor recirculation system malfunctions that pose threats of damage to the fuel barrier are described and evaluated in Sectio n G.6. It is shown in Section G.6 that none of the malfunctions results in fuel damage. The recirculation system has sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients.

G.3.3.3 System Component Design G.3.3.3.1 General G.3.3.3.1.1 Pressure Integrity (External Loop Piping, Recirculation Pump Bowl, Block Valve Body, Flow Control Valve Body)

Design pressures for the reactor recirculation system are based on peak steam pressure in the reactor dome, appropriate pump head allowances, and the evaluation head above the lowest point in the recirculation loop. Piping and related equipment pressure parts are chosen in accordance with applicable codes. Use of the code design criteria assures that a system designed, built, and operated within design limits has an extremely low pro bability of failure ca used by any known failure mechanism.

G.3.3.3.1.2 Inspection and Testing

Quality control methods are used during fabrication and assembly of the reactor recirculation system to assure that design specifications are met. Inspection and testing are carried out as described in Section G.4. The reactor coolant system is thoroughly cleaned and flushed before initial fuel loading.

LSCS-UFSAR G.3.3-2 REV. 13 During the preoperational test program, the reactor recirculation system was hydrostatically tested at 125% reactor vess el design pressure. Preoperational tests on the reactor recirculation system also included checking operation of the pumps, flow control system (including valves), and gate valves.

During the startup test program, horizontal and vertical motions of the reactor recirculation system piping and equipment were observed; supports were adjusted, as necessary, to assure that components were free to move as designed. Nuclear and recirculation system responses to recirculation pump trips at rated temperature and pressure were evaluated during the startup tests, and plant power response to recirculation flow control was determined.

G.3.3.3.2 Jet Pumps The core flooding capability of a jet pump design plant is discussed in detail in the emergency core cooling systems document filed with the NRC as a General Electric topical report. The ability to reflood the BWR core to the top of the jet pump applies to all jet pump BWR's and does not depend on the plant size or product line.

G.3.3.3.3 Recirculation Pump and Motor In order to assure functional performance of the recirculation pump and motor, the following additional requirement is met:

The pump and motor bearings shall have su fficient dynamic load capability at rated operating conditions to withstand the safe shutdown earthquake and to be able to coast down to 40% of rated speed on loss of power.

G.3.3.3.3.1 Analytical Methods for Evaluation of Pump Speed and Bearing Integrity Tests and procedures used to evaluate critical speed problems in pumps and to assure the integrity of the bearings for th e transient conditions are briefly discussed in the following subsections.

G.3.3.3.3.1.1 Pump Shaft Critical Speed The first critical speed of the recirculation pump shaft has been calculated to be above 130% of the operating speed. The abse nce of shaft vibration has been verified by testing the pump under rated speed conditions in the supplier's test loop. The absence of vibration was further verified in the plant during preoperational testing.

G.3.3.3.3.1.2 Pump Bearing Integrity

LSCS-UFSAR G.3.3-3 REV. 14, APRIL 2002 Adequacy of the bearing design has been verified by full temperature and pressure tests in the supplier's test loop.

G.3.3.3.4 Recirculation System Piping G.3.3.3.4.1 Pipe Rupture Protection against dynamic effects of pipe rupture is described in Section 3.6 of the UFSAR. Protection has been provided for the postulated break locations and break types that are specified in Subsection 3.6.1.2.

G.3.3.3.5 Suction and Discharge Block Valves

G.3.3.3.5.1 Safety Requirement Valves shall not close such as to affect the coastdown flow rate of the pump.

G.3.3.3.5.2 Design Description The 24-inch motor-operated gate valves provide pump and flow control valve isolation during maintenance. The operators take 2 minutes to ei ther fully open or close the valve. The suction valve is capable of closing with up to a 50-psi differential, while the discharge valve can close with up to a 400-psi differential. Both valves are remote-manually operated.

G.3.3.3.5.3 Safety Evaluation

An FMEA performed on the block valves is presented in Attachment G.B.1. In addition, an analysis was made to determine the effect of block valve closure on recirculation pump coastdown. The analysis postulates that coincident with a recirculation pump trip, the block valves begin to close. It was concluded that any closure time greater than 1 minute will have no effect on coastdown times. The consequence of an inadvertent closure without a coincident pump trip is covered in the FMEA, Attachment G.B.1.

G.3.3.3.6 Flow Control Valve, Actuator and Hydraulic Power Unit G.3.3.3.6.1 Safety Requirements Worst single failure or operator error shall result in an effective stroke rate of less than 30% of stroke per second in the open direction.

G.3.3.3.6.2 Design Description G.3.3.3.6.2.1 Flow Control Valves

LSCS-UFSAR G.3.3-4 REV. 14, APRIL 2002 This is a ball valve with a linkage connected to the actuator shaft (see Subsection G.2.1.2.2).

G.3.3.3.6.2.2 Flow Control Valve Actuator

The actuator is a modified commercial hydraulic cylinder. Deviations from the standard unit include special seals, special rod bearings and the necessary interface dimensional modifications. The cylinde r is rated for operation at 3000 psig.

A linear velocity transformer is mounted on the actuator rod and provides the velocity feedback to the flow control sy stem. The position feedback signal is provided by a transducer (Unit 1 only) connected to the ball shaft of the valve.

Hydraulic sealing of the actuator rod is pr ovided by a primary seal which provides the full pressure containment and a secondary seal which prevents any leakage outside the unit. Between the seals is a cavity which has a drain line connected back to the supply reservoir. A flow switch mounted in this line will visually indicate and annunciate a high drain flow condition to the operator.

G.3.3.3.6.2.3 Hydraulic Power Unit Units are proven commercial devices rated for operation up to 3000 psig. Actual operating pressure is 1800 psig. Qualificat ion of the units required a hydrostatic pressure test at 3000 psig on the high pressure circuits.

The units are located outside the drywell. Only the hydraulic lines, the pilot-operated check valves, and the hydraulic actuator are located inside the drywell.

Refer to Subsection G.5.2.3.4.9 for a complete description of the hydraulic power unit. G.3.3.3.6.2.4 Hydraulic Line Design

All high-pressure skid piping is ASTM A106, Grade B, Schedule 80. Actual operating pressure is 1800 psig (units are rated by the manufacturer for service up to 3000 psi). The skid equipment and actu ator are interconnected using piping and flexible metallic hose. The piping is ASME SA312 Type 304, Schedule 80. The flexible hose is ASTM A240 Type 316L with ASTM A312 Type 304, Schedule 80 welding nipples.

G.3.3.3.6.2.4.1 Design Pressure Rating of Interconnecting Piping

LSCS-UFSAR G.3.3-5 REV. 14, APRIL 2002

a. Drain: 150 psig

b. Remainder: 3,000 psig

G.3.3.3.6.2.5 Hydraulic Line Isolation Valves Two solenoid operated automatic globe isolation valves are provided on each hydraulic line, inside and outside the containment, as close to the containment as practical. These valves form part of the containment isolation system for lines that penetrate the primary containment boundary and are neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere; as such they prevent or limit the escape of fission products from postulated accidents.

G.3.3.3.6.2.6 Hydraulic Fluid The hydraulic fluid used is Fyrquel EHC.

It has an autoignition temperature of 1150° F. When heated to 455° F, the vapor will flash with an ignition source. If heated to 700° F, burning of vapor will la st at least five seconds after ignition.

G.3.3.3.6.3 Safety Evaluation G.3.3.3.6.3.1 Flow Control Valve G.3.3.3.6.3.2 Flow Control Valve Se izure: Pressure Boundary Integrity The Flow Control Valve (FCV) with the enclosed topworks consists of the body, the bonnet, shaft/ball assembly, the upper cover, the packing cartridge, the yoke and the actuator with the hydraulic lines.

The forces required to actuate the valve vary with the position:

LSCS-UFSAR G.3.3-6 REV. 13 a. To open the valve, the extend force varies from ~19,000 lb to

~11,500 lb.

b. To close the valve, the force vari es from ~11,000 (retract) to ~8,000 lb (extend) due to the flow torque force.

Should the valve ball seize, the hydraulic actuator is capable of producing a maximum force of 50,000 lb. The most severely stressed part would be:

a. The link. At the maximum actuator force the stress would be about equal to the yield stress.

All other parts are stressed below their yield points, and no failure is therefore expected.

The result of the above failures would not affect the pressure integrity of the valve or allow the ball to brea k away from the shaft.

G.3.3.3.6.3.3 Hydraulic Actuator and Hydraulic Power Unit

G.3.3.3.6.3.3.1 Actuator Velocity Limits Actuator velocity is normally controlled by the electronic circuits to limit any sustained motion to 10%/sec. A "hard-ov er" servovalve malfunction and/or an electronic malfunction which applies a hard-over signal to the servovalve will result in higher velocities, referred to as "excess" velocity.

When this occurs, velocity is limited initially only by the restriction of the hydraulic circuit. Once the accumulator has discharged all the fluid stored in it, velocity is limited by the pump flow capacity. Due to high friction forces, the FCV will remain motionless unless driven by the actuator.

Accordingly, two velocities are of interest, the peak instantaneous velocity which occurs while the accumulator is dischargin g, and the average velocity (i.e., full stroke/time to travel full stroke).

Since the actuator incorporates unequal areas, velocities must be calculated for both direction of motion (all analysis assumes failure of all interlocks which might shut the unit down).

The worst single failure, servovalve hard-over, will result in the following actuator transients:

a. peak velocity during accumulator discharge;

LSCS-UFSAR G.3.3-7 REV. 13 b. peak extending velocity 62%/sec (extending is in the direction to open the FCV); and

c. peak retracting velocity 54%/sec (retracting is in the direction to close the FCV).

For the rod extending event, the accumulator discharge lasts for 0.14 seconds, while the retracting duration is 0.22 seconds.

Average velocity over full stroke (including accumulator discharge):

a. extending 18%/sec (open FCV)

b. retracting 25%/sec (close FCV)

G.3.3.3.6.3.3.2 Summar y of Failure Analysis

A failure mode and effects analysis (FMEA) for the flow control valve and the hydraulic power unit are given in Attachment G.B.2.

For the flow control valve, there are two major limiting single features. One is valve failure which is bounded by the LOCA event; the second is a momentary stroke rate that exceeds an effective 30%/sec average rate. This latter value is the average continuous rate used in the transient analysis such that the transient MCPR is greater than the safety limit MCPR. The single failure that results in the highest stroke rate was determined to be a hard-over servovalve. A conservative design analysis produces a stroke transient as shown in Figure G.3.3-1. This transient was analyzed and found to be le ss severe than the effective 30%/sec rate.

Multiple failures, such as high pump pressure, servovalve hard-over, or loss of pressure, can result in instantaneous velo cities exceeding 60%/sec. These events are considered to be in the same category as accident events.

G.3.3.3.7 Flow Control System G.3.3.3.7.1 Safety Requirements Worst single failure or operator error shall result in an effective stroke rate of less than 30% of stroke per second in the open direction.

The control system shall not close the flow control valve in the unbroken loop at a rate that will affect the LOCA recirculation pump coastdown.

LSCS-UFSAR G.3.3-8 REV. 15, APRIL 2004 G.3.3.3.7.2 Design Description During normal operation, the recirculati on flow control system provides the recirculation flow demand signal to the hydraulic power unit (HPU) servovalves.

The flow control system has been designed to limit the maximum demand signal to an actuator stroke rate of less than 11%/sec. This is to ensure that the reactor protection system will not initiate scram.

Various interlocks are installed to ensure that this limit is not exceeded.

The design is discussed in more detail in Subsection G.5.2.

G.3.3.3.7.3 Safety Evaluation G.3.3.3.7.3.1 Flow Control Valve Stroke Rate A failure modes and effects analysis (FMEA) for the flow control system is given in Attachment G.B.3.

Single failures can result in an actuator stroke rate exceeding 11%/sec, which may result in reactor protection system activation. However, there are no single failures of the flow control system which would result in an effective stroke rate exceeding 30%/sec. This value is the average continuous rate used in the transient analysis.

The worst single failure would be a single demand for a hard-over servovalve. A conservative design analysis produces a st roke transient as shown in Figure G.3.3-

1. This transient was analyzed and found to be less severe than the effective 30%/sec rate.

G.3.3.3.7.3.2 Loss-of-Coolant Accident For postulated pipe breaks in the recirculation system, only the response of the valve in the unbroken loop is of concern.

The event would have the following effects on the system:

LSCS-UFSAR G.3.3-9 REV. 15, APRIL 2004 The recirculation flow loops would not interact with one another. Thus, the unbroken loop would see no significant change.

The valve in the unbroken recirculation line will not move in response to a recirculation loop pipe break.

G.3.3.3.8 Recirculation Pump Trip (RPT) System

G.3.3.3.8.1 Safety Requirements The RPT system is required to trip both recirculation pumps from their normal power source within 190 msec after a turbine/generator trip or load rejection event occurs with reactor power level greater than or equal to 25% of rated core thermal power. This trip cannot be prevented by any single component failure in the system.

G.3.3.3.8.2 RPT Design The RPT logic is derived from relay's in the reactor protection system (RPS) which are activated by turbine stop valve limit switches and turbine control valve oil pressure switches. Contacts from these relays are arranged in a logic scheme so that deenergizing various combinations of the relays will energize the RPT trip coil in each of two main power source circu it breakers (see Figure G.3.3-2).

Each main circuit breaker has two trip coils. One trip coil is activated by an RPT signal only. The other trip coil is activate d by all other breaker trip functions. The dual trip coils serve to separate the breaker safety function from its non-safety functions.

LSCS-UFSAR G.3.3-10 REV. 13 The design of the sensor/logic/circuit breaker chain is such that the response time between initiation signal and suppression of circuit breaker arc is less than 190 msec.

G.3.3.3.8.3 Failure Mode and Effects Analysis (FMEA)

An FMEA was performed for each power supply, fuse, sensor/relay pair, test switch, diode, test lamp, indicating lamp and circuit breaker in the RPT logic circuit. The detailed FMEA sheets are presented in Attachment G.B.4. The results of this FMEA are as follows:

a. A fail-open or fail-closed single random failure of individual fuses, trip sensors or relays in Channel A1, A2, B1, or B2 will not prevent or initiate RPT.

b. A fail-open or fail-closed single random failure of power supplies, test switches, diodes or test lamps in trip channels A or B will not prevent or initiate RPT.

c. A fail-open in an individual circui t breaker status indicating lamp or a fail-closed of an individual circuit breaker will not prevent RPT when required.

d. A short circuit in an individual circuit breaker status indicating lamp or fail-open of an individual circuit breaker will initiate RPT in a single loop.

G.3.3.3.9 Low-Frequency Motor-Generator (LFMG) Set G.3.3.3.9.1 Special Design Requirements G.3.3.3.9.1.1 Safety The LFMG set is nonessential equipment an d has no safety f unction. However, design requirement is that the LFMG set cannot interfere with a recirculation system safety function (e.g., RPT).

G.3.3.3.9.1.2 Power Generation The basic design requirements for the LFMG set are power generation oriented, that is, the purpose of the LFMG set is to prevent cavitation of the flow control valve at low recirculation flows while maintaining enough flow to prevent reactor pressure vessel temperature stratification. Thus, most events that cause the recirculation pumps to trip from the main power source will also start the LFMG set and close it in at 25% pump speed to maintain circulation.

LSCS-UFSAR G.3.3-11 REV. 13 G.3.3.3.9.2 Design Parameters G.3.3.3.9.2.1 General A single-line diagram of the recirculation pump motor power sources is shown in Figure G.3.3-3. The pump motors can be supplied from a nonessential 60-Hz power source through two essential class 1E, Seismic Category I, Safety Class 3 circuit breakers or from the 15-Hz output of the nonessential LFMG set. Interlocks are provided to prevent powering of the pump motor from both power sources simultaneously.

G.3.3.3.9.2.2 Interlocks

During plant startup, if it is necessary to start a recirculation pump and reactor power is less than 25% of rated powe r/20% feedwater flow (FCV cavitation interlock), a start signal to the pump motor closes the main power source breaker, accelerating the pump from zero to 100% speed, starts the LFMG set, trips the main power breaker at 100% speed, and closes in the LFMG set generator breaker when the pump coasts down to 25% speed.

Once reactor power is greater than 25%, the pumps may be transferred to the main power source. The transfer ca n only be accomplished if th e flow control valve is at its minimum position. This prevents a large flow increase transient. The power transfer signal trips the LFMG set and, once the generator breaker is open, closes the main power source breaker which causes the pump to accelerate from 25% to 100% speed. The FCV can then be used to control flow.

During a plant shutdown, when reactor power level drops below 25%, with recirculation flow less than 60%, operating procedures and automatic interlocks require that the recirculation pumps be tripped from their main power source and powered from the LFMG set.

In addition, the LFMG set will automati cally start and close in at 25% speed following a recirculation pump trip due to RPT, normal trip of both recirculation pumps, low temperature differential between main steam and recirculation pump suction, or vessel low level.

The LFMG set will not automatically start and close in if the recirculation pump trip is due to suction or discharge valve less than 90% open, vessel high pressure, normal trip of one of two operating pumps, or if the LFMG set electrical protection logic is activated.

LSCS-UFSAR G.3.3-12 REV. 13 G.3.3.3.9.3 Failure Mode and Effects Analysis (FMEA)

An FMEA was performed on the LFMG compon ents and interlocks. The details of the component FMEA are presented in Subsection G.3.3.3.11. The details of the interlock FMEA are presented in Subsection G.3.3.3.10.

These FMEA's show that the LFMG set ca nnot interfere with the RPT safety function and that the system interlocks are sufficient to prevent FCV cavitation event with a single random failure (i.e., operator error or equipment failure).

G.3.3.3.10 Control Interlocks G.3.3.3.10.1 Power-Generation Design Requirements Operating conditions have been identified which could result in cavitation of the flow control valve, recirculation pump, and/or jet pumps. Op erating procedures require the plant operator to avoid these conditions. In the event that, due to operator error, these conditions are not avoided, automatic interlocks exist which will prevent operation at these conditions.

G.3.3.3.10.2 Design Description G.3.3.3.10.2.1 Reactor Low Powe r Level - LFMG Set Auto Sequence When reactor power level is not high enou gh (~25%) to provide sufficient subcooling to prevent cavitation of the flow control valve, a start of the recirculation pumps will result in closure of the main power source breaker to accelerate the pump from zero to 100% speed, start of the LFMG set, trip of the main power source breaker when the pump reaches 100% speed and closure of the LFMG output breaker when the pump reaches 25% speed. The start can only occur if the recirculation suction and discharge valves are open, the FCV is at minimum position, the M/A station is in manual, control power is available to the LFMG set breakers and the electrical protection relays on the LFMG set have not actuated.

G.3.3.3.10.2.2 Reactor Normal Power Level Low-Speed to High-Speed Transfer Once reactor power level is greater than 25%(20% feedwater flow

), the operator may transfer the recirculation pumps to the main power source from the LFMG set or, if the pumps are tripped, they may be starte d by the main power source and will not transfer back to the LFMG set. These sequences can only occur if the suction and discharge valves are fully open, the FCV is at minimum position, the M/A station is

in manual, pump speed is less than 20%, feedwater flow is normal, reactor water level is normal, vessel water saturation temperature to recirculation suction T is normal, and RPT has not occurred.

G.3.3.3.10.2.3 High-Speed to Lo w-Speed Auto Transfer Sequence

LSCS-UFSAR G.3.3-13 REV. 13 During plant shutdown, when reactor power level approaches 30% decreasing, the operator will transfer the recirculatio n pumps from high-speed to low-speed operation. Interlocks require this transfer to be performed on both pumps simultaneously, thus avoiding an undesirable loop flow mismatch. Placing the control switch for both pumps into the transfer position automatically starts both LFMG sets, trips the main power source breakers, and closes the LFMG generator breakers when the pumps reach 25% speed. High to low speed transfer will automatically occur on RPT, less than 20% feedwater flow, reactor water low level, and low T between vessel water saturation temperature to recirculation suction temperature. During high-to-low speed transfer, a tripped pump will not be energized by the LFMG set at 25% speed if the other pump has failed to trip.

G.3.3.3.10.3 Failure Mode and Effects Analysis (FMEA)

An FMEA was performed on the recirculation system control interlocks. The detailed FMEA sheets are presented in Attachment G.B.5. The results of this FMEA are as follows:

No single random failure (i.e. operator action or equipment failure) will result in sustained operation in a region which will produce cavitation of the flow control valve, jet pumps, and/or recirculation pumps.

G.3.3.3.11 Power Supplies G.3.3.3.11.1 Design Requirements The recirculation system power supplies are designed so that no single failure of a

power supply can prevent the action of th e RPT safety function, nor can the single failure of a power supply prevent the prot ection of equipment required under the power-generation design-basis.

G.3.3.3.11.2 Power Supply Arrangement The arrangement of the various power supplies needed to power the recirculation system is shown in Figure G.3.3-4. A single-line diagram of the auxiliary power system showing the recirculation pump motor and LFMG set power supplies, the 120-Vac instrument bus, and the 125-Vdc co ntrol power to the 4.16-kV and 6.9kV circuit breakers is presented on Figure G.3.3-5.

G.3.3.3.11.3 Failure Mode and Effects Analysis (FMEA)

An FMEA was performed for each power supply or circuit breaker which has an effect on the safety design and power generation design bases for the recirculation system. The detailed FMEA sheets are presented in Attachment G.B.6. The results of this FMEA are as follows:

LSCS-UFSAR G.3.3-14 REV. 13 No single random failure of a power supply or circuit breaker will prevent the action of the RPT safety function, nor ca n the single failure of a power supply or circuit breaker prevent the protection of equipment required under the power generation design-basis.

Multiple failures are required to parallel the LFMG set with the main power source. Even if these multiple failures occurred, the ability of RPT to function would not be impaired, although the LFMG set could be damaged.

LSCS-UFSAR G.4.1-1 REV. 13 G.4 TESTING This section describes the development tests, preoperational tests and startup tests conducted on individual recirculation sy stem components and the entire system.

G.4.1 Development Tests G.4.1.1 Five-Nozzle Jet Pumps G.4.1.1.1 Development of Jet Pumps G.4.1.1.1.1 Improvement and Evaluation of Performance

Initial jet pump feasibility tests were performed in 1962 when the first quarter-scale jet pump was designed, built, and tested at reactor operating conditions. This jet pump was designed using information available in literature and performed with a peak efficiency of 30% at reactor cond itions. Subsequent work was directed toward increasing jet pump efficiency by developing improved components using quarter-scale models. A streamlined nozzle and a streamlined suction inlet with low losses were developed. In addition, nozzle-to-suction inlet spacing and the mixing throat length were optimized for the geometry intended for reactor application.

This development program produced a jet pump that performed with 32% equivalent reactor peak efficiency. In mid-1967 a full-size prototype jet pump was tested at reactor conditions to demonstrate that it performed as well as or better than the quarter-scale model previously tested. The quarter-scale model was scaled linearly to full size. The full-size prototype performed with a peak application efficiency of 34%. Four production models for the Dresden-2 plant were tested extensively at reactor conditions to confirm that performance fully met requirements and to calibrate the diffuser pressure differential for use in measuring reactor core flow.

The prototype jet pump design was refined somewhat (partly to improve performance) when the Dresden-2 jet pump was designed. Thus, the design used for the Dresden-2 jet pump is considered as the reference, first-generation design for BWR's. It is used here for comparison with the improved, second-generation jet pumps. Jet pumps for reactors are designed with substantial margins to avoid cavitation during normal operation. To acquire a thorough understanding of jet pump

operation, tests were performed with the full-size prototype and Dresden-2 jet pumps at off-design conditions which produced cavitation. This was accomplished by reducing the subcooling in steps to zero and then injecting measured amounts of LSCS-UFSAR G.4.1-2 REV. 13 steam into the recirculation flow upstream of the jet pump suction inlet to simulate steam carryunder from steam separators at zero subcooling.

The tests show that jet pump efficiency is not significantly reduced as cavitation is increased. No abrupt change in performance is produced by cavitation, and suction flow rate may be increased when cavitation is present. Even when suction flow contained 0.06% steam by weight (in thermodynamic equilibrium), the dropoff in efficiency was less than 10% of its value at design conditions. No changes in the well-polished jet pump surfaces exposed to flow were observed even after many hours of cavitation testing.

G.4.1.1.1.2 First-Generation Jet Pumps

The main features of the Dresden-2 jet pump are: an inlet pipe equipped with a turning vane in the 180° bend; an efficien t single-jet converging drive flow nozzle with a 3.3-inch outlet tip diameter; an elliptical suction inlet blended into the mixing throat; a mixing throat 8.1-inch ID and 10.5 diameter long; a conical diffuser with an included angle of 8° and an area ratio of 6:1 which is blended into the throat and into the tailpipe; and a tailpipe approximately one diameter in length.

The 180° bend is equipped with a single, fu ll passage width turning vane (based on GE development tests) to minimize bend losses. The nozzle and suction inlet are finished from a single casting, including the three struts which maintain the drive nozzle and the suction inlet in alignment. These struts are streamlined to minimize losses at the suction inlet. This combination and arrangement of components results in a jet pump that performs well at reactor conditions.

When the performance of the Dresden-2 jet pump, as well as all other jet pumps, was applied to plant design, sufficient margin was allowed to provide for the possible influence in such factors as manufacturing and installation tolerances, flow distribution patterns in a reactor, and surface finish changes during reactor lifetime. Confirmatory data from Dresden-2 and other plants using similar jet pumps show that jet pump performance in plants meets or exc eeds the performance requirements established from production jet pump tests of the type discussed here.

G.4.1.1.1.3 Second-Generation Jet Pumps There was substantial incentive to impr ove the efficiency of BWR jet pumps, because as their efficiency is increased, plant heat rate is improved, and the cost and size of recirculation pumps and motors used to drive the jet pumps are reduced.

Therefore, there has been a continuing development effort to improve the efficiency of jet pumps. This effort has been directed toward optimizing jet pump components and introducing new components or g eometry configurations. One of the components investigated, for example, has been the multijet drive nozzle.

LSCS-UFSAR G.4.1-3 REV. 13 The multijet nozzle provides more surface area for mixing of the drive and suction flow streams than does a single-jet nozzle of the same cross-sectional area. The mixing of the two streams can therefore be accomplished in a shorter throat. Efficiency is improved because comparatively the friction losses in the throat are reduced to a greater extent than they are increased by using the multijet nozzle. The multijet nozzle also produces a wake-type velocity profile at the entrance of the

diffuser, which enhances diffuser performance.

Development tests with multijet nozzles showed that a five-jet nozzle, with the jet outlets located on a circle and spaced equally apart, produces the best performance. The circle on which the jet outlets are located is concentric with the jet pump throat, and its diameter has been optimized experimentally with respect to the throat diameter. The jets are designed to disc harge parallel to the axis of the mixing throat as a result of experimental investigation of discharge directions.

An outline (Figures G.4.1-1 and G.4.1-2) of the second-generation jet pump used in BWR/5 and BWR/6 plants consists of: an inlet bend with a turning vane used to convey the drive flow from the inlet pipe to the drive nozzle; a five-jet drive nozzle; five streamlined struts for maintaining the nozzle and throat in alignment; a streamlined suction inlet to minimize suction flow inlet losses; a constant-area

mixing throat in which the high-velocity drive stream mixes with the low-velocity suction stream; a 3° included-angle diverging throat where both mixing and pressure recovery take place; a 5° incl uded-angle diffuser in which the kinetic energy of the mixed streams is converted to pressure energy; and the tailpipe, which further converts kinetic energy to pressure and leads to the high-pressure discharge flow from the jet pumps to the core inlet plenum.

The performance of a full-size prototype of this jet pump was thoroughly examined at reactor design and off-design conditions in a reactor downco mer annulus mockup.

The performance obtained at design drive flow rate and subcooling gives a peak efficiency of 41.5%. Performance at other test conditions was essentially the same within the accuracy of the expe riment, which is estimated at

+/-2% (two-sigma confidence limits), i.e., (41.5%

+/- 0.9)% efficiency.

The effect of varying drive flow rate from 50% to 100% of design flow on performance was studied. The pump was operated near the peak efficiency value with about 25 Btu/lb subcooling. The result s indicate that no significant change in performance occurs.

Comparison shows that the second-generation jet pump performs with substantially higher efficiency than the first-generation jet pump.

LSCS-UFSAR G.4.1-4 REV. 13 G.4.1.1.2 Cavitation Tests G.4.1.1.2.1 Effects of Cavitation Visual examination of the jet pumps of three BWR/5's (Laguna Verde, ZPS-1, and LaSalle) and one BWR/6 (Perry) revealed no damage after being subjected to cavitation for 1 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in the test loop.

G.4.1.1.2.2 Cavitation Tests of the Laguna Verde Production Jet Pumps

a. Introduction Due to changes in operating and startup conditions for General Electric's BWR/5 and /6, a need aros e to obtain rather extensive jet pump cavitation data.

It was decided to integrate the cavitation tests with the performance and diffuser calibration tests of the Laguna Verde (201-BWR/5) production jet pump performed in March 1976 at GE's facilities located in the Pacific Gas and Electric (PG&

E) Power Plant at Moss Landing, California.

The cavitation tests were performed for a range of dr ive flow rates. Onset of cavitation was determined by monitoring changes in jet pump performance, vibration, and hydraulic noise level. The purpose of tests at reactor conditions was to verify that cavitation will not reduce jet pump performance or cause damaging vibration levels.

b. Summary and Conclusions A cavitation coefficient (Kc), defined as the ratio of net jet pump suction head to jet pump suction ve locity head, was used to correlate jet pump cavitation tests results. Cavitation inception was not characterized by a constant value of Kc. The cavitation coefficient depended on jet pump suction-to-drive-flow ratio and the type of cavitation.

Tests were performed at 540° F, wh ich is the rated water temperature. For these conditions, cavitation did not raise the acoustic noise level, and jet pump acceleration levels actually decreased. Jet pump performance started dropping at Kc value from ~0.62 to 0.85, with the average being Kc = ~0.75. These Kc values are essentially the same as those obtained during cavitation tests at 540° F with the ZPS-1 and the La Salle production jet pumps. This illustrates, as expected, that all BWR/5 jet pumps have the same cavitation characteristics.

LSCS-UFSAR G.4.1-5 REV. 13

c. Discussion The start of cavitation is defi ned through its effect on jet pump performance. It is well known that cavitation in the jet pump reduces jet pump efficiency. Therefore, the start of cavitation is defined as the point where jet pump efficiency starts dropping below its value measured far away from possible cavitation conditions (i.e., at high subcooling values).

It is realized that some cavitation may exist before any influence on performance starts showing up. Howe ver, "noise cavitation" is most likely limited to the shear layer between the drive and the suction

streams. Since it does not extend to the walls, it does not affect performance and should not produce any material damage on the jet pump mixer walls.

The start of jet pump cavitation can usually be conveniently correlated using a jet pump cavitation parameter, Kc defined as

Kc = NPSH / [V 2/2g] Cavitation or formation of vapor bubbles first occurs in the shear layer between the jet and suction flow. When these vapor bubbles collapse, acoustic noise is generated, resulting in "noise cavitation." If the pressure is reduced further, the vapor bubbles occupy more and more volume until the bubbles touch the jet pump wall. At this point, the suction flow will begin to be choked and the jet pump performance will decrease. Also, when the bubbles to uch the wall, more intense acoustic noise is generated due to the solid boundary and vapor-fluid interface interacting. Therefore, the drop in performance coincides with a large increase in acoustic noise.

G.4.1.1.3 Production Tests Hydraulic testing of production jet pump models for specific nuclear plants is performed at the Moss Landing Large Steam-Water Test Facility to establish efficiency and flow-measurement characteristics. Determination of efficiency performance is necessary, since the capability to produce rated core flow with minimized auxiliary power consumption is essential for achieving warranted power levels and heat rates. Flow measurements must be obtained because of the dependence of permissible core power levels upon direct measurement of core flow rate.

LSCS-UFSAR G.4.1-6 REV. 13 The method of determining core flow rate in a reactor is based upon measuring the differential pressure applied to all jet pump diffusers in the reactor. Air calibration of four of the jet pump diffusers is re quired because variations in performance result from complex but subtle manufactu ring variations which develop between seemingly identical units. Hydraulic calib ration of only one jet pump unit of a particular design is performed because related air-test data from the four calibrated diffusers is available. Such data estab lish the probable water calibrations within acceptable accuracy limits after one unit has undergone both air and water testing. These four calibrated diffusers are then used to calibrate the remaining diffusers after installation in the reactor.

G.4.1.2 Flow Control Valve Testing

A series of tests was run in 1970 at the GE Moss Landing Facility on an 8-inch scale model to establish the hydraulic performance of the valve. The valve flow characteristic C v vs. position and cavitation index K c vs. position were established.

Tests were run at the Byron Jackson Pump Facility in July and August 1974 on the hydraulic power and servocontroller system to establish and verify velocity, step response linearity, dead-band, interlock, etc. These tests were run on a 24-inch valve. Tests were run at the Bingham-Willamette Pump facility from June to September 1974 in Portland, Oregon, in their recirculation loop on a 24-inch flow control valve, consisting of a "Floor Test" and "Hot Loop Test" to establish: valve feedback function delay time, step response time, flow as a function of position, and other parameters.

Flow tests were run on the present design of the flow control valve at the Alden Research Laboratory, Worster Polytechnic Institute, Holden, Massachusetts. These tests were run on a 20-inch valve in July 1975 and on a 24-inch valve in June 1975. The series of tests was directed and wi tnessed by ITT-Hammel-Dahl engineering personnel to determine the percentage of linear travel versus the valve coefficient C v. Tests were run at Bingham-Willamette again in the latter part of 1976 consisting of "Performance Tests" and "Operational Tests." The performance tests verified the function generator, flow vs. position, and also verified that the valve did not cavitate under operational conditions. The operational tests consisted of operating the valve for over 300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months under full flow and temperature simulating daily load following.

G.4.1.3 Low Frequency Motor-Generator (LFMG) Set Testing In 1976, a production LFMG set was tested in conjunction with a production recirculation pump and flow control valve. These were done with the recirculation LSCS-UFSAR G.4.1-7 REV. 13 pump in a pumping loop simulating the hydr aulic characteristics of an actual plant recirculation loop and simulating the conditions (flow, pressure and temperature) under which the LFMG set will be required to power the recirculation pump. Tests were performed to determine the following:

a. recirculation pump normal coas tdown time to 40%, 25%, and 15% of rated speed;

b. recirculation pump motor residual voltage decay time;

c. time for LFMG set to accelerate and be ready to accept load;

d. capability of LFMG set to accept the load when energizing the pump motor during pump coastdown at pu mp residual speeds of 40%, 25%, and 15% of rated speed; and

e. stability of the flow control valve system with the pump operating at 25% speed.

The results were quite satisfactory. There is a margin of approximately 3 to 1 between the time when the LFMG set must accept load and the time when the LFMG is ready to accept load. As expected, the optimum residual pump speed to perform the transfer was determined to be near 25% of rated speed. The LFMG set accepts the transfer with almost imperceptible change in MG set speed. The flow control valve is stable with 25% flow with the pump loop at normal pressure.

The production control circuit for the LFMG set was not included as part of the LFMG tests. However, pre-op tests on a production circuit have proven quite successful.

G.4.1.4 Hydraulic Power Unit and Actuator Testing In 1974 the actuator, hydraulic power unit (HPU), and complete analog control circuit were tested to prove the design and operability of the equipment. An analog computer with simplified programs was used to simulate the master and flux feedback inputs to allow full automatic closed-loop operation of the system. Tests were done with production recirculation pump, flow control valve, HPU, actuator and controls. With the exception of velocity tests were conducted to simulated plant conditions. These tests determined the following:

a. dynamic performance for each closed loop control mode (position, flow, flux, and master), b. stability, LSCS-UFSAR G.4.1-8 REV. 13 c. valve position versus flow characteristics, d. linearization of position versus flow characteristics, e. valve deadband, and

f. maximum actuator velocity under worst-case single failure.

Velocity tests were done with the hydraulic pump flow adjusted for maximum flow capability. The remainder of the system was adjusted for optimum stable operation. Velocity tests were also done with the actuator connected to the valve but without pressure in the valve and without flow through the valve. This was conservative, since actual velocities will be less than measured under conditions of flow and pressure in the valve.

Tests gave satisfactory results, and it wa s anticipated that performance would meet contractual requirements. Maximum averag e velocity measured was 21%/sec in the actuator retract (fastest) direction.

With the addition of the LFMG set, it wa s necessary to change the active stroke length of the flow control valve. In 1976 another series of te sts was conducted.

These tests were similar to those in 1974 except that closed-loop tests for the master, flux, and flow loops were not repe ated, and shorter lines were used between the HPU and actuator. Length of the hydr aulic lines was selected based on length of lines in an actual plant installation.

Similar results were obtained except that, as expected, dynamic performance was improved due to the shorter hydraulic lines. Also the actuator velocity was greater. Due to change in criteria, the actuator short-term peak velocity was measured as well as the average velocity. Short-term peak velocity was 27%/sec for approximately 5% of the active stroke, and average velocity was 22%/sec in the retract (fastest) direction.

LSCS-UFSAR G.4.2-1 REV. 13 G.4.2 Preoperational Tests G.4.2.1 Recirculation System Tests were performed on the recirculation system after construction and installation were complete and before fuel loading to confirm correct installation and that all process equipment was operational.

G.4.2.1.1 Test Conditions The test conditions vary depending on the state of vessel internals assembly and the flow rate needed for the test. The vessel internals establish backpressure for the jet pumps and thereby the system operating co nditions. Given the flow rate needed, the required water level, water temperature, and reactor pressure needed to prevent system cavitation were established.

G.4.2.1.2 Test Requirements G.4.2.1.2.1 Functional Tests

Functional tests are conducted on individual components or assemblies of components such as interlocks, indicators, recorders, switches, alarms, annunciators, hydraulic power unit - actuator -

flow control valve, et c., to verify that they function properly.

G.4.2.1.2.2 System Performance Tests

System performance tests are conducted in both one- and two-pump operating modes to verify that the performance is within expectations and within design criteria. Jet pump flows of at least 50% of rated are needed to verify that no construction debris is plugging the jet pump nozzles.

G.4.2.2 Vessel Internals Vibration Tests

Vibration tests are conducted to verify the structural integrity of core support structure and reactor internals in response to Regulatory Guide 1.20. The jet pumps are part of this program.

G.4.2.2.1 Vibration Test Conditions Flow testing includes operation of the recirculation system in hot, pressurized conditions without fuel, at flows up to 100%

of rated volumetric flow. The results of vibration measurements in prototype plants (Note: Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants, NEDE-24057-P, November 1977) show that this flow condition produces vibration responses in core support LSCS-UFSAR G.4.2-2 REV. 13 structures which are greater than those at normal operating conditions and are thus conservative for testing purposes. The test includes both balanced and one-loop recirculation system operation of sufficient duration to subject major components to a minimum of 10 6 cycles at their lowest dominant response frequencies and at maximum response amp litudes. Planned test times are 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months of two-pump operation at balanced flow, 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months of A-loop operation and 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months of B-loop operation at maximum flow condition without fuel in the reactor.

G.4.2.2.2 Inspection of Internals The inspection of reactor internals was conducted following preoperational testing of the reactor system. The inspection was conducted prior to fuel loading and reactor criticality. This inspection ha s been performed previously on prototype BWR's following preoperational testing.

After completion of flow testing, the vesse l head and the shroud head were removed and the vessel was drained. Access to the lower plenum is provided by opening a manhole in the shroud support plate. Reactor internal structures and components, including those in the lower plenum region, are given a close visual inspection to detect possible wear, cracking, loosening of bolts, and the presence of debris and loose parts. The inspection covered the following components:

a. Peripheral control rod drive and incore guide tubes, housing, and their lower joints.

b. Incore guide tube stabilizer connections and stabilizer bars. Plenum region for evidence of l oose and/or failed parts.

c. Inside surfaces of the jet pump adapter to shroud support welds and jet pump diffuser to jet pump adapter welds. (Unit 1 only.)

d. Shroud support plate support columns. (Unit 2 only.)

e. Liquid control and delta pr essure line and bracket welds.

f. The shroud-to-shroud weld.

g. Jet pump instrument lines and brackets.

h. The jet pump annulus for evidence of loose parts.

i. Jet pump beams, beam bolts, wedges, and locator screws.

j. Jet pump riser braces and welds.

LSCS-UFSAR G.4.2-3 REV. 13 k. Shroud and shroud head flange locating pins for evidence of deleterious motion marks other than those caused from normal installation.

l. Core support plate bolt keepers.

m. Steam separators and standpoints, shroud head bolt support ring brackets and supports.

n. Feedwater sparger and attachments.

o. Core spray lines, bracket, and core spray spargers.

p. Surveillance sample program specimen holder assemblies.

The La Salle preoperational test of re actor internals vibration conforms to Regulatory Guide 1.20 as a confirmatory test and inspection program. It pre-dated but fully meets the objectives of the NRC's approved generic program for BWR/4 and BWR/5 as accepted by the R. L. Te desco letter of Octobe r 28, 1980 to G. G.

Sherwood (which references NEDO-24057A).

Results of the preoperational test PT-PV-102 are included in the permanent station records.

LSCS-UFSAR G.4.3-1 REV. 13 G.4.3 Startup Tests G.4.3.1 Recirculation System G.4.3.1.1 General Description

Startup tests are those tests performed after fuel is loaded. The goals of the tests are as follows:

a. Verify the operational limits of the plant by establishing the power flow operating map.

b. Generate operating procedures that ensure rapid power ascension and maximize plant availability.

c. Obtain performance data needed to confirm the SAR analysis and conduct the surveillance required by the technical specifications and for detecting problems by establishing system performance bench marks where a deviation later in plant life indicates a potential problem. d. Ensure that the flow control system performs as specified.

e. Verify that all interlocks work properly.

G.4.3.1.2 Specific Tests

The subsections which follow specify the purposes of the specific tests.

G.4.3.1.2.1 Cavitation Tests Determine cavitation zones and establis h operating procedures to preclude operation in these zones. Verify adequacy of the automatic cavitation interlocks.

G.4.3.1.2.2 Stratification Test Determine the reactor operating modes wher e the stratification limits are likely to occur so that plant operating procedures can be established to avoid these modes.

G.4.3.1.2.3 Flow Increase Operation Test

Determine the operating procedures necess ary to avoid a scram when transferring from 25 to 100% speed and when restar ting an idle recirculation pump.

LSCS-UFSAR G.4.3-2 REV. 13 G.4.3.1.2.4 Steady-State Performance Test Obtain operating data for future reference and for jet pump surveillance required by Technical Specifications.

G.4.3.1.2.5 Jet Pump and Recirc ulation Pump Flow Measurement Calibrate the core flow measurement system and adjust the APRM flow bias instrumentation.

G.4.3.1.2.6 Recirculation Control System Test Tune up the flow control subsystem to achieve the warranted response requirements. Measure plant response to one and two recirculation pump trips.

G.4.3.1.2.7 Recirculation Pump Trip Test Record and verify acceptable performance of the RPT system.

LSCS-UFSAR G.5.1-1 REV. 13 G.5 RECIRCULATION PUMP TRIP AND FLOW CONTROL FEATURES G.5.1 Recirculation Pump Trip (RPT) System G.5.1.1 Identification

The RPT system includes power supplies, circuit breakers, trip circuitry, sensors and bypass switches to trip the recirculation pump motors from their main power supplies in response to a turbine/generator trip or load rejection event at greater than or equal to 25% power. Isolated outputs to the process computer and annunciators are provided for monitoring and are not a part of the RPT system.

The RPT is classified as Class IE, Seismic Category I, and Safety Class 3. The RPT system is designed to aid the reactor protection system (RPS) in protecting the integrity of the fuel barrier. Turbine stop valve closure or turbine control valve fast closure will initiate a scram and concurrent recirculation pump trip in order to keep the core within the thermal-hydraulic safety limits during abnormal operational transients.

G.5.1.2 Power Supplies The RPT system utilizes two types of power from the same sources as the reactor protection system (RPS); 120-Vac from nonessential RPS motor-generators is supplied for the sensor channels and esse ntial 125-Vdc from station batteries is supplied for the logic trip circuits for RPT.

G.5.1.3 Equipment Design G.5.1.3.1 Circuit Description The RPT system circuit is derived from the reactor protection trip sensor channels and combined into logic to perform the recirculation pump trip function (see Figure G.5.1-1). The fail-safe circuits are re lays energized during normal operation. A failure of any one channel will not trip the recirculation pump motor circuit

breakers or prevent tripping the breakers with the occurrence of an event requiring RPT. Turbine control valve fast closure signals to the RPT system, from the reactor protection system, are derived from oil line pressure switches located on each of the four fast-acting control valve hydraulic mechanisms. When hydraulic oil line pressure is lost, the RPT system trips the recirculation pump motors from their power supplies whenever the re actor is operating at greater than or equal to 25% of rated core thermal power (see Figure G.5.1-2).

LSCS-UFSAR G.5.1-2 REV. 13 Turbine stop valve closure signals to the RPT system from the reactor protection system, are derived from valve stem positi on switches mounted on the four turbine stop valves. The switches op en before the valve is more than 10% closed to provide the earliest positive indication of valve closure (see Figure G.5.1-3).

Turbine first-stage pressure signals to the RPT system, from the reactor protection system, are derived from four pressure swit ches. The pressure switches trip at a pressure setpoint corresponding to 25% of rated core thermal power to bypass the turbine stop valve and turbine control valve fast closure trips below this value (see Figure G.5.1-2).

Loss of one power supply will not cause nor prevent an RPT action.

The RPT, when initiated, completes the recirculation pumps circuit breakers trip. Restarting the pumps for normal power op eration requires deliberate operator actions. Channel and logic relays are fast-response, high-reliability relays from the reactor protection system. The relays are selected so that the continuous load will not exceed 50% of the continuous duty rating. The total response time from start of the turbine control valve fast closure signal or the turbine stop valve closure signal to complete suppression of the electric arc be tween the fully open contacts of the pump motor circuit breaker is less than 190 msec.

The RPT logic is illustrated in Figure G.5.

1-1. The system is arranged as two separately powered trip systems. Each logic trip system has at least two channels of the monitored variable. Either of the two automatic trip systems will trip both of the two recirculation pump motors.

G.5.1.3.2 Bypasses and Interlocks With the reactor power under 25% of rated core thermal power, the RPT is automatically bypassed by four pressure switches associated with the turbine first stage. Any one of the four channels in a bypass state initiates main control room annunciation. A manual keylocked bypass sw itch is provided for each of the two RPT systems, located on the relay panels, for logic testing. Placing the manual switch in bypass position initiates a control room annunciator. RPT is not inhibited, since the redundant system is operable.

The RPT system initiation logic is composed of contact from the RPS sensor channel relays. The interlock is performed using separate relay contacts so that no failure in the RPT system can prevent an RPS sc ram. Isolated contacts from the RPT system interlock with the recirculation pump low-frequency motor-generator set (LFMG) autostart circuit. A failure in the LFMG circuit does not affect the operation of the RPT system.

LSCS-UFSAR G.5.1-3 REV. 13 G.5.1.3.3 Redundancy The RPT is divided into two divisions. Each division duplicates the function of the other to the extent that either system performs the pump motor power supply trip regardless of the state of operation or failure of the other trip system. The turbine stop valve closure and turbine fast valve closure signals are diverse inputs to the RPT for pump motor trip. Each pump motor is provided with two circuit breakers in series tripped from redundant division logic.

G.5.1.3.4 Testability The RPT system logic can be tested during reactor operation without pumps trip and without inhibiting the pump-trip function by the redundant system.

A keylocked bypass switch for system circuit testing is provided for each system located at the RPS relay cabinets.

Because the RPT system logic consists of relay contacts actuated in the RPS, the RPS surveillance tests include the testing of the RPT logic system. See Section 7.2 of the UFSAR for RPS testing.

The recirculation pump motor power supply circuit breakers will be tripped by the RPT system during the reactor refueling period.

G.5.1.4 Environmental Considerations

The electrical equipment and devices of the RPT system are located in the control room and in the electrical equipment rooms. These areas have a controlled environment, isolated from accident environment, assuring reliable operation. The logic channel sensors located at the turbine-generator are discussed in Section 7.2 of the UFSAR.

G.5.1.5 Operational Considerations The sensor channel logic relays are normally energized with contacts open in the RPT system logic. Loss of both RPS M-G bu ses or sensor variables out of tolerance will cause recirculation pump trip.

The RPT system has no manual trip feature. The pump motor power supply circuit breakers have their normal manual control switches at the control room panels for normal pump start and stop operation. These controls are isolated from the RPT circuits. Circuit breaker status indications for the operator consist of a green pilot light for breaker open position and a blue p ilot light for breaker closed. In addition there is a red pilot light to monitor the RPT trip coil.

LSCS-UFSAR G.5.1-4 REV. 13 G.5.1.6 Analysis Presented in the following are analyses to demonstrate RPT system conformance to functional requirements and specific regulatory guides.

G.5.1.6.1 Conformance to Functional IEEE Requirements IEEE 279 1971 - Criteria for Protection Systems for Nuclear Power Generating Stations

a. General Functional Requirement (IEEE 279, Para 4.1). The function of the RPT system is to reduce th e fuel thermal consequences during turbine or generator trip transients by tripping the recirculation pumps early in the event.

The valve stem position of each of the four turbine stop valves is monitored by limit switches with se tpoint at 10% or less valve motion from the open position. The trip channel signals to the RPT system anticipated imminent closure of the stop valves. The response time of

the trip switch contacts is less than 10 msec after the valve has reached the setpoint. The logic arra ngement is established to enhance frequent testing of these valves without causing an RPT for each valve test. The logic is one-out-of-two-twi ce stop valve closure signals from system A or B to produce RPT. The turbine control valve fast closure is monitored by pressure switches mounted at the turbine acceleration relay EH coil line pressure. The pressure switches provide RPT channel signals in less than 30 msec after decay of oil line pressure to the set point. The logic is two-out-of-two for each trip system A and system B. Testing the logic circuit is available without recirculation pump trip.

The turbine stop valve and control valve fast-closure-trip bypass consists of sensing turbine first-stage pressure with two trips and four pressure switches so as to activate a trip bypass if the reactor is operating below 25% of rated core thermal power. This bypass is provided to permit continued reacto r operation at lower power levels when the turbine valves are closed. This bypass is automatically removed as the reactor power an d turbine first stage pressure increases above the setpoint value equivalent to 25% of rated core thermal power.

b. Single-Failure Criterion (IEEE 279, Para. 4.2). The RPT meets the single failure criterion. Sensors are electrically and physically separated, with conduit provided to the RPS cabinets. RPT signals to LSCS-UFSAR G.5.1-5 REV. 13 recirculation pump motor circuit breaker trip coils are separated into Division I and Division II and circuit breaker to pump motors are divisionally separated.

c. Quality and Components and Modules (IEEE 279, Para. 4.3

). The quality of components and modules of the RPT system are of equal quality and high standards of the Class IE RPS as described in NEDO 10139, "Compliance of Protection Systems to Industry Criteria: GE BWR Nuclear Steam Supply System." d. Equipment Qualification (IEEE 279, Para. 4.4

). This paragraph is satisfied by qualification testing and certification of active RPT components. Complete records covering all essential components are maintained by GE.

e. Channel Integrity (IEEE 279, Para. 4.5). The RPT system logic consists of RPS channels as descri bed in Section 7.2 of the UFSAR. Those components unique to the RPT system are located in the control room and electric equipment rooms environment.

f. Channel Independence (IEEE 279, Para. 4.6). The trip channels are physically separated and electrically isolated to meet this design requirement.

g. Control and Protection System Interaction (IEEE 279, Para. 4.7). The RPT system has no interface with co ntrol systems. Consequently, no failure or combination of failures in a control system will have any effect on the RPT system.

h. Derivation of System Inputs (IEEE 279, Para. 4.8). The RPT system is required to trip the recirculation pump motors on turbine/generator trip or load rejection. The sensing of turbine stop valve closure or turbine control valve fast closure are the actuated variables for the RPT function.

i. Capability of Sensor Checks (IEEE-279, Para. 4.9). The sensors are designed into the reactor protection system and are testable as described in Section 7.2 of the UFSAR.

j. Capability for Test and Calibration (IEEE 279, Para. 4.10). The turbine stop valves are tested individually by closing a stop valve and verifying RPT relay operation before the control room lights indicate the valve is closed. Calibration of the limit switches is possible only during shutdown and by physical observation. The turbine control LSCS-UFSAR G.5.1-6 REV. 13 valve closure pressure switches may be valved out, tested, and calibrated during periodic testing.

k. Channel Bypass or Removal from Operation (IEEE 279, Para. 4.11). The RPT system meets this design requirement as given above in Subsection G.5.1.3.2.

l. Operating Bypasses (IEEE 279, Para. 4.12). The recirculation pump motors are not required to trip below 25% of rated core thermal power.

The four bypass channels comply with all requirements of IEEE 279 criteria.

m. Indication of Bypasses (IEEE 279, Para. 4.13). Automatic or manual test bypasses are annunciated in the control room.

n. Access to Means for Bypassing (IEEE 279, Para. 4.14). Automatic channels can be removed from operation, one at a time, for periodic testing under administrative control. Manual test bypass is accomplished with a keyed switch for periodic testing, under administrative control.

o. Multiple Setpoints (IEEE 279, Para. 4.15). Multiple setpoints are not a design requirement for RPT.

p. Completion of Protective Action Once it is Initiated (IEEE 279, Para. 4.16). The RPT system initiates to completion circuit breaker trips in maximum of five cycles when rated reactor power is above 40%. The trip is sealed in by swit chgear action. Deliberate operator action is required to return the recirculation pumps to normal reactor operation.

q. Manual Actuation (IEEE 279, Para. 4.7). Manual initiation of RPT is not a design requirement since it is not possible to manually activate RPT in time to perform its safety function within 190 msec.

r. Access to Setpoint Adjustments, Calibration and Test Points (IEEE 279, Para. 4.10). Access to the turbine stop valve limit switches is not anticipated during reactor operation due to radiation and temperature environmental conditions. Turbine stop valve test controls are accessible. The turbine control valve fast closure pressure switches are available for periodic calibration.

s. Identification of Protective Actions (IEEE 279, Para. 4.19). RPT actions are identified to the channel level by control room annunciation.

LSCS-UFSAR G.5.1-7 REV. 13

t. Information Readout (IEEE 279, Para. 4.20). Control room annunciation complies with this design requirement.

u. System Repair (IEEE 279, Para. 4.21). The RPT system components are designed to facilitate maintenance, with the exception of the turbine stop valve limit switches. The redundancy of the eight switches permits plant operation with a defective switch until access can be gained to the switches for repair.

v. Identification (IEEE 279, Para. 4.22). The RPT system logic relays are located in divisionally separated RPS cabinets with marker plate identification. The pump motor circuit breakers have distinctive

markings and external cables are color coded. The RPT sensors are part of the reactor protection system and are discussed in Section 7.2 of the UFSAR.

IEEE 323-1971 The RPT system complies with this industry standard by type qualification testing and certification of all essential components. Complete records are maintained covering essential components.

IEEE 336-1971 The extent of conformance to this industry standard on the installation, inspection and testing requirements for instrumentation and electric equipment during construction is described in the Applicant's QA Program.

IEEE 338-1971 The RPT system is testable up to but not including actual recirculation pump trip circuit breaker during periodic testing of the sensor channels and logic systems. The pump trip circuit breaker testing is performed during a refueling outage.

IEEE 344-1971 This standard for seismic qualification is complied with for all RPT system Class 1E components.

LSCS-UFSAR G.5.1-8 REV. 13 G.5.1.6.2 Conformance to Regulatory Guides Conformance of the RPT system to Regula tory Guide 1.22 is demonstrated under IEEE 338-1971.

Annunciation is provided in the control ro om to indicate a system or part of a system is inoperable or bypassed. As a re sult of design, preoperational testing, and startup testing, no erroneous bypass indication is anticipated. Therefore, Regulatory Guide 1.47 conformance exists.

Conformance to Regulatory Guide 1.53 for single failures is by compliance with IEEE 279-1971, Paragraph 4.2.

The requirements of the RPT system design do not include manual initiation, hence Regulatory Guide 1.62 is not applicable.

G.5.1.6.3 General Design Criteria Conformance 10 CFR 50 Appendix A Criterion 13 - Each RPT system input is monitored and annunciated.

Criterion 19 - The RPT does not require controls or instrumentation in the control room. Criterion 20 - The RPT system monitors the appropriate variables and initiates automatically when the variables exceed the design setpoints.

Criterion 21 - The system is designed with four independent and separated sensor channels and two independent and separated system divisions. No single failure or operator action can prevent RPT. The syst em can be tested during plant operation to verify its availability.

Criterion 22 - The redundant portions of RPT are separated such that no single failure or credible natural disaster can prevent RPT. Functional diversity is accomplished by sensing the turbine stop valve position for turbine trip and turbine control valve fast closure for load rejection.

Criterion 23 - The logic system is fail-safe. Loss of one of the two divisions of logic electrical power will not prevent the RPT syst em function because the logic will trip.

The RPT system itself is energize-to-operate.

Loss of one division of 125-Vdc control power will not prevent the RPT system func tion. Postulated adverse environments will not prevent an RPT.

Criterion 24 - The RPT has no control function. The system has interlocks to control systems through isolation devices.

LSCS-UFSAR G.5.1-9 REV. 13 Criterion 29 - No anticipated operationa l occurrences will prevent the RPT from accomplishing its safety function.

LSCS-UFSAR G.5.2-1 REV. 15, APRIL 2004 G.5.2 Recirculation Flow Control G.5.2.1 Identification G.5.2.1.1 Function

The objective of the recirculation flow control function is to control reactor power level, over a limited range, by controlling the flow rate of the reactor recirculating water. The recirculation flow control system consists of the electrical circuitry, switches, indicators, motor, and alarm devices provided for operational manipulation of the recirculation flow control valves, the LFMG set, and the surveillance of associated equipment. Recirculation flow control is either by individual manual operation or ganged manual if the power level is above 65% of rated, when the plant is operating on a rod pattern where rated power is produced with rated recirculation flow. During periods of low power level such as plant startup and shutdown, the recirculation pump and motor is powered by the LFMG set and is operated at approximately 25% rated full load speed. Refer to the cycle-specific COLR to determine if automatic flow limits ar e provided for a particular cycle.

G.5.2.1.2 Classification This control function fulfills a power-generation criterion and is classified as nonessential for safety.

G.5.2.2 Power Sources G.5.2.2.1 Control System The recirculation flow control system instrument power is supplied by two nonessential 120-Vac instrument buses. Flow control loop A is powered from bus A and flow control loop B from bus B. That portion of the control system which is common to both loops A and B (master controller, etc.) is powered from either bus A or B. Each bus receives its normal power supply from the appropriate 480-Vac normal auxiliary power system.

Alternate On loss of normal auxiliary power, on e of the reserve auxiliary transformers provides backup power to the 480-Vac normal auxiliary power system.

LSCS-UFSAR G.5.2-2 REV. 15, APRIL 2004 G.5.2.2.2 Pump Drive Motor and Controls The recirculation pump drive motor is an a-c 4-pole induction motor. The recirculation pump/motor operates from the normal plant electrical supply during normal plant power operation. At low-power levels, the recirculation pump/motor operates from the electrical output of th e LFMG set. Since the LFMG set electrical output frequency is at approximately one-fourth the normal plant electrical frequency, the recirculation pump/motor is driven at approximately one-fourth of its rated speed.

The LFMG set consists of a 16-pole a-c induction motor driving a 4-pole a-c synchronous generator through a flexible coupling. This arrangement provides one-quarter normal plant frequency at the output of the generator.

The generator exciter is directly connected to the generator to provide a brushless excitation system. The voltage regulator for the excitation system is located in the auxiliary relay panel which is separate from the LFMG set.

G.5.2.3 Equipment Design

G.5.2.3.1 General Reactor recirculation flow is varied by throttling the recirculation flow control valve. The recirculation pumps operate at constant speed, on either LFMG or normal 60-cycle power. By adjusting the position of the flow control valve, the change in recirculation flow automatically changes the reactor power level.

Control of core flow is such that, at various control rod patterns, different power level changes can be automatically accommodated. For a rod pattern where rated power accompanies 100% flow, power can be reduced to approximately 65% of full power by manual flow variation. At other rod patterns, manual power control is possible over a range of approximately 30% of the maximum operating power level for that rod pattern.

An increase in recirculation flow temporarily reduces the vo id content of the moderator by increasing the flow of coolant through the core. The additional neutron moderation increases the reactivity of the core, which causes reactor power level to increase. The increased steam ge neration rate increases the steam volume in the core with a consequent negative reactivity effect, and new steady-state power level is established. When recirculation flow is reduced, the power level is reduced in the reverse manner.

LSCS-UFSAR G.5.2-3 REV. 15, APRIL 2004 Each flow control valve has its individual manual control system and a ganged control mode which allows both valves to be manually controlled simultaneously.

In the ganged control mode, each loop has an individual flow controller that causes adjustment to valve position to meet a demanded change in valve position and, hence, core flow and core power. This process continues until both the errors existing at the inputs of the individual flow controllers are driven to zero.

The reactor power change resulting from change in recirculation flow causes the turbine EHC control system to reposition the turbine control valves.

G.5.2.3.2 LFMG Set

The LFMG set is not intended to be capable of starting the recirculation pump/motor with the pump/motor initially at zero speed. At low reactor power levels, the pump/motor start is initiated on the normal plant electrical power supply; as the pump/motor speed approaches rated full load speed, it is tripped automatically. When the coastdown pump/m otor speed is about 25% of rated full load speed, the pump/motor is reenergized from the LFMG set and driven at about 25% rated full-load speed. Preceding initiation of the pump/motor, the plant operator may manually start the LFMG set. If the LFMG set is not operating when the pump/motor start is initiated, the LFMG set will be started automatically.

If pump/motor start is initiated at high er reactor power levels, the LFMG set will not start automatically, and the pump/motor will continue to operate at rated full load speed.

Certain trip functions, as defined in Table G.5.2-1, will trip the pump/motor and automatically transfer it to the LFMG set. Other trip functions will trip the pump/motor without transfer to the LFMG set.

LSCS-UFSAR G.5.2-4 REV. 15, APRIL 2004 G.5.2.3.3 Block Valves (Suction/Discharge)

The block valves are motor-operated gate valves for the recirculation pump and FCV isolation. The following interlocks are required for pump/motor startup and trip control function:

a. The power supply to the pump/motor will be tripped if either block valve is less than 90% open.

b. The pump/motor will not start unless the block valves are greater than 90% open.

G.5.2.3.4 Flow Control Valve Position Control Components The flow control valves can be controlled individually or jointly. The control configuration is shown in Figure G.5.

2-1. Each FCV has two independent controllers (AC70) and group S800 I/O modules supporting each redundant half of the hydraulic power unit (HPU) and associated process instrumentation. There is also an AC70 that controls the jet pump instrumentation flat panel display via a MODBUS connection.

Each loop flow controller M/A statio n communicates with both controllers corresponding to one recirculation loop.

The control and logic of each HPU subloop is implemented in each individual controller without redundancy. The redundancy is achieved by the existence of two independent HPU subloops (servo valve, solenoid operated 4-way valve, hydraulic pump, etc.). In case of failure in one controller, an automatic transfer is performed to the backup controller and HPU. A failure in one of the HPU subloops will force a transfer to backup HPU subloop and controller. The controllers are balanced to allo w bumpless transfer between them.

The ganged control algorithm is implemente d in one of the controllers. In ganged control mode, both recirculation flow loops are controlled in parallel by a ganged position setpoint (common position setpoint M/A station).

LSCS-UFSAR G.5.2-5 REV. 15, APRIL 2004 THIS PAGE INTENTIONALLY LEFT BLANK LSCS-UFSAR G.5.2-6 REV. 15, APRIL 2004 G.5.2.3.4.6 Manual/Automatic Transfer Stations Setting the individual M/A station flow controllers to automatic provides "ganged" parallel manual operation of the flow control loops.

G.5.2.3.4.7 Flow Controller The individual flow controller (one for each valve) transmits the signal that adjusts the valve position. Each flow-regulatin g valve can be manually positioned from individual M/A stations or in parallel from the ganged position setpoint station.

G.5.2.3.4.8 Limiter A limiting function is provided for one feedwater pump trip flow runback. An electronic limiter with reasonable range adju stment is provided in each main flow control loop. This limiter is normally held bypassed by controller logic. When one feedwater pump trip coincides with re actor low water level alarms, the main regulating valve control signal is limited to close the valve to the desired position.

G.5.2.3.4.9 Hydraulic Power Unit G.5.2.3.4.9.1 General Description The hydraulic power unit is an electromechanical device that moves the actuator on the flow control valve in response to flow demand changes. Each flow control valve has a separate hydraulic power unit.

Each unit has two subloops to provide redundant operation.

LSCS-UFSAR G.5.2-7 REV. 14, APRIL 2002 A failure in the "A subloop" would automatically place the "B subloop" into operation. The main components of the unit are shown on Figure G.5.2-2 (Unit 1 only). The units are proven commercial devices rated for operation up to 3,000 psig.

Actual operating pressure is 1800 psig.

Acceptance of the units requires a hydrostatic pressure test at 3,000 psig.

Generally, mechanical failure in a subloop will lock up the actuator. If the redundant subloop starts and is operational, the valve will unlock in less than 1 second. Some failures that cause a transfer to the standby subloop, such as high fluid temperature, will not result in a lockup.

The hydraulic power units are located outside the drywell. Only the hydraulic lines, pilot-operated check valves and the hydr aulic actuator are located inside the drywell. G.5.2.3.4.9.2 Operation A brief description of the operation of th e hydraulic power units (HPU) is given to aid in the understanding of how the system functions. A diagram showing the main components of the HPU is given in Figure G.5.2-2.

Each subloop has a total inventory of about 75 gallons of Fyrquel EHC hydraulic fluid, 50 gallons of which is stored in a reservoir tank. A 15-hp, 1800-rpm motor is coupled to a variable displacement pump. The displacement is controlled by a pressure compensator which maintains desired hydraulic pressure. Maximum capacity of the pump is 9.7 gpm.

Actuator rod movement is accomplished by a 4-way servovalve that can supply the pump pressure to either side of the actuator piston. When the servovalve is "centered," flow to the actuator piston is blocked, which results in no movement.

The servovalve movement is governed by the flow controller as described in Subsection G.5.2.3.4.7.

Downstream of the servovalve is a 4-way pilot-operated valve that isolates the standby HPU from the operating unit. This valve also serves to isolate (close) the actuator hydraulic lines for lockup purposes.

Mounted on the actuator, inside the drywell, are pilot-operated check valves. These valves can isolate the hydraulic cylinder in the event of a flow control lockup or hydraulic pressure loss. Supply pressure to both the pilot-operated check valve and the pilot-operated isolation valve is contro lled by a solenoid-operated 4-way valve.

This valve will cut off pressure to the pilot check valves and supply pressure to the isolation valve when actuated by a lockup signal from the flow controller. With this LSCS-UFSAR G.5.2-8 REV. 13 arrangement, actuator lockup is accomplished by both the pilot-operated check valves and the isolation valve.

A shuttle valve, located between the lockup valves of each subloop and the pilot-operated check valves, isolates the standby subloop from the operating unit.

On the discharge side of the pump is an accumulator that acts as a surge tank to improve the response time of the HPU. Inside the tank is a collapsible bladder that can be precharged with inert gas. By varying the initial precharge pressure, the surge fluid volume can be altered.

In addition to the pressure compensator on the pump, a pressure relief valve, set at 110% of operating pressure, is located near the pump discharge.

For clarity, several components were omitted from the HPU diagram. There are filters located in-line on the pump suction and discharge, on the servo return lines, and on the pump bypass drain line. Manual isolation valves are located between the pilot-operated check valves and the isolation valve, between the shuttle valve and the lockup valve, and on the pump suction. Since the reservoir tank is above the pump, the pump suction isolation valve permits maintenance on the HPU without draining the reservoir. To main tain hydraulic fluid at normal operating temperatures (below 145° F), an air-cool ed heat exchanger is located on the reservoir return line.

G.5.2.3.4.9.3 Protection Interlocks Various interlocks are provided to shutdown a malfunctioning subloop. The

following conditions will result in an automatic transfer to the standby HPU:

a. low pump discharge pressure;

b. low reservoir tank level (near empty); and

c. high-temperature hydraulic fluid (150° F).

Other conditions which will not cause automatic transfer, but will be displayed and annunciated are as follows:

a. low reservoir level (different from Item b);

b. warm temperature hydraulic fluid;

c. discharge pump filter dirty;

d. pump bypass drain filter dirty; and LSCS-UFSAR G.5.2-9 REV. 15, APRIL 2004

e. high actuator seal leakage.

G.5.2.3.4.9.4 Hydraulic Fluid Description The hydraulic fluid used is Fyrquel EHC. It has an auto-ignition temperature of 1150° F. When heated to 455° F, the vapor will flash with an ignition source. If heated to 700° F, burning of the vapor la sts at least five seconds once ignited.

G.5.2.3.4.10 Hydraulic Line Design Actual operating pressure is 1800 psig, although units are rated for service up to 3000 psig.

G.5.2.3.4.11 Actuator Design The actuator is a double-acting cylinder rated for 3,000 psi severe operation, 5,000 psi non-shock. Actual operating pr essure is 1800 psi.

Position stops are located inside the cylinder to limit the travel of the piston. The stops have been set to provide the correct stroke travel for the flow control valve.

A linear velocity transformer is mounted on the actuator rod and provides the velocity feedback to the flow control sy stem. The position feedback signal is provided by redundant transducers connected to the ball shaft of the valve.

Hydraulic sealing of the actuator rod is pr ovided by a primary seal which provides the full pressure containment, and a secondary seal which prevents any leakage outside the unit. Between the seals is a cavity which has a drain line connected back to the supply reservoir. A flow switch mounted in this line will visually indicate and annunciate to the operator a high drain flow condition.

G.5.2.3.5 Inspection and Testing The controller, valve position limiter, hydraulic power unit, and valve actuator are functioning during normal power operation. Any abnormal operation of these components can be detected during oper ation. The components that do not continually function during normal oper ation can be tested and inspected for calibration and operability during scheduled plant shutdown. All the recirculation flow control components are tested and inspected according to the component manufacturers' recommendations. The LFMG set can be operated, tested, and maintained during normal power operation.

LSCS-UFSAR G.5.2-10 REV. 15, APRIL 2004 G.5.2.4 Environmental Considerations The recirculation flow control system is not required for safety purposes, nor required to operate during or after the design-basis accident. The system is required to operate in the normal plant environment for power-generation purposes only.

G.5.2.5 Operational Considerations G.5.2.5.1 General Information Controllers for positioning the flow control valve are located in the main control room. The individual loop controllers have provisions for either ganged or manual operation. Control switches for pump isolation valves, LFMG set and pump/motor, and interlock reset functions are also located in the main control room. Switches and indicators for control of the flow cont rol valve hydraulic system are located on an operator station in the main control room for easy accessibility.

Except for the equipment protective interlocks, controls are manual, requiring operator action.

The LFMG set is required to supply powe r to the recirculation pump/motor only during plant low-power conditions. Provis ions are made to allow operation of the LFMG set independent of pump/motor operation during normal plant power operations as well as during plant shutdown.

G.5.2.5.2 Reactor Operator Information Indications, trends and alarms are provided to keep the operator informed of the status of system and equipment and to quickly determine location of malfunctioning equipment.

Visual display consists of loop flow, recirculation drive flow, valve position, and controller output and velocity feedback meters. Alarms are provided to alert the operator of malfunctioning control signals, inability of the valve position control loop, condition of hydraulic system, pump mo tor, and temperatures of cooling water. In most cases, alarms are supplemented by alarm indication provided by operator station indicators and messages to more closely define the problem area.

Indicating lights are provided to indicate the status of the LFMG set and the pump/motor control breakers. A pump/motor speed indicator is provided to indicate (in addition to the breaker indicating lights) to the operator which power supply is driving the pump/motor. Alarms are provided to alert the operator of automatic trips and transfers of the pump/motor, ma lfunctions, and availability of automatic control circuitry.

LSCS-UFSAR G.5.2-11 REV. 15, APRIL 2004 G.5.2.6 General Functional and Specific Regulatory Requirements Conformance General Functional Requirements Conformance The recirculation control system is designed so that no recirculation failure caused by the recirculation control system failure would result in violation of MCPR safety limit. The detailed description and analys is of the recirculation system failure events are in Section G.6.

The valve actuator has an inherent rate-limiting feature that will keep the resulting rate of change of core flow and power to within safety limit in the event of an upscale or downscale failure of the va lve position or velocity controller.

Specific Regulatory Requirements Conformance The interface signal and components that are used for both safety protection function and control function are designed in compliance with the following general design criteria (GDC) of Appendix A to 10 CFR 50, and IEEE 279, Section 4.7.

a. GDC 13 for Instrumentation and Control Requirements

Conformance to this requirement is shown in the process instrument diagram figure. The recirculation system is capable of monitoring and controlling all the important processes and control variables over their anticipated range for normal operation and for anticipated operational occurrences.

b. GDC 24 for Separation of Protection and Control System The end-of-cycle recirculation pump trip (RPT) is the only safety-related control signal in the recirculation system. Failure of the recirculation control system or any component thereof would not affect the capacity of the protection system to perform its safety function.

c. IEEE-279, Section 4.7 for Control and Protection System Interface Any failure in the recirculation control system will not prevent the protection system from performing its intended protective function.

LSCS-UFSAR G.6.1-1 REV. 13 G.6 TRANSIENT ANALYSIS Refer to Section 15.3.

LSCS-UFSAR G.7.0-1 REV. 15, APRIL 2004 G.7 REFERENCES

1. R. B. Linford, "Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor" (NEDO 10802), February 1973.

2. "General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K" (NEDO 20566).

Flows - vessel water saturation temperature and recirculation pump inlet is less than the setpoint value.

A High accuracy RTD is provided to measure recirculation pump in let temperature. A pressure transmitter measuring vessel pressure in conjunction with a function generator is provided to measure ve ssel water saturation temperature.

3. GE Document, "SAFER/GESTR-LOCA, Loss-of-Coolant Accident Analysis, LaSalle County Statio n Units 1 & 2," NEDC-32258P, October 1993.

4. GE Document, "Evaluation of Lower Recirculation Flow Control Valve Cavitation Interlock Setpoint for LaSalle County Station Unit 1 and 2", GE-NE-AOO-05525-01, May 1994.

5. GE Document, "LaSalle 1 and 2 Jet Pump Cavitation Protection Logic," GE-NE-B3500504-01 DRF B 35-00504, Class 3 June 1996.

6. "Exxon Nuclear Methodology For Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," XN-NF 19(P)(A) Volumes 2, 2A, 2B, 2C Exxon Nuclear Company, Inc., September 1982.

7. LaSalle LOCA-ECCS Analysis MAPLHGR Limits For Atrium-9B Fuel," EMF-2175(P), Siemens Power Corporation, August 1999.

8. Westinghouse Report PNTC 00-261, LaSalle Unit 1 - Reactor Recirculation Flow Control, RRFC Functional Description (Proprietary)

9. Westinghouse Report PNTC 00-129, LaSalle Unit 2 - Reactor Recirculation Flow Control, RRFC Functional Description (Proprietary)

10. "LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM-10 Fuel", EMF-2641(P), Revision 0, Framatome ANP, November 2001.

LSCS-UFSAR G.A-i REV. 15, APRIL 2004 ATTACHMENT G.A REACTOR RECIRCULATION SYSTEM DESIGN DRAWINGS The following drawings are included by Refe rence Only in this attachment. Copies of drawings are located in the UFSAR Volumes entitled "Drawings".

1. Piping and Instrumentation Drawings (P&ID)

M-93, Unit 1 M-139, Unit 2

2. Electrical Schematic Diagrams 1E-1-4205AA-AZ, Unit 1 1E-1-4205BA-BX, Unit 1 1E-1-4205ZA-ZC, Unit 1 1E-1-4205CA-CR, Unit 1 1E-2-4205AA-AZ, Unit 2 1E-2-4205BA-BZ, Unit 2 1E-2-4205ZA-ZC, Unit 2 1E-2-4205CA-CK, Unit 2

3. Control and Instrumentation Drawings M-2093, Unit 1 M-2139, Unit 2

LSCS-UFSAR G.B-i REV. 13 ATTACHMENT G.B FAILURE MODES AND EFFECTS ANALYSES

LSCS-UFSAR G.B-ii REV. 15, APRIL 2004 ATTACHMENT G.B FAILURE MODES AND EFFECTS ANALYSES TABLE OF CONTENTS PAGE G.B.1 Suction and Discharge Block Valves Failure Modes and Effects Analysis G.B.1-1 G.B.2 Flow Control Valve, Ac tuator and Hydraulic Power Unit Failure Modes and Effects Analysis G.B.2-1 G.B.3 Flow Control Instrumentation and Controls Failure Modes and Effects Analysis G.B.3-1

G.B.4 Recirculation Pump Trip (RPT) Failure Modes and Effects Analysis G.B.4-1 G.B.5 Control Interlocks Failure Modes and Effects Analysis G.B.5-1 G.B.6 Power Supplies Failure Modes and Effects Analysis G.B.6-1 G.B.7 References G.B.7-1

LSCS-UFSAR G.B-iii REV. 13 ATTACHMENT G.B FAILURE MODES AND EFFECTS ANALYSES LIST OF TABLES NUMBER TITLE G.B.1-1 Suction and Discharge Block Valves Failure Modes and Effects Analysis G.B.2-1 Flow Control Valve Failure Modes and Effects Analysis

G.B.2-2 Flow Control Valve Actuator an d Hydraulic Power Unit Failure Modes and Effects Analysis G.B.3-1 Flow Control Instrumentation and Controls Failure Modes and Effects Analysis G.B.4-1 Recirculation Pump Trip (RPT) Failure Modes and Effects Analysis G.B.5-1 Control Interlocks Failure Modes and Effects Analysis G.B.6-1 Power Supplies Failure Modes and Effects Analysis LSCS-UFSAR G.B-iv REV. 13 ATTACHMENT G.B FAILURE MODES AND EFFECTS ANALYSIS LIST OF FIGURES NUMBER TITLE G.B.2-1 Valve Actuator Hydraulic Power Unit

LSCS-UFSAR G.B.1-1 REV. 13 G.B.1 Suction and Discharge Block Va lves Failure Modes and Effects Analysis

LSCS-UFSAR G.B.2-1 REV. 13 G.B.2 Flow Control Valve, Actuator and Hydraulic Power Unit Failure Modes and Effects Analysis G.B.2.1 Flow Control Valve Failure Modes and Effects Analysis G.B.2.2 Flow Control Valve Actuator and Hydraulic Power Unit Failure Modes and

Effects Analysis (FMEA)

G.B.2.2.1 Purpose The purpose of this FMEA is to present the consequences of malfunctions within the hydraulic equipment which controls the reci rculation flow control valve in response to signals from the associated electronic equipment.

G.B.2.2.2 Discussion G.B.2.2.2.1 General The attached tabulation presents all conceivable single failures irrespective of their probability of occurrence. In addition, the "Remarks and Other Effects" column presents second, third, etc., failures which could increase the severity of the single failure. A diagram showing the main co mponents of the hydraulic power unit (HPU) is given in Figure G.B.2-1. Th e Roman numerals used correspond to the equipment numbers in the FMEA.

G.B.2.2.2.2 "Name" and "Failure Mode" Columns

The "Name" column lists the various compon ents by name and, where necessary for identification, the function. The "Failure Mode" column lists, for each entry in the name column, all failure modes which have unique effects. The exception to this is that failure modes which are common to all components (example: leakage) are listed only for the last entry, "ENTIRE CIRCUIT", in the "Name" column.

G.B.2.2.2.3 "Cause" Column The most likely cause(s) for each failure is listed in this column. In addition to those causes listed, nearly all failure mode s could be the result of high earthquake loads. G.B.2.2.2.4 "Symptoms and Local Effects" Column

Effects in terms of variations in hydraulic parameters (temperature, pressure, flow, etc.), operation of other components, reli ability, etc., are listed in this column.

Unless specified otherwise, this column has been based on the assumption that the failure has occurred while the equipment is OPERATIONAL (i.e., controlling the flow control valve).

LSCS-UFSAR G.B.2-2 REV. 13 G.B.2.2.2.5 "Method of Detection" Column The following symbols and de finitions are applicable:

A = Condition is annunciated in the control room.

V = Condition may be detected by visual inspection of electrical control panels and/or hydraulic power units (Note: Inspection herein is defined as visual check of indicators, listening for excessive noise, checking valve handles for proper position, etc.).

U = Condition is undetectable by above methods.

G.B.2.2.2.6 "Inherent Compensating Provision" Column Interlocks and other circuit features which reduce the severity of failure effects (local or system) are listed in this column. Interlocks which might not be tripped or which have no effect are not listed (for ex ample, since a "jammed" actuator cannot move anyway, an interlock which shuts down the equipment when this occurs would not be listed).

G.B.2.2.2.7 "Effect Upon System" Column This column presents the most critical effect on control of the flow control valve for each failure listed. The following symbols and phrases are used:

D = Degradation of transient performance (slow delay times, large overshoots, etc.).

L = Flow control valve position fixed due to initiation of hydraulic "lockup" circuits.

N = No effect on performance or "lockup" capability.

T = Flow control valve position fixed momentarily while operation is transferred to redundant circuits.

"Actuator will not respond....." = Flow control valve fixed in position even though "lockup" circuits are not engaged.

"Oscillation" = High frequency oscillation of flow control valve position.

"Excess Velocity" = Flow control valve position changes at rates exceeding normal limit of electronic controls (11%).

= Indicates sequence of effects. Examples:

LSCS-UFSAR G.B.2-3 REV. 13 T L Indicates that operation will be transferred to redundant circuits but a "lockup" will occur shortly after.

Only qualitative effects are listed herein. Quantitative results are presented in Section G.B.3.3.3.

G.B.2.2.2.8 "Remarks" Column Numbers in column refer to list of remarks at end of tabulation.

Key to "Remarks" Column

1. If redundant circuits are di sabled, effect will be "L".

2. If listed interlock(s) are disabled, flow control valve will not respond to changes in control signal and transfer will be initiated by servo error interlock.

3. If degradation of performance is se vere enough, servo error interlock will initiate a transfer to redundant circuits.

4. If both relief valve and pump co mpensator have failed (double failure) pressure increases or oscillation may be sufficient to cause oscillation of flow control valve.

5. Excess temperature may also occur, in which case, temperature interlock would initiate transfer to redundant circuits.

6. If pressure interlock disabled, degradation of performance will be affected and Comment 3 is applicable.

7. This is an extremely unlikely type of failure due to the construction of this valve. (In order for contaminants to prevent opening, they must first pass through one or more small orifices which, if blocked, cause the valve to fail open.) 8. The only single failure which can totally disable the lockup circuit is a loss of pressure integrity between the actuator piston and the pilot-operated check valve. Since the flow control valve cannot drive itself, additional failures must occur to allow the hydraulic circuit to drive the actuator. One example of simultaneous failures which will do this follows:

LSCS-UFSAR G.B.2-4 REV. 13 a. Electrical controls fail to stop pump.

b. Solenoid valve fails in "A" position or electrical controls fail to de-energize solenoid A.

c. Servovalve fails off center or electronic controls provide erroneous control signal to servovalve.

Other combinations will accomplish the same thing but all require a minimum of three simultaneous failures, at least one of which must be in the electrical control circuits.

9. If servo-error interlock does not function, transfer will not occur.

10. This is an extremely unlikely mode of failure for a check valve.

11. Shuttle valve construction is such that it is virtually impossible for shuttle valve to block pilot line from both subloops simultaneously. (Note: Even if it did, pressure in pilot line would leak off through pilot-operated check valves to allow them to close).

12. If pressure gauge was damaged during initial calibration of equipment or during an overhaul, all components in the affected subloop could be set incorrectly. This would be detectable, however, when transient performance was checked by a pronounced difference in the performance of the two subloops. If, on the other hand, damage occurred during operation, the fact that all pressure devices (pressure switch, relief valve, and pump compensator) appeared to be out of adjustment in the same direction would provide indication of gauge damage.

13. Based on the design of the servovalve, this is the least likely mode of servovalve failure. This equipment utiliz es a jet-pipe pilo t stage which, if plugged by contamination, causes the valve to center. (Conventional flapper-nozzle pilot stages used on most equipment tend to fail "hard-over" as a

result of contamination). Servo "hard-over" can result from a flow controller failure, e.g., short.

14. A servovalve failed "hard-over" is the only single failure which can cause a sustained, excess velocity. This failure is discussed in Section G.B.3.3.3. A number of multiple failure cases exis t which may make the consequences more severe. (Without the servovalve failure to initiate and maintain the velocity, other combinations of malfunctions will not initiate or maintain the excess velocities). Other coincident failures which may increase the severity (i.e., result in higher velocities than would result if only the servovalve failed) are as follows:

LSCS-UFSAR G.B.2-5 REV. 13 a. compensator failure - high;

b. relief valve failure - high and compensator failure;

c. excess pump flow capacity;

d. accumulator precharge low; and

e. loss of pressure integrity.

15. Failure of the servovalve to center and simultaneous failure of the pilot-operated valve to center when operation is transferred to the alternate subloop will result in flow loss from the operational (alternate) subloop and a degradation of transient performance.

16. Leakage will usually be slow enough to allow air to bleed to tank through internal leakage before sufficient amounts accumulate to cause oscillation.

17. If temperature interlock does not function, pump and/or servovalve may be damaged. Also life of all seals and actuator may be reduced.

18. If low-level interlock does not functi on, pressure will be lost and pressure interlock will cause transfer or lockup. Some pump seal damage may occur, however.

LSCS-UFSAR G.B.3-1 REV. 15, APRIL 2004 G.B.3 Flow Control Instrumentation and Controls Failure Modes and Effects Analysis The Failure Modes and Effects Analysis (FMEA) and reliability analysis of the digital Reactor Recirculation Flow Control (RRFC) system is provided in Reference G.B.7-1. NOTE 1 No input signal failure or internal component failure can have an effect more severe than sudden failure of the controller output to an extreme high- or low-value.

Failure in a flow controller affects only one recirculation loop in the same way that a flux controller failure affects both simultaneously, thus a flow controller failure is less severe than a flux controller failure.

NOTE 2 No input signal failure or internal component failure can have an effect more severe than sudden failure of the hydraulic servovalve to admit full hydraulic fluid flow to open or close the recirculation flow control valve. The servovalve "hard-over" failure effect is discussed in Subsection G.3.3.3. A failure in the opening direction may cause neutron flux to peak above the scram setpoint and trip the reactor; otherwise, flux will peak and settle to a normal maxi mum flow operating condition. A failure in the closing direction will cause flow to decrease to a minimum flow operating condition.

LSCS-UFSAR G.B.4-1 REV. 13 G.B.4 Recirculation Pump Trip (R PT) Failure Modes and Effects Analysis

LSCS-UFSAR G.B.5-1 REV. 13 G.B.5 Control Interlocks Failure Modes and Effects Analysis

LSCS-UFSAR G.B.6-1 REV. 13 G.B.6 Power Supplies Failure Modes and Effects Analysis

LSCS-UFSAR G.B.7-1

REV. 15, APRIL 2004 G.B.7 References

1. Westinghouse Report PNTC 01-031, Reactor Recirculation Flow Control (RRFC) System Reliability Analysis

LSCS-UFSAR TABLE G.2.5-1 TABLE G.2.5-1 REV. 14 - APRIL 2002 SYSTEM DESIGN DATA

a. Core thermal power, MW 3489 b. Dome pressure, psia 1020 c. Core flow, 10 6 lb/hr 114 d. Separator inlet quality, % 0.135 e. Core pressure drop, psi 25.4 f. Recirculation pump flow, % of rated 95.3 g. Recirculation pump flow, gpm 44,983 h. Recirculation pump flow, 10 6 lb/hr 17.0 i. Recirculation pump head, ft 830.7 j. Recirculation pump available NPSH, ft 345 k. Recirculation pump required NPSH, ft 108 l. Recirculation pump brake horsepower 8189 m. Flow control valve position, % open 95

n. Flow control valve C v, gpm/psi 6931 o. Valve pressue drop, psi 31.8 p. Downcomer enthalpy, Btu/lb 528.2 q. Downcomer specific volume, ft 3/lb 0.02123 r. Jet pump nozzle velocity, ft/sec 217.2 s. Jet pump M-Ratio 2.356

LSCS-UFSAR TABLE G.2.5-2 SHEET 1 OF 2 G.2.5-2 REV. 15, APRIL 2004 PRESSURES AND TEMPERATURES AT SELECTED CONDITIONS 105. % ROD LINE - VALVES WIDE OPEN NORMAL DOME PRESS. = 1030. PSIA NORMAL DOME TEMP. = 548. F NORMAL FLOW = 15554. GPM JP EFF. = 41.30 JP M RATIO = 2.326 PRESSURE IMPOSED ON PIPING MAY BE UP TO 13. PSI ABOVE NOMINAL TABLES FOR POINTS 5 THROUGH 16. POINT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 W 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 20.00 20.00 20.00 10.00 332.60 665.19 PDOME 6.58 0.45 6.72 5.03 2.74 271.52 271.52 248.87 247.25 245.56 235.43 230.23 226.47 216.91 214.89 45.83 13.34 TDOME - 14.08 -14.09 -14.08 -14.08 -14.09 -12.79 -12.79 -12.82 -12.82 -12.82 -12.83 -12.84 -12.84 -12.86 -12.86 -13.88 1.57 100. % POWER - LOWER FLOW CONTROL LIMIT NORMAL DOME PRESS. = 1020. PSIA NORMAL DOME TEMP. = 547. F NORMAL FLOW = 42845. GPM JP EFF. = 41.06 JP M RATIO = 2.350 PRESSURE IMPOSED ON PIPING MAY BE UP TO 14. PSI ABOVE NOMINAL TABLES FOR POINTS 5 THROUGH 16. POINT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 W 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 20.00 20.00 20.00 10.00 334.95 669.91 PDOME 6.61 1.62 8.50 6.99 4.89 283.89 283.89 222.43 221.00 219.50 209.69 205.08 201.45 192.47 188.56 41.91 11.92 TDOME -14.42 -14.42 -14.42 -14.42 -14.42 -13.09 -13.09 -13.16 -13.16 -13.16 -13.17 -13.17 -13.18 -13.19 -13.19 -14.21 1.42 100. % POWER AND 100. % FLOW NORMAL DOME PRESS. = 1020. PSIA NORMAL DOME TEMP. = 547. F NORMAL FLOW = 42845. GPM JP EFF. = 41.06 JP M RATIO = 2.350 PRESSURE IMPOSED ON PIPING MAY BE UP TO 14. PSI ABOVE NOMINAL TABLES FOR POINTS 5 THROUGH 16. POINT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 W 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 20.00 20.00 20.00 10.00 334.95 669.91 PDOME 6.61 1.62 8.50 6.99 4.89 283.89 283.89 222.43 221.00 219.50 209.69 205.08 201.45 192.47 190.68 41.91 11.92 TDOME -14.42 -14.42 -14.42 -14.42 -14.42 -13.09 -13.09 -13.16 -13.16 -13.16 -13.17 -13.17 -13.18 -13.19 -13.19 -14.21 1.

42 LSCS-UFSAR TABLE G.2.5-2 SHEET 2 OF 2 G.2.5-2 REV. 14 - APRIL 2002

20. % ROD LINE - VALVES WIDE OPEN NORMAL DOME PRESS. = 956. PSIA NORMAL DOME TEMP. = 539. F NORMAL FLOW = 46160. GPM JP EFF. = 37.32 JP M RATIO = 2.622 PRESSURE IMPOSED ON PIPING MAY BE UP TO 13. PSI ABOVE NOMINAL TABLES FOR POINTS 5 THROUGH 16. POINT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 W 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 20.00 20.00 20.00 10.00 362.20 724.41 PDOME 6.36 0.13 6.25 4.51 2.19 267.96 267.96 244.76 243.11 241.38 231.19 225.86 222.08 212.41 208.20 39.27 9.65 TDOME -3.63 -3.64 -3.63 -3.63 -3.63 -2.34 -2.34 -2.37 -2.37 -2.37 -2.38 -2.39 -2.39 -2.41 -2.41 -3.44 1.21 HOT STANDBY - LOW SPEED, FLOW CONTROL VALVE WIDE OPEN NORMAL DOME PRESS. = 947. PSIA NORMAL DOME TEMP = 538. F NORMAL FLOW = 11485. GPM JP EFF. = 38.82 JP M RATIO = 2.528 PRESSURE IMPOSED ON PIPING MAY BE UP TO 1. PSI ABOVE NOMINAL TABLES FOR POINTS 5 THROUGH 16. POINT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 W 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 20.00 20.00 20.00 10.00 352.80 705.59 PDOME 6.36 9.91 20.88 20.77 20.01 36.04 36.04 34.61 34.51 34.40 27.17 26.84 24.17 19.53 19.40 16.85 5.47 TDOME -0.33 -0.33 -0.31 -0.31 -0.32 -0.22 -0.22 -0.22 -0.22 -0.22 -0.23 -0.23 -0.23 -0.24 -0.24 -0.80 0.69

LSCS-UFSAR TABLE G.2.5-3 RECIRCULATION SYSTEM TRIP AND START FUNCTIONS TABLE G.2.5-3 REV. 11 - APRIL 1996 ACTION EVENT A B C D E 1 X X X 2 X X X 3 X X 4 X 5 X 6 X X 7 X X 8 X X X 9 X X 10 X Events 1. Suction or discharge block valves less than 90% open. 2. Vessel high pressure (1135 psig; ATWS)

3. Turbine trip or generator load rejection.

4. 100% speed power supply trip of one of two operating recirculation pumps. 5. 100% speed power supply trip of all operating recirculation pumps. 6. Total feedwater flow less than approximately 20% nuclear boiler rated. 7. Temperature difference between the vessel water saturation temperature and recirculation pump suction temperature is less than 10.1 o F. 8. Pump motor or LFMG set electrical protection logic is activated. 9. Vessel low level (Level 3).

10. Vessel low-low level (Level 2).

Actions A. Normal control trip 100% speed power supply.

B. RPT engineered Safety Class 3 trip of 100% speed power supply. C. Trip of 25% speed power supply.

D. Automatic start of LFMG set during coastdown following a 100% speed trip.

E. No automatic start of LFMG set during coastdown following a 100% speed trip.

LSCS-UFSAR TABLE G.2.5-4 SHEET 1 OF 2 TABLE G.2.5-4 REV. 14 - APRIL 2002 REACTOR RECIRCULATION SYSTEM DESIGN CHARACTERISTICS EXTERNAL LOOPS Number of loops - two Pipe Sizes (nominal OD):

Pipe suction - 24 in. Pump discharge - 24 in.

Discharge minifold - 16 in. Recirculation inlet line - 12 in.

Design Pressure (psig)/Design Temperature (°F)

Suction piping and valve up to and including pump Suction nozzle - 1250/575 Pump, discharge valves, and piping between - 1650/575 Piping after discharge blocking valve up to vessel - 1550/575 Pump auxiliary piping and cooling water piping - 150/212 Vessel bottom drain - 1275/575 OPERATION AT RATED CONDITIONS

Recirculation Pump: Flow - 44,983 gpm Flow- 17.0 x 10 6 lb/hr Total developed head - 831 ft Suction pressure (static) - 1023 psia Required NPSH - 108 ft Water temperature (max) - 535°F Pump brake hp - 8189 Flow velocity at pump suction - 39.5 ft/sec

Pump Motor:

Voltage rating - 6600 Speed - 1780 rpm Motor rating - 8900 Phase - 3 Frequency - 60 Hz Motor rotor inertia - 21,500 lb/ft 2

LSCS-UFSAR TABLE G.2.5-4 SHEET 2 OF 2 TABLE G.2.5-4 REV. 14 - APRIL 2002 Jet Pumps:

Number - 20 Total jet pump flow - 114 x 10 6 lb/hr Throat ID - 6.4 in.

Diffuser ID - 19 in. Nozzle ID (5 each) - 1.30 in. Diffuser exit velocity - 17.1 ft/sec Jet pump head - 98.3 ft

Flow Control Valve:

Type - Ball Material - Austenitic S/S Type actuator - Hydraulic Valve wide open C v - 7770 Valve actuator opening and closing rate (min/max) Percent of stroke per second - 10 to 11 Valve size diameter - 24 in.

Recirculation Loop Block Valves:

Type - Gate valve Actuator - Motor Material - Austenitic S/S Shutoff leakage - 2 cm 3/in/hr Valve size diameter - 24 in.

LSCS-UFSAR TABLE G.2.5-5 PLANT CONFIGURATION DATA TABLE G.2.5-5 REV. 0 - APRIL 1984 FLOW PATH LENGTH (in.) HEIGHT AND LIQUID LEVEL (in.) ELEVATION OF BOTTOM OF EACH VOLUME* (in.) MINIMUM FLOW AREAS (ft 2) A. Lower Plenum 216 216 216 -172.5 71.5 B. Core 164 164 164 44.0 142.0 C. Upper Plenum and Separators 178 178 178 208.0 49.5 D. Dome (Above Normal Water Level) 312 312 0 386.0 343.5 E. Downcomer Area 321 321 321 -51.0 79.5 F. Recirculation Loops and Jet Pumps 108.5 ft (one loop) 403 403 -394.5 132.5 in.

2 _________________________________

Reference point is recircul ation nozzle outlet centerline

LSCS-UFSAR TABLE G.3.3-1 RECIRCULATION SYSTEM EQUIPMENT CALASSIFICATION TABLE G.3.3-1 REV. 15, APRIL 2004 COMPONENT SAFETY FUNCTION SAFETY CLASS SEISMIC CATEGORY FUNCTION External loop piping NA 1 I Pressure integrity Jet pumps Passive 1 I Core reflood height Recirculation Pump Passive 1 I Pressure, Integrity, bearings Recirculation Pump motor

Passive 2 I Bearings only Block valves F023/F067 Passive 1 I Pressure integrity Block valve motor operator Passive NA NA Flow control valve F060 Passive 1 I Pressure integrity Flow control valve actuation:

a. hydraulic actuator; Passive NA NA b. hydraulic lines (inside containment);

Passive NA NA c. hydraulic power unit; Passive NA NA d. control system Passive NA NA Hydraulic line isolation valves Active 2 I Containment isolation RPT breakers and controls Active 3 I Turbine trip All other controls Passive NA

NA ATWS, system starts, trips LFMG Passive NA NA

Digital control components are seismically mounted.

LSCS-UFSAR TABLE G.5.2-1 REV. 11 - APRIL 1996 TABLE G. 5.2-1 REACTOR RECIRCULATION PUMP TRIP FUNCTIONS ACTION EVENT A B C D E 1 X X X 2 X X X 3 X X 4 X 5 X 6 X X 7 X X 8 X X X 9 X X 10 X Events 1. Suction or discharge block valves less than 90% open. 2. Vessel high pressure (ATWS).

3. Turbine trip or generator load rejection.

4. 100% speed power supply trip of one of two operating recirculation pumps. 5. 100% speed power supply trip of all operating recirculation pumps. 6. Total feedwater flow less than approximately 20% nuclear boiler rated. 7. Temperature difference between the vessel water saturation temperature and recirculation pump suction temperature is less than the minimum permissible value. 8. Pump motor or LFMG set electrical protection logic is activitated. 9. Vessel low level (Level 3).

10. Vessel low-low level (Level 2).

Actions A. Normal control trip of 100% speed power supply.

B. RPT engineered Safety Class 3 trip of 100% speed power supply. C. Trip of 25% speed power supply.

D. Automatic start of LFMG set during coastdown following a 100% speed trip.

E. No automatic start of LFMG set during coastdown following a 100% speed trip.

LSCS-UFSAR TABLE G.B.1-1 (SHEET 1 OF 2)

TABLE G.B.1-1 REV. 13 SUCTION AND DISCHARGE BLOCK VALVES FAILURE MODES AND EFFECTS ANALYSIS NO NAME AND FUNCTION FAILURE MODE CAUSE SYMPTOMS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS Suction & Discharge Block Valve 24 Motor Operated Gate Valve. Provides Recirc Pump & Flow Control Valve Isolation During Maintenance Leakage Corrosion Thermal Shock High Stress Due to SSE Improper Installation

Maintenance Depending on the Leakage Rate, a High Drywell Pressure Trip May Occur, Causing Reactor Scram and Initiation of HPCS. Repair or Replace Seal Packing. High Drywell Pressure Alarm Drywell Sump Alarms Cooler Condensate Alarms Drywell Rad Monitors Drywell Area Temperatures Emergency Make-up Systems Activated May Cause Reactor Scram Rupture LOCA Corrosion, Loads SSE, Weld Defect High Drywell Pressure Will Cause Reactor Scram and Initiation of HPCS. Coincident With Above, Low Reactor Water Level (3) Will Cause Trip of Recirc Pumps To Low Speed. At Level (2) Pumps Will Trip Off. Repair Valve and Requalify. Alarm On High Dry- well, Low RX Water Level Emergency Make-up Systems

Activated Reactor Scram LOCA Event Fail Close Logic Error Motor Failure Improper ASM Maintenance Shaft Seizure Unable to Restart Recirc Loop Loop or Plant Shutdown to Repair.

No Effect on Safety as Plant Should be in Hot Standby or Cold Start Mode. Close Position Switch On Valve Loop Unavailable Fail Open Logic Error, Motor Failure, Improper ASM Maintenance, Shaft Seizure Inability to Isolate Recirc Pump and FCV. Loop or Plant Shutdown to

Repair Inability to use shutdown cooling through loop since cannot prevent backflow through the recirc pump.

Open Position Switch-On Valve Open Position Switch-

On Valve Recirc Pump Flow and Head Indication Due to Backflow Low Jet Pump Flow Readings None Shutdown Cooling through Other Loop or through LPCS Lineup None None Both Valves

have to fail in a Loop to Prevent RHR

Operation.

LSCS-UFSAR TABLE G.B.1-1 (SHEET 2 OF 2)

TABLE G.B.1-1 REV. 0 - APRIL 1984 NO NAME AND FUNCTION FAILURE MODE CAUSE SYMPTOMS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS Suction & Discharge Block Valve (Cont) Inadvertent Closure Logic Error Operator Error At Low Speed, Pump Will Trip At 90% Open Limit Switch On Valve.

At High Speed Pump Will Trip At 90% Open Limit Switch On Valve. Position Limit Switch On Valve

Position Limit Switch On Valve Loop or Plant Shutdown if Scram has Occurred.

Loop or Plant Shutdown if Scram has

Occurred Motor Operator Two Minutes to close LSCS-UFSAR TABLE G.B.2-1 REV.0 - APRIL 1984 TABLE G.B.2-1 (SHEET 1 OF 2)

FLOW CONTROL VAVLE FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS Flow Control Valve 24 Ball Type Allows Power Changes by Varying Recirc Loop Flow Leakage Weld Defect Corrosion Seal Defect No Loss of Flow Control. Need to Shut One Loop Down, Isolate with Block Valves and Repair.

High Drywell

Pressure, High Level in Drywell Sump Emergency Make-up Systems Activated.

Loop Un-available Rupture (LOCA) Earthquake SSE High Thermal Stress. Bounded by LOCA Event. Plant Unavailable. High Drywell Pressure, Low Reactor Water Level Emergency Make-up Systems Activated. Plant Shutdown See Discussion in Section G.3.3.3 Opening or Closure Rate

>11%/Sec but

Less than Effective 30% Sec (see Section G.3.3.3) Logic Error Hydraulic Actuator Failure. Reactor Protection System Activated. MCPR Above Safety Limit. High Flux Transient. Reactor Scram. Most Flow Control Single Failures Result in Valve Lockup. Plant Un-available Opening or Closure Rate

>30%/ Sec Effective (see Section G.3.3.3)

Multiple Hydraulic Power Unit (HPU)

Failures. Accident Event. Multiple Failures Required in Hydraulic Power Unit.

High Flux Transient. Reactor Scram. Plant Un-available Multiple HPU Failures Can Result in Instantaneous Velocities Exceeding 60%/Sec. Lockup Logic Error Hydaulic Power

Unit Failure. Valve Locked in Position. Repair Can Be Made While Plant

is Operating. Lockup Annunciated and Displayed.

If Fault Lies in Hydraulic Power Unit, Standby Unit, Will Start.

Power Change By Control Rod Movement and Flow Control Using Remaining Recirculation Loop.

Valve Will Automatically Lockup on Abnormal Control Signal and HPU Malfunction LSCS-UFSAR TABLE G.B.2-1 REV.0 - APRIL 1984 TABLE G.B.2-1 (SHEET 2 OF 2)

FLOW CONTROL VAVLE FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS Seizure Wear Corrosion Seizure Could Result in High Actuator Loads Leading to

1) Link Failure 2) Level Shaft Key Failure 3) Shaft-Ball Shrink Fit Failure Loss of Flow Control.

Displayed and Annunciated if Flow Controller Demands Movement.

After Detection, Required to Shut Loop

HPU Down Flow Control with One Loop Remaining LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 1 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS I Electric Motor Will Not Run Power Loss, Electrical Fault, Mechanical Fault Loss of Pressure V, A Overload/Undervolt and Pressure Interlocks T 1, 2 II Coupling Decouples Mechanical Fault Loss of Pressure V, A Pressure Interlock T 1, 2 III Pump A. Pressure Control Set High or Compensator Jammed at Maximum Position Incorrectly Adjusted, Loss of Adjustment Contamination High Pressure V Relief Valve Limits Pressure to 110% Normal D 3, 4, 5, 14 B. Pressure Control Set Low, But Greater Than Interlock Set Point Incorrectly Adjusted, Loss of Adjustment Low Pressure V None D 3 C. Pressure Control Set Less Than Interlock Set Point Same as III.B Low Pressure V, A Pressure Interlock T 1, 6 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 2 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS D. Pressure Compensator Sticking or Fully Jammed in Mid Position Contamination and/or Wear Pressure Oscillations, or High Pressure with FCV Stationary, Low with FCV in Motion V Relief Valve Limits Peak Pressures, Pressure Interlock Will Trip if Oscillations Are Severe Enough D 3, 9, 14, 5 E. Excess Flow Capacity Incorrectly Adjusted or Mechanical Fault or Loss of Adjustment None U Compensator Reduces Flow to Required Value (i.e., Value Demanded by Servo Valve)

N 14 F. Inadequate Flow Capacity Sufficient to Maintain Pressure Incorrectly Adjusted, Loss of Adjustment, Internal Leakage Due to Wear None While Actuator is Stationary, More Extreme Pressure Decrease Transients When Actuator is in Motion U None D 3 G. No Flow Mechanical Failure Pressure Loss V, A Pressure Interlock T 1, 2 IV Pilot Operated Check Valve (Either) A. Fails Closed Mechanical Fault or Contamination Fluid Trapped in Actuator A None Actuator Will Not Respond to

Changes in Position Demand Signal (One Direction Only)

T L

LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 3 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS B. Fails Open Mechanical Fault, Contamination None During Normal Operation. Fluid Not Trapped in Actuator During Lockup U Pilot Operated 4-Way Valves Trap Fluid During Lockup. Also, Pumps Stop During Lockup and FCV Cannot Drive Itself N 8 V Relief Valve A. Failed Closed Contamination Increased Pressure Spikes During Transients, Reduced Reliability Due to Pressure Spikes U Steady State Pressure Maintained at Normal Value by Pump Compensator N 4, 7, 14 B. High Setting Improperly Adjusted; Loss of Adjustment Same as V.A., But Less Severe U Same as V. A. N 4, 14 C. Low Setting But Greater Than Interlock Setpoint

Incorrectly Adjusted; Loss of Adjustment Low Pressure V None D 3, 5 D. Low Setting and Less Than Interlock Set-Point

Incorrectly Adjusted; Loss of Adjustment Low Pressure V, A Pressure Interlock T 1, 6, 5 E. Failed Open Contamination, Mechanical Fault Loss of Pressure V, A Pressure Interlock T 2, 1 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 4 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS VI 4-Way Valve, Sol. Op.

A. Fails Centered or at "B" Position Loss of Control Signal, Incorrect Signal, Coil Failure, or Contamination (Start-up Only for Last Cause)

Pilot-Operated Check Valves and Pilot Operated 4-Way Valves Close, Locking Actuator A Servo Error Interlock Actuator Does not Respond to Control Signals

T 1, 9 B. Fails at "A" Position Incorrect Control Signals or Contamination None While Subloop is Operational. Delays Speed of Subloop Isolation During Transfer to Alternate Subloop, Delays Actuator Lockup During Shutdown U During Transfer or Shutdown, Subloop's Pump is Stopped. Pilot Operated Check Valves and Pilot Operated 4-Way Valve Close Upon Pressure Loss Providing Lockup N 8 VII Pilot Operated 4-Way Valve A. Fails Centered (Closed) Contamination Loss of Adjustment, or Improper Adjustment Servo Valve Isolated From Actuator A Servo Error Interlock Actuator Does Not Respond to Control Signals

T B. Does Not Center (i.e.,

Remains in "X" Position or Travels Past Center to "Y"

Position).

Contamination Loss of Adjustment, or Improper Adjustment. No Effect when Subloop is Operational. Servo Valve Not Isolated from Interconnecting Piping when subloop is Not Operational. U Servo Valve is Centered (Closed) When Subloop is Not Operational and Thus Acts to Isolate Remainder of Subloop. Pilot Operated Check Valves Still Provide Actuator Lockup Function. N 8, 15 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 5 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS VIII Check Valve, Pressure Line A. Fails Closed Contamination Loss of Pressure V, A Pressure Interlock T 1, 2, 10 B. Fails Open Contamination No Effect When Subloop is Operational. Allows Potentially Damaging Backflow from Accumulator to Pass Through Filter and Pump Whenever Pump Stops. V (During Shut Down) None N IX Servo Vale A. Reduced Gain Open Circuit in One Coil, Contamination Decreased Flow Through Servo Valve for any Specific Control Signal U None D 3 B. Fails Centered Contamination Loss of Cont. Signal, Short in Coil, or Internal Mechanical Fault. Servo Valve Flow does not Respond to Control Signals. A Servo Error Interlock Actuator Fails to Respond to Control Signals Actuator May "Drift" Slowly From Desired Position.

T 9, 1 C. Excessive Null Shift Loss of Adjustment, Improperly Adjusted Internal Mechanical Fault. Servo Valve Partially Open With No Signal, Responds More to Signals of One Po larity than the Other. U None D 3, 8, 15 D. Oscillates Loose or Broken Internal Parts.

Pressure Oscillations, Noise, Vibration. V None Oscillation LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 6 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS E. Fails Hard Over Contamination Internal Fault, Incorrect Control Signal. Full Pump and Accumulator Flow Directed to Actuator.

A Servo Error Interlock. Also, sizing and Adjustment of Other Components Limits Velocity.

Excess Actuator Velocity T 9, 13, 14 8, 15 F. Degradation of Flow Characteristics and

or Excess Null Leakage. Erosion Or Scoring of Spool and Sleeve. Increased Non-Linearities in Flow Gain. U None D 3 X Shut-Off Valve (Manual), Servo Valve Isolation A. Failed Open Contaminants Stem Broken No Effect During Operation, Subloop Should not be Maintained While Other Subloop

is Operating.

U (While Operating) None N B. Failed Closed Inadvertently Operated. Subloop Isolated From Portion of Field Piping V Servo Error Interlock Actuator Will not Respond to Control Signals.

T 9, 1 XI Check Valve, Servo Valve, Drain A. Failed Open Contamination Broken Spring No Effect During Operation. Fluid Loss from Reservoir During Servo Valve Replacement.

U Line is Small and Easily Plugged for Maintenance.

N B. Failed Closed Contamination Servo Valve will Fail Centered and Possibly Be Permanently Damaged. Possible Loss of Pressure Integrity in Low-Pressure Circuits

A Servo Error Interlock Actuator Fails to Respond to Control Signals.

T 9, 10, 1 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 7 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XI Check Valve, Servo Valve Drain A. Failed Open Contamination Broken Spring.

No Effect During Operation. Fluid Loss from Reservoir During Servo Valve Replacement.

U Line is Small and Easily Plugged for Maintenance.

N B. Failed Closed Contamination Servo Valve will Fail Centered and Possibly Be Permanently Damaged. Possible Loss of Pressure Integrity in Low-Pressure

Circuits A Servo Error Interlock Actuator Fails to Respond to Control Signals.

T 9, 10, 1 XII Check Valve, Pump Drain Filter Bypass A. Fails Open Contamination Broken Spring. Pump Drain Flow Bypass Fuller's Earth Filter. Reliability Decreased by Loss of Filtration. U Protection Against Particulate Contamination Provided by Other Filters. Introduction of Water and Chloride is Minimal if Good Operational Practices are followed.

N B. Failed Closed Contamination Pump Drain Flow Cannot Bypass Fuller's Earth Filter. May Result in Excess Pressure Damaging Pump Shaft Seal and/or Fuller's Earth Filter.

U High Pressure Across Element is Alarmed to Protect Element of Filter From Damage.

N 10 XIII Check Valve, Heat Exchange Bypass A. Failed Open Contamination Broken Spring. Flow Bypasses Heat Exchanger, May Result in Excess Temperature. Will also Result in Reservoir Fluid Loss During Maintenance of Return Line Components.

V, A Temperature Interlock T 1, 17 B. Failed Closed Contamination Excessive Pressure at Inlet to Temperature Control Valve.

Potentially Damaging to Temperature Control Valve, Return and Drain Filters, and Pump Shaft Seal. Possible Loss of Pressure Integrity in Low Pressure Circuits.

A Will Result in Drain Filter Alarm, However; Damage Could Occur Too Rapidly for Alarm to be of Use.

N 10 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 8 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XIV Check Valve, Heat Exchanger Outlet A. Failed Open Contamination Broken Springs No Effect During Operation, Loss of Fluid From Reservoir During Maintenance of Return Circuit Components.

U None N B. Failed Closed Contamination Flow Bypasses Heat Exchanger. May Cause Excessive Temperature V, A Temperature Interlock T 10, 1, 17 XV Shuttle Valve A. Fails to Shuttle (Can Only Occur on Startup or Transfer) Fully or Partially in One Direction Only. Contamination, Mechanical Fault Pressure from Operational Subloop Blocked or Partially Blocked from Pilot Line Vented to Tank Via Alternate Subloop Resulting in Closure of Pilot Operated Check Valves. A Servo Error Interlock Actuator Fails to Respond to Control Signals

T 9, 11, 1 B. Fails to Shuttle Fully in Either

Direction (Can Only Occur on Startup or Transfer).

Contamination, Mechanical Fault Same as XI.A. A Does Not Prevent Lockup Since Pilot Line is Vented.

Actuator Fails to Respond to Control Signals.

T

L 11 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 9 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XVI Needle Valve Used as Bypass Valve A. Failed Closed Handle Broken Off.

No Effect During Operation. Maintainability Degraded.

U During Operation None N B. Failed Open Contamination, Improperly Adjusted, Loss of Adjustment

Pump Flow Bypassed to Return V None D 3 XVIa Needle Valve Used as Gauge Shut Off Valve A. Open Too Far. Loss of Adjustment, Improperly Adjusted.

Excessive Shock and Vibration at Pressure Gauge. Reliability of Gauge Degraded.

V None N B. Closed Loss of Adjustment, Improperly Adjusted Pressure Gauge Isolated from Circuit V None N XVIb Needle Valve Used as Pressure Switch Shutoff Valve. A. Failed Open. Contamination No Effect During Operation.

Maintenance of Pressure Switch Less Convenient.

V None N B. Failed Closed Before Subloop Pressurized

Improperly Adjusted Pressure Interlock will be Tripped. A None T 1 C. Failed Closed After Subloop Pressurized

Loss of Adjustment Pressure Interlock Disabled. V None N XVII Ball Valve Used as Suction Line Shutoff Valve A. Failed Closed Inadvertent Operation. Improper

Setting. Pump Suction Isolated From Reservoir. Loss of Pressure.

Possible Pump Damage.

A, V Pressure Interlock T 1, 2 B. Failed Open Contamination, Stem Broken No Effect During Operation. Reservoir Must be Drained to Service Pump.

V None N LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 10 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XVIIa Ball Valve Used as Drain Shutoff Valve A. Failed Closed Inadvertent Operation. Improper Setting Subloop Reservoirs Isolated from Each Other and All Actuator Drain Flow Directed to other Reservoir.

Eventual Low Fluid Level in Reservoir.

V, A Low Level Interlock. Also, Reservoirs Sized such that Interlock will be Tripped Prior to Over Filling of Alternate Reservoir.

T 1, 18 XVIII Temperature Control Valve A. Fails Open Contamination, Broken Spring Return Fluid Always Flows Through Heat Exchanger. Low Temperature

V None D 3 B. Fails Closed Leakage Vapor in Capillary, Pinched

Capillary.

Flow Blocked from Heat Exchanger. High Temperature V, A Temperature Interlock T 1, 17 XIX Heat Exchanger A. Does Not Cool Fluid. Fouling Air or Fluid Flow Blocked.

High Temperature V, A Temperature Interlock T 1, 17 XX Accumulator A. Pre-Charge High Improper Adjustment Less Fluid Stored in Accumulator. More Severe Pressure Spikes During Transients, However, Negligible Effect on Life or

Reliability.

U None D 3, 14 B. Pre-Charge Low Improper Adjustment or Leakage More Fluid Stored in Accumulator.

U None D 3 C. Leakage of Pre-Charge to System Fluid (Hydraulic).

Faulty Bladder Oil Column Stiffness Reduced, Pressure Oscillations, Noise V Equipment Design is such that Small Amounts of Air Will be Bled Back to Tank Before They Can Initiate Oscillation.. Oscillation 16 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 11 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XXI Filter, Pressure Line Bypassing Fluid Faulty Bypass Valve, Torn Element, Dirty Element. Reduced Reliability Due to Lack of Filtration. (Primarily Effects Servo Valve.)

A (If Due to Dirty Element or Faulty Bypass) None (If Due to Torn Element)

Return and Fuller's Earth Filters Provide Some Protection (Sump Strainer

Also). N XXII Strainer A. Blocked Dirty Loss of Pump Output Pressure or Erratic Output Pressure If Blockage is Extreme. Otherwise, Excessive Noise from Pump.

V, A, If Extreme Pressure Interlock (If extreme) T 1, 6 B. By-passing Fluid Torn Reduce Protection for Pump, Thus, Reduced Pump Reliability. None Primary Control of Contamination Provided by 10µ Filters.

N XXIII Filter, Fuller's Earth, Pump Drain Line Blocked (Causes Fluid to Bypass Via Check Valve) Dirty Element, Contamination in

Orifice. Reduced Reliability Due to Loss of Protection Against Chlorides, Water, Etc. A, V if Element is Dirty, N One

if Orifice is Blocked Protection Against Particulate Contamination Provided by Other Filters. Introduction of Water and Chlorides is Minimal if Good Operational Practices Followed.

N XXIV Pressure Switch, Pressure Line A. Fails Open Electrical Fault, Broken, Spring, Jammed Actuator, Loss of Adjustment

Faulty Indication of Pressure Loss and Initiation of Interlock Action. V, A None T 1 B. Fails Closed Electrical Fault, Jammed Actuator, Loss of Adjustment.

Disables Pressure Interlock and Alarm. U, V (When Subloop is Not

Operational). None N LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 12 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XXV Pressure Switch, Drain Filter A. Fails Open Electrical or Mechanical Fault.

"Drain Filter Dirty" Alarm Disabled Thus Reduced Reliability.

V (When Filter is Dirty)

Bypass Check Valve Protects Filter From Damage. N B. Fails Closed Electrical or Mechanical Fault.

Trips "Filter Dirty" Alarm. V None N XXVI Temperature Switch A. Fails Open Electrical or Mechanical Fault. Disables Temperature Alarm and/or Interlock. V if Temperature is High Otherwise, U.

None N 17 B. Fails Closed Electrical or Mechanical Fault. Trips Temperature Alarm and/or Interlock V None T 1 XXVII Level Switch A. Fails Open Electrical or Mechanical Fault. Disables Fluid Low Level Alarm and/or Interlock V if Level is Low, Otherwise, U.

None N 18 B. Fails Closed Electrical Fault or Hole in Float. Trips Low Level Interlock and/or alarm V, A None T 1 XXVIII Flow Switch A. Fails Open Electrical or Mechanical Fault. Actuator Drain Flow Alarm Disabled.

U None N B. Fails Closed Electrical or Mechanical Fault. Actuator Drain Flow Alarm Tripped. U None N XXIX Sight Glass A. Glass Broken Someone Hit It. Fluid Lost from Reservoir V, A Sight Glass Located High Enough to Limit Fluid Loss N XXIXa Thermometer B. Incorrect Reading Incorrect Visual Reading, No Effect Since Device is Not Used to Adjust Components

V if Error Significant None N LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 13 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XXX Filler A. Leaks Loose Increased Chance for Contamination U Filters Protect Against Contamination.

N XXXI Pressure Gauge Incorrect Reading Zero Shift, Failed Movement, Deformed Bourden Tube. No Effect Unless Components Are Readjusted Without Checking Gauge Calibration.

See Remarks See Remarks N 12 XXXII 4-Way Valve, Manual A. Failed in Center or Unused Position Inadvertent Operation of Valve. Vents Pilot Line Causing Pilot Operated Check Valves to Close. V, A Servo Error Interlock Actuator Will Not Respond to Control Signal.

T 1, 9 B. Failed in Operational Position Contamination Broken Stem. No Effect During Operation. Equipment Should not be Maintained While Alternate Subloops Operational.

V None N XXXIII Filter A. Bypassing Dirty Fluid Faulty Bypass, Torn or Dirty Element. Reduced Reliability Due to Lack of Filtration. V Other Filters Provide Sufficient Protection, Thus, Reliability is minimal.

N XXXIV Filter Reservoir Saturated for Damaged Element. Poor Maintenance Allows water Vapor and/or Particulate Contamination In Reservoir, Reducing Reliability.

U System Filters Will Protect Critical Components.

N LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 14 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS XXXV Hydraulic Cylinder A. Excess Friction or "Stiction".

Worn Bearings, Contamination Greater Differential Pressure Required to Drive Actuator.

None None D 3 B. Actuator Totally Jammed Worn Bearings Seized Actuator Will Not Move A None Actuator Will Not Respond to

Changes in Control Signals.

T L XXXVI Entire Circuit A. Internal Leakage

Between High- and Low- Pressure Circuits

or Internal Leakage Between Actuator Extend and Retract Lines. 1. Worn Valve, Worn Seal, Etc. Decrease in Flow Available to Drive Actuator. U None D 3 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 15 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS B. Restriction Between Pump and Servo Valve, Servo Valve and Actuator, or Servo Valve and Tank.

Partially Closed Valve. Decrease in Differential Pressure Available to Drive Actuator U None D 3 C. Restriction Between Pump and Pilot Circuits. Partially Closed Valve. Increase in Time Required to Unlock Actuator. No Effect During Operation But May Cause a Transfer or Shutdown During Startup.

U (If Operating) V, A (If Startup Affected). None N (If Operating L (During Startup).

D. Restriction Between Pilot Circuits and Tank. Partially Closed Valve. Increased Time Required to "Lock" Actuator. U Pump (s) Are Stopped Whenever Lockup is Initiated. After Pump is Stopped, Normal Internal Leakage Will Cause Pressure to Decay Allowing Springs to Close Pilot op. Check Valves and 4-Way Valves.

N 8 E. External Leakage From

Redundant Circuits Seal Worn, Loose Fitting, Etc.

Earthquake Low Fluid Level V, A Low Level Interlock T 1, 18 LSCS-UFSAR TABLE G.B.2-2 REV.0 - APRIL 1984 TABLE G.B.2-2 (SHEET 16 OF 16)

FLOW CONTROL VALVE ACTUATOR AND HYDRAULIC POWER UNIT FAILURES MODES AND EFFECTS ANALYSIS

NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS F. External Leakage From Common Circuits.

Seal Worn, Loose Fitting, Etc. Earthquake Low Fluid Level. V, A None T L G. Loss of Pressure Integrity in Redundant

Circuits. Rupture of Component or Piping. Earthquake Low Fluid Level, Possibly Low Pressure Also, Flow and/or Pressure May be Unavailable to Drive Actuator or Maintain Pilot Valves in Open Position. V, A Low Level Interlock, Low Pressure Interlock, Servo Error Interlock. Actuator May Not Respond to Changes in Control Signal.

T 1, 18, 14 H. Loss of Pressure Integrity in Common Circuits. Rupture of Component or Piping Earthquake Low Fluid Level, Possibly Low Pressure Also, Flow and/or Pressure May be Unavailable to Drive Actuator or Maintain Pilot Valves in Open Position. If Rupture Occurs in Open and Closed Lines, Less Restriction in Circuit. V, A None Actuator May Not Respond to Changes in Control Signal.

T

L 8, 14 I. Air Leakage Into Circuit Loose Suction Line Fitting, Worn Pump Shaft Seal, Etc. Earthquake Loss of Fluid Stiffness, Noise, Pressure Fluctuations. V Circuit Design is Such that Pump Suction is Usually Flooded (No Vacuum) to Prevent Occurrence. Also, Small Volumes of Air Will be Bled Back to Tank Before Problems are

Caused. Oscillation 16

LSCS-UFSAR TABLE G.B.3-1 REV 15, APRIL 2004 TABLE G.B.3-1 FLOW CONTROL INSTRUMENTATION AND CONTROLS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTOMS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 Flow Control (A or B Loop) (Ganged Control) Output Signal

High/Low Electrical/Electronic Failure See Note 1 Recirc System Instrumentation 11% Rate Limiter by Valve Positioning Electronics Flow Increase or Decrease 2 Valve Positioning Electronics Drive Hydraulic Servo Valve to Extreme Position Electrical/Electronic

Failure See Note 2 Recirc System Instrumentation None Flow

Increase, Decrease, or Lockup LSCS-UFSAR TABLE G.B.4-1 REV.0 - APRIL 1984 TABLE G.B.4-1 (SHEET 1 OF 4)

RECIRCULATION PUMP TRIP (RPT) FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 Sensor Fuse F8A (Typ for F8B, C or D)

Fail Open Broken Wire, Loose Connection, Etc. Fuse Open Without Short Circuit or Overcurrent. Relay K8A Deenergized. Failure in Trip (Safe) Direction. RPT is 1:1 Instead of 2:2 for Turbine Control Valve Fast Closure Trip.

Annunc & Comp Fail-safe Logic Design Negligible 2 Turbine Control Valve Fast Closure Sensor C71-N005A (Typ for N005C, N005B, N005D) Fail Closed

Fail Open Mechanical Failure, Contact Shorted or Fused, Etc.

Broken Wire, Loose Connection, Etc. Relay K8A Engergized Contact 11-12 Open, Turbine Control Valve Trip is Inhibited.

Relay K8A Deenergized Fail-Safe, Contact 11-12 Closed. RPT is 1:1 Instead of 2:2 for Turbine Control Valve Fast Closure Trip Periodic Test

Annunc & Comp Redundant RPT System B Trip is

Fail-Safe Logic Design Minor, Pump Trip is 1:1

Negligible 3 Sensor Fuse F9A (Typ for F9B, C, or D)

Fail Open Broken Wire, Loose Connection, Etc. Fuse Open Without Short Circuit or Overcurrent Relay K9A Deenergized RPT Bypass is 1:1 Instead of 2:2 for Reactor Power Above 30% Trip Permissive

Annunc & Comp Fail-Safe Negligible 4 Reactor Power Sensor C71-N003A, (Typ for N003B, C, or D)

Fail Closed

Fail Open Mechanical Failure, Contact Shorted or

Fused, Etc.

Broken Wire, Loose Connection, Etc. Relay K9A Energized, Contact 11-12 Open, Reactor Power Trip Permissive is Inhibited RPT is 1:2 Instead of 2:2

Relay K9A Deenergized, Contact 11-12 Closed, RPT Bypass Failure in Safe Direction

Periodic Test is Available

Annunc & Comp Redundant RPT System. B Trip is

Available

None Minor, Pump Trip is 1:1

Negligible 5 Sensor Fuse F10A (Typ for F10E, C, G, B, F, D, or

H) Fail Open Broken Wire, Loose Connection, Etc. Fuse Open Without Short Circuit or Overcurrent, Relay K10A Deenergized RPT is 1:2 Instead of 1:2 Twice for Turbine Stop Valve Trip; Failure in Safe Direction

Annunc & Comp Fail-Safe Logic Negligible LSCS-UFSAR TABLE G.B.4-1 REV.0 - APRIL 1984 TABLE G.B.4-1 (SHEET 2 OF 4)

RECIRCULATION PUMP TRIP (RPT) FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 6 Turbine Stop Valve Sensor (C71-N006A) (Typ for N006E, C, G, B, F, D, or H)

Fail Closed Fail Open Mechanical Failure, Contact Shorted or Fused Etc.

Broken Wire, Loose Connection, Etc.

Relay K10A Energized, Contact 9-10 Open, Turbine Stop Valve Trip is 1:1 and 1:2 Instead of 1:2 Twice

Relay K10A Deenergized Contact 9-10 Closed, Stop Valve Trip is 1:2 and 2:2 Instead of 1:2 Twice; Fail-Safe.

Periodic

Test Annunc & Comp Redundant RPT System B Operational

Fail-Safe Logic Design Negligible

Negligible 7 Logic Fuse F31A, (Typ for F31B, F32A, F32B)

Fail Open Broken Wire, Loose Connection, Etc. Fuse Open Without Short Circuit or Overcurrent RPT is 1:1 Instead of 1:2 BKR Trip Coil Status Light is Not on Bus B Trip is Available Minor Pump Trip is 1:1 8 Test Bypass Switch S13A (Typ for S13B)

Fail Closed In Bypass Position Fail Open in

'Normal' Position Mechanical Failure, Contact Shorted or

Fused, Etc.

Mechanical Failure, Loose Connection, Etc. Normally Open Contacts Closed in 'Bypass' Position

Contacts Open, RPT System A is Disabled Periodic Test

Annunc Logic Design

Redundant RPT System B Trip is

Available Both Recirc Pumps Might Trip Surveillance Test of Logic Minor, Pump Trip is 1:1 9 Diode CR3A (Typ For CR3B, CR4A, CR4B)

Fail Shortened

Fail Open Mechanical Failure, Fused, Etc.

Broken Wire, Loose Connection, Etc. No Symptoms

Diode Open, Pump BKR CB3A Cannot Trip, System A is Disabled Periodic Test

Periodic Test (Shutdown)

Redundant RPT System B Trip is Available No Safety Impact - If Both Diodes Short, Will Not Be Able To Ascertain Individual Trip Coil Continuity. RPT Still Operative.

Minor, Pump Trip is 1:1 LSCS-UFSAR TABLE G.B.4-1 REV.0 - APRIL 1984 TABLE G.B.4-1 (SHEET 3 OF 4)

RECIRCULATION PUMP TRIP (RPT) FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 10 RPT Aux Relay K30A (Typ for K30B) Contacts

RPT Relay K30A Coil Fail Closed Fail Open Fail Short Fail Open Mechanical Failure, Contacts Fused, Etc

Broken Wire, Loose Connection, Etc..

Contacts Closed, Isolated Signals for Start Permissive to LFMG Set Start Circuit

Contacts Open, LFMG Auto Start Circuit Inhibited

On Trip, Fuse Will Blow, Disabling System A.

No Annunciation, Computer Indication.

Computer

Periodic Test

Shutdown Test Only Shutdown Test None

None Redundant RPT System B Operative

Redundant RPT System B Operative None

None

11 Circuit Breaker Status Indicating Lamp DS27A (Typ for DS27B, DS28A, DS228B) Fail Shorted

Fail Open Mechanical Failure, Contacts Shorted, Etc.

Mechanical Failure, Loose Connection, Etc. Circuit Breaker CB3A Trips, Reactor on One Loop Flow

Loss of CB3A Indication Status of

RPT Trip Coil Speed Indication

Decreased Reactor Power Redundant Indicating Light None

None None

Negligible Extremely Remote Postulated

Failure 12 Circuit Test Lamp DS25A (Typ for DS25B, C, D, DS26A, B, C, D) Fail Closed Fail Open --

Mechanical Failure, Loose Connection, Etc. --

Loss of Periodic Logic Test Indication

--

Periodic Test Redundant Trip System B Operative

None --

None LSCS-UFSAR TABLE G.B.4-1 REV.0 - APRIL 1984 TABLE G.B.4-1 (SHEET 4 OF 4)

RECIRCULATION PUMP TRIP (RPT) FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 13 RPT Circuit BKR CB3A (Typ for CB3B, CB4A, CB4B)

Fail Closed Fail Open Mechanical Failure, Loose Connection, Etc.

Mechanical Failure Hot Short To Trip Coil CB3A Does Not Trip Open.

RPT System A is Defeated.

CB3A Trips Without Trip Signal, Reactor on One Loop Flow Indicating Lights Speed Indication.

Decreased Reactor Power Redundant System B Trip

is Available

None Minor, Pump

Trip is 1:1

None 14 Trip Channel A1, A2, Sensor 120 VAC Power Supply (Typ for Channel B1, B2)

Fail Deenergized Feeder Breaker Open, Loose Connection, Etc. Turb Cont Valve Trip is 1:2 Instead of 2:2 for BUS A&B: Turb Stop Valve Trip is 1:2 Instead of 1:2 Twice for BUS A&B; 30% Power Bypass is 1:1 Instead of 2:2 for BUS A&B. Annunc & Comp None Logic BUS is in Half Trip for BUS A&B LSCS-UFSAR TABLE G.B.5-1 REV. 0 - APRIL 1984 TABLE G.B.5-1 (SHEET 1 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 A Logic Circuit Reactor Low-Power LMFG Auto Transfer Sequence (Typical for B Logic Circuit) A. F067A (F023A Limit Switch Stack 10-10C Fails Closed and 11-11C Fails

Open with Valve Closed Limit Switch Stack Mechanical Allows Recirc Pump Start with Discharge (Suction) Valve Closed Pump Energy Will Heat Up Dead Headed Recirc Water Flow Indicator If 10-10C Only Fails, Pump Will Trip on Lockout When

Breaker Closes Possible Pump Damage F067A (F023A) Limit Switch Stack 10-10C

Fails Open with Valve Open Stack Mechanical Failure, Open

Circuit Prevents Recirc Pump Start or Transfer to High Speed Failure to Start as Seen on Ind Lights

Speed Ind Limits Capacity F067A (F023A) Limit Switch Stack 11-11C

Fails Closed with Valve Open Stack Mechanical Failure, Short

Circuit Recirc Pump Breaker CB3A Trips Alarm, Recirc Pump Motor-A Trip Limits Capacity B. FCV F060A Min Position Limit Switch

Fails Closed with Valve Full Open Short Circuit, Mechanical

Failure Allows Recirc Pump to be Started to 25% Speed or 100% Speed with the Flow Control

Valve Fully Open Valve Position Indicator, Loop Flow Ind. Operating Procedures Require Closing Valve Before

Initiating Start Sequence Possible Time Delay Overcurrent Trip Reactor Subjected to Flow Increase

Transient FCV F60A Min Position Switch

Fails open with Valve Closed to Min Position

Open Circuit, Mechanical

Failure Prevents Recirc Pump Start or Transfer to High Speed Speed Indication, Ind. Lights Limits Capacity LSCS-UFSAR TABLE G.B.5-1 REV. 0 - APRIL 1984 TABLE G.B.5-1 (SHEET 2 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 C. K103 Relay Fails Closed When Control Power is not Available to

CB3A Short Circuit on Contacts Recirc Pump Gets Start Signal, LFMG Set Starts But Main Breaker Fails to Close. During Low to High Speed Transfer, LFMG Set Trips But Main Breaker Fails to Close Ind. Lights Speed Ind.

Alarm - Incomplete Sequence Limits Capacity This Occurrence During Low to High Speed Transfer Results in Recirc Pump Trip from 25%

Speed K103 Relay Fails Open When Control Power is Available to CB3A Open Circuit to Coil or Contacts Prevents Recirc Pump Start Transfer to High Speed Ind. Lights, Speed Ind. Alarm -Circuit Not

Available Limits Capacity D. Contacts K138A, 4-8 Fail to Open on Incomplete Sequence Short Circuit Allows Operator to Initiate Start/Transfer Sequence Without First Resetting Incomplete Sequence Logic After an Incomplete

Sequence Start Breaker Protective Relays, Operator Action Possible Pump Motor Over-heating from

Excessive Starts Operator Would Have to Ignore Incomplete Sequence Alarm Contacts K138A, 4-8 Fail Open Open Circuit Prevents Recirc Pump Start or Transfer to High Speed Ind. Lights Speed Indicator Limits Capacity E. 1, 1T Contacts of S101A Control Switch Fail Closed Short Circuit Recirc Pump Starts or Transfer from Low to High Speed Occurs Providing all Other Interlocks are Satisfied Unexpected Pump Start or

Low High Speed Transfer Operating Procedures Required Operator to Trip Pump if Loop Flow Imbalance Occurs. 1, 1T Contacts of S101A Control Switch Fail Open Open Circuit Prevents Recirc Pump Start or Transfer to High Speed LSCS-UFSAR TABLE G.B.5-1 REV. 15 - APRIL 1984 TABLE G.B.5-1 (SHEET 3 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 F. K111A Remains Energized When Control Power

is not Available to LFMG Set 1) Motor Breaker 2) Gen Breaker Short Circuit Attempted Start of Recirc Pump Results in Pump Accelerating to 100% Speed as LFMG Set Fails to 1) Start 2) Close in at 25% Speed Speed Ind. Ind. Lights Alarm - Recirc Pump Motor A Auto Transfer Incomplete Alarm - Circuit

Not Available Minor Flow Increase/Decrease Transient

<30%/Second LSCS-UFSAR TABLE G.B.5-1 REV. 15 - APRIL 2004 TABLE G.B.5-1 (SHEET 4 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 2 A Logic Circuit Reactor Normal Power Level Low Speed to High Speed Transfer (Typical for B Logic Circuit) A. Contacts 3-4 of K113A Circuit Fail Closed Short Circuit This Seal-Out Contact Prevents Sustained High Speed Operation When Recirc Pump is Started to 25% Speed and is Redundant to the digital Reactor Water Level Control System The Digital Reactor Water Level Control System Loss of Redundancy B. Contacts 3-4 of K128A Circuit Fail

Closed Short Circuit Allows Attempted High Speed Transfer with Low Feedwater Flow. Trips LFMG Set and Pump Speed to Zero. Operating Procedures Prevent Transfer Low FW Flow Indicating Light. Alarm Auto Trip or Transfer to

Low Speed 5-6 Contacts of K128A Will Trip Pump on Low FW Flow if Operator Error Initiates Transfer Failure of K128A to Energize on Low FW Flow May Result in Cavitation Without Pump Trip C. Contacts 3-4 of K129A Short Circuit Allows Attempted High Speed Transfer With Vessel Low Level. Trips LFMG Set and Pump Speed to zero. Operating Procedures Prevent Transfer

Low Level Indicating Light Alarm

- Auto Trip or Transfer to Low Speed 5-6 Contacts of K129A Will Trip Pump on Low Level if Operator Error

Initiates Transfer Failure of K129A to Energize on Low Level May Results in Cavitation Without Pump Trip D. Contacts 3-4 of K130A Short Circuit Allows Attempted High Speed Transfer With Low Steamline to Recirc Loop T. Trips LFMG Set and Pump Speed to Zero.

Operating Procedures Prevent Transfer Low T Indicating Light Alarm

- Auto Trip or Transfer to Low Speed

5-6 Contacts of K130A Will Trip Pump on Low T if Operator Error Initiates Transfer Failure of K130A to Energize on Low T May Result in Cavitation Without Pump Trip E. Contacts 3-4 of K132A Short Circuit Allows Attempted High Speed Transfer Following RPT Signal. Trips LFMG Set and Pump Speed to Zero. Pump Cannot Start Due to Trip Signal Directly from RPS System Alarm - Auto Trip or Transfer to

Low Speed 3-4 Contacts of K132A Will Trip Pump on RPT None Low Reactor Power Level and Low FW Flow Also Prevent

High Speed Operation Following RPT/SCRAM Event

LSCS-UFSAR TABLE G.B.5-1 REV. 0 - APRIL 1984 TABLE G.B.5-1 (SHEET 5 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 3 A Logic Circuit High Speed to Low Speed Auto Transfer Sequence (Typical for B Logic Circuit) A. Contacts 3-3T of S101A Fail Closed Short Circuit Placing Control Switch S101B in Transfer M/G Will Not Change Status of Pump or LFMG Set Ind. Lights None B. Contacts 8-8T of S101A Fail Closed

Short Circuit Placing Control Switch S101A in Transfer MG Will Not Change Status of Pump or LFMG Set Ind. Lights C. Contacts 3-3T of S101A

Fail Open Open Circuit Placing Control Switches S101A and S101B in Transfer M/G Will Result in Recirc Pump

'B' Transferring from High to Low Speed and Recirc Pump A Tripping to Zero Speed Speed Ind. Must Manually Restart Recirc Pump A to 25%

Speed D. Contacts 3-3T of S101B Fail Open

Open Circuit Same as 3.C Above Except Pump 'A' Transfers to Low Speed and Pump 'B' Trips to Zero Speed Same as 3.C Above Same as 3.C Above E. Contacts 1-2 of K132A Fail

Open Open Circuit LFMG Set Will Fa il to Close in at 25% Speed Following RPT Speed Indication After RPT Alarm - Auto Trip or Transfer to Low

Speed Must Manually Restart Recirc Pump to 25%

Speed F. Contacts 1-2 of K128A Circuit Fail Open Open Circuit LFMG Set Will Fa il to Close in at 25% Speed on Low Total Feedwater Flow Speed Indication Same as 3.E above Must Manually Restart Recirc Pump to 25%

Speed G. Contacts 1-2 of K129A Circuit Fail

Open Open Circuit LFMG Set Will Fa il to Close in at 25% Speed on Low Vessel Water Level Speed Indication Same as 3.E Above Must Manually Restart Recirc Pump to 25%

Speed LSCS-UFSAR TABLE G.B.5-1 REV. 0 - APRIL 1984 TABLE G.B.5-1 (SHEET 6 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 3 H. Contacts 1-2 of K130A Fail

Open Open Circuit LFMG Set Will Fail to Close in at 25% Speed on Low Steamline to Recirc Loop T Speed Indication Same as 3.E Above Must Manually Restart Recirc Pump to 25%

Speed I. Contacts 1-2 of K132A or 1-2 of K128A or 1-2 of K129A or 1-2 of K130A Fail Closed Open Circuit No Transfer of Pump to Low Speed will Result. J. Contacts 5-6 of K125A Fail

Open Open Circuit Failure of LFMG Set to Close in at 25% Speed Following RPT, Low FW Flow, Low Level or Low T Speed Ind. Alarm -

Trip or Transfer to Low Speed Must Manually Restart Recirc Pump to 25%

Speed K. Contact of Breaker 3B

Fails Closed With Breaker Closed Short Circuit Initiation of High Speed to Low Speed Transfer Sequence will Result in Normal Transfer of Both Pumps to Low Speed L. Contact of Breaker 3B Fails Open with Breaker Open Open Circuit Initiation of High-Low Speed Transfer Sequence Will Result in 'A' Pump Running at 25% Speed and 'B' Pump Trip to Zero Speed Speed Ind. Must Manually Restart Pump to 25% Speed M. Contacts Open 3-4 of K111A Fail Open Open Circuit Initiation of High-Low Speed Transfer Sequence Will Result in 'B' Pump Running at 25% Speed and 'A' Pump Trip to Zero Speed Speed Ind. Must Manually Restart Recirc Pump to 25%

Speed LSCS-UFSAR TABLE G.B.5-1 REV. 0 - APRIL 1984 TABLE G.B.5-1 (SHEET 7 OF 7)

CONTROL INTERLOCKS FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 3 N. Contact 3-4 of K111A Fail Closed With no Control Power to LFMG Set 1) Motor Bkr 2) Gen Bkr Short Circuit Initiation of High-Low Speed Transfer Sequence Will Result in 'B' Pump Running at 25% Speed and 'A' Pump Trip to Zero Speed Alarm - Low Speed Circuit not Available Must Manually Restart Recirc Pump Once Breaker Control Power is Restored

LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 1 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 LFMG set motor Breaker 1A (Typical for Bkr 1B*) A. Breaker Trips Overcurrent short circuit, relay failure Opening Breaker 1A* causes Breaker 2A* to rip if initially closed. Recirc pump trips if initially powered by LFMG set. LFMG set not available to run recirc at 25% speed. Ind. Lights, alarm, "Recirc MG

Set-A Protective Relay Trip" Loss of ability to start recirc pump at low reactor power level and to transfer from high speed to low speed.

Recirc pump trip from 25% speed is less severe than trip from 100% speed. B. Breaker will not close Blown fuse or open circuits Cannot start MG set. Ind. Lights alarm, "Low Speed Auto Transfer Circuit" not available Loss of ability to start recirc pump at low reactor power level and to transfer from high speed to low speed.

C. Breaker will not trip Blown fuse or open circuit Cannot stop MG set Ind. Lights LFMG set interlocks with main power source are through gen output breaker ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 2 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 1 System reacts normally in a low speed to high speed transfer but MG set continues to run.

2 LFMG set gen Breaker 2A* (Typical for Bkr 2B*)

A. Closes when Bkr 3A* and Bkr 4A* are closed and Bkr 1A* is open 11, 12 contacts of K113A or 9, 10 Contacts of K121A fail-closed plus K126A energizes, plus

K122A energizes, plus K134A energizes plus 1E005/b deenergizes plus 1A021/ a energizes OR Jumper installed Contact 11 of K113A to Contact 7 of K134A in Panel Bkr 2A* trips from Bkr 1A*/b contact in trip circuit Breaker indi- Bkr 1A*/b cating lights contact green

to red Momentary in- rush of currents to generator.

Possible rotation of LFMG set, possible over heating of generator windings.

________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 3 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 2 P001A plus jumper of Bkr 3A*/b in swgr 151 plus jumper of Bkr 1A*/a in swgr 14ly OR 125-Vdc hot short to closing coil of Bkr 2A*

OR Mechanical failure of closing mechanism in Bkr 2A* B. Same as 2.A plus Bkr 2A* fails to trip from Bkr 1A*/b

contact in trip circuit Same as 2.A plus open circuit on Bkr 1A*/b

contact Bkr 2A* senses overcurrent and trips on lockout. Alarm - Recirc MG set "A" protective relay trip, breaker ind. Lights 86G lockout relay Momentary overcurent, not visible effect. ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 4 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 2 C. Same as 2.C Plus Bkr 2A*

fails to trip on overcurrent Same as 2.B plus simultaneous

failure of instantaneous overcurrent relays on A and C or failure of 86G to energize.

Bkr 3A* senses overcurrent and trips Bkr 3A* and Bkr 2A* on lockout. Alarm - recirc pump motor "A" trip, breaker ind. Lights 86 lockout relay Momentary overcurrent, no visible effect. Reactor under goes recirc pump trip

transient. D. Same as 2.C plus Bkr 3A*

fails to trip on overcurrent Same as 2.C plus simultaneous

failures of instantaneous overcurrent relays on A and C or failure of 86 to energize. Generator subjected to 25-50 times normal in-rush current, tends to overspeed to 4 times normal speed, torque values 625 times normal, excessive winding heating, winding de-flection forces 625 times normal, overvoltage - 10 times normal. Estimate motor and generator wave. Same as 2.A, B and/or plus

speed indication Manual trips available. LFMG set positioned such that high velocity projectiles emanating from the LFMG set will not damage plant equipment.

Major damage to LFMG set.

Reactor under goes recirc pump trip

transient with coastdown time within the limits of normal trip and pump seizure. 50 percent probability of sustaining overspeed. Exciter would not sustain overspeed. Above conditions would exist until LFMG set ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 5 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 2 D. (Cont'd) electrically or mechanically separated itself from main power source.

E. Closes when Bkr 3A* and Bkr 4A* are closed and Bkr 1A* is closed 11, 12 Contacts of K113A or 9, 10 Contacts of K121A fail closed plus

K126A energizes plus K122A energizes plus Bkr 3A*/b deenergizes

OR Jumper installed Contact 11 of K113A to Contact 7 of K134A in Panel P001A Plus jumper at Bkr 3A*/b in Swgr 151 OR Bkr 2A* senses overcurrent tripping Bkr 1A* and Bkr 2A* on lockout. Alarm - recirc MG set "A" Protective relay trip, breaker ind. Lights 86G lockout relay Momentary overcurrent, no visible effect ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 6 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 125-Vdc hot short to closing

coil to Bkr 2A*

OR Mechanical Failure of

closing mechanism in Bkr 2A*

F. Same as 2.E plus Bkr 2A*

fails to trip on over current Same as 2.E plus simultaneous failure of instantaneous overcurrent relays on A and C or failure of 86G to energize Bkr 3A* senses overcurrent and trips Bkr 3A*, Bkr 1A* and Bkr 2A* on lockout. Alarm - Recirc pump motor "A" trip, breaker ind. Lights 86 lockout relay Momentarily overcurrent.

No visible effect Reactor under goes recirc pump trip

transient G. Same as 2.F plus Bkr 3A*

fails to trip on overcurrent Same as 2.F plus simultaneous

failure of instantaneous over-current LFMG set subjected to conditions described in 2.D but as motor begins to overspeed, Bkr 1A* senses overcurrent, Bkr 1A* and Bkr 2A* trip on lockout Alarm - Recirc MG Set "A" protective relay trip, breaker ind. Lights 86 lockout relay Momentarily LFMG set overspeed and overcurrent

________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 7 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS G. (Cont'd) Relays on A and C or failure of 86 to energize H. Same as 2.G plus Bkr 2A* and Bkr 1A* fail to trip on 86G

lockout Same as 2.E plus simultaneous failure of instantaneous overcurrent relays on A and C or failure of 86G to energize LFMG set subjected to conditions as described in 2.D until the LFMG set electrically or mechanically separates itself from the 6.9-kV power source and the 4160V power source. Same as 2D I. Trips Overcurrent, or relay failure or short circuit Recirc pump trips, if initially powered by LFMG set. LFMG set not available to run recirc pump at 25% speed Ind. Lights alarm - MG set a protective

relay trip Loss of ability to start recirc pump at low reactor power

level and to transfer from high speed to

low speed Recirc pump trip from 25% speed is less severe than a trip from 100% speed ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 8 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS J. Will not close Blown fuse, or open circuit Cannot start recirc pump at low reactor power or transfer from high speed to low speed operation Ind. Lights, alarm -

"Recirc Pump A Low Speed Auto Transfer Circuit Not Available"

Prolonged condition limits reactor operating power level K. Will not trip Blown fuse, or open circuit Cannot stop recirc pump when powered from LFMG set, cannot transfer low speed to high speed, pump can be tripped from LMFG Breaker 2A* Ind. Lights, alarm -

"Recirc Pump A motor auto transfer incomplete Prolonged condition limits reactor operating power level 3 Recirc pump motor Breaker 3A* (typical for Breaker 3B*)

A. Closes when Bkrs 2A, 4A and 1A are closed (recirc pump motor powered from LFMG set) 2, 8 Contacts of K-116A or 1, 2 Contacts of K133A and 3, 4 Contacts of K123A and 6, 6T Contacts of S101A fail closed and Contacts of Bkr

2A*/b fail closed BKR 2A* senses overcurrent and trips Bkr 1A* and Bkr A* on lockout Alarm - Recirc MG set "A" protective relay trip, breaker ind.

Lights 86G lockout erlay Recirc pump speed increases from 25% to 100%

Reactor undergoes flow increase transient 30%/sec, possible scram ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 9 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS OR Jumper installed 2 to 8 on K116A in Panel P001 plus jumper on Bkr 2A*/b in swgr 151-1.

OR Hot short 125-Vdc to 1A aux relay or closing

coil in Bkr 3A*

OR Mechanical failure of closing mechanism in Bkr 3A* B. Same as 3.A plus Bkr 2A* Same as 3.A plus simultaneous Bkr 3A* senses overcurrent and trips Bkr 1A*, 2A* and 3A* on lockout Alarm - Recirc pump motor "A" 86 lockout relays Momentarily overcurrent None ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 10 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 3 Fails to trip on overcurrent Failures of instantaneous over current relays on A and C OR Failure of 86G to energize Trip, breaker ind. lights rent. No visible effect C. Same as 3.B plus 3A* fails to trip on over-current Same as 3.B plus simultaneous failures of

instantaneous over-current relays on A and C OR Failure of 86 to energize LMFG set subjected to Conditions described in 2.G but as motor overspeeds Bkr 1A* senses overcurrent and trips Bkr 1A* and Bkr 2A* on lockout Alarm - recirc MG set "A" protective relay trip, breaker ind.

lights 86G lockout Recirc pump speed increases from 25% to 100%

Reactor under goes flow increase transient <30%/ sec possible scram D. Same as 3.C plus Bkr 2A*

and Bkr 1A* fail Same as 3.C plus simultaneous

failures of instantaneous Same as 2.H

Plus recirc pump speed increases from 25% to 100T Same as 2.G Same as 2.G Same as 2.G Reactor under goes flow increase

transient <30%/ sec followed by recirc pump trip from 100%

or less

________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 11 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS to trip on 86C lockout overcurrent relays on A and C OR Failure of 86C to energize speed, coastdown time within the limits of normal trip and pump seizure E. Breaker Trips Overcurrent, relay failure, short Recirc pump trips from 100% to 0% speed Alarm - "Recirc Pump A motor Trip" ind. lights, loop flow Reactor subjected to flow decrease

transient F. Breaker will not close Blown fuse, open circuit Cannot start recirc pump, cannot transfer from low speed to high speed operation Alarm -"Recirc Pump A" low speed auto transfer circuit not available

Prolonged conditions limits reactor operating power level G. Breaker will not trip from

Coil No. 1 (TC-1) Blown fuse, open circuit TC-1 will not trip breaker from turbine trip or load rejection. TC-2 will trip breaker with an additional time delay of a few milliseconds Indicating lights Bkr 4A* (TC-1) will trip recirc pump 1/2 trip logic becomes 1/1

________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 12 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS H. Breaker will not trip from trip

coil No. 2 (TC-2) Blown fuse, open circuit Recirc pump will not auto trip from logic interlocks or from fault protection relays Surveillance test Manual trip from Bkr 4A* is available Loss of Recirc loop as operational interlocks and functions.

Safety systems not jeopardized.

4 Breaker 4A* (Typical for Bkr 4B*) A. Breaker trips Relay failure short circuit Recirc pump trips from 100% to 0% speed, Bkr 3A* trips Alarm "Recirc Pump Motor - 1A Trip" ind. Lights, loop flow Reactor subjected to flow decrease

transient B. Breaker will not close Blown fuse, Open Circuit Cannot start recirc pump, cannot transfer from low speed to high speed operation Ind. Lights Prolonged condition limits reactor operating power level C. Breaker will not trip from trip

coil No. 1 (TC-1) Blown fuse, open circuit TC-1 will not trip breaker from turbine trip or load rejection Indicating lights Bkr 3A* (TC-1) will trip recirc pump 1/2 trip logic becomes 1/1

________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 1984 TABLE G.B.6-1 (SHEET 13 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS D. Breaker will not trip from

trip coil No. 2 (TC-2) Blown fuse, open circuit Breaker will not open from control switch. Indicating lights Bkr 3A* not Bkr 4A* is normally used to trip the recirc pump from the main power source

Somewhat reduced operating/maintenance

flexibility until fault repaired 5 240-Vac 1 supply to gen. voltage regulator (Typical for loop B) A. Loss of power Feeder breaker trip Gen field voltage decays to zero. Recirc pump motor decays to zero % speed Speed ind. Flow ind. Loss of ability to start recirc pump at low reactor power level and to transfer from high speed to low

speed B. Over-current Short ground Overload trips 86G relay, locking out LFMG set. LFMG set not available to power recirc pump, recirc pump motor trips from 25% to 0% speed Alarm, "Recirc MG Set". A protective relay trip, speed inc. Same as I ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 15, APRIL 2004 TABLE G.B.6-1 (SHEET 14 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 6 120-Vac dist. Panels at MCC 131A-2 or 132B-1, Redundant (Non-Div) (Typical for 120-Vac Dist. PNL at MCC 132B-1 (Non-Div) Loss of Power Feeder breaker trip None Alarm - "Control Signal Abnormal" Loss of Redundancy Unable to change recirculation flow with FCV 7 125-Vdc dis. Panels 111X (Non-Div) and 11Y (Div 1)

(Typical for 125-Vdc dist panels 112X (Non-Div) and 112Y (Div 2) Loss of Power Main breaker or selected output

breakers trip Loss of control powers to Bkrs 1A*, 2A*, 3A* and 4A*. These breakers will not close or trip if required to do so manually or in response to automatic logic. Alarms - "Feed Breaker Trip" "Switchgear DC Control

Power Failure" Recirc pump trip safety function maintained by redundant trip coil in Bkr 4A* which is powered by division II control power Normal recirc breaker operations inhibited pump start, transfer high speed, low speed to high speed. Pump trip on RPT signal not

inhibited but LFMG set will not start in this

case.

8 6.9-kV swgr 151 (Typical for 6.9-kV swgr 152) Loss of Power Overcurrent feeder breaker

faults Bus undervoltage relay trips Bkr 3A*. Alarm - "6.9-kV Swgr 151.

Voltage Low," "Rx, Recirc Pump 1A Auto Trip." Recirc pump trips from 100% to 0%

speed Reactor subjected to recircc flow decrease

transient. ________________________ *GE Designation LSCS-UFSAR TABLE G.B.6-1 REV. 0 - APRIL 2004 TABLE G.B.6-1 (SHEET 15 OF 15)

POWER SUPPLIES FAILURE MODES AND EFFECTS ANALYSIS NO. NAME AND FUNCTION FAILURE MODE CAUSE SYMPTONS AND LOCAL EFFECTS INCLUDING DEPENDENT FAILURES METHOD OF DETECTION INHERENT COMPENSATING PROVISION EFFECT UPON SYSTEM REMARKS AND OTHER EFFECTS 9 4.6-kV swgr (Div 1) (Typical for 4.16-kV swgr 142Y (Div II). Loss of Power Overcurrent Simultaneous feeder breaker faults LFMG set lockout trips recirc pump is powered from LFMG at 25% speed.

Operating hydraulic pump trips. Bus undervoltage relay trips Bkr 1A* Alarms - "4.16-kV Swgr Voltage Low, " - "LFMG set 1A Bkr 1A auto Trip" FCV lockup circuit Unable to change recirc flow

________________________ *GE Designation

LSCS-UFSAR H.0-i REV. 15, APRIL 2004 APPENDIX H - FIRE HAZARDS ANALYSIS TABLE OF CONTENTS The entire contents of UFSAR Appendix H - Fire Hazards Analysis, are relocated to the LaSalle County Station Fire Protection Report (FPR). References to UFSAR Appendix H remain valid.

LSCS-UFSAR I.0-1 REV. 14, APRIL 2002 APPENDIX I INFORMATION RELEVANT TO KEEPING LEVELS OF RADIOACTIVITY IN EFFLUENTS TO UNRESTRICTED AREAS AS LOW AS REASONABLY ACHIEVABLE Programs and requirements for control of ra dioactive effluents, ca lculation of offsite doses, and radiological environmental monitoring are included in the Offsite Dose Calculation Manual.

LSCS-UFSAR J.0-i REV. 14, APRIL 2002 APPENDIX J ANALYSIS OF THE EFFECTS OF MODERATE ENERGY LINE THROUGH WALL LEAKAGE CRACKS OUTSIDE PRIMARY CONTAINMENT TABLE OF CONTENTS PAGE J.0 INTRODUCTION J.0-1 J.1 MODERATE ENERGY LINES J.1-1 J.2 PLANT AREAS AND EQUIPMENT AFFECTED BY MELB J.2-1 J.3 SPRAY SHIELD J.3-1 J.4 EFFECTS OF FLOODING J.4-1 J.5 CONCLUSION J.5-1

LSCS-UFSAR J.0-ii REV. 14, APRIL 2002 APPENDIX J ANALYSIS OF THE EFFECTS OF MODERATE ENERGY LINE THROUGH WALL LEAK AGE CRACKS OUTSIDE PRIMARY CONTAINMENT LIST OF TABLES NUMBER TITLE J.2 Plant Areas and Equipment Affected by MELB

LSCS-UFSAR DRAWINGS CITED IN THIS APPENDIX*

J.0-iii REV. 14, APRIL 2002 DRAWING* SUBJECT M-69 Service Water, Unit 1 M-71 Fire Protection M-72 Fire Protection Sheets (1, 2, and 5) M-72 Fire Protection Control Room Breathing Air (Sheet 3) M-72 Fire Protection High Pressure Breathing Air (Sheet 4) M-74 Cycled Condensate Storage, Unit 1 M-75 Clean Condensate Storage M-78 Hydrogen & Carbon Dioxide System (Sheet 1) M-78 Stator Cooling System (Sheets 2 and 3) M-78 HWC System (Sheet 5) M-85 Diesel Oil System, Unit 1 (Sheet 1) M-85 Security Diesel Generator Building (Sheet 2) M-86 Primary Containment Chilled Water Coolers, Unit 1 M-87 Core Standby Cooling System Equipment Cooling Water System, Unit 1 M-90 Reactor Building Closed Cooling Water, Unit 1 M-94 Low Pressure Core Spray (LPCS), Unit 1 M-95 High Pressure Core Spray (HPCS), Unit 1 M-96 Residual Heat Removal System (RHRS), Unit 1 (Sheets 1-4) M-96 Suppression Pool Clean Up and Transfer System (Sheet 5) M-101 Reactor Core Isolation Cooling System, Unit 1 M-102 Station Heat Recovery M-210 Reactor Building Piping Plan El. 710'-6" Area 3, Unit 1 M-214 Reactor Building Piping Plan El. 710'-6" Area 3, Unit 2 M-219 Reactor Building Piping Plan El. 694'-6" Area 4, Unit 1 M-223 Reactor Building Piping Plan El. 694'-6" Area 4, Unit 2 M-225 Reactor Building Piping Plan El. 673'-4" Area 2, Unit 1 M-226 Reactor Building Piping Plan El. 673'-4" Area 3, Unit 1 M-229 Reactor Building Piping Plan El. 673'-4" Area 2, Unit 2 M-230 Reactor Building Piping Plan El. 673'-4" Area 3, Unit 2 M-414 Diesel Generator Room Piping Plan Basement Floor Area 5, Unit 1 M-415 Diesel Generator Room Piping Plan Basement Floor Area 1, Unit 2 M-419 Diesel Generator Room Piping Plan Basement Floor Area 5, Unit 2

The listed drawings are included as "General References" only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.

LSCS-UFSAR J.0-1 REV. 14, APRIL 2002 APPENDIX J - ANALYSIS OF THE EFFECTS OF MODERATE ENERGY LINE THROUGH WALL LEAKAGE CRACKS OUTSIDE PRIMARY CONTAINMENT J.0 INTRODUCTION

This Appendix summarizes the Moderate Energy Line Break (MELB) Analysis, which was performed in accordance with the guidance provided in:

NUREG 0800(Standard Review plan), Section 3.6.1

Branch Technical Position (BTP) ASB-3-1 (provided as Attachment B to NUREG 0800, Section 3.6.1).

J. F. O'Leary Letter of July 12, 1973 (provided as Attachment C to NUREG 0800, Section 3.6.1).

Branch Technical Position (BTP) MEB-3-1.

The MELB analysis demonstrates the capability to shutdown the plant and maintain in a cold shutdown condition assuming a MELB (i.e., through wall crack in a moderate energy line) and a single active component failure. The focus of the MELB analysis is the protection of safe shutdown equipment from spray and flooding resulting from a postulated MELB. In accordance with the guidance of Branch Technical Position (BTP) ASB 3-1, the analysis is based upon:

The postulated MELB occuring during no rmal plant conditions. It does not occur concurrently with other desi gn basis accidents or transients.

Offsite power being available. A Loss-of Offsite Power (LOOP) concurrent with the postulated MELB is not consid ered. The analysis demonstrates that a MELB in a safety-related building will not result a trip of the RPS system or the FW/turbine trip system. A crack in a moderate energy line in the turbine building could possibly result in a trip of the FW/Turbine. However, such a crack would not affect equipment required to achieve cold shutdown.

LSCS-UFSAR J.1-1 REV. 14, APRIL 2002 J.1 MODERATE ENERGY LINES Moderate energy fluid systems are those that meet the following criteria during normal plant conditions:

a) Maximum operating temperature or pressure is:

Maximum operating temperature less than or equal to 200

°F, or

Maximum operating pressure le ss than or equal to 275 psig. b) High energy piping systems wh ose pressure exc eeds 275 psig or temperature exceeds 275

°F for short periods only (1% of the normal operating life span of the plant, or 2% of the time period required to accomplish its system design function). c) Nominal pipe size is greater than 1 inch. d) Minimum pressure is greater than 10 psig.

Moderate energy lines specifically exclud ed from MELB consideration are those whose stress has been demonstrated by analysis meet the exclusion criteria of BTP MEB 3.1, paragraph B.2.c.1. This is where the stress is less than or equal to: 1.2 Sm for Class 1 piping 0.4(1.2 S h + S A) for Class 2, 3, and B31.1 piping Criterion d was chosen on the following basis: a) Since the flow rate and the maximum spray distance from a postulated through -wall crack are related to pressu re by a square root relationship, the 10 psig is very small.

b) The pipe wall thickness is many ti mes the required minimum wall thickness for such low pressure s and therefore, the probability of a rupture occurring is acceptably small.

The moderate energy systems consist of:

Clean Condensate (MC)

Cycled Condensate storage (CY)

Core Standby Cooling System (DG, FC, HP/E22, RH/E12)

Diesel Oil System (DO)

Fire Protection (FP)

Hydrogen and Carbon Di oxide System (HY, CO)

High Pressure Core Spray (HP/E22)

Low Pressure Core Spray (LP/E21)

Reactor Building Equipment Drains (RE)

Residual Heat Removal (RH/E12)

Reactor Core Isolation Cooling (RI/E51)

Station Heat Recovery (SH)

Primary Containment Chiller Water Coolers (VP)

Reactor Building Closed Cooling Water (WR)

Service Water (WS)

LSCS-UFSAR J.2-1 REV. 14, APRIL 2002 J.2 AFFECTED SAFETY-RELATED EQUIPMENT Table J.2 identifies each area of the plan t that could potentially be affected by a MELB. For each of these areas the Table then identifies whether a MELB is postulated. If yes, the safe shutdown equipment located in the area, if no, the basis for excluding the area. For each area that a MELB is postulated, the Table identifies whether there is any safe s hutdown equipment, which requires spray protection. If spray protection is not required, the basis for this conclusion is identified.

LSCS-UFSAR J.3-1 REV. 14, APRIL 2002 J.3 SPRAY SHIELDS Safe shutdown can be achieved without spray shields being required to protect equipment from the spray from postulated cracks in moderate energy lines. As discussed in Section J.2, the basis for this is that sufficient separation and redundancy exists.

LSCS-UFSAR J.4-1 REV. 17, APRIL 2008 J.4 EFFECTS OF FLOODING Due to the presence of floor drains and stairwells, the safe shutdown equipment that could potentially affected by postulated flooding resulting from a crack in a moderate energy line is that located in lowest elevations. This is the RHR pump motors and the RHR service water pump motors.

The Division 1 and Division 2 RHR pumps are located in separate watertight rooms with bulkhead doors that open to the raceway. These doors protect these rooms from the postulated failure of the suppression pool. Should the postulated crack occur outside these rooms, the safe shutdown equipment in the room is protected.

Should the postulated crac k occur inside one of these rooms, any flooding would be limited to that room. As discussed in the above table, if such spray disabled the equipment, sufficient redundancy exists, to ensure the capability to accomplish the safe shutdown function assuming a single active equipment in a redundant division or system. Each of these rooms have sump pumps that alarm upon high water level. Thus the operator would receive indication of a leak and take action to isolate it.

The Division 1 and Division 2 RHR service water pumps are also located in separate watertight rooms with bulkhead doors. These doors protect any flooding from propagating to the adjacent rooms. Only the Division 2 room has a door that communicates with the Turbine Building. This is acceptable because no credible flood source has been identified in the Turbine Building.

Should the postulated crack occur inside one of these rooms, any flooding would be limited to that room. As discussed in the above table, if such spray disabled the equipment, sufficient redundancy exists, to ensure the capability to accomplish the safe shutdown function assuming a single active equipment in a redundant division or system. Each of these rooms have sump pumps that alarm upon high water level. Thus th e operator would receive indication of a leak and take action to isolate it.

LSCS-UFSAR J.5-1 REV. 14, APRIL 2002 J.5 CONCLUSION The MELB analysis postulates that a mo derate energy line outside of the primary containment develops a through wall crack which results in spray. This crack is postulated to occur during normal operatin g conditions. The capability to shut the plant down and maintain it in a cold shutdown condition has been evaluated. This evaluation concluded that sufficient separa tion and redundancy exists such that if the resultant spray were to disable equipment and a single active equipment were to occur, the affected Unit would still be able to achieve and maintain cold shutdown.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 1 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Primary Containment N/A Drywell No N/A Excluded per BTP ASB 3-1, paragraph B.1 Primary Containment N/A Wetwell No N/A Excluded per BTP ASB 3-1, paragraph B.1 Reactor Building 673'-4" Raceway No N/A The Raceway at Elev. 673'-4" and 694'-0" represents a common area. Grating at Elev. 694'-0" separates the two elevations. Therefore, environmentally they are one area. This area contains high energy RI steam lines.

Reactor Building 673'-4" NW corner room (A-RHR) Yes No The NW corner rooms at Elev. 673'-4" and 694'-0" and the RHR heat exchanger room at Elev. 710'-6" comprise a single area with multiple entry points. Therefore, environmentally they are one area. At this elev, the safe shutdown equipment located in this area is the "A" RHR pump 1(2)E12-C002A. During conditions when a MELB is postulated to occur (i.e., normal operating conditions) the RHR equipment functions to support decay heat removal and vessel inventory control. For vessel inventory control refer to discussion under the HPCS pump (Reactor building Elev. 673'-4", SW corner room). For decay heat removal, there are multiple methods available to accomplish this function. They are Div 1 RHR, Div 2 RHR, and several alternate decay heat removal methods. Since the equipment associated with these methods is physically separated, should spray from a moderate energy line crack disable one Division of RHR, and a single active equipment failure disable other Division of RHR, an alternate decay heat removal method would be available. Reactor Building NE corner room (LPCS & RCIC)

No N/A The NE corner rooms at Elev. 673'-4" and 694'-0" are connected via the stairwell. Therefore, environmentally they are one area. This area contains high energy RI steam lines.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 2 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Reactor Building 673'-4" SE corner room (B & C RHR)

Yes No The SE corner rooms at Elev. 673'-4" and 694'-0" and the RHR heat exchanger room at Elev. 710'-6" comprise a single area with multiple entry points. Therefore, environmentally they are one area. At this elev., the safe shutdown equipment located in this area is the "B" RHR pump 1(2)E12-C002B, and the RHR water leg pump 1(2)E12-C003. During conditions when a MELB is postulated to occur (i.e., normal operating conditions) the RHR equipment functions to support decay heat removal and vessel inventory control. For vessel inventory control refer to discussion under the HPCS pump (Reactor building Elev. 673'-4", SW corner room). For decay heat removal, there are multiple methods available to accomplish this function. They are Div 1 RHR, Div 2 RHR, and several alternate decay heat removal methods. Since the equipment associated with these methods is physically separated, should spray from a moderate energy line crack disable one Division of RHR, and a single active equipment failure disable other Division of RHR, an alternate decay heat removal method would be available.

Reactor Building 673'-4" SW corner room (HPCS) Yes No The SW corner rooms at Elev. 673'-4" and 694'-0" are connected via the stairwell. Therefore, environmentally they are one area. At this elev., the safe shutdown equipment in this area is associated with the HPCS system (i.e., HPCS pump 1(2)E22-C001, HPCS water leg pump) 1(2)E22-C003, and HPCS instrument panel 1(2)H22-P024). PER BTP ASB 3-1 a MELB is only postulated to occur during normal operating conditions. Since a crack in a moderate energy line is not postulated to occur concurrent with an accident, the accident mitigation function of HPCS is not required. However, HPCS is one of the systems that can be used to provide water inventory to the core. Should HPCS become disabled, other systems are available to provide water to the core. These other systems are RCIC & ADS/LPCS, ADS/RHR-A, ADS/RHR-B & C. Since the components associated with these methods are physically separated, if spray from a moderate energy line crack disabled HPCS and a single active equipment failure disables another method, at least one means would remain available.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 3 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Reactor Building 673'-4" Rx Bldg drain tank room Yes No Area has no SR equipment Reactor Building 673'-4" Raceway No N/A The Raceway at Elev. 673'-4" and 694'-0" represents a common area. Grating at Elev. 694'-0" separates the two elevations. Therefore, environmentally they are one area. This area contains high-energy RI steam lines.

Reactor Building 694'-0" NW corner room Yes No The NW corner rooms at Elev. 673'-4", 694'-0", and the RHR Heat Exchanger room at Elev. 710'-6" comprise a single common area with multiple entry points. Therefore, environmentally they are one area. At this elev, the only safe shutdown equipment are the RHR Heat Ex and MOVs which are excluded Reactor Building 694'-0" NE corner room No N/A The NE corner rooms at Elev. 673'-4" and 694'-0" are connected via the stairwell. Therefore, environmentally they are one area with high energy RI lines.

Reactor Building 694'-0" SE corner room Yes No The SE corner rooms at Elev. 673'-4", 694'-0", and the RHR Heat Exchanger room at Elev. 710'-6" comprise a single common area with multiple entry points. Therefore, environmentally they are one area At this elev., the only safe shutdown equipment are the RHR Heat Ex and MOVs which are excluded Reactor Building 694'-0" SW corner room Yes No The SW corner rooms at Elev. 673'-4" and 694'-0" are connected via the stairwell. Therefore, environmentally they are one area. The area at this Elev. contains no safe shutdown equipment Reactor Building 694'-0" Main Steam Tunnel No N/A Area contains high energy FW & MS lines.

Reactor Building 710'-6" Drywell equip drain room Yes No Area contains no safe shutdown equipment Table J.2 Plant Areas and Equipment Affected by MELB Sheet 4 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Reactor Building 710'-6" "A" RHR heat ex room Yes No The RHR Heat Exchanger room at Elev. 710'-6" and NW corner rooms at Elev. 673'-4", 694'-0", comprise a single common area with multiple entry points. Therefore, environmentally they are one area. At this elev., the only safe shutdown equipment are the RHR Heat Ex and MOVs which are excluded Reactor Building 710'-6" "B" RHR heat ex room Yes No The RHR Heat Exchanger room at Elev. 710'-6" and SE corner rooms at Elev. 673'-4", 694'-0", comprise a single common area with multiple entry points. Therefore, environmentally they are one area. At this elev., the only safe shutdown equipment are the RHR Heat Ex and MOVs which are excluded.

Reactor Building 710'-6" General area Yes No Only safe shutdown equipment are MCC 480V 1(2)35Y-2 (1(2)AP76E)

MCC 480V 1(2)36Y-1 (1(2)AP82E) These MCCs provide electrical power to Div 1 and Div 2 respectively. The specific components being fed are valves associated with the HG, LPCS, RHR, RHR-WS, and MSIV-LCS; water leg pumps associated with LPCS & RI; DO fuel oil transfer pump; Cubicle coolers associated with the RHR, LPCS, RHR-WS; and the MSIV-LCS heaters. As discussed previously, the only components that are required during conditions in which a MELB is postulated to occur are those associated with the decay heat removal and vessel inventory control. For vessel inventory control refer to discussion under the HPCS pump (Reactor building, Elev. 673'-4", SW corner room). For decay heat removal, there are multiple methods available to accomplish this function. They are Div 1 RHR, Div 2 RHR, and several alternate decay heat removal methods. Since the equipment associated with these methods is physically separated, should spray from a moderate energy line crack disable one MCC, and a single active equipment failure disable other MCC, an alternate decay heat removal method would be available.

Reactor Building 740'-0" General area No N/A Area contains high energy RI steam lines.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 5 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Reactor Building 761'-0" General area No N/A Area contains high energy RD lines.

Reactor Building 786'-6" General area No N/A Area contains high energy RT lines.

Reactor Building 807'-0" RWCU valve room No N/A Area contains high energy RT lines.

Reactor Building 820'-6" RWCU area No N/A Area contains high energy RT lines.

Reactor Building 820'-6" SGTS Area Yes No The safe shutdown equipment in this area is 480V MCC 136X-1 (Unit 1), 236X-1 (Unit 2) [1(2)AP78E]. The important loads during postulated MELB conditions are associated with the VC and VX systems. Should spray im pair this MCC, sufficient redundancy exists to accomplish the function of the VC equipment even if a single active equipment failure occurs. Procedurally controlled compensatory actions ensure that the function of the VX system can be accomplished should VX equipment be impaired by a MELB.

Reactor Building 832'-0" Instrument Storage Room Yes No Area has no SR equip Reactor Building 843'-6" Refuel Floor Yes No Area has no shutdown equip Table J.2 Plant Areas and Equipment Affected by MELB Sheet 6 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Diesel Gen. Building 674'-0" Div 1 diesel fuel tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 674'-0" Div 2 diesel fuel tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 674'-0" HPCS diesel fuel tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 674'-0" Div 1 RHR-WS pump room Yes No The safe shutdown equipment is the Div 1 CSCS equipment. Specifically the RHR-WS pumps (1(2)E12-C300A,B), Div 1 DG cooling water pump (0DG01P). Since neither an accident or a LOOP is not postulated concurrent with a crack in a moderate energy line, only the RHR-WS pumps which support decay heat removal, are required to achieve and maintain cold shutdown. There are multiple methods available to accomplish this function of decay heat removal. They are Div 1 RHR & CSCS, Div 2 RHR & CSCS, and several alternate decay heat removal methods. Since the equipment associated with these methods is physically separated, should spray from a moderate energy line crack disable Div 1 of CSCS, and a single active equipment failure disabled other Division of CSCS, an alternate decay heat removal method would be available.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 7 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Diesel Gen. Building 674'-0" Div 2 RHR-WS pump room Yes No The safe shutdown equipment is the Div 2 CSCS equipment. Specifically the RHR-WS pumps (1(2)E12-C300C,D, Div 2 DG cooling water pump (1(2)DG01P). Since neither an accident nor a LOOP is not postulated concurrent with a crack in a moderate energy line, only the RHR-WS pumps which support decay heat removal, are required to achieve and maintain cold shutdown. There are multiple methods available to accomplish this function of decay heat removal. They are Div 1 RHR & CSCS, Div 2 RHR & CSCS, and several alternate decay heat removal methods. Since the equipment associated with these methods is physically separated, should spray from a moderate energy line crack disable Div 2 of CSCS, and a single active equipment failure disabled other Division of CSCS, an alternate decay heat removal method would be available.

Diesel Gen. Building 674'-0" HPCS cooling water pump room Yes No The only equipment in this area that is potentially required for safe shutdown is the HPCS diesel generator cooling water pump 1(2)E22-C002. As discussed previously, HPCS is one of the systems that can be used to provide vessel inventory control. Should HPCS become disabled, sufficient other systems are available to provide water to the core. These other systems are RCIC & ADS/LPCS, ADS/RHR-A, ADS/RHR-B & C. Since the components associated with these methods are physically separated, if spray from a moderate energy line crack disabled HPCS and a single active equipment failure be postulated to occur, at least one means would be available.

Diesel Gen. Building 710'-6" Div 1 DG room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 710'-6" Div 1 diesel day tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 710'-6" Div 2 DG room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 8 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Diesel Gen. Building 710'-6" Div 2 diesel day tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown. Diesel Gen. Building 710'-6" HPCS DG room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 710'-6" HPCS diesel day tank room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown. Diesel Gen. Building 736'-6" Div 1 diesel vent room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Diesel Gen. Building 736'-6" Div 2 diesel vent room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown. Diesel Gen. Building 736'-6" HPCS diesel vent room Yes No A Loss-of Offsite Power (LOOP) is not assumed. Therefore the emergency diesel generators are not required for safe shutdown.

Auxiliary Building 687'-0" Div 3 Essential Swgr room Yes No In Unit 1, area contains no moderate energy lines Unit 2 contains moderate energy line 2RTA1A-3" The only equipment in this area that is potentially required for safe shutdown is 4.16KV switchgear 143 (2AP07E), 480V MCC 143-1 (2AP79E), HPCS Battery (2DC18E), HPCS battery Charger (2DC19E), HPCS relay panel (2H13-P625). This equipment controls HPCS components. As discussed previously, HPCS is one of the systems that can be used to provide water inventory to the core. Should HPCS become disabled, other systems are available to provide water to the core. These other systems are RCIC & ADS/L PCS, ADS/RHR-A, ADS/RHR-B & C. Since the components associated with these methods are physically separated, if spray from a moderate energy line crack disabled HPCS and a single active equipment failure be postulated to occur, at least one means would be available.

Table J.2 Plant Areas and Equipment Affected by MELB Sheet 9 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Auxiliary Building 710'-6" Div 1 Essential Swgr room No N/A Area contains no moderate energy lines Auxiliary Building 710'-6" Labs & Offices Yes No Area contains no safe shutdown equipment Auxiliary Building 731'-0" Div 2 Essential Swgr room No N/A Area contains no moderate energy lines Auxiliary Building 731'-0" AEER No N/A Area contains no moderate energy lines Auxiliary Building 749'-0" Cable spreading room Yes No Area contains no safe shutdown equipment Auxiliary Building 749'-0" Elec equip room Yes No Area contains no safe shutdown equipment Auxiliary Building 768'-0" Main control room No N/A Area contains no moderate energy lines Auxiliary Building 768'-0" Computer room Yes No Area contains no safe shutdown equipment Auxiliary Building 768'-0" Security control center Yes No Area contains no safe shutdown equipment Auxiliary Building 768'-0" Offices Yes No Area contains no safe shutdown equipment Table J.2 Plant Areas and Equipment Affected by MELB Sheet 10 of 10 J.2 REV. 14, APRIL 2002 Bldg. Elev. Area Description MELB Postulated in Area Safe shutdown equipment requiring protection Exclusion Justification Auxiliary Building 786'-6" and 802'-0" Lower vent equip floor Yes No The safe shutdown equip is:

0PL15J; 0PL16J; 0PL42J; 0PL43J; 0VC02AA,AB; 0VC01CA,CB; 0VC02CA,CB; 0VC04CA,CB; 0VC05CA,CB; 0VE01AA,AB

0VE01CA,CB; 0VE02CA,CB; 0VE03CA,CB; 0VE04CA,CB; Various dampers. The fans are vaneaxial fans where the motor is located internal to the fan housing and therefore is protected from spray. Correspondingly the damper actuators are enclosed, protecting electrical items from spray. The cooling coils have no electrical items. Therefore, the safe shutdown components that could potentially be affected by spray from a crack in a moderate energy line are the control panels and the motor associated with the refrigeration compressor. Since a design basis accident is not postulated concurrent with a crack in a moderate energy line, this equipment is not required for to maintain the habitability of the control room envelope (control room and AEER). The equipment is however required to maintain the temperature in the control room envelope to the design temperature for the equipment. The plant may be shutdown from the control room or from the AEER via the remote shutdown panel. There are redundant systems to provide cooling to the control room (Unit 1 VC system and Unit 2 VC system) and to the cool the AEER (Unit 1 VE system and Unit 2 VE system). Therefore, a total of four systems are available to accomplish the function of maintaining a suitable environment for equipment in the control room envelope that is used to safely shut down the plant and maintain it in a safe shutdown condition. Due to physical separation of the equipment, spray from a moderate energy line may affect one or two of these systems, but the other systems will remain available. Therefore, safe shutdown can be achieved, assuming a single active equipment failure.

Auxiliary Building 815'-0" Upper vent equip floor Yes No Area contains no safe shutdown equipment

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LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE J-1.0 ASSOCIATED CONTROLLED DRAWINGS* FOR J-1 FIGURES

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LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE J-2.0 ASSOCIATED CONTROLLED DRAWINGS* FOR J-2 FIGURES LSCS-UFSAR REV. 14, APRIL 2002

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LA SALLE COUNTY STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE J-3.1 TYPICAL CROSS-SECTION DETAIL OF SPRAY SHIELD

LSCS-UFSAR K.0-1 REV. 15, APRIL 2004 APPENDIX K INTERIM OFFSITE POWER ARRANGEMENT Because of transmission right-of-way proble ms during construction of the station, a temporary, interim supply of 138 KV had to be brought into LaSalle during construction of the station. This was in lieu of one of the dual, 345 KV lines that would permanently tie into the Edison grid from each generator unit at LaSalle. A description of this interim configuration was provided in Appendix K of the FSAR. Section 8.2 of the UFSAR describes the offsite power system currently connected to the LaSalle station.

The connection between the 345 kV and 1 38 kV systems at LaSalle was made permanent after the second pair of 345 kV lines was installed to support the 138 kV transmission system.

LSCS-UFSAR APPENDIX L REQUIREMENTS RESULTING FROM TMI-2 ACCIDENT TABLE OF CONTENTS L-i REV. 15, APRIL 2004 PAGE L.0 Introduction L-1 L.1 Shift Technical Advisor (I.A.1.1) L-1 L.2 Shift Supervisor Administrative Duties (I.A.1.2) L-1 L.3 Shift Manning (I.A.1.3) L-1 L.4 Immediate Upgrading of Operator and Senior Operator Training and Qualification (I.A.2.1) L-1 L.5 Administration of Training Programs for Licensed Operators (I.A.2.3) L-1 L.6 Scope and Criteria for Licensing Examinations (I.A.3.1) L-1 L.7 Evaluation of Organization and Management Improvements of Near-Term Operating License Applicants (I.B.1.2) L-2 L.8 Short-Term Accident Analysis and Procedure Revision (I.C.1) L-2 L.9 Shift Relief and Turnover Procedures (I.C.2) L-2 L.10 Shift Supervisor Responsibilities (I.C.3) L-2 L.11 Control Room Access (I.C.4) L-2 L.12 Procedures for Feedback of Operating Experioence to Plant Staff (I.C.5) L-2 L.13 Guidance on Procedures for Verifying Correct Performance of Operating Activities (I.C.6) L-2 L.14 NSSS Vendor Review of Procedures (I.C.7) L-2 L.15 Pilot Monitoring of Selected Emergency Procedures for Near-Term Operating License Applicants (I.C.8) L-3 L.16 Control Room Design (I.D.1) L-3 L.17 Plant Safety Parameter Display Console (I.G.1) L-3 L.18 Training During Low-Power Testing (I.G.1) L-3 L.19 Reactor Coolant System Vents (II.B.1) L-3 L.20 Plant Shielding (II.B.2) L-3 L.21 Post-Accident Sampling L-3 L.22 Training for Mitigating Core Damage (II.B.4) L-4 L.23 Relief and Safety Valve Test Requirements (II.D.1) L-4 L.24 Relief and Safety Valve Position Indication (II.D.3) L-4

LSCS-UFSAR APPENDIX L TABLE OF CONTENTS (Cont'd)

L-ii REV. 15, APRIL 2004 PAGE

L.25 Auxiliary Feedwater System Feasibility Evaluation (II.E.1.1) L-4 L.26 Auxiliary Feedwater Initiation and Indication (II.E.1.2) L-4 L.27 Emergency Power for Pressurized Heaters (II.E.3.1) L-4 L.28 Containment Dedicated Penetrations (II.E.4.1) L-4 L.29 Containment Isolation Dependability (II.E.4.2) L-4 L.30 Additional Accident-Monitoring Instrumentation (II.F.1) L-4 L.31 Inadequate Core Cooling Instruments (II.F.2) L-5 L.32 Emergency Power for Pressurized Equipment (II.G.1) L-5 L.33 Assurance of Proper ESF Functioning (II.K.1.5) L-5 L.34 Safety-Related System Operability Status (II.K.1.10) L-5 L.35 Pressurizer Low-Level Coincident Signal Bistables (II.K.1.17) L-5 L.36 Operator Training for Prompt Manual Trip (II.K.1.20) L-5 L.37 Automatic Safety Grade Anticipatory Trip (II.K.1.21) L-5 L.38 Proper Functioning of Heat Removal Systems (II.K.1.22) L-5 L.39 Reactor Vessel Level Instrumentation (II.K.1.23) L-5 L.40 Control of Auxiliary Feedwate r Independent of the Integrated Control System L-6 L.41 Failure Mode Effects Analysis on the Integrated Control System L-6 L.42 Safety-Grade Anticipatory Reactor Trip L-6 L.43 Thermal Mechanical Report - Effort of High-Pressure Injection on Vessel Integrity for Small-Break Loss-of-Coolant Accident with no Auxiliary Feedwater L-6 L.44 Power Operated Pressurizer Relief Valves (II.K.2.14) L-6 L.45 Effects of Slug-Flow on Steam Generator Tubes L-6 L.46 Reactor Coolant Pump Seal Damage L-6

LSCS-UFSAR APPENDIX L TABLE OF CONTENTS (Cont'd)

L-iii REV. 15, APRIL 2004 PAGE

L.47 Potential for Voiding in the Reactor Coolant System During Transients L-6 L.48 Sequential Auxiliary Feedwater Flow Analysis (II.K.2.19) L-6 L.49 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System (II.K.3.1) L-6 L.50 Report on Overall Safety Effe ct of Power-Operated Relief Valve Isolation System (II.K.3.2) L-6 L.51 Failure of PORV or Safety Valve to Close (II.K.3.3) L-7 L.52 Automatic Trip of Reactor Coolant Pumps During Loss-of-Coolant Accident (II.K.3.5) L-7 L.53 Evaluation of Power-Operated Relief Valve Opening Probability During Overpressure Transient (II.K.3.7) L-7 L.54 Proportional Integral Derivative Controller Modification L-7 L.55 Proposed Anticipatory Trip Modification L-7 L.56 PORV Failure Rate (II.K.3.11) L-7 L.57 Confirm Existence of Anticipatory Reactor Trip Upon Turbine Trip L-7 L.58 Separation of HPCI and RCIC System Initiation Levels L-7 L.59 Modify Break Detection Logic to Prevent Spurious Isolation of HPCI and RCIC L-7 L.60 Reduction of Challenges and Failures of Relief Valves-Feasibility Study and System Modification L-8 L.61 Report on Outages of ECCS (II.K.3.17) L-8 L.62 Modification of ADS Logic-Feas ibility for Increased Diversity for Some Event Sequences L-8 L.63 Restart of Core Spray and LPCI System (II.K.3.21) L-8 L.64 Automatic Switch Over of RCIC System Suction-Verify Procedures and Modify Design L.65 Conform Adequacy of Space Cooling for HPCI and RCIC Systems L-8 L.66 Effect of Loss of AC Power on Pump Seals (II.K.3.25) L-8 L.67 Provide Common Reference Level for Vessel Level Instrumentation L-8

LSCS-UFSAR APPENDIX L TABLE OF CONTENTS (Cont'd)

L-iv REV. 15, APRIL 2004 PAGE

L.68 Verify Qualification of Accumulators on ADS Valves L-9 L.69 Revised Small-Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K L-9 L.70 Plant-Specific Calculations to Show Compliance with 10 CFR 50.46 L-9 L.71 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure (II.K.3.44) L-9 L.72 Evaluation of Depressurization with Other Than ADS (II.K.3.45) L-9 L.73 Response to List of Concerns from ACRS Consultant (II.K.3.46) L-9 L.74 Emergency Preparedness (III.A.1.1) L-9 L.75 Emergency Support Facilities (III.A.1.2) L-9 L.76 Improving Licensee Emergency Preparedness - Long-Term (III.A.2) L-10 L.77 Primary Coolant Sources Outside Containment (III.D.1.1) L-10 L.78 In-Plant Radiation Monitoring L-10 L.79 Control Room Habilitability (III.D.3.4) L-10 LSCS-UFSAR L-1 REV. 15, APRIL 2004 L.0 INTRODUCTION The full text of CECo responses to NU REG-0737 action items for LaSalle County Station are included in Appendix L of th e FSAR. Where applicable, the effects of evaluations, analyses, or modifications completed in response to specific action items have been incorporated in the text of the UFSAR. This appendix lists all action items and provides a brief discussion of the current status of each item along with references to UFSAR sections and Te chnical Specification sections where the effects of applicable responses have been included.

L.1 Shift Technical Advisor (I.A.1.1)

All NRC commitments on this topic ha ve been met. Responsibilities and qualifications of the shift technical adviso r are given in the Technical Specifications.

L.2 Shift Supervisor Administrative Duties (I.A.1.2)

All NRC commitments on this topic have been met. Responsibilities of the shift supervisor (Shift Manager at LaSalle County Station) are delineated in Section 13.1.2.2 of the UFSAR and the Technical Specifications.

L.3 Shift Manning (I.A.1.3)

All NRC commitments on this topic have been met. The Technical Specifications reflect current NRC-approved LSCS requirements for shift manning and administrative procedures on working hours.

L.4 Immediate Upgrading of Operator and Senior Operator Training and Qualification (I.A.2.1)

All NRC commitments on this topic have been met. Additional information on training is provided in Section 13.2 of the UFSAR.

L.5 Administration of Training Programs for Licensed Operators (I.A.2.3)

All NRC commitments on this topic have been met. Additional information on the training program is provided in Section 13.2 of the UFSAR.

L.6 Scope and Criteria for Li censing Examinations (I.A.3.1)

All NRC commitments on this topic have been met. Additional information on the training program is provided in Section 13.2 of the UFSAR.

LSCS-UFSAR L-2 REV. 15, APRIL 2004 L.7 Evaluation of Organization and Management Improvements of Near-Term Operating License Applicants (I.B.1.2)

All NRC Commitments on this topic have been met. The current LSCS organization is reflected in the Quality Assurance Topical Report.

L.8 Short-Term Accident Analysis and Procedure Revision (I.C.1)

All NRC Commitments on this topic have been met. Emergency procedures are discussed in Section 13.5.2 of the UFSAR.

Emergency planning is discussed in Section 13.3 of the UFSAR.

L.9 Shift Relief and Turnover Procedures (I.C.2)

All NRC Commitments on this topic have been met. Section 13.5.1.3.6 of the UFSAR provides procedures for log book usage and control.

L.10 Shift Supervisor Responsibilities (I.C.3)

All NRC Commitments on this topic have been met. Directives and procedures have been prepared and implemented to meet the requirements of this action item. Shift supervisor (Shift Manager at LaSalle County Station) responsibilities are discussed in Sections 13.1.2.2 and 13.5.1.3.1 of the UFSAR.

L.11 Control Room Access (I.C.4)

All NRC Commitments on this topic have been met. Procedures have been prepared and implemented.

L.12 Procedures for Feedback of Operating Experience to Plant Staff (I.C.5)

All NRC Commitments on this topic have been met. The site follows procedures to ensure operating experience information continues to be appropriately reviewed and disseminated to plant staff.

L.13 Guidance on Procedures for Veri fying Correct Performance of Operating Activities (I.C.6)

All NRC Commitments on this topic have been met. Equipment control procedures are described in Section 13.5.1.3.3 of the UFSAR.

L.14 NSSS Vendor Review of Procedures (I.C.7)

LSCS-UFSAR L-3 REV. 15, APRIL 2004 Requirement has been satisfied. Procedur es have been reviewed and accepted by General Electric.

L.15 Pilot Monitoring of Selected Emer gency Procedures for Near-Term Operating License Applications (I.C.8)

All NRC commitments on this topic have been met. Emergency planning is discussed in Section 13.3 of the UFSAR.

Emergency procedures are discussed in section 13.5.2 of the UFSAR.

L.16 Control Room Design (I.D.1)

All NRC commitments on this topic have been met.

L.17 Plant Safety Parameter Display Console (I.D.2)

All NRC commitments on this topic have been met. Design information is included in Section 7.8.2 of the UFSAR.

L.18 Training During Low-Power Testing (I.G.1)

All NRC commitments on this topic have been met. The low power training and testing programs provided sufficient operator training and accumulation of sufficient technical data during both normal and off-normal plant operation.

L.19 Reactor Coolant System Vents (II.B.1)

All NRC commitments on this topic have b een met. The existing reactor coolant system vent pathways are adequate as originally designed.

The elimination of the Steam Condensing Mode of Residual Heat Removal (RHR) operation has eliminated the low flow rate mode of RHR during post-LOCA conditions. This action has eliminated the possibility of having non-condensible gasses come out of solution in the top of the RHR heat exchangers, thereby, eliminating the requirement to maintain an available vent path via the E12-F073 and E12-F074 valves. Refer to UFSAR Section 5.4.7.2.2.3.

L.20 Plant Shielding (II.B.2)

All NRC commitments on this topic have b een met. The shielding design review results and associated modifications are described in Section 12.3.2.5 of the UFSAR.

L.21 Post-Accident Sampling (II.B.3)

LSCS-UFSAR L-4 REV. 15, APRIL 2004 All NRC commitments on this topic have been met. Descriptions of the LSCS high-radiation sampling system (HRSS) are pres ented in Section 11.5.5 of the UFSAR.

L.22 Training for Mitigating Core Damage (II.B.4)

All NRC commitments on this topic have been met. The training program is discussed in Section 13.2 of the UFSAR.

L.23 Relief and Safety Valve Test Requirements (II.D.1)

All NRC commitments on this topic have been met. Additional information is provided in Section 5.2.2.10 of the UFSAR.

L.24 Relief and Safety Valve Position Indication (II.D.3)

All NRC commitments on this topic have been met. Details of implementation plan are incorporated in Section 7.7.15.2.6 of the UFSAR.

L.25 Auxiliary Feedwater System Feasibility Evaluation (II.E.1.1)

Applicable to PWR designs only.

L.26 Auxiliary Feedwater Initiation and Indication (II.E.1.2)

Applicable to PWR design only.

L.27 Emergency Power for Pr essurized Heaters (II.E.3.1)

Applicable to PWR designs only.

L.28 Containment Dedicated Penetrations (II.E.4.1)

All NRC commitments on this topic have been met. Further discussions of the combustible gas control system are provid ed in section 6.2.5 of the UFSAR. Containment penetrations are listed in Table 6.2-21 of the UFSAR.

L.29 Containment Isolatio n Dependability (II.E.4.2)

All NRC commitments on this topic have been met. The containment isolation systems are discussed in Section 6.2.4 of the UFSAR.

L.30 Additional Accident - Monitoring Instrumentation (II.F.1)

LSCS-UFSAR L-5 REV. 15, APRIL 2004 All NRC commitments on this topic have been met. Discussions are incorporated in Sections 11.5.2.2 and 7.5.2 of the UFSAR.

L.31 Inadequate Core Cooling Instruments (II.F.2)

All NRC commitments on this topic have been met. Based on the BWR Owners Group evaluation, no other water level instruments provide as accurate and ready indication of adequate cooling as those installed. Additionally, in-core thermocouples could not provide unambiguous indication of inadequate core cooling.

L.32 Emergency Power for Pressurized Equipment (II.G.1)

Applies to PWR designs only.

L.33 Assurance of Proper ESF Functioning (II.K.1.5)

All NRC commitments on this topic have b een met. Valve position requirements, positive controls, and test and maintenance procedures associated with ESF systems were reviewed to assure proper ESF functioning.

L.34 Safety-Related System Operability Status (II.K.1.10)

All NRC commitments on this topic have been met. Prerequisites for removing safety-related equipment from service are controlled by the Technical Specifications.

L.35 Pressurizer Low-Level Coinci dent Signal Bistables (II.K.1.17)

Applies to PWR designs only.

L.36 Operator Training for Prompt Manual Trip (II.K.1.20)

Applies to PWR designs only.

L.37 Automatic Safety Grade Anticipatory Trip (II.K.1.21)

Applies to PWR designs only.

L.38 Proper Functioning of He at Removal Systems (II.K.1.22)

All NRC commitments on this topic have been met. Manual actions can be taken to mitigate the consequences of a loss of feedwater; however no manual actions are required. Sufficient systems exist to automatically mitigate these consequences.

L.39 Reactor Vessel Level Instrumentation (II.K.1.23)

All NRC commitments on this topic have been met. Reactor vessel level instrumentation is discussed in Section 7.7.1 of the UFSAR.

LSCS-UFSAR L-6 REV. 13 L.40 Control of Auxiliary Feedwater Indepe ndent of the Integrated Control System (II.K.2.2)

Applies to PWR designs only.

L.41 Failure Mode Effects Analysis on the Integrated Control System (II.K.2.9)

Applies to PWR designs only.

L.42 Safety-Grade Anticipatory Reactor Trip (II.K.2.10)

Applies to PWR designs only.

L.43 Thermal Mechanical Report--Effect of High-Pressure Injection on Vessel Integrity for Small-Break Loss-of-Coolant Accident with no Auxiliary Feedwater (II.K.2.13)

Applies to PWR designs only.

L.44 Power Operated Pressurizer Relief Valves (II.K.2.14)

Applies to PWR designs only.

L.45 Effects of Slug Flow on Steam Generator Tubes (II.K.2.15)

Applies to PWR designs only.

L.46 Reactor Coolant Pump Seal Damage (II.K.2.16)

Applies to PWR designs only.

L.47 Potential for Voiding in the Reactor Coolant System During Transients (II.K.2.17)

Applies to PWR designs only.

L.48 Sequential Auxiliary Feedwater Flow Analysis (II.K.2.19)

Applies to PWR designs only.

L.49 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System (II.K.3.1)

Applies to PWR designs only.

LSCS-UFSAR L-7 REV. 15, APRIL 2004 L.50 Report on Overall Safety Effect of Power-Operated Relief Valve Isolation System (II.K.3.2)

Applies to PWR designs only.

L.51 Failure of PORV or Safety Valve to Close (II.K.3.3)

All NRC commitments on this topic have been met. Equipment failures are reported in accordance with 10CFR50.72 and 10CFR50.73.

L.52 Automatic Trip of Reactor Coolant Pumps During Loss-of-Coolant Accident (II.K.3.5)

Applies to PWR designs only.

L.53 Evaluation of Power-Operated Relief Valve Opening Probability During Overpressure Transient (II.K.3.7)

Applies to PWR designs only.

L.54 Proportional Integral Derivative Controller Modification (II.K.3.9)

Applies to PWR designs only.

L.55 Proposed Anticipatory Trip Modification (II.K.3.11)

Applies to PWR designs only.

L.56 Power-Operated Valve Failure Rate (II.K.3.11)

Applies to PWR designs only.

L.57 Confirm Existence of Anticipatory Re actor Trip Upon Turbine Trip (II.K.3.12)

Applies to PWR designs only.

L.58 Separation of HPCI and RCIC System Initiation Levels (II.K.3.13)

All NRC commitments on this topic have been met. At LSCS both the HPCS and RCIC are initiated on low water level. Based on an evaluation performed by General Electric, the proposed separation of RCI C and HPCI initiation is unnecessary for safety considerations.

The RCIC system automatically shuts off on high reactor vessel level. In response to this item, the RCIC system logic was revised to allow RCIC to restart on subsequent low water level as discussed in Se ction 7.4.1.2.3.1 of the UFSAR.

L.59 Modify Break Detection Logic to Prev ent Spurious Isolation of HPCI and RCIC (II.K.3.15)

LSCS-UFSAR L-8 REV. 15 APRIL 2004 All NRC commitments on this topic have been met. No modification is required for HPCS. Modification of RCIC logic has been implemented and is incorporated in Section 5.4.6.2.3 of the UFSAR.

L.60 Reduction of Challenges and Failure s of Relief Valves--Feasibility Study and System Modification (II.K.3.16)

All NRC commitments on this topic have been met. Feasibility study and evaluation were completed. No modifications are required for LSCS.

L.61 Report on Outages of ECCS (II.K.3.17)

All NRC commitments on this topic have b een met. Reporting of plant operational history is performed per Technical Specification 5.6.

L.62 Modification of ADS Logic--Feasibilit y for Increased Diversity for Some Event Sequences (II.K.3.18)

All NRC commitments on this topic have been met. ADS logic is discussed in Sections 7.3.1.2.2 and 6.3.5.2 of the UFSAR.

L.63 Restart of Core Spray and LPCI System (II.K.3.21)

All NRC commitments on this topic have been met. Modification has been implemented. Section 7.3.1. of the UFSAR was revised to reflect the modification.

L.64 Automatic Switchover of RCIC System Suction--Verify Procedures and Modify Design (II.K.3.22)

All NRC commitments on this topic have been met. Modification has been implemented. Section 5.4.6.3 of the UFSAR was revised to reflect the modification.

L.65 Confirm Adequacy of Space Cooling for HPCI and RCIC Systems (II.K.3.24)

All NRC commitments on this topic have been met. An area cooling system that is designated as Engineering Safety Featur e (ESF) cools each ECCS cubicle. ECCS area cooling is discussed in Section 9.4.5 of the UFSAR.

L.66 Effect of Loss of AC Po wer on Pump Seals (II.K.3.25)

All NRC commitments on this topic have been met. Total loss of recirculation pump seals due to a loss of cooling will only result in a minimal loss of water that can be made up by normal or emergency water level controls.

L.67 Provide Common Reference Level for Vessel Level Instrumentation (II.K.3.27)

LSCS-UFSAR L-9 REV. 15, APRIL 2004 All NRC commitments on this topic have been met. A common reference level is used. L.68 Verify Qualification of Accu mulators on ADS Valves (II.K.3.28)

All NRC commitments on this topic have been met. ADS logic is discussed in Section 7.3.1.2.2 of the UFSAR and the accumulators are discussed in Section 9.3.1.2.2 of the UFSAR.

L.69 Revised Small-Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K (II.K.3.30)

All NRC commitments on this topic have been met. Additional discussion of LOCA analysis is presented in section 6.3 of the UFSAR.

L.70 Plant-Specific Calculations to Show Compliance with 10 CFR 50.46 (II.K.3.31)

All NRC commitments on this topic have been met. Results of LSCS-specific small-break LOCA calculations are provided in Section 6.3.3.7 and Table 6.3-8 of the UFSAR.

L.71 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure (II.K.3.44)

All NRC commitments on this topic have b een met. Detailed study and evaluation concluded that for anticipated transients combined with the worst single failure, the core remains covered.

L.72 Evaluation of Depressurizati on with Other than ADS (II.K.3.45)

All NRC commitments on this topic have been met. BWR Owners Group analysis determined that no appreciable improvement is gained by a slower depressurization rate. L.73 Response to List of Concerns from ACRS Consultant (Michelson Concerns) (II.K.3.46)

All NRC commitments on this topic have been met.

L.74 Emergency Prep aredness (III.A.1.1)

All NRC commitments on this topic have b een met. The current revision of the Emergency Plan is discussed in Section 13.3 of the UFSAR.

L.75 Emergency Support Facilities (III.A.1.2)

LSCS-UFSAR L-10 REV. 15, APRIL 2004 All NRC commitments on this topic have been met. Technical Support Center (TSC), Operational Support Center, and Em ergency Operations Facility have been established as described in the Emergency Plan.

Note: The TSC no longer has controlled hu midity or access to steam humidifiers.

L.76 Improving Licensee Emergency Preparedness--Long Term (III.A.2)

All NRC commitments on this topic have b een met. The current revision of the Emergency Plan is discussed in Section 13.3 of the UFSAR.

L.77 Primary Coolant Sources Outside Containment (III.D.1.1)

All NRC commitments on this topic have been met. Discussion of the Primary Coolant Sources Outside Containment Program is included in Section 5.5.2 of the Technical Specifications.

L.78 In-Plant Radiation Monitoring (III.D.3.3)

All NRC commitments on this topic have been met. Discussion of radiation monitoring is discussed in Section 12.5.3.

3.1 of the UFSAR and Section 5.5.3 of the Technical Specifications.

L.79 Control Room H abitability (III.D.3.4)

All NRC commitments on this topic have been met. Discussion of Control Room Habitability is included in Sections 6.4 and 2.2.3 of the UFSAR.

LSCS-UFSAR M.0-1 REV. 13 APPENDIX M ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT Appendix M has been incorporated into Section 3.11 of the UFSAR.

LSCS-UFSAR N.0-1 REV. 14, ARPIL 2002 APPENDIX N INTERIM SECONDARY CONTAINMENT During operation of Unit 1 while Unit 2 was still under construction, temporary structures and procedures were employed to prevent an uncontrolled release path to the environment. These temporary measures were necessary because the Unit 1 and Unit 2 reactor buildings have interconnecting air paths. These measures constituted an "interim" secondary containment. Current details of the design and performance of the secondary containment are presented in Section 6.2.3 of the UFSAR.

RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Appendix a Through Appendix N

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