LR-N17-0034, Salem Generating Station, Units 1 & 2, Revision 29 to Updated Final Safety Analysis Report, Appendix 3A, Pse&G Positions on Us NRC Regulatory Guides
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SGS-UFSAR APPENDIX 3A PSE&G POSITIONS ON USNRC REGULATORY GUIDES Revision 6 February 15, 1987 APPENDIX 3A PSE&G POSITIONS ON USNRC REGULATORY GUIDES (*) Regulatory Guide 1.1 -NET POSITIVE SUCTION HEAD FOR EMERGENCY CORE COOLING AND CONTAINMENT HEAT REMOVAL SYSTEM PUMPS Regulatory Guide 1. 1 requires that the Emergency Core Cooling and Containment Heat Removal Systems be designed so that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss-of-coolant accidents (LOCAs). The Westinghouse design of the Emergency Core Cooling System and Containment Spray System provides adequate NPSH to all system pumps. The NPSH for all of the pumps is evaluated for both the injection and recirculation modes of operation following a LOCA, except for the containment spray pumps, which are only used during the injection mode (see Section 6. 2) The evaluation has shown that the end of the injection mode of operation gives the limiting NPSH available for the centrifugal charging and safety injection pumps. At the end of the injection mode, the suction for these pumps is being provided from the refueling water storage tank. The NPSH available at this time is determined from the elevation head and vapor pressure of the water in the refueling water storage tank, which is at atmospheric pressure, and the pressure drop in the suction piping from the tank to the pumps. The NPSH evaluation is based upon all pumps operating at the design flow rates. The recirculation mode of operation gives the limiting NPSH requirement for the residual heat removal pumps, and the minimum NPSH available is determined from the following calculation: NPSH =(h) -(h) + (h) -(h) available containment vapor static loss pressure pressure head
- The original nomenclature "Safety Guide" was changed to "Regulatory Guide" in December 1972. SGS-UFSAR 3A-1 Revision 24 May 11, 2009 The containment pressure value will be equal to the initial air pressure in containment prior to the LOCA (i.e., the pre-accident partial air pressure in containment). However, when the containment sump vapor pressure exceeds the containment initial pressure then the following is assumed: (h) . =(h) contalnment vapor pressure pressure The containment air pressure value used in the NPSHa calculation is based on the containment conditions prior to the accident only and does not include any credit for accident pressure conditions is conservatively determined based on minimum containment initial pressure, and maximum temperature and relative humidity conditions. The calculation also accounts for further reduction of this initial air pressure based on possible maximum cooldown of the containment environment post-LOCA. The vapor pressure term used in the NPSHa for the sump water being pumped is based on the highest temperature of the sump fluid for the condition being evaluated. The static head term in the NPSHa is calculated using the minimum available water inventory in the containment for recirculation operations. This minimum water inventory ensures that the containment sump strainers are fully submerged prior to initiation of recirculation phase. It is believed that the methods utilized in calculating NPSH meet the intent of the Regulatory Guide, of ensuring adequate NPSH with adequate margin for the centrifugal charging, safety injection, residual heat removal, and containment spray pumps. Regulatory Guide 1.2 -THERMAL SHOCK TO REACTOR PRESSURE VESSELS Although NRC Regulatory Guide 1.2 was withdrawn by the NRC on July 31, 1991, SGS commitments, as stated below, are not affected by this withdrawal. Current Westinghouse research programs and pressure vessel design conform with the intent of the Regulatory Guide. 3A-2 SGS-UFSAR Revision 24 May 11, 2009 Westinghouse is continuing to obtain fracture toughness data through participation in the HSST Program at the Oak Ridge National Laboratory. The fracture toughness data recently obtained include tests on un-irradiated material using specimens up to 12 inches thick. In addition, new testing techniques have evolved which allow the measurement of valid fracture toughness data with much smaller specimens than have been used in the past. irradiated data correspond to startup or beginning-of-life of a plant. These un-Post-irradiation data were obtained from 2-inch thick specimens in 1970. Post-irradiation data on 4-inch thick fracture mechanics specimens are available from the HSST Program in 1974. Elastic plastic test procedure should greatly simplify the problem of obtaining irradiated fracture toughness data because of the associated reduction in required specimen size. Westinghouse is also engaged in an extensive materials irradiation surveillance program from which irradiated fracture toughness data are obtained for actual vessel material. The present data were used in a rigorous linear-elastic fracture mechanics analysis of the reactor vessel thermal shock problem. The results of this analysis show that under the postulated accident conditions, the integrity of the reactor vessel would be maintained throughout the life of a plant. Westinghouse's continuing participation in the HSST Program will yield confirmatory in formation of material properties and fracture mechanics analytical methods. If additional margin against brittle fracture is required, or if the remaining data from the HSST Program does not confirm the present analysis, the reactor vessel can be annealed at any point in its service life. Westinghouse is currently engaged in a research program to determine the optimum annealing time and temperature. No hardware for vessel annealing has yet been designed, but appropriately designed space heaters could be utilized as one conceivable method of annealing. The design of Westinghouse reactor vessels does not preclude post-irradiation heat treatment. 3A-3 SGS-UFSAR Revision 24 May 11, 2009 Regulatory Guide 1.3 -ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1.4 -ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A LOSS-OF-COOLANT ACCIDENT FOR PRESSURIZED WATER REACTORS The Salem Station complies with Regulatory Guide 1.183 instead. Regulatory Guide 1 . 5 -_A_S_S_U_M_P_T_I_O_N_S __ U_S_E_D ___ F_O_R ___ E_V_A_L_U_A_T_I_N_G ___ T_H_E ___ P_O_T_E_N_T_I_A_L RADIOLOGICAL CONSEQUENCES OF A STEAM LINE BREAK ACCIDENT FOR BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1. 6 -INDEPENDENCE BETWEEN REDUNDANT STANDBY (ONSITE) POWER SOURCES AND BETWEEN THEIR DISTRIBUTION SYSTEMS It is believed that the Salem Station design conforms with the intent of the Regulatory Guide. Regulatory Guide 1. 7 -CONTROL OF COMBUSTIBLE GAS CONCENTRATION INCONTAINMENT FOLLOWING A LOSS-OF-COOLANT ACCIDENT The Salem Station design conforms to the intent of the Regulatory Guide as described in Section 6.2. 3A-4 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.8 QUALIFICATION AND TRAINING OF PERSONNEL FOR NUCLEAR POWER PLANTS, revision 2, April 1987 Salem complies with Regulatory Guide 1.8, except as noted below. The Operations Director shall either hold an SRO license or have held an SRO license for a similar unit (PWR) or have been certified at an appropriate simulator for equipment senior operator knowledge. Licensed Operator qualifications and training shall be in accordance with 10CFR55. The Radiation Protection Manager shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The Director-Nuclear Oversight (NOS), and the Engineering Manager positions under the Site Engineering Director, which correspond to the Engineer in Charge, must meet or exceed the qualifications of ANSI/ANS 3.1
-1981. Qualification requirements for the Nuclear Safety Review Board personnel performing the offsite independent review function and PORC members are described in their associated program documents.
See Section 13 for further discussion of staffing of plant personnel.
Regulatory Guide 1.9 - SELECTION, DESIGN, AND QUALIFICATION OF DIESEL GENERATOR SET CAPACITY FOR STANDBY POWER SUPPLIES
The Salem Station design conforms to the intent of the Regulatory Guide as indicated in Section
- 8. Regulatory Guide 1.10 - MECHANICAL (CADWELD) SPLICES IN REINFORCING BARS OF CATEGORY I CONCRETE STRUCTURES Although NRC Regulatory Guide 1.10 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal.
The Salem Station design conforms to the intent of the Regulatory Guide as described in Section 3.8. Regulatory Guide 1.11 - INSTRUMENT LINES PENETRATING PRIMARY REACTOR CONTAINMENT The Salem Station design conforms with the intent of Regulatory Guide 1.11 and General Design Criterion (GDC) 55 for instrument lines. Both containment pressure and RVLIS isolation inside containment is provided by a sealed bellows arrangement. The containment pressure bellows are located immediately adjacent to the inside containment wall. The RVLIS bellows are located near the process pressure sources at the RCS hot legs, in the seal table room and in the reactor cavity. Outside containment isolation for containment pressure is provided by the diaphragm in the pressure transmitter connected to the bellows by a sealed, fluid filled tube. Outside containment isolation for RVLIS is provided by sealed, fluid filled isolators that convey RCS pressure to DP transmitters. The justification for these special arrangements results from the importance of containment pressure and RVLIS indication during accident conditions.
3A-5 SGS-UFSAR Revision 29 January 30, 2017
Regulatory Guide 1.12 -INSTRUMENTATION FOR EARTHQUAKES The Salem Station design conforms with the intent of the Regulatory Guide as described in Section 3.7. Regulatory Guide 1.13 -SPENT FUEL STORAGE FACILITY DESIGN BASIS The Spent Fuel Cooling System design conforms with the intent of the Regulatory Guide as discussed in Section 9. 1. The design of the Fuel Handling System conforms to the recommendations of Regulatory Guide 1.13. Regulatory Guide 1.14 -REACTOR COOLANT PUMP FLYWHEEL INTEGRITY The Salem Station design conforms with the intent of the Regulatory Guide. Regulatory Guide 1.15 -TESTING OF REINFORCING BARS FOR CATEGORY I STRUCTURES CONCRETE Although NRC Regulatory Guide 1.15 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station design generally conforms to the intent of the Regulatory Guide as discussed in Section 3. 8. However, instead of one full diameter specimen from each bar size tested for each 50 tons, or fraction thereof, of rebar produced from each heat as required by the Regulatory Guide, two specimens for each 25 tons or less of heat have been taken for testing. If any of the four specimens failed to meet the specification, the entire heat was rejected. It is believed that this procedure is as conservative as that of the Regulatory Guide. Regulatory Guide 1.16-REPORTING OF OPERATING INFORMATION Information will be reported as indicated in the Regulatory Guide, with the exception of the information provided in the Monthly Operating Report. The Monthly Operating Report information will be reported as indicated in Generic Letter 97-02. 3A-6 SGS-UFSAR Revision 17 October 16, 1998 Regulatory Guide 1.17 -PROTECTION AGAINST INDUSTRIAL SABOTAGE, 6/73 (endorses N18.17) Although NRC Regulatory Guide 1.17 was withdrawn by the NRC on July 5, 1991, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station security plan will conform with the intent of the Guide. Regulatory Guide 1.18 -STRUCTURAL ACCEPTANCE TEST FOR CONCRETE PRIMARY REACTOR CONTAINMENTS Although NRC Regulatory Guide 1.18 was withdrawn by the NRC on July 21, 1981, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station containment structural acceptance test will conform with the intent of the Regulatory Guide as described in Section 6.2. Regulatory Guide 1.19 -NONDESTRUCTIVE EXAMINATION OF PRIMARY CONTAINMENT LINER WELDS Although NRC Regulatory Guide 1.19 was withdrawn by the NRC on July 21, 1981, SGS commitments as stated below, are not affected by this withdrawal. The Salem Station containment liner examinations conform with the intent of the Regulatory Guide. Regulatory Guide 1.20 -VIBRATION MEASUREMENTS ON REACTOR INTERNALS Westinghouse will comply with the requirements of the Regulatory Guide. If, for some overriding reason, deviations from this guide are followed, non-compliance will be justified. 3A-7 SGS-UFSAR Revision 13 June 12, 1994 For each prototype reactor internals design, a program of vibration analysis, measurement, and inspection will be developed and reviewed by the NRC prior to the performance of the scheduled preoperational functional test. Westinghouse has prepared the vibrational analysis and test programs for prototype 2-, 3-, and 4-loop plants. Regulatory Guide 1.21 -MEASURING AND REPORTING OF EFFLUENTS FROM NUCLEAR PLANTS Salem Station's conformance to RG 1.21 is controlled via the Salem Offsite Dose Calculation Manual (ODCM) and its implementing procedures. of effluents is controlled by the ODCM. Routine reporting Emergency effluent reporting is controlled by the Emergency Plan. Regulatory Guide 1.22 -PERIODIC TESTING OF PROTECTION SYSTEM ACTUATION FUNCTIONS The Salem Station Protection System is designed in accordance with IEEE Standard 279-1971. Safety actuation circuitry is provided with a capability for testing with the reactor at power. The protection system design complies with the Regula tory Guide. Under the present design, there are protection functions which are not tested at power. These are described in Section 7. Additionally, the following manual functions are not tested at power: 1. Generation of a reactor trip by tripping the reactor coolant pump breakers 2. Generation of reactor trip by tripping the turbine 3A-8 SGS-UFSAR Revision 28 May 22, 2015
- 3. Generation of a reactor trip by use of the manual trip switch 4. Generation of a reactor trip by manually actuating the Safety Injection System 5. Generation of a safety injection signal by use of the manual safety injection switch 6. Generation of a containment spray signal by use of the manual spray actuation switch Exception is taken to testing the devices listed above, as allowed by the Regulatory Guide, where it has been determined that: 1. "There is no practicable system design that would permit operation of the equipment without adversely affecting the safety or operability of the plant." The present position is that it is not a "practicable system design" to provide equipment to bypass a device such as a reactor coolant pump breaker or a MSIV solely to test the device. In the case of manual initiation switches, the design for test capability would require that switches be provided on a train or sequential basis. This increases the operator action required to manually actuate the function. 2. "The probability that the protection system will fail to initiate the operation of the actuated equipment is, and can be maintained, acceptably low without testing the actuated equipment during SGS-UFSAR reactor operation." The probability of failure of the above devices is considered to be very low. 3A-9 Revision 16 January 31, 1998
- 3. "The actuated equipment can routinely be tested when the reactor is shut down." In all the cases discussed above, it is only the device function which is not tested. The logic associated with the devices has the capability for testing at power. Regulatory Guide 1.23 -ONSITE METEOROLOGICAL PROGRAMS The Salem Station meteorological program will conform with the intent of the Regulatory Guide. Regulatory Guide 1.24 -ASSUMPTIONS USED FOR EVALUATING THE POTENTIAL RADIOLOGICAL CONSEQUENCES OF A PWR RADIOACTIVE GAS STORAGE TANK FAILURE The assumptions used are in agreement with the Regulatory Guide, as described in Section 15. Regulatory Guide 1.26 -QUALITY GROUP CLASSIFICATIONS AND STANDARDS The Salem Station design meets the intent of the Regulatory Guide in that appropriate quality levels have been assigned to components, structures, and systems relative to the importance of their safety functions. Systems, structures, and components were designed in accordance with the codes and standards that were in effect at the time of design and at the equipment order dates, 3A-10 SGS-UFSAR Revision 20 May 6, 2003 which were prior to the inception of the NRC's quality group classification system. The Regulatory Guide was not issued until March 1972, at which time construction was well underway. The codes and standards which were used are presented in the appropriate sections of the FSAR. Regulatory Guide 1.27 -ULTIMATE HEAT SINK (Revision 2) The Salem Station design generally conforms with the intent of the Regulatory Guide (FSAR Section 9.2). Regulatory Guide 1.28 -QUALITY ASSURANCE PROGRAM REQUIREMENTS (DESIGN AND CONSTRUCTION) Salem Generating Station is committed to the requirements of NQA-1-1994 for Quality Assurance Program requirements. Regulatory Guide 1.29 -SEISMIC DESIGN CLASSIFICATION, 8/73 The Salem Station design conforms to the intent of the Regulatory Guide. Previously, the only area of non-conformance with the Regulatory Guide was in the classification of the Spent Fuel Pool Cooling (SFPC) System. SFPC piping and pipe supports are analyzed as seismic class I. SFPC components have been seismically evaluated under SQUG GIP methodology. The basis for this classification is provided in Section 9.1. Regulatory Guide 1.30 -QUALITY ASSURANCE REQUIREMENTS FOR THE INSTALLATION, INSPECTION, AND TESTING INSTRUMENTATION AND ELECTRIC EQUIPMENT, 8/72 (endorses N45.2.4) The Salem Station design conforms with the intent of the Regulatory Guide. 3A-11 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.31 -CONTROL OF FERRITE CONTENT IN STAINLESS STEEL WELD METAL The Regulatory Guide states that weld deposits should contain between 5 and 12 to 15 percent delta ferrite. It is not practical to specify "absolute minimum" or even maximum delta ferrite limits as a basis for acceptance or rejection of otherwise acceptable austenitic stainless steel welds. Westinghouse places control on the actual wire analysis for inert gas welding processes and on the final weld deposit for the fluxing weld process. In the case of the bare wire, when used with inert gas processes, although the wire may contain 5 percent ferrite, only about 1 or 2 percent ferrite will be developed in the resultant weld deposits. This is not the case in fluxing processes such as when using coated arc electrodes or submerged arc, since the flux is enriched with additional ferrite formers resulting in higher ferrite contents in the resultant weld deposits. Similarly, the amount of ferrite that may exist in any given weld will vary across the width of the weld deposit depending upon the base materials being joined. For example, when fully austenitic wrought product is welded, the interface regions will be practically zero percent ferrite because of the resultant base metal dilution, but it will progressively increase toward the weld centerline. Conversely, when a two-phase (austenitic + ferrite) cast product which normally contains over 15 percent ferrite is welded, the interface region will be high in delta ferrite content depending upon the amount of delta ferrite available and diluted from the casting base material. The ferrite distribution in a weld will also vary depending upon the weld position. That is: in areas of the downhand and horizontal position, weld deposit ferrite will be the highest; whereas, in the vertical and overhead position, weld deposit 3A-12 SGS-UFSAR Revision 6 February 15, 1987 ferrite will be the lowest in a given weld because of different welder manipulations necessary to overcome effects of gravity. In addition, types 310 and 330 weld materials are always fully austenitic, yet sound welds are being made every day with these alloys using fine tuned welding procedures. Also, welds are being made without the use of filler metal, such as electron beam welds and autogeneous gas shielded tungsten arc welds. Furthermore, the limits as set are arbitrary because various methods used to measure the percentage of delta ferrite yield widely differing results. The Welding Research Council has recognized this situation and have an organized approach which may result in an acceptable solution. The basis for classifying the low, medium, and high energy input ranges is not given in the Regulatory Guide. Using the Westinghouse conservative welding procedure parameters, the following energy inputs are being applied to produce high quality welds. They are: 1. SMAW 15.4 to 95 kJ/in. using 1/16 to 3/16 dia electrodes 2. GTAW 2.16 to 32.5 kJ/in. using .03 to 1/8 dia wires 3. GMAW 46 to 55 kJ/in. using .03 to 1/16 dia wires 4. SAW 74 to 79 kJ/in. using .09 to 1/8 dia wires Westinghouse has a large amount of evidence showing that the above energy input ranges produce fissure-free weldments in both shop and onsite welding. Westinghouse does not require in-process delta ferrite determination. When the welding material is tested (in accordance with the requirements of ASME Section III, NB2430, and includes delta ferrite determinations), sound welds displaying more than one 3A-13 SGS-UFSAR Revision 6 February 15, 1987 percent average delta ferrite content by any agreed method of determination will be considered unquestionable. All other sound welds which display less than 1 percent average delta ferrite will be considered acceptable provided there is no evidence of malpractice or deviation from procedure parameters. If evidence of the latter prevails, sampling will be required to determine the acceptability of the welds. in the system or component. The sample size shall be 10 percent of the welds If any of these weld samples are defective, that is, fail to pass bend tests as described by ASME,Section IX, all remaining welds shall be sampled and all defective welds shall be removed and replaced. Field welding of the nuclear steam supply system and other nuclear class components is performed using Public Service Electric and Gas (PSE&G) welding procedures. In some areas of austenitic stainless steel welding, these procedures call for use of the 16-8-2 electrode. This particular electrode composition was developed to provide fissure-free welds in austenitic systems without reliance on ferrite content, which is generally limited to 3 percent, and frequently the amount is less than 1 percent. Therefore, ferrite control and determination, which comprise the bulk of the Regulatory Guide, are not considered applicable to the 16-8-2 welding electrode. The 16-8-2 welding electrode was initially developed for service temperatures where delta ferrite exhibits a tendency to transform into the sigma phase, and embrittling condition in austenitic stainless steel. Service temperatures at the Salem Station are too low to support a need for this type of protection, but PSE&G's long service history with this welding composition (since 1955) in steam piping systems has provided a level of confidence and expertise which overrides the consideration of alternate materials. Service and inspection records show that numerous welds have been performed satisfactorily in high pressure steam service temperatures up to 1100°F for operating times exceeding 150,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> in the PSE&G generating systems. 3A-14 SGS-UFSAR Revision 6 February 15, 1987 PSE&G welding and inspection practices comply with the intent of the Regulatory Guide and Appendices A and B to 10CFR50 in the following manner: 1. Strict control is maintained over electrode chemistry and identification for procedure qualification, welder qualification, and production welding. This is accomplished through purchase specifications, untested lots certified mill test from approved lots, reports, segregation of locked storage of welding supplies on site, recorded allocation of electrodes to welders, and maintenance of lot identity from site receiving to completed weld joint. 2. The weld procedure qualification demonstrates the capability of producing welds free from unacceptable fissuring. This includes visual examination of procedure qualification bend bars and macrotech specimens with the unaided eye and under 10 power magnification. 3. Welder performance qualification bend bars, when made, are examined in the same manner to verify that the welder's technique maintains freedom from unacceptable fissures. 4. Welds for nuclear class systems are subjected to a liquid penetrant and radiographic examination where required. Heavy wall welds, such as in the reactor coolant piping, are subjected to in-process examinations by a liquid penetrant and radiography at one or more intermediate stages in the welding out of the groove. 5. SGS-UFSAR Ferrite content for each lot of austenitic stainless steel electrode is qualified by magnegage measurements of a test weld pad. For nuclear plant welds, ferrite 3A-15 Revision 6 February 15, 1987 outside the range of 3 to 12 percent for E-308, E-309, and E-316 is considered rejectable. 6. Production welding parameters are monitored on a spot-check basis by the field welding supervision and the Field Quality Control Group. Regulatory Guide 1.32 -USE OF IEEE STANDARD 308-1971, "CRITERIA FOR CLASS lE ELECTRIC SYSTEMS FOR NUCLEAR POWER GENERATING STATIONS" The Salem Station design satisfies the requirements of IEEE Standard 308-1971, with the exception that Class 1 diesel fuel oil storage capacity provides less than seven days of diesel operation under worst case loading. See Section 9. 5. 4 for a description of how long term Emergency Diesel generator fuel oil storage requirements are met. Regulatory Guide 1.33 -QUALITY ASSURANCE PROGRAM REQUIREMENTS 2/78 (endorses Nl8.7-1976/ANS 3.2) (OPERATION), The Salem Generating Station is committed to the requirements of NQA-1-1994. See the Quality Assurance Topical Report, Appendix C, Section 1.3.2.3 for further discussion. Regulatory Guide 1.34 -CONTROL OF ELECTROSLAG WELD PROPERTIES Electroslag welding of Nuclear Classes 1 and 2 components is confined to the area of reactor coolant piping elbows. These are made from cast clamshells of ASTM A351 Gr. CF-8M joined together on longitudinal seams by the electroslag process. Welding of these components was performed under specified weld procedure control monitored by Westinghouse. PSE&G also established that the shop production welds were in conformance to the procedure qualification. 3A-16 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.35 -_I_N_S_E_R_V_I_C_E _____ S_U_R_V_E_I_L __ LAN ___ C_E ____ O __ F ____ U_N_G_R_O_U __ T_E_D ___ T_E_N __ D_O_N_S ___ I __ N PRESTRESSED CONCRETE CONTAINMENT STRUCTURES This Regulatory Guide is not applicable to the Salem Station containment structures. Regulatory Guide 1.36 -NONMETALLIC THERMAL INSULATION FOR AUSTENITIC STAINLESS STEEL The Salem Station design conforms with the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.37 -QUALITY ASSURANCE REQUIREMENT FOR CLEANING OF FLUID SYSTEMS AND ASSOCIATED COMPONENTS OF WATER-COOLED NUCLEAR POWER PLANTS, 3/73 The Salem Station program for cleaning of fluid systems and associated components conforms to NQA-1-1994 and the intent of the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.38 -QUALITY ASSURANCE REQUIREMENTS FOR PACKAGING, SHIPPING, RECEIVING, STORAGE, AND HANDLING OF ITEMS FOR WATER-COOLED NUCLEAR PLANTS, 10/76 Quality Assurance requirements comply with the requirements of NQA-1-1994. Regulatory Guide 1.39 -HOUSEKEEPING REQUIREMENTS FOR WATER-COOLED NUCLEAR POWER PLANTS, 3/73 Salem Generating Station complies with the requirements of NQA-1-1994. 3A-17 SGS-UFSAR Revision 24 May 11, 2009 Regulatory Guide 1.40 -QUALIFICATION TESTS OF CONTINUOUS -DUTY MOTORS INSTALLED INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS The Salem Station design conforms to the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.41 -PREOPERATIONAL TESTING OF REDUNDANT ON-SITE ELECTRIC POWER SYSTEMS TO VERIFY PROPER LOAD GROUP ASSIGNMENTS The preoperational testing program for the onsi te electric power system will verify the independence of each redundant power source and its ability to supply its associated load group. The actual test methods being developed may not be identical to those specified in the Regulatory Guide for the following reasons. The Salem plant design includes the safeguards equipment controller which incorporates several testing features to assure that loading of the diesel-generator requirements. will be accomplished in accordance with the design It is expected that a series of overlapping tests will adequately substitute for the tests specified in the Regulatory Guide. The Unit 2 initial preoperational test program is in full conformance with the Regulatory Guide which functionally demonstrates the independence among redundant onsite power sources and their load groups. This is accomplished by the performance of the Integrated Safeguards Test. As stipulated in part c.1 of the Guide, isolation from the offsite transmission network will be accomplished by the direct actuation of the undervoltage sensing relays (opening the 4 kV ac undervol tage relay knife switches) . All loads off the Unit 2 group buses not required to maintain necessary and independent construction and testing activities, as well as backup power to Unit 1, will be de-energized to the maximum extent practical. The functional testing requirements covered under c.2 and c.3 of this guide are performed as part of the Integrated Safeguards Test. 3A-18 SGS-UFSAR Revision 6 February 15, 1987 Regulatory Guide 1.42 -INTERIM LICENSING POLICY ON AS LOW AS PRACTICABLE FOR GASEOUS RADIOIODINE RELEASES FROM LIGHT-WATER-COOLED NUCLEAR POWER REACTORS Although NRC Regulatory Guide 1.42 was withdrawn by the NRC on March 22, 1976, SGS commitments, as stated below, are not affected by this withdrawal. Gaseous radioiodine releases from the Salem Station will conform with the regulatory position set forth in the Regulatory Guide. Regulatory Guide 1.43 -CONTROL OF STAINLESS STEEL WELD CLADDING OF LOW ALLOY STEEL COMPONENTS The Salem Units 1 and 2 reactor vessel flange, shell course, head and nozzle surfaces in contact with primary coolant were clad with stainless steel weld metal. The original reactor vessel heads and shell courses were constructed of ASTM A-302 Grade B for the dome and peel sections, and ASTM A-508 Class 2 for the flange (Unit 1), and ASTM A-533 Grade B Class 1 for the dome and peel sections, and ASTM A-508 Class 2 for the flange (Unit 2) plate material made to fine grain practice, and clad by the 3-wire submerged arc process, which is a low heat-input process. This material and cladding process are not restricted by Regulatory Guide 1. 43 and consequently these portions of the vessels comply directly with the guide. The original head and vessel flanges and the primary nozzles for both units were constructed of A-508 Class 2 material, and clad by the manual metal arc (MMA) and manual inert gas (MIG) processes. In addition, the 3-wire submerged arc process was also used on the flanges for both units. The 3-wire submerged arc and MMA processes are low heat-input process to the same degree as MMA and submerged arc processes. The replacement RVCH on both units are constructed from a monoblock forging of SA-508 Gr. 3 Cl. 1 steel rather than by welding formed plates and a flange. The surfaces in contact with reactor coolant were clad with stainless steel weld overlay using either the submerged arc welding process or the shielded metal ark welding process following strict preheat and temperature requirements. Therefore, the Salem Units 1 and 2 replacement reactor vessel closure heads comply with Regulatory Guide 1.43. 3A-19 SGS-UFSAR Revision 22 May 5, 2006 However, as recognized by the Regulatory Guide, and as shown by the extensive metallurgical examinations and fracture mechanics evaluation performed by Westinghouse ( 1), underclad cracking, if present, would be of high integrity and would have no detrimental effect on the structural integrity of the affected components. Thus, the vessels are suitable for the use intended. For Unit 1 steam generators, stainless steel weld cladding is applied to the steam generator channel heads in contact with primary coolant. The heads, including the nozzles and manway openings, are cast ASTM A-216 grade WCC material. The head hemispherical surfaces are clad by the two-wire series submerged arc process with controlled dilution of the deposit, and the channel head nozzles and manway openings are clad by the oscillating submerged arc single wire technique. Both processes are low-heat input techniques. The material and the weld processes are not restricted by the Guide. For the Unit 2 steam generators, the heads, including the tubesheet, nozzles and manway openings are forgings, SA-508, Gr3, Cl.2. The ferritic base metals that are clad are procured to fine grain practice and are not considered susceptible to underclad cracking. Weld procedure qualification is performed on material of the same specification (or equivalent) as used in production. For the Unit 2 steam generators, all primary side ferri tic steel surfaces (primary side of the tubesheet and inside surfaces of the primary head) are clad to prevent corrosion. The tube sheet is clad with Inconels 82 and 182. Due to its sensi ti vi ty to reheat cracking, Inconel 52 is not used for the tubesheet cladding on which the tube-to-tubesheet welding is performed. The most stressed areas of the channel head in contact with the primary coolant are clad (buttering of the tubesheet on which the divider plate is welded) or welded with Inconel 152. These areas are associated with the divider plate to primary head and tubesheet junctions. and 309L stainless steel. The primary head is clad with Type 308L Stainless steel weld cladding is applied to the pressurizer shell courses, heads, spray nozzle, and manway opening surfaces in contact with primary coolant. The shell courses are constructed of SA-533 Class 1 plate, use of which is not restricted by the guide. (1) Westinghouse Nuclear Energy Systems Report WCAP-7733, "Reactor Vessels Weld Cladding-Base Metal Interaction," T. R. Mager, et al., April 1971. 3A-20 SGS-UFSAR Revision 24 May 11, 2009 In addition, the shell courses are clad by the two-wire series submerged arc process with controlled dilution of the deposit, which is a low-heat input technique. The pressurizer heads are cast SA-216 grade WCC material, applied by a low-heat input process, namely the two-wire series submerged arc process with controlled dilution of the deposit. Because of the materials and/or the low-heat input processes used for stainless steel weld cladding of the pressurizer and steam generators, no underclad cracking is expected, and the intent of the Regulatory Guide is met for these components. Regulatory Guide 1.44 -CONTROL OF THE USE OF SENSITIZED STAINLESS STEEL Treatment of sensitized stainless steel components of the Nuclear Steam Supply System (NSSS), particularly the reactor vessel nozzle safe ends, has been covered in detail in the Westinghouse Topical Report, WCAP-7477-L, "Sensitized Stainless Steel in Westinghouse PWR Nuclear Steam Supply Systems," March 1970. This report indicates that, where applicable, Westinghouse and their subcontractors have complied with the intent of the Regulatory Guide. Field erection procedures under the direct supervision of PSE&G and United Engineers and Constructors (UE&C) will comply with the regulatory positions wherever possible. Exposure to sensitizing temperatures in field welding operations will be monitored by testing weld procedure qualification specimens. Regulatory Guide 1.45 -REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS The Salem Station design conforms to the intent of the Regulatory Guide. Regulatory Guide 1.46 -PROTECTION AGAINST PIPE WHIP INSIDE CONTAINMENT Although NRC Regulatory Guide 1.46 was withdrawn by the NRC on March 11, 1985, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station is protected against pipe whip inside containment as described in Section 3.6. 3A-21 SGS-UFSAR Revision 24 May 11, 2009 Regulatory Guide 1.47 -BYPASSED AND INOPERABLE STATUS INDICATION FOR NUCLEAR PLANT SAFETY SYSTEMS The Salem Station design meets the requirements of the Regulatory Guide with the exception of regulatory position C. 4. The Salem Station design does not provide for manual initiation of individual alarms. The Overhead Alarm System has the capability to manually initiate all alarms for testing purposes. It is believed that this arrangement is a suitable alternative to regulatory position C.4. Regulatory Guide 1.48 -DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY I FLUID SYSTEM COMPONENTS Although NRC Regulatory Guide 1.48 was withdrawn by the NRC on March 11, 1985, SGS commitments, as stated below, are not affected by this withdrawal. As stated in Section 3. 9, the piping design commitment for the Salem Station was to utilize the design philosophy of ANSI B31. 1, but with allowances of 1.2 Shand 1.8 Sh for the summation of primary type longitudinal stresses under the OBE and DBE earthquake loadings, respectively. Although the design commitment was made early in the design process, these values conform closely with values published in paragraphs NC-3611.1 (b) (4) (c) (1) and (2) of the Winter 1972 Addendum to ASME Section III. Thus, the suggested stress limits for piping in the Regulatory Guide appear to be satisfied. Despite the above stress limit commitments, in a number of cases actual stresses encountered were below Sh, the code limit for the "normal" condition. It is believed that this fact provides a reasonable assurance of the continued operability of equipment connected to this piping, and, as such, satisfies the intent of the Regulatory Guide. 3A-22 SGS-UFSAR Revision 13 June 12, 1994 Regulatory Guide 1.49 -POWER LEVELS OF WATER-COOLED NUCLEAR POWER PLANTS The license application power levels and ultimate power levels of the Salem units are below those maximum power levels set forth in the Regulatory Guide. Regulatory Guide 1.50 -CONTROL OF PREHEAT TEMPERATURES FOR LOW-ALLOY STEEL WELDING Field welding of low-alloy steel on Nuclear Class systems conforms with the regulatory positions set forth in the Regulatory Guide. Regulatory Guide 1. 51 -INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 NUCLEAR PLANT COMPONENTS NRC Regulatory Guide 1.51 was withdrawn by the NRC on July 15, 1975. 10CFR50.55 a (g) (4) addresses the design, access considerations and Pre-Service Inspection exam requirements, for ASME Section XI Codes that become effective subsequent to the construction of those applicable components. Later editions of the Code have incorporated inspection requirements for Class 2 and 3 components. 3A-23 SGS-UFSAR Revision 16 January 31, 1998 Regulatory Guide 1.52 -DESIGN, TESTING AND MAINTENANCE CRITERIA FOR ATMOSPHERE CLEANUP SYSTEM AIR FILTRATION AND ABSORPTION UNITS OF LIGHT-WATER-COOLED NUCLEAR POWER PLANTS (Revision 1, The Salem Station atmosphere cleanup systems which fall within the scope of the Regulatory Guide are as follows: 3A-23a SGS-UFSAR Revision 16 January 31, 1998 THIS PAGE INTENTIONALLY LEFT BLANK 3A-23b SGS-UFSAR Revision 13 June 12, 1994 Primary Systems: 1. Containment Fan Cooler Units Secondary Systems: 1. Control Room Emergency Filtration Unit 2. Auxiliary Building Exhaust Units 3. Fuel Handling Building Exhaust Units All of these systems conform to the intent of the Regulatory Guide in many respects. The areas where the systems are at variance with the intent of the regulatory positions follow: Regulatory Position C.2.a. In the secondary systems, those portions of the systems designed for mitigation of accident doses are standby units and are not redundant. These filter banks are used in conjunction with other normally operating components which are maintained operable or fail-safe under an accident condition. High-efficiency particulate air (HEPA) or other types of after-filters are not provided downstream of charcoal filter banks. The use of such after-filters has not previously been practiced in Pressurized Water Reactor ( PWR) design. The Salem charcoal filters have been designed for relatively low levels of iodine deposition and the charcoal cells are to be blown free of "fines" during manufacture. Regulatory Position C.2.j. The primary and secondary systems are not "intact" units. components will be maintained or replaced individually. Individual It would be impractical to fabricate, ship, or install "intact" units of the size used in the Salem Station design. that are maintained SGS-UFSAR These units generally consist of modular sections 3A-24 Revision 6 February 15, 1987 individually. consideration. Exposure Regulatory Position C.2.1. of workers during maintenance is taken into Although the leakage rate from the housings in the primary system is designed to be 1 percent of the design flow rate, system ductwork leakage is permitted to be as much as 5 percent of the design flow rate. Because the primary system is a recirculation system entirely within the containment, it is unnecessary to require a low leakage rate from the ductwork. The 5-percent leakage rate allowed is reasonable and practical for sheet metal construction. Regulatory Position C.3.c. Prefilter materials have not been restricted to those on the UL Building Material List. Flammability, corrosion, erosion, radiation resistance, and other aspects of materials generally available have been considered. In general, the Westinghouse guidelines for materials acceptable within containment have been used, as well as recommendations appearing in ORNL-NSIC-65 ( 1) . Regulatory Position C.3.e. Filter and absorber mounting frames will be painted. An epoxy base paint will be used which provides resistance to general rusting and chemical attack, and allows for easy decontamination when necessary. Regulatory Position C.3.h. The use of zinc coated (galvanized) steel has not been discouraged in the Salem Station systems. It is primarily used for ductwork and is not discouraged for such use in ORNL-NSIC-65. SGS-UFSAR 3A-25 Revision 6 February 15, 1987 Regulatory Position C.4.c Vacuum breakers are not provided on the access doors of the filter trains, since it is not intended to allow any person to enter the housings while they are in use, except during performance of certain tests under controlled conditions. Regulatory Position C.5.b. The planned DOP testing for HEPA filters is: 1. For normally operating units, initially and upon replacement (estimated to be after 9 to 18 months' use) 2. For standby units, initially, 18 months, and upon replacement. Filter manufacturers have indicated that HEPA filters exposed to a relatively clean atmosphere can be expected to last for more than 3 years. Intermittent DOP testing of these long life filters will not necessarily detect an adverse change in filter efficiency. The use of silicone sealants is not prohibited in the atmosphere cleanup systems. ORNL-NSIC-65 suggests the use of silicone rubber as a general caulking or sealing compound (1). Regulatory Position C.5.c. The acceptance criteria for in-place DOP testing of HEPA filters is penetration less than 1% as per the guidance of Generic Letter 83-13 for assumed charcoal adsorber efficiency of at least 90% and assumed HEPA filter efficiency of at least 99%. Regulatory Position C.5.d. The acceptance criteria for leak testing of activated charcoal adsorber sections is bypass leakage less than 1% as per the guidance of Generic Letter 83-13 for assumed charcoal adsorber efficiency of at least 90% and assumed HEPA filter efficiency of at least 99%. ( 1) Burchsted, C. A.; and Fuller, A. B., "Design, Construction and Testing of High-Efficiency Air Filtration Systems for Nuclear Application," ORNL-NS1C-65, Oak Ridge National Laboratory, January 1970. 3A-26 SGS-UFSAR Revision 16 January 31, 1998 Regulatory Guide 1.53 -APPLICATION OF THE SINGLE-FAILURE CRITERION TO NUCLEAR POWER PLANT PROTECTION SYSTEMS The Salem Station Protection System design, in general, meets the requirements of the Regulatory Guide. The Salem Station protection system is designed in accordance with the requirements of IEEE Standard 279-1971 which requires that any single failure within the Protection System shall not prevent proper protective action at the system level. The design of the Protection System included the application of the single failure criterion to the logic and actuators. Testing provisions for the Protection System which assure the operability of equipment needed to perform a protective function include continuity checks for those cases, as described in Regulatory Guide 1.22, where there is no practicable system design that would permit testing of actuated equipment during power operation. The testing provisions assure that failures within the Protection System are detectable. Regulatory Guide 1.54 -QUALITY ASSURANCE REQUIREMENTS FOR PROTECTIVE COATINGS APPLIED TO WATER-COOLED NUCLEAR POWER PLANTS, 6/73 (endorses N101.4) ANSI NIOI.4 is being used to provide guidelines requirements for safety-related coatings. for Quality Assurance Regulatory Guide 1.55 -CONCRETE PLACEMENT IN CATEGORY I STRUCTURES Although NRC Regulatory Guide 1.55 was withdrawn by the NRC on July 21, SGS commitments as stated below are not affected by this withdrawal. 3A-27 SGS-UFSAR Revision 13 June 12, 1994 1981, I The concrete placement in Salem Category I structures generally conforms with the intent of the Regulatory Guide. ACI 318 as well as ACI 301 are followed as the basis of the concrete specification. Concrete compression tests are evaluated in accordance with ACI 301-66, Chapter 17. During concrete operations, an inspector was at the batch plant to certify the mix proportions for each batch of concrete. The Inspector took samples of the ingredients and ran tests periodically to determine moisture content of aggregates, applicable. gradation of aggregates, and temperatures of materials when An inspector was also at the construction site to inspect reinforcing and form placements, make slump tests, and take concrete test cylinders in accordance with specified procedures. The project specification for concrete also requires construction joints to be shown on the design drawings. Once the construction joints were established, concrete was poured in accordance with the drawings. Placement of concrete during cold and hot weather is as described in Section 3. 8 of the FSAR. Contact between designer and constructor was well maintained. Experienced field personnel supervised the concrete placement in accordance with the specifications and good practice. Regulatory Guide 1.56 -MAINTENANCE OF WATER PURITY IN BOILING WATER REACTORS This Regulatory Guide is not applicable to PWRs. Regulatory Guide 1.57 -DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS This Regulatory Guide is not applicable to the Salem Station which has a reinforced concrete containment with a steel liner. 3A-28 SGS-UFSAR Revision 13 June 12, 1994 Regulatory Guide 1.58 -QUALIFICATION OF NUCLEAR POWER PLANT INSPECTION, EXAMINATION, AND TESTING PERSONNEL, 9/80 NRC Regulatory Guide 1. 58 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994. 3A-29 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.59 -(Revision 2) DESIGN BASIS FLOODS FOR NUCLEAR POWER PLANTS The design basis floods and the protection requirements for the Salem Station are discussed in Sections 2.4 and 3.4. Regulatory Guide 1.60 -DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS Normalized Housner' s average ground response spectra were used for the Salem Station design. The design response spectra presented in the Regulatory Guide are based on three reference papers published in 1973, and are not applicable to the Salem Station design. Regulatory Guide 1.61 -DAMPING VALUES FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS The damping values used for seismic design of the Salem Station are more conservative than those presented in the Regulatory Guide. 3A-30 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.62 -MANUAL INITIATION OF PROTECTIVE ACTIONS The design of the protection systems for the Salem Station conforms with the intent of the Regulatory Guide. Regulatory Guide 1.63 -ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR WATER-COOLED NUCLEAR POWER PLANTS The Salem Station design conforms with the intent of the Regulatory Guide (See Section 8.1.5). Regulatory Guide 1.64 -QUALITY ASSURANCE REQUIREMENTS FOR THE DESIGN OF NUCLEAR POWER PLANTS, 10/73 SGS is committed to the requirements of NQA-1-1994. Regulatory Guide 1.65 -MATERIALS AND INSPECTIONS FOR REACTOR VESSEL CLOSURE STUDS The Regulatory Guide was published after procurement of the reactor vessel bolting material for the Salem Units 1 and 2. However, this material meets the intent of the Regulatory Guide as follows: The reactor vessel closure stud bolts for both units were machined from bars of SA 540 Grade B24 material. The closure nuts and washers were machined from tubes of SA 540 Grade B23 material. The bolting material qualification Section III Code and Addenda ( 1965 tests were performed per the Summer Addenda for Unit 1 and ASME 1966 Summer Addenda for Unit 2) in effect at the 3A-31 SGS-UFSAR Revision 25 October 26, 2010 time of the bolt procurement, which required meeting an average of 35 ft-lbs energy with no lateral expansion tests required and no maximum tensile strength limitation. code requirements. All bolting material for both units met the ASME Charpy tests were performed at 10°F on bolting material bar and tube specimens (three impact tests per bar or tube end) as required by the ASME Code. For the bar material from which the Unit 1 studs were made, all data from the bars tested were in excess of 45 ft-lbs, which conforms directly with the impact energy requirements of the guide. For Unit 2 material, six out of nine bars tested showed impact values of 45 ft-lbs or greater at both ends. Of the remaining three bars, the one with the lowest impact energy (at one end) showed values of 40, 43, and 45 ft-lbs. For the tube material from which the Unit 1 nuts and washers were made, none of the tubes that were tested showed impact energy values of 45 ft-lbs or greater on both ends of the tube. The lowest energy values, obtained on one end of one tube, were 38, 38, and 38 ft-lbs. For Unit 2, seven of eleven tubes tested showed impact energy values at both ends of 45 ft-lbs or greater. Of the remaining tubes, the one with the lowest impact data showed values of 40, 42, and 45 ft-lbs (at one end) . For the bars and tubes showing 10°F impact data averaging below 45 ft-lbs, Westinghouse believes that the intent of the guide is met, inasmuch as sufficient fracture toughness is expected at the preload temperature or at lowest service temperature, both of which are significantly above the 10°F Charpy test temperature. Also, the impact energies at the preload or lowest service temperatures will be higher than was obtained at the lower, actual Charpy test temperature. 3A-32 SGS-UFSAR Revision 13 June 12, 1994 The bolting materials for both units were inspected to the requirements of the ASME Code and Addenda requirements in effect at the time of bolt procurement. A radial scan covering 100 percent of the circumferential surface of the bars, based on a standard back reflection, was performed. The requirements for protection of the stud bolts and stud bolt holes against corrosion and in-service inspection requirements for the bolting material are discussed in Section 5. Regulatory Guide 1.66-NONDESTRUCTIVE EXAMINATION OF TUBULAR PRODUCTS Although NRC Regulatory Guide 1. 66 was withdrawn by the NRC on September 28, 1977, SGS commitments, as stated below, are not affected by this withdrawal. Tubular products and fittings have been nondestructively examined in accordance with the requirements of the applicable paragraphs of ANSI B31.7. Regulatory Guide 1.67 -INSTALLATION OF OVERPRESSURE PROTECTION DEVICES Although NRC Regulatory Guide 1.67 was withdrawn by the NRC on April 15, 1983, SGS commitments, as stated below, are not affected by this withdrawal. The Salem Station design conforms to the intent of the Regulatory Guide. Any deviations from the design recommendations of the Regulatory Guide have been analytically justified. Regulatory Guide 1.68 -PREOPERATIONAL AND INITIAL STARTUP TEST WATER-COOLED POWER REACTORS PROGRAMS The Salem Station Preoperational and Initial Startup Test Program 3A-32a SGS-UFSAR Revision 13 June 12, 1994 FOR conforms with the intent of Revision 0 of the Regulatory Guide with the following exceptions: Appendix A A.1.c-Vibration Tests -Tests will be performed as indicated in Section 14.4, Exception 3, and Section 3.9. 3A-32b SGS-UFSAR Revision 26 May 21, 2012 A.4.a -through A.4.h A. 5.p -A.6.e-A.l3 -B.l.j -D.l.a-D.l.n-D.l.p-Power Conversion System -System operability tests will be performed; expansion and restraint tests are not planned. Leak Detection Systems -Extensive testing on an individual basis of sumps, drainage system, and instrumentation will be accomplished during the testing program. Emergency Power Systems -Tests of this system as presently planned do not provide for the carrying of all required loads for several hours. Radioactive Waste Systems -No tests are planned at this time to verify the amount of plateout in sample system piping. This is a long-range program which is not adaptable to the preoperational and initial startup phases of testing. Vibration Monitoring -Reference FSAR Section 14.4, Exception 3, and Section 3.9. Natural Circulation Tests Reference Section 14. 4, I Exception 1. Dropped Rod-Reference Section 14.4, Exception 7. Vibration Monitoring -Reference Section 14.4, Exception 3, and Section 3. 9. The Unit 2 initial test program is in conformance with the Regula tory Guide with the following exceptions: Paragraph 2a -The shutdown margin shall be verified by boron analysis only to ensure the boron concentration is as required by the Technical Specifications. 3A-33 SGS-UFSAR Revision 10 July 22, 1990 Paragraph 2f -Vibration levels and piping reaction to transient conditions are evaluated prior to fuel loading as discussed in Sections 3.9 and 5.2. Differential pressure instrumentation is not part of the Salem Station design for monitoring the differential pressure across the reactor vessel and steam generators. The cold and hot leg temperatures are used for monitoring fouling of the major components. Pump flow measurements are calculated during the test program which will verify that the differential pressure across the major components is not excessive. Paragraph 4a -The boron coefficient will be determined at the normal hot zero power temperature. The boron coefficient over the temperature range allowed by the Technical Specifications for criticality does not vary significantly to require additional measurements to be made. Paragraph 4u -See Exception 6a, Section 14.4. Paragraph 5c -See Exception 1, Section 14.4. Paragraph 5c -This requirement is not applicable to the Salem Station design. Paragraph 5d-This requirement is only applicable to BWRs. Paragraph 5e -See Exception 5, Section 14.4. 3A-34 SGS-UFSAR Revision 10 July 22, 1990 Paragraph 5g -Adequately covered prior to Power Ascension Program plus Normal Technical Specification and/or operational surveillance. specified power levels provides no additional information. Demonstration at Paragraph 5i -The system used for determining the individual rod positions is the Rod Position Indication System. The Technical Specifications require that the system be operable to detect control rod position within the Technical Specification limits. Paragraph 5j -A specific test to verify that the reactivity control system functions in accordance with design will not be included in the test program. These systems are continuously being tested during the test program whenever power level changes are made. In addition, response to 5g applies. Paragraph 5m -No additional testing of the Reactor Coolant System is planned subsequent to initial criticality. See response to 5g. Paragraph 5q -Failed fuel detectors are calibrated, alarms verified operable, and are continuously monitored while in service. required. No special tests are Paragraph 5r applies. See Exceptions, Section 14.4. In addition, response to 4o Paragraph 5s -See response to 5g. Paragraph 5t -Relieving capacities not verified. to 5g. Paragraph 5u -See response to 5g. 3A-35 SGS-UFSAR Additionally, see response Revision 25 October 26, 2010 Paragraph Sv -Accomplished during various other power ascension procedures, but not necessarily conducted at designated power levels. response to Sg. Additionally, see Paragraph Sw -Station design does not permit measurement of component temperatures. Shielding and penetration cooling systems are demonstrated operable during preoperational testing. Paragraph Sx -Room coolers, Service Water System pumps, Core Cooling System, and integrated safeguards verify minimum operating components Engineering design specified flow rate. assures adequate See response to Sg. Paragraph Sec -See response to Sg. heat removal Paragraph Sdd-See response to Regulatory Guide 1.68.2. Paragraph See -See response to Sg. Paragraph Sff -See response to Sg. capacities available. at minimum Paragraph Sgg The demonstration of the operability of equipment for anticipated transients without a reactor trip is still under discussion between the NRC and Westinghouse and other NSSS vendors. This test requirement will not be performed until agreement is reached, if required. Paragraph Sii -See Exceptions, Section 14.4, and response to Sg. Paragraph Skk -No plans for this demonstration. The response of the station to bypassing a single feedwater heater that results in the most severe credible case of feedwater temperature reduction is analyzed in Section 1S. The consequences of this incident were shown to be more moderate than those considered for the Excessive Load to a 10-percent step change in are SGS-UFSAR Increase Accident, which is equivalent station power. The 10-percent step changes 3A-36 Revision 10 July 22, 1990 performed during the Phase III test program and are considered adequate to check station responses. Paragraph 511 -See Exception 6, Section 14.4. Paragraph 5mm The main steam isolation valves will close during power operation in response to any one of the following signals being generated by the station reactor protection system or operator initiation: 1. 2/4 high steam line flow in coincidence with 2/4 low-low Tavg or I 2/4 low steam line pressure. 2. Hi-Hi containment pressure. 3. Manual actuation from the control room console (4 separate pushbuttons only.) Signal 1 will also generate a safety injection signal which initiates a reactor trip. Signal 2 is generated after a Hi containment pressure signal is generated which will initiate a safety injection and reactor trip. For Signals 1 and 2, a reactor trip will precede the closing of the main steam isolation valves. Signal 3 will only be initiated when the operator manually initiates each signal to each main steam isolation valve from the control room console. is done by procedure, in which case a reactor trip will have occurred. This 3A-37 SGS-UFSAR Revision 16 January 31, 1998 The closing of the main steam isolation valves while at power will not be performed because in all cases a reactor trip will precede the closing of the main steam isolation valves during plant operation. Paragraph 5oo -See response to 2f. Regulatory Guide 1.68.2 -INITIAL STARTUP TEST PROGRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILITY FOR WATER-COOLED NUCLEAR POWER PLANT The initial startup test program conforms to Objectives C.1.a and C.1.b of the Regulatory Guide. The design of the plant, however, does not permit Objective C.1.c, verification of cold shutdown capability, to be performed. The Salem Station was designed for remote hot shutdown from outside the control room. This is described in Section 7.7 of the FSAR. Our capability to go to a cold shutdown condition through the use of procedures and temporary modification is described in Section 7. General Design Criterion 19 of 10CFR50 Appendix A requires a design capability for remote hot shutdown with a potential capability for subsequent cold shutdown through suitable procedures. A detailed procedure was written explaining the actions to be taken to bring the station from a hot shutdown to a cold shutdown condition from outside the control room. In order to demonstrate that the actions described in the operating procedure could be performed, a walk through encompassed a visual inspection of the areas that are required to be manned by the operators during various phases of remote cold shutdown. It demonstrated the availability and access to the equipment and also the communication required. No physical operation or changes will be made. The major steps for remote cold shutdown are the following: 3A-38 SGS-UFSAR Revision 6 February 15, 1987
- 1. Trip the reactor (this may be accomplished from the control room or locally -the reactor trip will also cause a turbine trip). 2. Reactor Coolant System temperature will decrease to the no-load value through automatic operation of the Steam Dump System or main steam atmospheric relief. Hot standby temperatures will be maintained by automatic system operation and can also be maintained by local manual control of the main steam atmospheric relief valves with termination of steam dump (closure of dump valves or MSIVs). 3. The Main Feedwater System isolation valves will close when Reactor Coolant System temperature reaches 554°F. Steam generator levels will be maintained by local manual operation of the Auxiliary Feedwater System. 4. Maintain station status from the remote hot shutdown panels. 5. Borate Reactor Coolant System to cold shutdown condition manually. 6. Take manual actions necessary to prevent inadvertent Safety Injection System operation as required in accordance with normal shutdown procedures. 7. Cooldown within appropriate limits through manual control of main steam atmospheric relief valves. 8. Maintain steam generator levels and pressurizer levels during cool down through manual control charging. 3A-39 SGS-UFSAR of auxiliary feedwater and Revision 16 January 31, 1998
- 9. Reduce pressurizer temperature and pressure through manual control of pressurizer spray valves within appropriate limits. 10. When Reactor Coolant System pressure reaches :c:; 1000 psig, manually close accumulator isolation valves. 11. Initiate Residual Heat Removal System operation locally when Reactor Coolant System temperature is below 350°F and pressure is below 375 psig. 12. Take actions necessary to arm Pressurizer Overpressure Protection System when Reactor Coolant System temperature is 312°F and pressure is less than 375 psig. 13. Bring unit to cold condition and maintain via local control. Regulatory Guide 1.69 -CONCRETE RADIATION SHIELD FOR NUCLEAR POWER PLANTS The Salem Station design and construction of the concrete radiation shields conforms with the intent of the Regulatory Guide. Regulatory Guide 1.70-STANDARD FORMAT AND CONTENT OF SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS The Salem FSAR was submitted prior to publication of this Regulatory Guide, and was not, therefore, prepared in the standard format suggested by the guide. The FSAR does, however, contain essentially all of the information suggested in the Regulatory Guide, either in the text or in the form of responses to specific questions from the regulatory staff. However, PSEG will utilize Regulatory Guide 1.181 in conjunction with Revision 1 of NEI 98-03, Guidelines for Updating Final Safety Analysis Reports, as guidance for maintaining the UFSAR in accordance with 10 CFR 50.71(e). 3A-40 SGS-UFSAR Revision 28 May 22, 2015 Regulatory Guide 1.71-WELDER QUALIFICATION FOR AREAS OF LIMITED ACCESSIBILITY The welding supervisor's knowledge of the area and his assignment of welders to the job was relied upon to assure that satisfactory welds are made in locations of limited or restricted accessibility. Welds with limited accessibility may require radiography beyond Code requirements at the discretion of the site welding department. Site welding department personnel audited the welding parameters on production welds to assure that they were within the parameters established in the "Welding Procedure Specifications." Regulatory Guide 1.72-SPRAY POND PLASTIC PIPING This Regulatory Guide is not applicable to the Salem Station. Re gu 1 at or y Guide 1 . 7 3 --"Q'-'-U-=-A=L:..:I:..:F:....:I=-C:..:A_:_T=-=-I -'-0-=-N----=T-=E:..:S-=T'-'-S'--_O:_:F=----E=L=E--'-C-=T-'-R-"'I'--'C'---V'-"A-=L=-V-'--E=---O-=-=-P=E-=-RA=-=-=T-=O:..::R_:c:_S INSTALLED INSIDE THE CONTAINMENT OF NUCLEAR POWER PLANTS Environmental qualification of motor operated valves located inside the containment is discussed in Section 3.11 and Section 7. Regulatory Guide 1.74-QUALITY ASSURANCE TERMS AND DEFINITIONS, 2/74 NRC Regulatory Guide 1.74 was withdrawn by the NRC. requirements of NQA-1-1994. SGS is committed to the Regulatory Guide 1.75-PHYSICAL INDEPENDENCE OF ELECTRIC SYSTEMS 3A-41 SGS-UFSAR Revision 23 October 17, 2007 As discussed in Section 7. 5. 3 .1, the Salem Station electric systems do not conform to the recommendations in Regulatory Guide 1.75, since this was not an original design criterion. New equipment will be integrated into our existing separation provisions. The Salem separation criteria have been approved by the NRC staff as described in Section 7. 8 of the original and Supplement 1 and Section 8.4.5 of Supplements 3 and 4 of the Safety Evaluation Report. Regulatory Guide 1.78, Revision 1, December, 2001-Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release The Salem Generating Station design conforms to the intent of the Regulatory Guide. 3A-42 SGS-UFSAR Revision 26 May 21, 2012 Regulatory Guide 1.79-PREOPERATIONAL TESTING OF ECCS FOR PWRs This guide is addressed in Section 6.3. Regulatory Guide 1. 82 -SUMPS FOR EMERGENCY CORE COOLING AND CONTAINMENT SPRAY SYSTEMS This guide is addressed in Section 6.3. Regulatory Guide 1.83 -INSERVICE INSPECTION OF PWR STEAM GENERATOR TUBES (Revision 1) Inservice inspection of steam generator tubing will be performed in accordance with the Technical Specifications, which are in general conformity with Regulatory Guide 1.83. Regulatory Guide 1.88 -COLLECTION, STORAGE, AND MAINTENANCE OF NUCLEAR POWER PLANT QUALITY ASSURANCE RECORDS, 10/76 NRC Regulatory Guide 1. 88 was withdrawn by the NRC on July 31, 1991. SGS is committed to the requirements of NQA-1-1994, Supplement 175-1, Section 4, Storage, Preservation, and Safekeeping, with the following specific exceptions for the Records Storage Room No. 145 in the Nuclear Administration Building: 1. Per NUGEG-0800, Records Storage Room No. 145 was built to comply with option ( 3) "a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> rated fire resistant file room meeting NFPA 232 ... ". Regulatory Guide 1.88 endorses NFPA 232-1975 and NQA-1-1994 endorses NFPA 232-1986; however, during construction, NFPA 232-1991 was utilized to provide an acceptable level of record protection, 3A-42a SGS-UFSAR Revision 24 May 11, 2009
- 2. A cable tray which passes through the room is enclosed with a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> rated symmetrical wrap system to assure its presence will not effect the room contents or fire protection features, and 3. The ceiling is pierced by several miscellaneous drainage lines and two ventilation ducts. A drip pan, with discharge outside the room, is provided for the miscellaneous drainage plumbing to minimize the potential for inadvertent wetting of records and fire dampers are installed in the ventilation ducts. Regulatory Guide 1.91 -EVALUATION OF EXPLOSIONS POSTULATED TO OCCUR ON TRANSPORTATION ROUTES NEAR NUCLEAR POWER PLANTS This subject is discussed in Section 2. Regulatory Guide 1.94 -QUALITY ASSURANCE REQUIREMENTS FOR INSTALLATION, INSPECTION, AND TESTING OF STRUCTURAL CONCRETE AND STRUCTURAL STEEL DURING THE CONSTRUCTION PHASE OF NUCLEAR POWER PLANTS, 4/76 Major modifications made to the Salem Station will comply with NQA-1-1994. 3A-42b SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.97 -INSTRUMENTATION OF LIGHT-WATER COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT Implementation of this Regulatory Guide is described in Section 7. Regulatory Guide 1.99 -EFFECTS OF RESIDUAL ELEMENTS ON PREDICTED RADIATION DAMAGE TO REACTOR VESSEL MATERIALS The basis as well as the scope of the Regulatory Guide for predicting adjustment of reference temperature are inappropriate since the data base used was incomplete and included some data which were not applicable. The primary data were obtained from the 6-inch A302B ASTM reference correlation monitor material reported to be irradiated at 550°F and do not appear to include significant operating plant data which have been collected by industry. Specifically, it appears from our evaluation of the Regulatory Guide that data obtained from various Westinghouse surveillance capsules which were irradiated at low fluences ( 1 x 1019 n/ cm2) were not included in the development of the slope of the curve. Also, data from the Yankee Rowe surveillance capsules, which were irradiated at 450°F to 525°F (1) were inappropriately included in the development of the slope for the ASTM reference material. As indicated in the Regulatory Guide itself, irradiation at 450°F has been shown to cause twice the adjustment of reference temperature when compared to irradiation at 550°F. Inclusion of the low fluence surveillance capsule data and detection of the high fluence (low temperature) Yankee Rowe capsule data in the development of the curve results in a curve which is significantly different in slope than the curve in the Guide which results in higher predicted adjustment of reference 3A-43 SGS-UFSAR Revision 6 February 15, 1987 temperature at high fluences (2). Preliminary analyses of irradiation data being performed by the Metal Properties Council Task Group, which has been assigned the task of developing radiation damage curves, confirm the Westinghouse contention that the slope of the curves presented in the Regulatory Guide would result in unrealistic adjustments in reference temperature. With reference to the Guide Position controlling residual elements to levels that result in a predicted adjusted reference temperature of less than 200°F at end-of-life, Westinghouse contends that the stresses in the vessel can be limited during operation in order to comply with the requirements of Appendix G to 10CFR50 even though the end-of-life adjusted reference temperature may exceed 2 0 0°F. By applying the procedures of Appendix G to ASME Section III, the stress limits including appropriate Code safety margin can be met. Westinghouse believes that Figure 2 of the Regulatory Guide is incorrect since upper shelf energy for the 6-inch thick ASTM A302B reference correlation monitor material reported by Hawthorne indicates essentially a constant upper shelf as fluences above approximately 19 2 1.0 x 10 n/cm (3). Concerning the Guide definition of upper shelf energy, Westinghouse believes, on the basis of extensive experience, that a curve should be fit to the existing data using best engineering judgment. Normally, at least three specimens would be included above the upper end of transition region; additional specimens would be included when the shelf level appears to be marginal. As an alternative to Regulatory Guide 1. 99, operating limits for Unit 2 have been determined by using the current radiation damage curves as shown on Figure 3A-1. These limits are provided in the Technical Specifications. 3A-44 SGS-UFSAR Revision 6 February 15, 1987 References 1. Letter to the Secretary of the Commission by C. E. Eicheldinger, NS-CE-748, Figure 1, September 22, 1975. 2. C.Z. Serpan, Jr., and Hawthorne, J. R., "Yankee Reactor Pressure Vessel Surveillance Notch Ductility Performance of Vessel Steel and Maximum Service Fluence Determined from Exposure During Cores II, III, and IV," NRL Report 6616, September 29, 1967. 3. Hawthorne, J. R., "Radiation Effects Information Generated on the ASTM Reference Correlation-Monitor Steels," to be published. Regulatory Guide 1.100 -SEISMIC QUALIFICATION OF ELECTRICAL NUCLEAR POWER PLANTS EQUIPMENT FOR The Unit 2 design does not fully comply with the Regulatory Guide which endorses IEEE Standard 344-1975, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations." Safety-related electric equipment for Salem Unit 2 (other than Westinghouse supplied NSSS electric equipment) was seismically tested or analyzed based on IEEE Standard 344-1971. Regulatory Guide 1.101 -EMERGENCY PLANNING AND PREPAREDNESS FOR NUCLEAR POWER PLANTS Salem conforms to Regulatory Guide 1.101, Revision 3, August 1992, and used as the planning basis "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", NUREG-0654/FEMA REP.-1, Rev. 1 (November 1980); and "Methodology for Development of Emergency Action Levels", NUMARC/NESP-007. The Emergency Plan Manuals, as revised, describe the total emergency program as described in Section 13.3. 3A-45 SGS-UFSAR Revision 16 January 31, 1998 THIS PAGE INTENTIONALLY LEFT BLANK 3A-46 SGS-UFSAR Revision 16 January 31, 1998 Regulatory Guide 1.102 -FLOOD PROTECTION FOR NUCLEAR POWER PLANTS (Rev. 1) Flood protection, as described in Section 3.4, conforms to this Regulatory Guide. Regulatory Guide 1.105 -INSTRUMENT SETPOINTS The instrument range, accuracy, and set points for the protection system at Salem Station were determined on the bases of system design, accident analyses, and Technical Specification requirements. Unit 2 meets the intent and is in general conformance with the Regulatory Guide subject to the following: 1. The Technical Specifications establish a setpoint and an allowable value providing a set margin allowance for inaccuracy of the instrument, calibration uncertainties, and instrument drift. Periodic functional tests and checks specified in the Technical Specifications are expected to detect any instrument drift exceeding the Technical Specification margins that may occur during the interval between required calibrations. 2. Instrument range is based on the span necessary for the instrument to perform its intended function. 3. The necessary qualifications testing of selected instruments to verify instrument performance and accuracy under adverse conditions was performed in general agreement with the requirements of IEEE Standard 323-1971, "IEEE Trial-Use Standard Guide for Qualifying Class 1 Electric Equipment for Nuclear Power Generating Stations," instead of IEEE Standard 323-1974 as recommended in Regulatory Guide 1.89. 4. Control of setpoints is provided through a combination of any of the following: administrative procedures, cabinet/panel locks, and setpoint locking devices. 5. The setpoints and the accuracy of the instruments are adequate to SGS-UFSAR meet the assumptions used in the safety analyses. The Technical Specifications provide sui table margins to account for the i terns in Regulatory Position 1. Documentation of drift rates and their relationship to testing intervals is not provided in the setpoint analyses. 3A-47 Revision 26 May 21, 2012 Regulatory Guide 1.106-THERMAL OVERLOAD PROTECTION FOR ELECTRIC MOTORS ON MOTOR-OPERATED VALVES The regulatory position requires ensuring that safety-related motor operated valves whose motors are equipped with thermal overload protection devices integral with the motor starter will perform their function. provided under regulatory position 1 and regulatory position 2. Two options are Regulatory Position 1 provides the option to bypass the thermal overload protection (a) continuously (except during testing) or (b) under accident conditions. This option can be implemented provided that the completion of the safety function is not jeopardized or that other safety systems are not degraded. Regulatory Position 2 provides the option to keep the thermal overload in service provided the trip setpoint of the thermal overload protection device is established with all uncertainties resolved in favor of completing the safety-related action and the thermal overload protection device should be periodically tested. The MOV control circuit original design bypassed the Thermal Overload Relay (TOR) contacts in order to comply with Regulatory Guide 1.106, position 1. In the mid 1990's, modifications were performed to the MOV control circuits to re-establish the TOR. The sizing of the Thermal Overload Heater (TOH) for the safety related MOVs is based on IEEE standard 741 guidance. Periodic testing of the TOR and TOH has been established. Current thermal overload protection for safety-related motors at Salem is in accordance with RG 1. 106 Regulatory Position 2. Regulatory Guide 1.108 -PERIODIC TESTING OF DIESEL GENERATORS USED AS ONSITE (Revision 1) ELECTRIC POWER SYSTEMS AT NUCLEAR POWER PLANTS The Unit 2 design meets the intent of the Regulatory Guide. Periodic testing requirements for the diesel-generators will be identified in the Technical Specifications. 3A-48 SGS-UFSAR Revision 26 May 21, 2012 The Unit 2 initial preoperational test program is in conformance with Regulatory Guide 1.108 with the following exceptions: Paragraph c2.a(4) -Compliance with the section will be by tripping the diesel output breaker at 2750 kW (2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating) verifying that the voltage regulation is maintained within the acceptable limits and the allowable overspeed does not trip out the diesel. We feel this transient is more severe than the load shedding requirements identified in the Regulatory Guide. Paragraph c. 2a ( 5) -The test described in this section will be performed, but due to the sequence of testing it may not be immediately after the test described in c.2.a(3) The generator systems will, however, be at full load temperatures. Paragraph c.2.a(6) -The station is not designed to perform the test described in this section. Paragraph c.2.a(9) To accomplish frequency of surveillance testing this will reliability demonstration, be increased to acquire the the 23 consecutive successful valid starts per diesel prior to proceeding beyond the Zero Power Physics Test Program. Credit will be taken for those diesel starts accomplished to date or scheduled during Integrated Safeguards Testing, as long as the diesels are loaded to a minimum of approximately 25 percent and the run durations are approximately 30 minutes or more. All additional starts will comply with the Regulatory Guide criteria for valid tests. Regulatory Guide 1.114 -GUIDANCE ON BEING OPERATOR AT THE CONTROLS OF (Revision 2) CONTROLS OF NUCLEAR POWER PLANT Salem Station conforms with the intent of this Regulatory Guide. 3A-49 SGS-UFSAR Revision 6 February 15, 1987 Regulatory Guide 1.115 -PROTECTION AGAINST LOW TRAJECTORY TURBINE MISSILES (Revision 1) The Unit 2 design conforms with the intent of the Regulatory Guide. Regulatory Guide 1.117 -TORNADO DESIGN CLASSIFICATION The Unit 2 design conforms to the requirements of the Regulatory Guide. Specific details of the missile protection methods used for various Unit 1 and Unit 2 components are discussed in Section 3.5.2.2. Regulatory Guide 1.118 -PERIODIC TESTING OF ELECTRIC POWER AND PROTECTION SYSTEMS Periodic testing requirements are identified in the Technical Specifications. Regulatory Guide 1.121 -BASIS OF PLUGGING DEGRADED PWR STEAM GENERATOR TUBES The Salem Station design conforms to the intent of this Regulatory Guide. Regulatory Guide 1.124 -DESIGN LIMITS, LOADING COMBINATIONS, AND (Revision 1) SUPPLEMENTARY CRITERIA FOR CLASS I LINEAR-TYPE COMPONENT SUPPORTS Component support design conforms to the intent of the Regulatory Guide. Specific provisions to determine allowable stress increase factors for faulted conditions are used. Regulatory Guide 1.127 -INSPECTION OF WATER-CONTROL STRUCTURES ASSOCIATED WITH NUCLEAR POWER PLANTS This Regulatory Guide does not apply to the Salem design. 3A-50 SGS-UFSAR Revision 23 October 17, 2007 Regulatory Guide 1.130 -DESIGN LIMITS AND LOADING COMBINATIONS PLATE-AND-SHELL-TYPE COMPONENT SUPPORTS FOR CLASS I The Unit 2 methodology and criteria for the design of the Reactor Coolant System piping and equipment supports is discussed in Sections 3. 7, 3. 9, and 5. 5. Regulatory Guide 1.137 -FUEL-OIL SYSTEMS FOR STANDBY DIESEL GENERATORS, 10/79 Diesel fuel oil sampling is performed as follows: a. A fuel oil sample is taken from each truck delivering fuel oil to Salem whenever possible. However, if several trucks arrive at once, a minimum of 1 in 4 trucks is sampled. b. All newly received fuel oil is pumped into the 20,000 barrel Fuel Oil Storage Tank. every 30 days. Fuel oil in this tank is sampled at least once c. A small percentage of the fuel oil in the 20,000 barrel tank is introduced into the diesel fuel oil storage system as necessary. This small percentage is added infrequently to the four 30,000 gallon Diesel Fuel Oil Storage Tanks (two for each unit) as necessary to maintain the minimum level above the 23,000 gallon limit in each Diesel Fuel Oil Storage Tank as specified by the Salem Technical Specifications. d. Fuel oil in the four 30,000 gallon Diesel Fuel Oil Storage Tanks is sampled as required by the Salem Technical Specifications. 3A-51 SGS-UFSAR Revision 16 January 31, 1998
- e. All fuel oil samples taken in actions a through d are sent to an independent lab ora tory within 4 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the time the sample is taken. The analysis performed is consistent with Regulatory Guide 1.137 and ASTM D975-77 and the analysis report is submitted to the Salem Station within 30 days of receipt of the sample at the laboratory. f. All fuel oil deliveries, samples taken, and related analysis reports are logged at the station. When reports indicate that fuel oil quality is not within acceptable limits, station management will take appropriate action to restore it to within acceptable limits. g. Actions a through f are subject to verification during routine monitoring and audits of the fuel oil program and procedures conducted by NQA personnel. Regulatory Guide 1.141 -CONTAINMENT ISOLATION PROVISIONS FOR FLUID SYSTEMS The Salem Station design utilizes the Regulatory Guide for guidance in acceptable containment isolation design configurations as detailed in Section 3.0 in ANSI Standard N271-1976. Specific guidance is provided in this Section for those penetration designs which fall under the heading of "Other Defined Basis" as detailed in ANSI Standard N271-1976. Regulatory Guide 1.143 -DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN LIGHT -WATER -COOLED NUCLEAR POWER PLANTS In accordance with guidance provided in this Regulatory Guide, the Contaminated Floor and Equipment Drain Systems and small portions of the Liquid Waste Disposal System that are designated with Piping Schedule 53D (Piping Specification SPS53) have been reclassified to be Non-Nuclear (Quality Group D). The Salem Station design meets the intent of the Regulatory Guide. Augmented quality assurance requirements have been imposed to ensure that the quality level recommended in the Regulatory Guide is maintained. 3A-51a SGS-UFSAR Revision 16 January 31, 1998 Regulatory Guide 1.144 -"AUDITING QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS", 9/80 (endorses N45.2.12) NRC Regulatory Guide 1.144 was withdrawn by the NRC on July 31, 1991. committed to the requirements of NQA-1-1994. SGS is Regulatory Guide 1.145 -"ATMOSPHERIC DISPERSION MODELS FOR POTENTIAL ACCIDENT CONSEQUENCE ASSESSMENTS AT NUCLEAR POWER PLANTS", 11/82 (reissued 2/83) The atmospheric dispersion calculations are consistent with the Regulatory Guide, as described in Section 2.3. Regulatory Guide 1.146-"QUALIFICATION OF QUALITY ASSURANCE PERSONNEL FOR NUCLEAR POWER PLANTS", N45.2.23) PROGRAM AUDIT 8/80 (endorses NRC Regulatory Guide 1.146 was withdrawn by the NRC on July 31, 1991. committed to the requirements of NQA-1-1994. SGS is Branch Technical Position APCSB 9.5-1, Appendix A, "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976". The QA Program is applied to the Fire Protection Program to an extent consistent with the requirements of Section C of Appendix A to Branch Technical Position APCSB 9.5-1. Regulatory Guide 1.155 -Salem Station complies Section 3.12. SGS-UFSAR "Station Blackout", August 1988. with Regulatory Guide 1.155 3A-51b as described Revision 23 October 17, 2007 in Regulatory Guide 8.8 -INFORMATION RELEVANT TO ENSURING THAT OCCUPATIONAL RADIATION EXPOSURES AT NUCLEAR POWER STATIONS WILL BE AS LOW AS IS REASONABLY ACHIEVABLE (ALARA) It is the philosophy of management that in the design and operation of the Salem Station, personnel exposure to radiation be kept to ALARA. Regulatory Guide 8. 8 will be followed when applicable and practical as judged by the personnel responsible for the ALARA program. The Salem Station design is in compliance with Regulatory Guide 8. 8 to the extent that a great deal of consideration was given to reducing radiation exposures and levels to ALARA by design, layout, and selection of materials. Extensive radiation shielding has been installed and a detailed radiation contamination survey program has been established to verify that the radiation levels are within the specified design limits. Periodic radiation surveys are taken to assure that the radiation levels are within the specified limits and to identify potential problem areas. Maximum equipment and compartment dose rates have been calculated and shielding installed to reduce radiation levels to the designated zone limits for the particular area. Separation of components by shielding, the use of reach rods, and separate valve compartments have been used to reduce exposures for operation and maintenance of the unit. Labyrinth compartment entrances and locked wire mesh doors are used in addition to administrative procedures to minimize exposure from high radiation sources. Pipes carrying radioactive fluids are routed through shielded pipe chases or hot equipment compartments to maintain low radiation levels in corridors and aisles. Instrument transmitters and direct readout radiation monitors are generally located in low radiation areas. A balanced ventilation system and floor and equipment drains are used to handle airborne contamination and radioactive liquids, respectively. Other design features employed to maintain ALARA dose rates and exposures include shielded remote filter handling equipment, decontamination room, back flush connection on major radioactive system piping, and automatic resin and sludge drumming and solidification. Local sample connections for most samples have been moved from hot compartments to lower radiation areas. Local portable shielding is available for areas where extended maintenance may have to be performed in high radiation areas. In addition to these design features, administrative procedures are employed for personnel exposure monitoring and reducing or maintaining low in-plant radiation exposures. 3A-52 SGS-UFSAR Revision 16 January 31, 1998 The Plant Manager has responsibility at the station level for the establishment and execution of the station Radiological Protection Program. He implements the Program to meet the requirements of 10CFR20, Regulatory Guides, and the Technical Specifications. The Radiation Protection Manager (RPM) is responsible for implementation of the Nuclear Business Unit Radiation Protection Program at the Salem and Hope Creek generating stations. The RPM reports to the Plant Manager in matters of radiation protection and ALARA. The RPM maintains liaison with other NBU departments and upper management regarding the program, and directs the Radiation Protection Program through the Radiation Protection Superintendents. The RPM is the member of the NBU management who is responsible for the ALARA program, and reviews the program to determine how personnel dose may be reduced. The Radiation Protection Supervisors are responsible to the RPM for the day-to-day functions in the ALARA, Operations and Support areas. The RPM (or his designate): 1. Ensures ALARA pre-job planning of major work items that are estimated to give large man-rem exposures 2. Establishes an effective Radiation Work Fermi t Program to control individual exposures 3. Ensures that adequate controls are in place to prevent the uncontrolled release of radioactive materials 4. Performs periodic formal assessments of the content and implementation of the NBU ALARA program including recommendations for improvements. 5. Ensures that ALARA requirements and concepts are appropriately included in the NBU ALARA program. 3A-53 SGS-UFSAR Revision 21 December 6, 2004 Design and equipment changes are reviewed by the Plant Operations Review Committee (PORC). A Radiation Protection Program has been established and implemented for both Salem and Hope Creek. Procedures are written and approved covering such subjects as access control, radioactive material control, dosimetry, surveys, and radiation exposure permits. Through the use of these established procedures, and with guidance and supervision by the Radiation Protection Superintendents and Supervisors, work will be performed in accordance with ALARA philosophy. The typical portable radiation survey instruments are described in Section 12. The criteria for selection of the portable instruments include: 1. Ability of instrument to perform in its intended use with reliability and accuracy 2. Ease of calibration and repair 3. Interchangeability of components 3A-54 SGS-UFSAR Revision 24 May 11, 2009
- 4. Weight and size for user acceptance 5. Standard readouts and controls/adjustments to simplify training of users The installed radiation monitoring instrumentation and locations are described in Sections 11 and 12. The access control for the areas where radiation and contamination are possible is described in Section 12. The single point of access to and exit from the controlled area is located at Elevation 100 feet (ground level) in the Service Building. The traffic pattern for male and female workers is basically the same with the exception of the locker room/change area. The facilities available and used to provide proper radioactive material and contamination control are established and are common to both units. REGULATORY GUIDE 1.181 -Content Of The Updated Final Safety Analysis Report in Accordance With 10 CFR 50.71(e) The PSEG Nuclear procedures for updating the UFSAR are based on NEI 98-03, Revision 1, which is endorsed by Regulatory Guide 1.181. The purpose of NEI 98-03 is to provide licensees with guidance for updating their FSARs consistent with the requirements of 10CFR50.71(e). Guidance is also provided for making voluntary modifications to UFSARs (i.e., removal, reformatting and simplification of information, as appropriate) to improve their focus, clarity and maintainability. REGULATORY GUIDE 1.183 -ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR EVALUATING DESIGN BASIS ACCIDENTS AT NUCLEAR POWER REACTORS The assumptions used are in agreement with the Regulatory Guide, as described in Section 15. 3A-55 SGS-UFSAR Revision 28 May 22, 2015