ML20195E246
| ML20195E246 | |
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| Issue date: | 06/08/1988 |
| From: | Office of Nuclear Reactor Regulation |
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?A DRAFT SAFETY EVALUATI6.1 REPORT ON CHAPTER 4 0F EPRI's UTILITY REQUIREMENTS DOCUMENT prepared by the Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission June 1988 l
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TABLE OF CONTENTS P_ age PREFACE ................................................ v
- 1. INTRODUCTION ...................................... 1
- 3. BWR REACTOR PRESSURE VESSEL AND INTERNALS .........- 5 4.- BWR CORE AND FUEL ................................. 7
- 5. BWR CONTROL R0D DRIVE SYSTEM ...................... 10
- 6. PWR REACTOR PRESSURE VESSEL AND INTERNALS ......... 13
- 7. PWR CORE AND FUEL ................................. 15
- 8. PWR CONTROL R0D DRIVE SYSTEM ...................... 18
- 9. CONCLUSIONS ....................................... 20
- 10. REFERENCES ........................................ 21 Appendix A - Generic Safety Issues Pertaining to ALWR URD Chapter 4 EPRI ALWR CH 4 DSER iii 06/08/88
l PREFACE This Draft Safety Evaluation Report (DSER) is the fourth in a series the U.S.
Nuclear Regulatory Commission (NRC) plans to issue on its reviews of the 13 chapters of the Advanced Light Water Reactor (ALWR) Utility Requirements Document (URD). These utility requirements apply to boiling-water-reactor (BWR) and pressurized-water-reactor.(PWR) plants in sizes up to 1350 MWe; however, an 1100 MWe plant size with a six-flow turbine was used in establishing some
- requirements that are based, in part, on economic considerations.
The first DSER in this series, issued in September 1987, addressed the URD Executive Summary and Chapter 1 regarding the overall objectives and require-ments of the ALWR program. Chapter 2, "Power Generation Systems," was reviewed in the second DSER, which was issued in February 1988. The third DSER, issued in May 1988, covered Chapter 3, "Reactor Coolant System and Reactor Non-Safety Auxiliary Systems." This DSER on Chapter 4, "Reactor Systems," is being issued in June 1988.
l The staff is currently reviewing Chapter 5, "Engineered Safety Systems." Based on the Electric Power Research Institute (EPRI) plan to submit the remaining eight chapters during 1988, the staff expects to complete the last DSER in the l series by September 1989. As a result of comments made by NRC and others, l EPRI will revise the 13 chapters. The staff will consider the changes made while it is preparing the final SER.
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EPRI ALWR CH 4 DSER v 06/08/88
DRAFT SAFETY EVALUATION REPORT ON CHAPTER 4, "REACTOR SYSTEMS"
- 1. INTRODUCTION On June 18, 1987 (Ref. 1), EPRI* submitted URD Chapter 4 for-staff review.
Chapter 4 defines the utility-generated requirements for design of the reactor systems for advanced light-water-reactor (ALWR) plants. The scope of this chapter covers the following for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs): reactor pressure vessel, nozzles and safe-ends, reactor internalc, in-vessel portions of fluid systems (including reactor internal pumps, emergency core cooling system (ECCS) piping, and spargers),
nuclear fuel, control rods, and the control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection, and testing are also covered, except for refueling and refueling-related tools, which are covered in Chapter 7.
On November 13 and December 11, 19P,7 (Refs. 2 and 3, respectively), the NRC staff asked EPRI to supply additional information relative to Chapter 4. EPRI provided the information in its responses dated January 25, 1988 (Ref. 4) and March 28, 1988 (Ref. 5). Earlier, EPRI had submitted letters dated July 8, 1986 (Ref. 6) and July 9, 1987 (Ref. 7) transmitting topic papers on generic safety issues pertaining to Chapters 1, 3, and 4 and one optimization subject paper for Chapter 3. Appendix A to this draft safety evaluation report (DSER) lists those issues that pertain to Chapter 4, it indicates where EPRI's elements of resolution for them are included in the utility requirements, and it sumarizes the status of NRC's review of EPRI's resolutions.
The staff's approach to reviewing the utility requirements document (URD) is described in Sections 1.3 and 10 of the DSER for Chapter 1. As noted therein, the staff is using the NRC standard review plan (5RP, Ref. 8) as guidance for the review. However, the URD places primary emphasis on preventing significant problems that have been experienced in existing plants; many details are
- For convenience, the term "EPRI" is used in this document to designate the Electric Power Research Institute and/or its ALWR Utility Steering Committee. I EPRI ALWR CH 4 DSER 1 06/08/88
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lacking that will have to be provided later for specific designs. The staff, therefore, has considered the utility requirements at the level of detail I presented in order to identify any actual or potential conflicts with NRC requirements, but not to determine their adequacy to meet all such requirements.
- 2. REQUIREMENTS COMMON TO BWRs AND PWRs Section 2 of Chapter 4 defines the utility requirements for the ALWR that are cominon to BWR and PWR reactor systems. As a general requirement, this section specifies that the reactor systems are to be designed to perform the following functions:
Generate the thermal power necessary to meet required plant electrical power output while not exceeding specified nuclear, thermal / hydraulic, and mechanical design limitations.
Serve as a pressure boendary and a barrier to prevent the release of radicactivity from the reactor core or reactor coolant.
Provide a flow path for the forced circulation of coolant to remove heat generated by the reactor core under all operat':ng conditions and facilitate removal of decay heat by natural or forced circulation from the core after shutdown.
Provide for control of core reactivity.
With respect to instrumentation and controls (I&C), Chapter 4 invokes require-ments that will affect their type, location, and configuration but notes that all requirements for instrumentation, including final sensors, will be covered in Chapter 10. The staff did not find any I&C requirements in Chapter 4.that conflict with NRC requirements; t,ut the staff will defer its conclusions until Chapter 10 is reviewed.
Under the heading of "Performance Requirements," Section 2.2 of Chapter 4 prescribes general design requirements that reflect the intent of the criteria in 10 CFR Part 50, Appendix A, as they apply to reactor systems. For example, paragraph 2.2.1 establishes the basic "defense-in-depth" principle for the ALWR by specifying two separate barriers against the release of fuel fission products: the fuel cladding and the reactor coolant pressurs boundary. The other topics addressed in this section are the following: pressure boundary integrity, negative power coefficient, freedom from power oscillations, margin EPRI ALWR CH 4 DSER 2 06/08/88
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operation and transients, reactivity control reliability, shutdown margin, and criticality margin.
The staff reviewed the requirements in Section 2.2 of Chapter 4 and concluded that none of them conflict with NRC requirements. However, the staff does not consider paragraph 2.2.4, "Freedom from Oscillations," to be adequate. It specifies that "the reactor core shall, as a minimum, be designed to be controllable to compensate for power oscillations without exceeding specified fuel design limits." The staff's position is that it is not sufficient that the core be controllable, particularly if this would require operator action.
The design must ensure that fuel safety limits will not be exceeded for any power oscillation, either because of physical limitations or because it is prevented by an automatic cafety-grade system, such as a reactor scram or a power runback. (open item)
Section 2.3 of Chapter 4, "Equipment Design Requirements , addresses naterials, the reactor pressure vessel (RPV) and its internals, and core and fuel design.
In response to staff comments, EPRI will modify several requirements in this section (Refs. 4 and 5), as discussed below.
Paragraph 2.3.1.7 presently specifies that the "RPV design shall be such that special protection systems and controls for low temperature overpressure protection (LTOP) are not required." However, the staff comented that main-taining a low nil ductility transition temperature (RTliDT) does not eliminate the need for special LTOP. In response, EPRI will replace the specification in par 6gnph 2.3.1.7 with a reference to paragraph 3.3.2.3 in Chapter 3, which states that LTOP shall be provided by the residual heat removal (RHR) pressure relief system. In addition, paragraph 3.3.2.4 in Chapter 3 and Chapter 5 will be modified to require that the minimum calculated end-of-life pressure set points for LTOP protection be considered in determining the size of the RHR relief capacity.
Also pertaining to the LTOP issue, paragraph 2.3.1.6 of Chapter 4 will be replaced by the following:
Except for the core belt region in the PWR, the reactor pressure vessel shall be designed and fabricated such that EPRI ALWR CH 4 DSER 3 06/08/88
o, the initial nil ductility transition temperature (RTNOT) at the most limiting location is not greater than 10*F. The '
initial RT NDT in the reactor vessel core belt region for the PWR shall not exceed -20 7. The calculated end-of-life (60 years of service) shift in RT caused by irradiation for NDT core belt materials (calculated RT HDT as specified in Regulatory Guide 1.99, Revision 2) shall not exceed 30 F for both the PWR and BWR.
In addition, paragraph 2.3.1.2 of Chapter 4 will be modified to require a maximum of 0.03 percent copper in the core beltline forging of the PWR.
As indicated in the staff's DSER for Chapter 3 (Section 3), EPRI's proposed changes relative to LTOP are reasonable; however, a final judgment on their adequacy cannot be made at this time because new NRC requirements are being considered by the staff as part of the resolution of generic safety issue (GSI)
- 94. (open item) l Paragraph 2.3.1.8 requires that a material surveillance program be established l to monitor reactor vessel irradiation and its effect on the vessel material properties. EPRI will expand this requirement to state that the surveillance program shall comply with 10 CFR Part 50, Appendix H, and American Society for Testing and Materials (ASTM) standard E-185. Adherence to the specifications in these documents provides assurance of the structural integrity of the reactor vessel throughout the plant life.
Paragraph 2.3.1.8.1 includes a requirement relative to fracture toughness specimens inserted into the reactor vessel. In response to a staff coment, EPRI will add the following sentence to the paragraph: "Fracture toughness shall be determined using the J IC Method in accordance with ASTM E-813."
In paragraph 2.3.2.1.1 the URD specifies that the RPV and its non-renovable internals shall have a design life of 60 years. Since the ALUR is expected to receive an additional 20 years of neutron irradiation beyond that of current operating plants, the staff commented that an irradiation dosage limit for the RPV internals (including bolting) should be stated. In response, EPRI will EPRI ALWR CH 4 DSER 4 06/08/88
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add a statement requiring the plant designer to specify such a dosage limit based on data from operating nuclear plants.
Paragraph 2.3.2.1.4 requires that RPV designs must provide significant margin in meeting fatigue design criteria without compromising other aspects of the design. This margin is required to account for uncertainties in predicting reactor service cycles and conditions that include extended life and load-follow duty. The actual margin will be evaluated for acceptability during the staff's review of each individual plant.
The connon requirements in Section 2.3.2 also cover items such as vessel fabrication, head seals and leakage monitoring, autcmated in-service inspections, the refueling cavity seal, reactor bolting, and insulation. None of these were found to conflict with regulatory requirements. The same was true of the common requirements in Section 2.3.3 relative to core and fuel design, fuel handling, and pellet cladding interaction resistance.
- 3. BWR REACTOR PRESSURE VESSEL AND INTERNALS The reactor pressure vessel is the major element of the reactor coolant system pressure boundary containing and supporting the reactor core and reactor internals. It also provides for reactor coolant supply and a floodable volume to keep the core covered. The internals provide the supporting elements and devices inside the RPV that, together with the reactor core, reactor coolant system, and RPV instrumentation, perform the function of nuclear steam generation.
Section 3 of Chapter 4, together with the applicable portions of Section 2, provides the utility requirements for the RPV and its internals and instrumen-tation for the BWR version of the ALWR plant. The staff reviewed these utility requirements and found that none of them conflict with NRC requirements.
However, several items deserve further discussion, as indiceted below.
The performance requirements in Section 3.2 specify that "the RPV shall be capable of satisfying all functional requirements under all normal and transient operating conditions as defined in Chapter 1." In addition, the RPV EPRI ALWR CH 4 DSER 5 06/08/88
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steam volume plus the steamline volume is required to be large enough that the changes in RPV pressure with time, during normal operations and events, do not exceed values acceptable for safety and operational analyses. EPRI has informed the staff that the RPV will be larger in diameter than current BWR pressure vessels in order to accomodate the reactor internal pump (RIP) motors Consequently, the RPV water inventory will be increased and its wall fluence will be lower.
4 With respect to the RPV internals, the performance roquirements call for a configuration design that will provide stable natural circulation in the normal operating regions of the power-flow map. In addition, this design is to have the capability to transfer smoothly to the forced circulation regime following startup of the RIPS and to reestablish natural circulation following a trip of the RIPS. The purpose of these requirements is to minimize dependence on active systems and to provide greater margin for accomodating operating transient conditions.
Section 3.3 covers equipment design requirements. Among these is paragraph 3.3.1.2 which specifies that nozzles, safe-ends, thermal sleeves, and spargers shall be designed for the full number of design life cycles without replacement.
The safe-ends are expected to eliminate pipe cracking as a result of inter-granular stress corrosion cracking (IGSCC). For the feedwater nozzles and the core cooling nozzles, the safe-ends are to be of a "tuning-fork" design that has proven to be effective where thermal cycling occurs.
Paragraph 3.3.1.4.2 specifies that the main steamline flow limiters shall be part of the reactor steam outlet nozzles. One advantage of this combination is that the rate of steam flow into the containment following a steamline rupture would be reduced. It would also reduce the dynamic loads on reactor internals and containment structures.
Paragraph 3.3.1.9.1 requires that a power-assisted machine be provided that can be placed on the RPV closure head. This machine would be remotely controlled to disconnect and remove vessel stud nuts. The objective of this requirement is to reduce the critical path refueling time by reducing the head removal time.
EPRI ALWR CH 4 DSER 6 06/08/88
Paragraph 3.3.2.4.3 specifies that the feedwater sparger shall be designed with top exit holes followed by flow guides to aim the feedwater radially inward. The URD states that this design prevents reactor coolant from flowing back into the feedwater spargers and pipes, prevents temperature cycling with resultant cracking at low feedwater flow rates, and prevents water hammer in the feedwater piping.
Paragraph 3.3.3.2.1 requires that several sets of RPV water level instrumentation be provided. One would be a wide-range set, consisting of four divisions with instrument taps located in each of four quadrants, which will provide signals for reactor protection and safety systems. This will permit the use of any two-out-of-four logic and eliminate "l/2 scram" situations (e.g., during instrument testing).
Paragraph 3.3.3.3 specifies that temperature measurement instrumentation shall be provided only if it is necessary for plant operating procedures (e.g., to monitor metal temperature differentials during plant startup and shutdown).
Because of the improved RT of the ALWR, the URD states that thermocouples NDT would not be required for protecting the RPV from brittle fracture; therefore, it may be possible to eliminate the thermocouples completely. However, the staff recommends that protection of tne RPV from brittle fractures not be eliminated because of improved material; hence, thermocouples and a materials surveillance program are necessary.
- 4. BWR CORE AND FUEL The core and the fuel are required to generate heat up to a rated value, throughout planned operating cycles, with sufficient margin and control to accommodate normal operations and the safety analysis events listed in Table 3.1 of URD Chapter 1.
The performance requirements specify (in Section 4.2.1.2) that the core characteristics shall allow stable operation (decay ratio < l.0) for all expected op /ating conditions. This indicates all areas of the operating map except the "excluded region" shown in Figure 4.4-2 of Chapter 4. (Figure 4.4-2 is reproduced on page 8 of this DSER.)
EPRI ALWR CH 4 DSER 7 06/08/88
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The staff recognizes that instabilities could possibly exist in the "excluded region" because of off-normal operating conditions (e.g., very low moderator temperature) or large errors in the analysis methods. However, the staff maintains that allowing an operating region that is thermal-hydraulically unstable by design is highly undesirable. Attenpts should, therefore, be made to prevent unstable operation by design, even in the "excluded region."
The performance requirements also specify that the plant designer shall calculate a margin of at least 15 percent to the above stability limits.
This is acceptable if the margin is determined after including calculational uncertainties. Some current methods used to calculate decay ratios for thermal hydraulic stability have been judged by the staff to have 20 to 25 percent uncertainties.
Paragraph 4.2.1.A.2, in the section on load following and maneuvering capability, specifies that no preconditioning shall be required of the fuel for maneuvering.
The URD explains that the intent of this requirement is to remove previous related limitations on plant maneuvering so that components and systems other than the fuel establish the plant maneuvering limits. The staff recognizes that this is a desirable objective; whether it can be approved for a specific plant will depend largely on the fuel design.
In the section on nuclear and thermal hydraulic design, paragraph 4.2.1.6.2 specifies that the core design shall provide for extended cycle operation at reduced power or with reduced feedwater temperature. Operation with reduced feedwater temperature would result in less thermal-hydraulic stability.
Therefore, the benefits of extending cycle operation by reducing the feedwater temperature must be weighed against the undesirability of decreasing core stability. This will be a matter for consideration of specific plant designs.
Section 4.2.2 establishes performance requirements for fuel reliability, burnup and lifetime. Included is a requirement that premature fuel failure that results from manufacturing defects shall be less than 1 per 30,000 fuel rods, with a goal of 1 per 50,000. The URD also states that recent industry experience has shown that 1 failure per 30,000 is an achievable reliability.
In addition, Section 4.2.2 specifies that the basic fuel mechanical design is EPRI ALWR CH 4 DSER 9 06/08/88
to be capable of peak bundle-average burnups of at least 45,000 MWD /MTU. This burnup capability is stated to be consistent with present or expected near-term experience. On the basis of the specified burnup rate, the URD requires the fuel rods and fuel assembly structural components to be designed for a minimum core residence time of seven years.
EPRI proposes (Ref. 7) that its requirements in Section 4.2.2, and similar requirements for PWRs in Section 7.2.2 of Chapter 4, are sufficient to consider generic issue B-22 to be resolved for the ALWR. This issue was established to track industry efforts to improve the reliability of predictions of fuel performance during normal operations and postulated accident conditions. On the basis of current industry experience, there has been a substantial improvement in fuel reliability since B-22 was initiated. The staff is continuing to monitor fuel performance; however, the staff agrees with EPRI that generic issue B-22 is resolved for the ALWR.
Section 4.3 of Chapter 4 covers BWR equipment design requirements for the core and fuel, neutron sources, and nuclear instrurrentation. None of these were found to conflict with current regulatory requirements.
- 5. BWR CONTROL R0D DRIVE SYSTEM The control rod drive (CRD) systen includes the electrohydraulic control rod drives, rod drive motors, hydraulic control units, hydrualic supply system, scram and scram pilot solenoid valves, air header dump valves, interconnected piping, and associated instrumentation, including rod position and separation sensors. In a BWR, the CRD system is required to nerform the following functions:
Withdraw and insert the control rods at normal rate for operetional control.
Control and indicate the positions of the control rods throughout the full stroke.
Insert the control rods for shutdown (scram) at the high rate required to maintain fuel integrity.
j EPRI ALWR CH 4 DSER 10 06/08/88
Control position of sel e ted rods for core thermal-hydraulic stability control.
Control position of ganged-rod groups for faster rod position changes.
Provide for insertion of control rods by an alternative and diverse method on receipt of ATHS (anticipated transient without scram) signals.
Supply measured purge wcter to the reactor internal pumps.
Section 5.2 of Chapter 5 establishes performance requirements and Section 5.3 describes equipment design requirements for the CRD system. The staff reviewed these utility requirements and found that none of them, with the possible exception of a requirement in paragraph 5.3.5.3 relative to scram pilot solenoid valves, conflict with NRC requirements. That exception and several other items that deserve special mention are discussed below.
In paragraph 5.2.1.2 there i a requirement that limiting conditions for operation be developed to define the acceptable number and arrangement of CRDs that are found to exceed the maximum scram times during test or operation.
The rationale for allowing the use of such CRDs recognizes that a more precise calculation may be carried out by considering the actual performance of nearby control rods, measurement errors, 6.d the current core operating conditions.
EPRI anticipates that thir use of CRDs with scram times in excess of the maximum may help to meet the plant availability goals (e.g., 87 percent annual average over the life of the plant) without reducing safety.
Paragraph 5.2.1.4 specifies that the scram performance and design of the CRD and hydraulic system shall accommodate either an all-hafnium type or a boron carbide type of control rod. The hafnium-type control rods have slower scram times because they are heavier.
Electric motor drives are specified in paragraph 5.3.1.1 for withdrawal and insert notion at nornal speed. The rationale in the URD indicates that this specification is based upon favorable experience with electric motor drives in BWRs overseas. The ability to move rods in small increments permits more precise core power shaping and reduces the tendency for fuel cladding cracking associated with large increments. In addition, the use of electric motor drives enables simpler seals to be used and allows changing these without removing the CRDs.
EPRI ALWR CH 4 DSER 11 06/08/88
Paragraph 5.3.1.2 specifies that the scram action of the CRD shall be achieved by water hydraulic pressure taken fron gas-charged accumulators. It also allows each accumulator to provide scram pressure for several CRDs if the concept is adequately supported by a safety evaluation. EPRI stated that this concept is being utilized successfully in overseas plants.
Paragraph 5.3.5.3 states that the scram pilot solenoid valves shall be designed for continuous operation at the minimum and maximum voltages and frequencies required by the reactor protection system. The rationale stated for this is to avoid overheating and consequent damage to valve materials observed in earlier plants. The staff questioned EPRI whether this requirement was meant to replace the electric protective assemblies (EPAs) that were used in past BWR designs to prevent an overvoltage, undervoltage, or underfrequency condition from failing the scram pilot solenoid valves in a non-fail-safe state.
EFRI's response described a failure mode that had already been eliminated for the advanced boiling-water-reactor (ABWR) design, but it was not the original cause for the addition of the EPAs. In a subsequent discussion with the staff, EPRI stated that the intent of this section of the URD was te preclude the overheating and binding of the solenoids by requiring that they be designed to operate continuously over the full range of voltages and frequencies that could be put out by the RPS power supplies. They intend to deoonstrate this through a failure modes analysis of the power supplies. The staff expressed its skepticism of being able to demonstrate that there are no failure modes of the reactor protection system (RPS) power supplies that would resdit in a non-fail-safe failure of the scram pilot solenoids. This item will be considered further during review of the RPS pcwer supply requirements included in Chapters 10 and 11 of the URD. (open item)
Provision for an alternate means of rod insertion using a separate and diverse means (from normal scram by the reactor trip system) is required by paragraph 5.3.6. This is consistent with the NRC regulations in 10 CFR 50.62, "Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants."
EPRI ALWR CH 4 DSER 12 06/08/88 1
Several requirements in Section 5.3.9 of Chapter 4 have the objective of reducing personnel radiation exposure during maintenance and reducing critical path time during refueling. These include providing a machine for automated removal and reinstallation of the CRDs and convenient location of the CRD servicing area to the CRD removal machine. This section also requires the provision of adequate working space below the CRD mechanisms to permit removal of the motors and maintenance on the shaft seals without removing the mechanisms from the reactor.
- 6. PWR REACTOR PRESSURE VESSEL AND INTERNALS The PWR reactor pecssure vessel (RPV) and its internals provide a high integrity pressure boundary containing the reactor coolant, reactor core, and fuel fission products. In addition, the RPV and internals perform the following functions:
(1) Provide support for fuel assemblies and maintain their orientation and position within the reactor core.
(2) Provide the necessary structure that will result in a flow path for the reactor core to adequately remove heat generated by the core while:
assuring proper reactor flow distribution resisting upward flow-induced movement of the fuel assemblies avoiding flow-induced vibration (e.g., fuel rods, hold-down springs, and control rod assembly fingers) assuring positive location and guidance of control rod. assemblies assuring that heat generated by each fuel assembly is removed by the reactor coolant (3) Provide information regarding the RPV water level during shutdown Section 6.2 of Chapter 4 includes a requirement that "apprcpriate analysis shall be performed to demonstrate the adequacy of the RPV for a natural circulation cooldown of the reactor from full power." As written, this requirement can be construed to require the RPV to be adequate for a single natural circulation cooldown; whereas, the RPV should be able to withstand multiple cooldowns. This requirement should, therefore, be modified to EPRI ALWR CH 4 DSER 13 06/08/88 i
recognize that fact. As so modified, the requirement will be sufficient for the staff to consider GSI 79, "Unanalyzed Reactor Vessel Thermal Stress During Natural Convection Cooldown," to be resolved for the ALWR VRD. However, the staff cautions EPRI that when the generic resolution of GSI 79 is issued, it may require more specific actions by future plant designers and operators.
(cpen item)
Section 6.3 of Chapter 4 covers equipment design requirements for the RPV, reactor internals, and RPV i 1 mentation. The staff did not find that any of these utility requirements conflict with NRC's requirements. However, some of the items are of particular interest, as indicated below.
In paragraph 6.3.1.1, the URD specifies that the RPV shall be supported by support pads welded to or forged integrally with the vessel at the flange or at the primary coolant nozzle elevation. Supporting the RPV in this manner is a design improvement, as compared with support from a lower head. EPRI states that thermal expansion stresses in the vessel nozzles and in the RCS piping will be reduced. Thermal expansion displacement for the permanent refueling flange seal will also be reduced. And, if the support pads are integrally forged, the attachment welds will be eliminated, thus reducing in-service inspection requiremer,ts.
Reductions in refueling titre and worker radiation exposure are expected to result from two requirements in Section 6.3.1.6. One of these requirements specifies that means shall be provided to remotely detension RPV. studs and to remotely remove and cover stud holes. Another requirement specifies an integrated head disassembly capability that will enable the entire head package cid all related components to be lifted as a single unit.
i Paragraph 6.3.1.8 requires the PWR refueling cavity seal to be located as high as practical on the reactor vessel so as to minimize the area of vessel material exposed to borated water. The seal is to be designed so that it is not susceptible to any single failure that could result in a rapid drain down of the refueling cavity.
EPRI ALWR CH 4 DSER 14 06/08/88 1
Paragraph 6.3.2.2.3 requires the fuel assembly hold-down force to be sufficient to permit operation of all main coolant pumps at any temperature acceptable for running one or more main coolant pumps. The rationale given for this requirement is t: 4 temporary reductions in coolant temperature can then be made during startup operations without the need to shut down a main coolant pump. This will reduce the possibilities of a reactor coolant pump seal failure. The operation of all main coolant pumps at any temperature can also reduce plant heatup time. In Section 6.3.2.3 the hydreulic design requirements specify that the flow in the core peripheral region shall be upwards during all normal operating conditions. This will preclude an upward pressure gradient from the core baffles to peripheral fuel rods and thus will reduce the possibility of jet impingement on peripheral fuel assemblies.
Paragraph 6.3.2.6.1 requires the reactor upper internals and the lower internals (core support structures) each to be removable as a unit by a vertical lift with no in-vessel disassenbly required. This will simplify removal of the internals and allow for their naintenance, inspection, and repair. ,
Paragraph 6.3.3.1 includes a requirement for permanently installed piping betteen the RCS and the pressurizer to eliminate inaccuracies in level l monitoring during plant shutdowns. Such inaccuracies have been experienced in current plants that have temporarily installed tygon tubes; when the tubes kink, air is entrapped in the tube. During normal operation, the piping for l
the shutdown level monitoring system would be disconnected from the pressurizer ana RCS by means of blind flanges.
- 7. PWR CORE AND FUEL Section 7 of Chapter 4 defines the utility requirements for the PWR core and fuel. The components covered in this section include fuel assemblies, fuel rods, reactivity control devices, neutron sources, and core instrumentation, i
The core and fuel are to generate heat up to a rated value, throughout planned operating cycles, with sufficient margin and control to accommodate anticipated plant transients and planned maneuvers, all within defined EPRI ALWR CH 4 DSER 15 06/08/88
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limits, Section 3 of Chapter 1 specifies the transients and maneuvers to be considered.
i The performance requirements state (in phagraph 7.2.1.1) that "calculations S
for regulatory licensing shall use methods accepteble to the NRC and the most severe identified transients and acceptance criter la specified in Chapter 1, Section 3." The staff cautions that many of its appravals of correlations and codes have specific ranges of acceptability and ha w only been approved for limited applications. It is important to stress thct the calculational methods used must be approved by the NRC.
Paragraph 7.2.1.2 states that:
The core operating characteristics shall provide for stable operation-for all expected operating conditions. They shall also permit the operator to recognize the iminence of unstable operation during system disturbances or adjustments and take action to avoid it. However, fuel thermal limits shall not be exceeded even with no operator action.
The staff offers the following general comments on the above requirement which
-should be considered in preparing future drafts of the URD:
(1) For the current generation of plants the staff has found that thermal-hydraulic -
oscillations are not a problem for PWRs. This finding may have to be fj reva11 dated if the thermal-hydraulic characteristics of the advanced PWR 0I are sufficiently different fron current designs.
(2) The staff allows PWRs to have h ially unstable, but not radially unstable, xenon oscillations. ,(
l i (3) The staff requires that the thermal-hydraulic stability ar;d radial xenon stability characteristics of new designs be esteb11shed by testing.
Paragraph 7.2.1.4.1 specifies that the core shall be designed with the capability for load following and programed load cycling without ad,ius\ ing the soluble boron concentration during the maneuver. Moderator temperature changes would be used for rapid response to small load changes, and low-worth control rods would be used to assist in programned load cycling centrol. Both 3
EPRI ALWR CH 4 DSER 16 06/08/88
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[ t i of these mechanisms have been proven in operating PWRs. EPRI's rationale for this requirement states that "rodded maneuvering control without use of
- soluble boron is assessed to be feasible and practical with increased use of e
- proven, low-worth control rod des hns." Thi is not forbidden by current NRC requirements. !
Paragraph 7.2.*.4.2 specif 9s that "the fuel shall be designed to avoid limitation of .naneuvering capaL711ty and the rate of power increase for hot startups of
~ t b plant. Cold startup hower restrictions due to fuel shall be climinated."
e The staff recognhes.the benefits of meeting this objective and will review the specific fuei designs when they are submityed.
Section 7.2.2 of Chapter 4 addresses PWR fuel, reliability, burnup, and lifetime in essentially the same terms as those in Section 4.2.2 for BWR t'uel. The only apparent differences ure associated with the specification of a minimum of 95,000 MWD /MTUaverageburnupforBWRfuel. These utility objectives are not i prehibited regulatory requirements. '
< The mechanical design requirements for the PWR core ar.d fuel are presented in Section 7.3.1.1. They reflect the intent of the ALWR Program to employ proven l y , des igt.s. They also include ';ch matters as designing the fuel to b3 debris resistant. None of these items are incompatible with regulatory requirements.
In paragraph 7.3.1.2.3 the URD specifies that the fuel cycle design shall have a non-positive moderator temperature coefficict.t (liTC) above 50 percent power at the beginning of life and for operation over the entire p;wer range later in life. This requirement permits a positive MTC below 50 percent power of beginning of life, which provides flexibility to permit long fuel cycles, but t retains tne operational benefits of a non-positive MTC in the 50 to 100 e ', percent poIver load cycling range.
\ !
Positive moderator coefficients are not prohibited by NRC requirements.
However, the effect of allowing positive moderator coefficients must be evaluated throughout the fuel cycle for all transients and accider.ts, including ATUS. Paragraph 7.3.1.2.3 is not very prescriptive in defining when in the fuel cycle the positive moderator coefficients wou'd be allowed above 50 percent power and what requirements will be used to decide when this is acceptable.
( EPRI ALWR CH 4 DSER 17 06/08/88
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. , f Section 7.3.1.4 addresses materials requirements for the fuel assemblies, fuel rod cladding, and control rods. To the extent this section refers to Chaptor 1, the staff's DSER on that chapter also applies here. However,inreviewingk
r the utility specifications in paragraph 7.3.1.4.1, the, staff has not identifuc' ,
any other potential conflicts with NRC requirements.
Other topics addressed by Section 7.3.1 include: ccntrol of hydriding, fretting corroshn, fuel assembly hold-down force, hold-down springs, cladding collapse, fuel red bow, and fuel assembly bow. The staff reviewed all of the UhD requirements pertaining to these topics and did not identify any that are incompatible with NRC requirements. 1 Section7.3.2coverstheutilityrequirementsforneutronsourcesagd instrumentation. None of them were found to conflict with the NRC '
requirements; however, several of particular interest are discussed below.
In paragraph 7.3.2.1.1, the URD specifies that the reactor core shall be /
designed so that the initial startup and subsequent startups can ,be performed with an adequate neutron level signal on the out-of-core source range ,
instruments. This avoids the need for special startup procedures which resul -
in reduced plant capacity factor and provider, adequate margin to allow for J extendeo shutdown periods. ,
J Paragraph 7.3.2.3.1requirestheuseohfixedin'corepeutrondetectors instead of moveable in-core detectc7s for monitoring core power distribution.
This will simplify plant equipmenhnd rc' uce d the maintenance effort.
Paragraph 7.3.2.4 specifies thau thomotouples to monitor core outlet temperature should be placed integrally with the neutron detector string. This arrangement will simplify the instrumentatio'n by eliminating the need for separate thermocouple penetrations and conduits.
- 8. PWR CONTROL 30D Dh!VE SYSTEtt The PWR control rod drive (CRD) system is defined as the CRD mechanisms, position indicators, drive shafts, and electrical connectors. The control l
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rods are covered in Section 7 of Chapter 4. The power supplies, power cables, and breakers will be covered in Chapter 11.
/ (
In a PWR, the CRD system is required to perform the following functions:
i Position (withdraw and insert) the control rods in the core in response to comands from the rod control system.
Release the control rods for gravity insertion into the core upon power interruption in response to a reactor trip initiated from either manual or
,, automatic reactor protection system controls at the required rate to maintain fuel integrity.
Permit the latching and unlatching of the connection between the drive rod and the control rod assemblies.
Provide for control of position of individual control rods and control rod banka.
In Section 8.2 of Chapter 4, the URD states that the CRD system pressure boundary shall be designed for a service life of 60 years and establishes the following design and test criteria:
Operating basis earthquake 300 full load cycles l Safe shutdown earthquake 1 event Operational Scrams 1500 l Test Scrams 450 Pressure Test 1 per year The above criteria appear to be appropriate for a service life of 60 years.
However, this proposed lifetime exceeds that of existing CRD mechanisms (CRDMs); hence, it is not known whether that goal can be achieved.
Paragrapn' 8.2.2 requires the CRD system to be designed n Geat, if power is interrupted to the CRD coils or motors, the control roa ,re inserted by EPRI ALWR CH 4 DSER 19 06/08/88
r 1
. l gravity from the fully withdrawn position to the fully inserted position within a '
predetermined scram time. That scram time must be such that the total time from sensor activatien to completion of rod insertion satisfies the most restrictive accident analysis. This utility requirement is compatible with NRC requirements.
The performance requirements also address the CRD response tine, positioning control, and verification of rod positions. Safety and reliability are addressed in Section 8.2.4, which specifies that the CRD system shall be designed so that no single failure of a component, structure, system function, or service function will prevent the CRD system from performing its safety-related function of preventing rod drop and rod ejection. In addition, the CRDMs are to be designed to operate without coolant flow (air or water) for a minimum of 30 minutes. This capability is to provide a reasonable period of time for restoring the system after a loss of the CRDM cooling ',ystem.
Section 8.3 of Chapter 4 covers equipment design requirements including structural and mechanical considerations, materials, electrical and instru-mentation design, and maintenance and testing. These utility requirements specify a variety of details intended to achieve high quality design, reliable operatit,n, and simplified maintenance. For example, paragraph 8.3.1.3 specifies that all CRDM seals are to be seal welded to prevent leakage of reactor coolant and the CRDM seals are to be accessible for repair without removing any adjacent CRDs. Similarly, the CRDM stator coils and all electrical parts are to be replaceable without breaking the primary system pressure boundary and without removing any adjacent CRDMs.
The staff reviewed the equipment design requirements and concluded that none of them are inccmpatible with NRC requirements.
- 9. CONCLUSIONS On the basis of the evaluations sunnarized above, the staff concludes that the utility requirements included in Chapter 4 are in general agreement with the NRC guidelines and regulatory requirements for the systems involved. Subject EPRI ALWR CH 4 DSER 20 06/08/88
l to confirmation tnat EPRI has met its comitments to nodify various items in Chapter 4 and has satisfactorily addressed the open items in this DSER, the staff believes it will be possible to make the following determination in the final SER:
Chapter 4 of the Utility Requirements Document contains require-ments that, if properly translated into a design in accordance with the NRC regulations in force at the time of submittal, should
- l. generate a nuclear power plant that will have all the attributes I required by the regulations to ensure that there is no undue risk to the health and safety of the public or to the environment.
- 10. REFERENCES
- 1. Letter from E. E. Kintner (GPU Nuclear), Chairman, AWR Utility Steering Committee, to T. Murley, Director, Office of Nuclear Reactor Regulation NRC, enclosing Chapters 3 and 4 of the ALWR Utility Requirements Document, June 18, 1987.
- 2. Letter from P. H. Leech, NRC, to E. E. Kintner (GPU Nuclear), Chairman, ALWR Utility Steering Committee, requesting additional information relative to Chapters 3 and 4 of the ALWR Utility Requirements Document, November 13, 1987.
- 3. Letter from P. H. Leech, NRC, to E. E. Kintner (GPU Nuclear), Chairman, ALWR Utility Steering Comittee, requesting additional information relative to Chapters 3 and 4 of the ALWR Utility Requirements Document,
! December 11, 1987.
- 4. Letter from E. E. Kintner (GPU Nuclear), Chairman, ALWR Utility Steering Comittee, to L. S. Rubenstein, NRC, responding to References 2 and 3, January 25, 1988.
- 5. Letter from E. E. Kintner (GPU Nuclear), Chairman, ALWR Utility Steering Comittee, to L. S. Rubenstein, NRC, further responding to References 2 and 3 March 28, 1988.
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- 6. Letter from C. F. Sears (Northeast Utilitier), Chairman, ALWR Utility Steering Committee, to H. R. Denton, NRC, transmitting information on four safety and licensing issues, July 8, 1986.
- 7. Letter from E. E. Kintner (GPU Nuclear), Chairman, ALWR Utility Steering Comnittee, to T. Murley, Director, Office of Nuclear Reactor Regulation.
NRC, transmitting six topic papers on generic safety issues, July 9, 1987.
- 8. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," Rev. 3, July 1981.
- 9. Letter from T. P. Speis, NRC, to C. F. Sears (Northeast Utilities),
Chairman, ALWR Utility Steering Committee, responding to Reference 6, December 9, 1986.
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. ., v APPENDIX A: GENERIC ISSUES PERTAINING TO ALWR URD CHAPTER 4 Identification Title Chapter 4 Sections / Paragraphs Review Status GSI 79 Unanalyzed RV Thermal Stress 6.2 Qualified acceptance During Natural Convection of resolution Cooldowr, GSI 94 Additional Low Temperature 2.3.1.2, 2.3.1.6, & 2.3.1.7 Qualified acceptance Overpressure Protection of resolution B-22 LWR Fuel 4.2.2, 7.2.2 Resolution acceptable for ALWR EPRI ALWR CH 4 DSER 06/08/88