Resolution of Generic Safety Issues: Issue 119: Piping Review Committee Recommendations (Rev. 4) ( NUREG-0933, Main Report with Supplements 1–34 )
In an August 1983 memorandum,834 the EDO requested a comprehensive review of NRC requirements in the area of nuclear power plant piping. In response to this request, the NRC Piping Review Committee (PRC) was formed to review and evaluate existing regulatory requirements to: (1) provide recommendations on where and how the NRC should modify requirements; and (2) identify areas requiring further action. The scope of the PRC review covered piping in safety-related systems and high energy lines important to safety in new and operating plants. With respect to postulated pipe breaks, the scope covered all high energy lines.
An NRC steering committee consisting of members from RES, NRR, OIE, and ELD was formed to review and develop a plan for implementing the changes recommended in the PRC report.611 The steering committee agreed to focus its attention on the recommended research and regulatory changes designated in the PRC report611 as Category A (high priority) recommendations. The PRC-recommended research and regulatory changes were restructured by the steering committee (combining of research and regulatory recommendations) to form 9 tasks to be addressed by the NRC implementation plan,835 5 of which are addressed below. These 5 tasks consist primarily of NRR regulatory actions and some closely-related research efforts. The remaining 4 tasks of the NRC implementation plan related only to research activities and were excluded from this issue.
The five parts of this issue primarily involve revisions to Regulatory Guides and the SRP.11 No significant change in public safety was expected to result from resolution of this issue; however, resolution of the various tasks was expected to result in less complex and more realistic approaches to piping design and operation in nuclear power plants. The results were expected to yield more efficient regulatory practices, improve plant piping systems design, increase plant reliability, and decrease ORE associated with inspections and repairs. The NRC steering committee agreed that, based on the information provided in NUREG-1061,611 this work should continue on a schedule consistent with high-priority issues. Therefore, this issue was classified as a Regulatory Impact issue. RES took the lead responsibility for resolution of this issue with assistance from other NRC Offices.835 The following is an evaluation of the 5 parts of this issue.
ITEM 119.1: PIPING RUPTURE REQUIREMENTS AND DECOUPLING OF SEISMIC AND LOCA LOADS
This task combined two PRC Category A regulatory recommendations with one PRC Category A research recommendation. The designations of the three PRC recommendations were: (1) leak-before-break (A-1); (2) decoupling of seismic and LOCA loads (A-5); and (3) completing research on decoupling (A-4).
One part of the task involved rulemaking changes to GDC-4 in Appendix A of 10 CFR 50 to redefine the need to consider the dynamic effects of pipe breaks. A proposed rule to modify GDC 4 was published1087 in July 1985 and codified leakbefore-break technology, but was limited only to the primary loop piping of PWRs; the final rule was published1340 in April 1986. A proposed broad scope rule dealing with all high energy piping in LWRs was published1341 in July 1986; the final rule was published1342 in October 1987. With the issuance of these revised rules, revisions to SRP11 Sections 3.6.1 and 3.6.2 were needed to eliminate the postulation of arbitrary intermediate breaks. The second part of this task involved relaxation of the requirement to consider LOCA and seismic loads simultaneously. A revision to SRP11 Section 3.9.3 was to be pursued to decouple seismic and pipe rupture loads in the mechanical design of components and their supports.
The existing GDC-4 requirement and SRP11 Section 3.6.2 pertaining to postulated double-ended guillotine breaks (DEGB) of the largest pipes and postulated arbitrary intermediate pipe breaks needed to be changed to include more realistic criteria and to allow consideration and acceptance of validated analysis methods. The requirements of GDC-4 led to a situation where protective devices were added to forestall events that are extremely unlikely. These protective devices that were designed for the extremely unlikely events could, however, reduce safety and increase worker radiation exposure under normal operations and design basis events.
SRP11 Section 3.9.3 requires that piping systems and associated components be designed for the combined effects of an SSE and a LOCA. The evolution of seismic design requirements and the calculations of pipe rupture loads have significantly increased the resultant loads obtained by combining these effects. However, field evaluations of piping at conventional power plants and petrochemical facilities indicated that ruptures in piping of the type found in nuclear power plants do not occur during severe earthquakes. Therefore, the staff believed that relaxation of these requirements at all LWRs would not affect plant or public safety.
This task was classified as a Regulatory Impact issue that resulted in revisions1343,1344 to SRP11 Sections 3.6.1 and 3.6.2. In addition, Generic Letter No. 87-111345 was issued to licensees on the relaxation in arbitrary intermediate pipe rupture requirements (SRP Section 3.6.2). In 1986, the staff terminated1345 all work on a proposed revision to SRP11 Section 3.9.3. Thus, this issue was resolved.
ITEM 119.2: PIPING DAMPING VALUES
This task combined PRC regulatory recommendation A-2 (modify seismic damping values used in seismic designs) and PRC research recommendation B-3 (complete research on damping tests). It constituted a two-level approach that could affect all LWRs: a short-term plan and a long-term plan. The short-term action called for a revision to Regulatory Guide 1.841347 as the vehicle for NRC endorsement of ASME Code Case N-411. The long-term action called for revisions to Regulatory Guide 1.611348 and SRP11 Section 3.9.2 to incorporate, not only ASME Code Case N-411, but also new positions on pipe damping for high-frequency loads and for time-history analyses.
The short-term endorsement of the ASME Code Case N-411 was to be restricted to seismic response analysis, but not time-history analysis. The long-term action was to result in extensive changes to SRP11 Section 3.9.2 and Regulatory Guide 1.611348 to provide more comprehensive guidance on pipe damping for both seismic and BWR hydrodynamic loadings. Criteria for other non-seismic dynamic loads could also be addressed in the SRP11 Section 3.9.2 revision.
In general, dynamic piping response could be more accurately predicted if use was made of higher piping damping values than those identified in the existing regulatory guide. The use of higher damping values would result in nuclear plant piping systems having significantly less snubbers and supports and an overall better balance of design, considering all piping loads. A decrease in the number of snubbers and supports could allow better inspection of equipment and components at significantly reduced ORE.
The staff originally planned to take the lead in developing improved pipe damping values and classified the task as a Regulatory Impact issue. However, with the cooperative effort of EPRI, ASME, and the NRC in pursuing the concern, the staff concluded that the most effective approach to the use of more realistic damping values for dynamic piping analysis was through ASME III, Appendix N. When this appendix is completed, the staff will make a decision on its endorsement. As a result, the issue was dropped from further pursuit.1336
ITEM 119.3: DECOUPLING THE OBE FROM THE SSE
This task corresponds to PRC regulatory recommendation A-3 (decouple OBE from SSE). 10 CFR 100, Appendix A, Section V(a)(2), stipulates that "(t)he maximum vibratory ground acceleration of the OBE shall be at least one-half the maximum vibratory ground acceleration of the SSE." Therefore, the existing requirement implies the coupling of the two earthquake design levels: SSE and OBE. In developing the existing regulations, it was assumed that the SSE would control the design in nearly all aspects and that the OBE would serve as a separate check of those systems where continued operation was desired at a lower level of ground motion. However, in practice, the assumed load factors, damping, stress levels, and service limits have caused the OBE, rather than the SSE, to control the design for many systems including concrete and steel structures and nuclear piping. In addition, seismic design for OBE accounts for certain safety-related factors such as fatigue and seismic anchor movement that are not considered in the design for the SSE.
Decoupling of the OBE from the SSE or modification of the associated load factors, etc., would impact the design of new plants and would extend well beyond piping considerations. The actions required to resolve this task include: (1) rulemaking to amend and revise Appendix A to 10 CFR 100 to permit decoupling of the OBE and SSE and to incorporate the use of probabilistic methodology in earthquake design; (2) revising and developing Regulatory Guides; (3) updating pertinent sections of the SRP11; and (4) advising various industry code committees to revise appropriate codes and guides to reflect changes in the regulations.
A complete listing of the Regulatory Guides and SRP Sections that may be affected by this task were to be identified during the review phase of this task and the related tasks contained in the NRC implementation plan835 which is of much broader scope.
There is no technical basis for coupling the OBE with the SSE. Designing the piping systems to the SSE is the primary means of ensuring safety. Additional margin is provided by specifying the OBE and thus the level at which inspections will be required before continued operation would be permitted. The more realistic approach of using specific probabilities (return periods) for OBE and the decoupling of the OBE levels and frequencies from those of the SSE will allow assurance of public safety to be placed on a more rational basis.
This item is a Regulatory Impact issue that was integrated775 by RES into a revision to 10 CFR 100, Appendix A.
ITEM 119.4: BWR PIPING MATERIALS
This task corresponds to PRC regulatory recommendation A-4 to replace regular grade 316SS and 304SS materials in BWR recirculation piping with an alloy resistant to IGSCC. The NRR action related to this task involved preparation of Revision 2 to NUREG-0313750 and evaluation of each licensee's actions in compliance with this revision.
IGSCC in BWR piping has occurred in a range of piping sizes over the last 25 years and has resulted in major reactor outages. The risk studies reported611 indicate that pipe failures, even assuming the higher rates due to IGSCC, would not be a major contributor to core-melt and public risk. However, use of materials more resistant to IGSCC should significantly reduce levels of ISI and reactor outage times. Therefore, plant outages and recurring ORE could be significantly reduced by resolution of this task.
This item is a Regulatory Impact issue that required1506 updating of Regulatory Guide 1.441507 by RES to reflect the staff's findings in NUREG-0313,750 Revision 2, as recommended925 by NRR. As a part of the improvements to NUREG-0933, the NRC staff clarified in SECY-11-0101, “Summary of Activities Related to Generic Issues Program,” dated July 26, 2011,1967 that the Generic Issues Program will not pursue any further actions toward resolution of licensing and regulatory impact issues. Because licensing and regulatory impact issues are not safety issues by the classification guidance in the legacy Generic Issues Program, these issues do not meet at least one of the Generic Issues Program screening criteria and do not warrant further processing in accordance with Management Directive 6.4, “Generic Issues Program,” dated November 17, 2009.1858 Therefore, this issue will not be pursued any further in the Generic Issues Program.
ITEM 119.5: LEAK DETECTION REQUIREMENTS
This task corresponds to PRC regulatory recommendation A-6 (leak detection requirements). To accomplish this task, additional data are necessary to further validate and improve existing leak-rate prediction analyses. Of particular interest would be investigation and improvement of local leak detection systems such as acoustic emission monitors or moisture-sensitive tapes. These latter techniques may be important for establishing the validity of leak-before-break at specific locations in certain piping systems. The task requires a combination of two approaches: (1) the surveying of operating plants to determine the adequacy of existing leak detection systems; and (2) completion of the research recommended by the PRC and applying the results of the research to regulatory requirements. Subsequent to the completion of key elements of the research effort, the regulatory actions may include the following:
|(1)||Identify required TS changes such as: (a) unidentified leakage limits for BWRs and PWRs in the context of locating and detecting leakage from cracks with margin; (b) adequacy of surveillance requirements and calibration of systems; (c) alarms; (d) TS consistency; (e) new systems or different detection system combinations; and (f) forward-fit and backfit considerations.|
|(2)||Revise SRP11 Section 5.2.5 and Regulatory Guide 1.45.603|
|(3)||Issue NUREG-0313,750 Revision 2.|
It was believed that resolution of this task could affect all LWRs to varying degrees.
No direct safety significance could be attributed to this task. However, knowledge of the leak rates associated with various postulated through-wall crack lengths and confidence in the ability to detect leakage in a timely manner are important elements of the leak-before-break concept that eliminates the postulated DEGB.
This item is a Regulatory Impact issue. As a part of the improvements to NUREG-0933, the NRC staff clarified in SECY-11-0101, “Summary of Activities Related to Generic Issues Program,” dated July 26, 2011,1967 that the Generic Issues Program will not pursue any further actions toward resolution of licensing and regulatory impact issues. Because licensing and regulatory impact issues are not safety issues by the classification guidance in the legacy Generic Issues Program, these issues do not meet at least one of the Generic Issues Program screening criteria and do not warrant further processing in accordance with Management Directive 6.4, “Generic Issues Program,” dated November 17, 2009.1858 Therefore, this issue will not be pursued any further in the Generic Issues Program.
- NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, (1st Ed.) November 1975, (2nd Ed.) March 1980, (3rd Ed.) July 1981.
- Regulatory Guide 1.45, “Reactor Coolant Pressure Boundary Leakage Detection Systems,” U.S. Nuclear Regulatory Commission, May 1973. 
- NUREG-1061, “Report of the U.S. Nuclear Regulatory Commission Piping Review Committee,” U.S. Nuclear Regulatory Commission, (Vol. 1) August 1984, (Vol. 2) April 1985, (Vol. 3) November 1984, (Vol. 4) December 1984, (Vol. 5) April 1985.
- NUREG-0313, “Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,” U.S. Nuclear Regulatory Commission, July 1977, (Rev. 1) July 1980, (Rev. 2) January 1988.
- Memorandum for R. Emrit from A. Murphy, “Generic Issue Management Control System, Issue No. 119.3, Decouple OBE from SSE,” February 21, 1992. 
- Memorandum for H. Denton and R. Minogue from W. Dircks, “Review of NRC Requirements for Nuclear Power Plant Piping,” August 1, 1983. 
- Memorandum for W. Dircks from R. Minogue, “Plan to Implement Piping Review Committee Recommendations,” July 30, 1985. 
- Memorandum for E. Beckjord from T. Murley, “Regulatory Guide 1.44,” April 30, 1992. 
- Federal Register Notice 50 FR 27006, “10 CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures,” July 1, 1985.
- Memorandum for B. Morris from L. Shao, “Resolution of Generic Issue 119.2,” July 16, 1990. 
- Federal Register Notice 51 FR 12502, “10 CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures,” April 11, 1986.
- Federal Register Notice 51 FR 26393, “10 CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures,” July 23, 1986.
- Federal Register Notice 52 FR 41288, “10 CFR Part 50, Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures,” October 27, 1987.
- Federal Register Notice 53 FR 1968, “Standard Review Plan Revision,” January 25, 1988.
- Federal Register Notice 52 FR 23376, “Standard Review Plan Issuance,” June 19, 1987.
- Letter to All Operating Licensees, Construction Permit Holders, and Applicants for Construction Permits from U.S. Nuclear Regulatory Commission, “Relaxation in Arbitrary Intermediate Pipe Rupture Requirements (Generic Letter 87-11),” June 19, 1987. 
- Memorandum for Distribution from G. Arlotto, “Termination of Proposed Revision to SRP 3.9.3,” October 2, 1986. 
- Regulatory Guide 1.84, “Design and Fabrication Code Case Acceptability—ASME III, Division 1,” U.S. Nuclear Regulatory Commission, (Rev. 30) October 31, 1994. 
- Regulatory Guide 1.61, “Damping Values for Seismic Design of Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, October 1973. 
- Memorandum for W. Minners from L. Shao, “Closeout of GSI 119.4,” July 17, 1992. 
- Regulatory Guide 1.44, “Control of the Use of Sensitized Stainless Steel,” U.S. Atomic Energy Commission, May 1973. 
- Management Directive 6.4, “Generic Issues Program,” U.S. Nuclear Regulatory Commission, November 17, 2009.
- SECY-11-0101, “Summary of Activities Related to Generic Issues Program,” July 26, 2011. [ML111590814]