U.S. NUCLEAR REGULATORY COMMISSION December 2008 

OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1

DRAFT REGULATORY GUIDE 

Contact: K. Nguyen

(301) 251-7453 

DRAFT REGULATORY GUIDE DG-1178 

(Proposed Revision 1 of Regulatory Guide 1.151, dated July 1983) 

INSTRUMENT SENSING LINES 

A. INTRODUCTION 

This guide describes a method that the staff of the U.S. Nuclear Regulatory Commission (NRC) 

considers acceptable for use in complying with the Commission’s regulations with regard to the design 

and installation of safety-related instrument sensing lines in nuclear power plants. To meet these 

objectives, the sensing lines must serve a safety-related function to prevent the release of reactor coolant 

as a part of the reactor coolant pressure boundary and provide adequate connection to the reactor coolant 

system for measuring process variables (e.g., pressure, level, and flow). The term “safety-related” refers 

to those structures, systems, and components necessary to ensure (1) the integrity of the reactor coolant 

pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe-shutdown 

condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in 

potential offsite exposures comparable to the guideline exposures in Title 10, Part 100, “Reactor Site 

Criteria,” of the Code of Federal Regulations (10 CFR Part 100) (Ref. 1). 

The regulatory framework that the NRC has established for nuclear power plants consists of a 

number of regulations and supporting guidelines, including but not limited to the general design criteria 

(GDC) of Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50, 

“Domestic Licensing of Production and Utilization Facilities” (Ref. 2). The provisions of 10 CFR Part 50 

require that design criteria be established for structures, systems and components important to safety to 

provide reasonable assurance that the facility can be operated without undue risk to public health and 

safety. GDC 1, “Quality Standards and Records,” requires that structures, systems, and components 

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory 

position in this area. It has not received final staff review or approval and does not represent an official NRC final staff position. 

Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory 

analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may 

be submitted to the Rulemaking, Directives, and Editing Branch, Office of Administration, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555-0001; emailed to nrcrep.resource@nrc.gov; submitted through the NRC’s interactive 

rulemaking Web page at http://www.nrc.gov; faxed to (301) 415-5144; or hand-delivered to Rulemaking, Directives, and Editing 

Branch, Office of Administration, US NRC, 11555 Rockville Pike, Rockville, Maryland 20852. Between 7:30 a.m. and 4:15 

p.m. on Federal workdays. Copies of comments received may be examined at the NRC’s Public Document Room, 11555 

Rockville Pike, Rockville, MD. Comments will be most helpful if received by February 6, 2009. 

Electronic copies of this draft regulatory guide are available through the NRC’s interactive rulemaking Web page (see above); the 

NRC’s public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the NRC’s Electronic 

Reading Room at http://www.nrc.gov/reading-rm/doc-collections/; and the NRC’s Agencywide Documents Access and 

Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML080510453. 
DG-1178, Page 2 

important to safety be designed and installed to quality standards commensurate with the importance-tosafety of the functions to be performed. GDC 13, “Instrumentation and Control,” requires that 

instrumentation be provided to monitor variables and systems to assure adequate safety. GDC 24, 

“Separation of Protection and Control Systems,” requires, in part, that the interconnection of the 

protection and control systems be limited so as to ensure that safety is not significantly impaired. GDC 

55, “Reactor Coolant Pressure Boundary Penetrating Containment,” requires, in part, that each line that is 

part of the reactor coolant pressure boundary and that penetrates the primary reactor containment shall be 

provided with containment isolation valves, unless it can be demonstrated that the containment isolation 

provisions for a specific class of lines (such as instrument lines) are acceptable on some other defined 

basis. Regulatory Guide 1.53, “Application of the Single-Failure Criterion to Nuclear Power Plant 

Protection Systems” (Ref. 3), provides guidance on acceptable methods for satisfying the Commission’s 

regulations with respect to the separation and independence of the electrical power, instrumentation, and 

control portions of nuclear power plant safety systems. 

The NRC issues regulatory guides to describe to the public methods that the staff considers 

acceptable for use in implementing specific parts of the agency’s regulations, to explain techniques that 

the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to 

applicants. Regulatory guides are not substitutes for regulations and compliance with them is not 

required. 

This regulatory guide contains information collection requirements covered by 10 CFR Part 50 

that the Office of Management and Budget (OMB) approved under OMB control number 3150-0011. 

The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information 

collection request or requirement unless the requesting document displays a currently valid OMB control 

number. 

B. DISCUSSION 

Background 

Committee SP67.02 of the Instrument Society of America (ISA) prepared American Nuclear 

Standards Institute (ANSI)/ISA-S67.02.01-1999, “Nuclear Safety-Related Instrument-Sensing Line 

Piping and Tubing Standard for Use in Nuclear Power Plants” (Ref. 4). On December 9, 1974, this 

committee received an approved scope and project, Charter N677, from ANSI. ISA approved ANSI/ISAS67.02.01-1999 on November 15, 1999, following approval by the ISA Standards and Practices Board on 

July 15, 1999. The standard provides design, physical protection, and installation requirements for 

instrument sensing lines and sampling lines for nuclear power plant applications. The standard 

establishes the applicable code requirements and code boundaries for the design and installation of 

instrument sensing lines interconnecting safety-related piping and vessels with both safety-related and 

nonsafety-related instrumentation. The supplementary portions of the regulatory position of Regulatory 

Guide 1.151, Revision 0, “Instrument Sensing Lines,” issued July 1983, have now been included in the 

updated ANSI/ISA-S67.02.01-1999 and, consequently, are no longer needed. 

The Power Generation Committee of the Institute of Electrical and Electronics Engineers (IEEE) 

Power Engineering Society developed IEEE Std 622-1987, “Recommended Practice for the Design and 

Installation of Electric Heat Tracing Systems for Nuclear Power Generating Stations” (Ref. 5), and the 

IEEE Standards Board approved it on December 11, 1986. ANSI approved it on November 23, 1987. 

The standard provides recommended practices for designing and installing electric heat tracing systems in 

nuclear power generating stations to prevent them from freezing in cold weather and to prevent certain 
DG-1178, Page 3 

concentrations of chemicals, such as boric acid solutions, from crystallizing or solidifying within an 

instrument piping system. The recommendations include identification of requirements, heater design 

considerations, power systems design considerations, temperature control considerations, alarm 

considerations, finished drawings and documents, installation of materials, startup testing, temperature 

tests, and maintenance of electric pipe heating systems. 

Operational events have occurred in which evolved gases in sensing lines have affected measured 

water levels in operating boiling-water reactors (BWRs). Concern over these events has resulted in staff 

guidance to ensure that plants are operated safely and conservatively. The potential problem is that 

dissolved gases can evolve in the reference leg for level measurements in the reactor vessels of BWRs as 

the solubility of the gases decreases during depressurization. Such gases can be trapped in the sensing 

line and affect differential pressure measurements, particularly level measurements. Such events have 

been reported in licensee event reports with significant level measurement errors. Since level 

instrumentation plays an important role in plant safety and is required for both normal and accident 

conditions, NRC Bulletin 93-03, “Resolution of Issues Related to Reactor Vessel Water Level 

Instrumentation in BWRs,” dated May 28, 1993 (Ref. 6), recommended that each utility implement 

corrective actions to ensure that the level instrumentation design is of high functional reliability for longterm operation. In response to the bulletin, the majority of BWR licensees decided to install a reference 

leg backfill system to supply a continuous flow of water from the control rod drive (CRD) hydraulic 

system through the reference legs to preclude migration of dissolved noncondensable gases into the legs. 

However, NRC Information Notice 93-89, “Potential Problems with BWR Level Instrumentation Backfill 

Modifications,” dated November 26, 1993 (Ref. 7), reported on several potential design problems with the 

retrofit backfill system by which a single failure in the backfill system would lead to a severe transient. 

The design measures that respond to this potential problem should ensure that the features and systems 

that mitigate or preclude evolved gases do not themselves introduce additional single-failure mechanisms 

in the protection system. The following regulatory provisions support taking additional measures to 

address the potential for evolved gases in sensing lines: 

• GDC 13 (in Appendix A of Part 50) states, “Instrumentation shall be provided to monitor 

variables and systems over their anticipated ranges for normal operation, for anticipated 

operational occurrences, and for accident conditions as appropriate to assure adequate safety.” 

• GDC 21, “Protection System Reliability and Testability” (in Appendix A of Part 50), states, “The 

protection system shall be designed for high functional reliability commensurate with the safety 

function to be performed.” 

• GDC 22, “Protection System Independence” (in Appendix A of Part 50), states, “The protection 

system shall be designed to assure that the effects of natural phenomena, and of normal operating, 

maintenance, testing, and postulated accident conditions, do not result in loss of the protection 

function.” 

• 10 CFR 50.55a(h)(2), “Protection systems,” requires that protection systems for plants with 

construction permits issued after January 1, 1971, but before May 13, 1999, meet the 

requirements stated in either IEEE Std 279, “Criteria for Protection Systems for Nuclear Power 

Generating Stations,” or in IEEE Std 603-1991, “IEEE Standard Criteria for Safety Systems for 

Nuclear Power Generating Stations” (Ref. 8). 

Additionally, the staff notes the following concerns that support these measures: 

• Level-sensing instrumentation may not accurately monitor reactor vessel water level under 

normal cooldown or accident conditions. 

• Instrumentation may not be reliable during and following normal and rapid depressurization. 
DG-1178, Page 4 

• Degassing may cause a loss of the reactor vessel water-level indication function during and 

following normal depressurization and rapid depressurization. 

• Section 5.5, “System Integrity,” of IEEE Std 603-1991 states: “The safety system shall be 

designed to accomplish their safety function under the full range of applicable conditions 

enumerated in the design basis.” While Section 4.20 of IEEE Std 279-1971 (Ref. 9) states: “The 

protection system shall be designed to provide the operator with accurate, complete, and timely 

information pertinent to its own status and to generating station safety.” However, using this 

approach, the water level instrumentation for the reactor vessel may not be accurate during and 

following normal depressurization and rapid depressurization. 

C. REGULATORY POSITION 

The requirements of ANSI/ISA-S67.02.01-1999 provide a basis acceptable to the NRC staff 

for designing and installing safety-related instrument sensing lines in nuclear power plants subject to 

the following: 

1. The endorsement of ANSI/ISA-S67.02.01-1999 is limited to instrument sensing lines and does 

not include Section 6, “Sample-Line Fabrication, Routing, Installation, and Protection.” The 

original ANSI/ISA-S67.01 covered only sensing lines, while ANSI/ISA-S67.10 addressed 

sampling lines. The revised standard combines the two. This regulatory guide addresses only the 

sensing line guidance. 

2. Root valve and accessible isolation valves shall be provided, unless it can be demonstrated that 

the containment isolation provisions for sensing lines are acceptable on some other defined basis. 

The root valve and accessible isolation valve may be the same valve if the arrangement meets all 

other requirements for isolation and accessibility. Figures 1 through 6 of ANSI/ISA-S67.02.01-

1999 show sketches depicting root valve and accessible isolation valve locations. Although not 

specifically addressed in the standard’s Section 4.2, “Mechanical Design Requirements for 

Sensing,” these drawings and guidance are applicable to sensing lines. The provision for root 

valves and access is derived from GDC 55. 

3. The requirements of IEEE Std 622-1987 provide a basis acceptable to the NRC staff for 

designing and installing electric heat tracing systems in nuclear power generating stations. 

4. Provisions shall be made to mitigate the potential effects of trapped, evolved gases in sensing 

lines during or following depressurization events as long as the associated measurements are 

required for monitoring the plant or for operating the safety system. This position is based on 

GDC 13, GDC 21, GDC 22, and 10 CFR 50.55a(h). 

D. IMPLEMENTATION 

The purpose of this section is to provide information to applicants and licensees regarding the 

NRC’s plans for using this draft regulatory guide. The NRC does not intend or approve any imposition or 

backfit in connection with its issuance. 

The NRC has issued this draft guide to encourage public participation in its development. 

The NRC will consider all public comments received in development of the final guidance 

document. In some cases, applicants or licensees may use proposed or previously established acceptable 

alternative methods for complying with specified portions of the NRC’s regulations. Otherwise, 
DG-1178, Page 5 

the methods described in this guide will be used in evaluating compliance with the applicable regulations 

for license applications, license amendment applications, and amendment requests. 

REGULATORY ANALYSIS

1. Statement of the Problem 

The current revision of Regulatory Guide 1.151 (Revision 0), issued July 1983, is based on 

ANSI/ISA-S67.02, issued in 1973. A number of reactor events occurring between 1973 and 1983 led to 

supplementary guidance in Revision 0 of the regulatory guide for concerns not addressed in the ISA 

standard. The current revision of the standard, ANSI/ISA-S67.02.01-1999, includes all the 

supplementary guidance, thus obviating the need for most of the current supplementary guidance in 

the regulatory position. 

The revised standard, ANSI/ISA-S67.02.01-1999, also includes guidance regarding sampling 

lines that was previously included in a separate standard. Regulatory Guide 1.151 only applies to sensing 

lines, so this change must be addressed. 

Additional events have occurred since 1983 that warrant supplementary guidance in the 

regulatory position that the current ANSI/ISA standard does not include. One is the issuance of NRC 

Bulletin 93-03 regarding evolved gas in liquid reference legs following depressurization events. 

IEEE Std 622-1987 was issued, and is applicable to sensing lines. 

2. Objective 

The objective of this regulatory action is to review the new ISA standard, reactor operating 

history, and regulatory basis to determine if any changes should be made to Regulatory Guide 1.151, 

Revision 0. 

3. Alternative Approaches 

The NRC staff considered the following alternative approaches: 

• Do not revise Regulatory Guide 1.151. 

• Update Regulatory Guide 1.151. 

3.1 Alternative 1: Do Not Revise Regulatory Guide 1.151

Under this alternative, the NRC would not revise this guidance, and the original version of this 

regulatory guide would continue to be used. This alternative is considered the baseline or “no action” 

alternative and, as such, involves no value/impact considerations. If chosen, this alternative would result 

in referencing an industry standard that is no longer supported and made available by ISA. 

3.2 Alternative 2: Update Regulatory Guide 1.151

Under this alternative, the NRC would update Regulatory Guide 1.151, taking into consideration 

the content of the revised standard ANSI/ISA-S67.02.01-1999 and the events in operating reactors and 
DG-1178, Page 6 

additional standards that are applicable to sensing lines. The four position statements in the draft of the 

revised standard may be considered individually for inclusion in the approved version without affecting 

inclusion of the others. 

One benefit of this action is that it would enhance reactor safety by addressing the most current 

ANSI/ISA and IEEE standards on safety systems endorsed by the NRC. 

A further review of regulatory guidance that references sensing lines or Regulatory Guide 1.151 

finds that there are direct references to material in the regulatory positions of the current regulatory guide. 

The revised regulatory guide now notes that this material is included in ANSI/ISA-S67.02.01-1999. The 

staff would also have to update these references in the standard review plan and elsewhere. 

The costs to the NRC would be the one-time cost of issuing the revised regulatory guide (which is 

expected to be relatively small), and applicants would incur little or no cost. The staff would also need to 

revise the other regulatory documents that reference materials in the previous version of the regulatory 

guide or the previous referenced standard. 

4. Conclusion 

Based on this regulatory analysis, the staff recommends that the NRC revise Regulatory 

Guide 1.151. The staff concludes that the proposed action will enhance reactor safety by referencing the 

latest ANSI/ISA and IEEE standards on safety systems endorsed by the NRC. It could also lead to cost 

savings for the industry, especially with regard to applications for standard plant design certifications and 

combined licenses. 

REFERENCES 

1. 10 CFR Part 100, “Reactor Site Criteria,” U.S. Nuclear Regulatory Commission, Washington, 

DC. 

2. 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” U.S. Nuclear 

Regulatory Commission, Washington, DC. 

3. Regulatory Guide 1.53, “Application of the Single-Failure Criterion to Nuclear Power Plant 

Protection Systems,” U.S. Nuclear Regulatory Commission, Washington, DC. 
DG-1178, Page 7 

4. ANSI/ISA-S67.02.01-1999, “Nuclear Safety-Related Instrument-Sensing Line Piping and Tubing 

Standard for Use in Nuclear Power Plants,” American National Standards Institute/Instrument 

Society of America, Research Triangle Park, NC, 1999. 

5. IEEE Std 622-1987, “Recommended Practice for the Design and Installation of Electric Heat 

Tracing Systems for Nuclear Power Generating Stations,” Institute of Electrical and Electronics 

Engineers, Piscataway, NJ, 1987. 

6. Bulletin 93-03, “Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in 

BWRs,” U.S. Nuclear Regulatory Commission, Washington, DC, May 28, 1993. 

7. Information Notice 93-89, “Potential Problems with BWR Level Instrumentation Backfill 

Modifications,” U.S. Nuclear Regulatory Commission, Washington, DC, November 26, 1993. 

8. IEEE Std 603-1991, “Standard Criteria for Safety Systems for Nuclear Power Generating 

Stations,” Institute of Electrical and Electronics Engineers, Piscataway, NJ, 1991 (and the 

correction sheet dated January 30, 1995). 

9. IEEE Std 279-1971, “Criteria for Protection Systems for Nuclear Power Generating Stations,” 

Institute of Electrical and Electronics Engineers, Piscataway, NJ, 1971. (withdrawn) 

10. Regulatory Guide 1.26, “Quality Group Classifications and Standards for Water-, Steam-, and 

Radioactive-Waste-Containing Components of Nuclear Power Plants,” U.S. Nuclear Regulatory 

Commission, Washington, DC.