U.S. NUCLEAR REGULATORY COMMISSION July 2010

Revision 1

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

The NRC issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable 

for use in implementing specific parts of the agency=s regulations, techniques that the staff uses in evaluating specific problems or 

postulated accidents, and data that the staff needs in reviewing applications for permits and licenses. Regulatory guides are not 

substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in 

regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a 

permit or license by the Commission. 

This guide was issued after consideration of comments received from the public. 

Regulatory guides are issued in 10 broad divisionsC1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials 

Facilities; 4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 

9, Antitrust and Financial Review; and 10, General. 

Electronic copies of this guide and other recently issued guides are available through the NRC=s public Web site under the 

Regulatory Guides document collection of the NRC=s Electronic Reading Room at 

http://www.nrc.gov/reading-rm/doc-collections/ and through the NRC=s Agencywide Documents Access and Management 

System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML092850042. The regulatory analysis 

may be found in ADAMS under Accession No. ML101870472. 

REGULATORY GUIDE 1.141

(Draft was issued as DG-1213, dated June 2009)

CONTAINMENT ISOLATION PROVISIONS 

FOR FLUID SYSTEMS 

A. INTRODUCTION 

This guide describes updated methods that the U.S. Nuclear Regulatory Commission (NRC) staff 

considers acceptable for use in complying with the Commission’s requirements for isolation of fluid 

systems that penetrate containment. 

Title 10, of the Code of Federal Regulations, Part 50, ADomestic Licensing of Production and 

Utilization Facilities@ (10 CFR Part 50), Appendix A, AGeneral Design Criteria for Nuclear Power Plants,@

General Design Criteria (GDC) 54, 55, 56, and 57 establishes that piping systems that penetrate the 

primary reactor containment be provided with isolation capabilities that reflect the importance to safety of 

isolating these piping systems. The principal design criteria found in licensees’ final safety analysis 

report (FSAR) should be similar in philosophy and meet the intent of the GDC. 

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. 
Rev. 1 of RG 1.141, Page 2 

B. DISCUSSION 

Working Group ANS-56.2 of the American Nuclear Society (ANS) Standards Committee 

ANS-50, Nuclear Power Plant Systems Engineering, prepared a standard that specifies the minimum 

design requirements for containment isolation of fluid systems that penetrate the primary containment 

boundary of light-water-cooled reactors. The American National Standards Institute (ANSI) 

Committee N18, Design Criteria for Nuclear Power Plants, approved this standard and designated it 

ANSI N271-1976, “Containment Isolation Provisions for Fluid Systems” (Ref. 2). The provisions of 

ANSI N271-1976 include minimum design, testing, and maintenance requirements for the isolation of 

fluid systems that penetrate the primary containment of light-water-cooled reactors. Requirements for the 

design and testing of power supplies, the qualification of Class 1E equipment, and the design and testing 

of protection systems are outside the scope of this standard. 

ANSI N271-1976 contains requirements (indicated by the word “shall”) and recommendations 

(indicated by the word “should”). The recommendations and the requirements of the standard were 

evaluated with respect to importance to safety. The NRC considers all recommendations to be of 

sufficient importance to safety to be endorsed along with the requirements given in the standard. 

The previous guide was issued in 1978. The proposed 2009 revision incorporates operating 

experience from the past three decades. Significantly, the proposed revision includes improved 

regulatory guidance as a result of the NRC staff’s review of the lessons learned from the accident at Three 

Mile Island Nuclear Generating Station, Unit 2. It also updates the NRC guidance on acceptable design, 

testing, and maintenance requirements that licensees may use to comply with their plant-specific design 

criteria, which meet the intent of GDC 54, 55, 56, and 57 of Appendix A to 10 CFR Part 50 for the 

isolation of fluid systems that penetrate the primary containment of light-water-cooled reactors. 

C. REGULATORY POSITION 

The requirements and recommendations for the containment isolation of fluid systems that 

penetrate the primary containment of light-water-cooled reactors, as specified in ANSI N271-1976, are 

generally acceptable and provide an adequate basis for use, subject to the following: 

1. Section 3.6.4 of ANSI N271-1976 states, “The closed system shall be leak tested in accordance 

with 5.3 of this standard unless it can be shown by inspection that system integrity is being 

maintained for those systems operating at a pressure equal to or above the containment design 

pressure.” The system integrity inspections may be applied to closed systems inside the 

containment in lieu of leak testing. 

2. Section 3.6.6 of ANSI N271-1976 states “Relief valves in the backflow direction may be 

employed as isolation valves provided they satisfy the requirements of this standard.” The 

licensee may use relief valves in the backflow direction (discharge terminates inside primary 

containment). Relief valves may also be used in the forward (relief) flow direction (discharge 

terminates outside containment) as isolation valves provided that: 

a. there are no satisfactory alternatives either by system design or selection of valve type; 

b. the relief set-point is greater than 1.5 times the containment design pressure in a manner 

consistent with NRC SRP 6.2.4, Subsection SRP Acceptance Criteria Item #7 (Ref. 3); 

and 

c. the valve reseats at a pressure at least 1.425 times the containment design pressure. 
Rev. 1 of RG 1.141, Page 3 

3. The licensee should provide thermally induced overpressure protection for liquid-filled piping 

between containment isolation barriers inside containment to prevent damage when the piping is 

isolated unless the licensee can demonstrate that the pressure between the isolation barriers 

cannot exceed the design pressure of the isolation barriers or the design pressure of the piping. 

Any thermally induced overpressure protection method that the licensee uses should provide such 

protection inside containment at the maximum back-pressure condition that could exist during a 

loss-of-coolant accident. 

4. Section 4.2.3 of ANSI N271-1976 states, “Sealed closed isolation valves are under administrative 

controls and do not require position indication in the control room for valve status.” Because the 

containment isolation valves are components of the containment isolation system, which is an 

engineered safety feature system, all power-operated valves should have position indication in the 

control room. 

5. Section 4.2.4 of ANSI N271-1976 states, “Isolation valve closure shall be completed when an 

isolation signal is received, and the valve shall not be opened until the signal is removed and 

deliberate operator action is taken (reset switch).” The reactor operator should not be able to 

override a containment isolation signal in such a way that would return any isolation valve to its 

normal (pre-accident) condition by a single action. More specifically, neither the reset/override 

of the safety injection actuation signal nor the reset/override of a containment isolation actuation 

signal for a group of valves should cause the reopening of any isolation valve. The licensee 

should not consider the use of procedural controls to prevent the reopening of a valve upon 

reset/override as an acceptable design alternative. The design of the reset/override capability 

should require a deliberate separate operator action, in addition to the reset/override of the signal, 

for the reopening of each isolation valve. Written, approved procedures should control the 

reopening of each containment isolation valve. Regulatory Guide 1.33, “Quality Assurance 

Program Requirements (Operation),” (Ref. 4), provides additional guidance on procedures. 

6. Section 4.2.5 of ANSI N271-1976 states, “Diversity in means of actuation of automatic isolation 

valves in series should be considered to preclude common mode failure.” The NRC staff’s 

position is that the licensee should provide diversity in the parameters sensed (i.e., types of 

isolation signals) for the initiation of containment isolation. The licensee may design the 

containment isolation logic to automatically initiate containment isolation upon the occurrence of 

an isolation signal derived from the individual coincidence logic of any of the continuously 

monitored parameters, such as those given in Section A.2 of Appendix A to ANSI N271-1976 for 

boiling-water reactors or in Section B.2 of Appendix B to ANSI N271-1976 for pressurized-water 

reactors. As a minimum, the licensee should monitor the following parameters, each with the 

capability of initiating containment isolation: 

a. high containment pressure; 

b. high radiation level within containment; and 

c. any manual, automatic, or coincident actuation of an engineered safety feature system or 

subsystem. 

7. Section 4.4.2 of ANSI N271-1976 states, “For power-operated isolation valves, which do not 

receive a containment isolation signal, the primary mode shall be a remote manual initiation 

signal from the main control room.” However, a containment isolation signal should 

automatically isolate all nonessential systems, as required in 10 CFR 50.34(f)(2)(xiv). 

8. Section 4.4.8 of ANSI N271-1976 gives general design requirements for closed systems. In 

addition, all branch lines and their isolation valves in closed systems both inside and outside the 
Rev. 1 of RG 1.141, Page 4 

containment should meet the design criteria of Section 3.5 or Section 3.6.7 of ANSI N271-1976 if 

the branch lines constitute one of the containment isolation barriers. 

9. Section 4.6.3 of ANSI N271-1976 cites Regulatory Guide 1.7, “Control of Combustible Gas 

Concentrations in Containment following a Loss-of-Coolant Accident” (Ref. 5), for guidance in 

determining radiation exposures for a loss-of-coolant accident. Regulatory Guide 1.89, 

“Qualification of Class 1E Equipment for Nuclear Power Plants” (Ref. 6), gives more appropriate 

guidance to determine radiation exposures for a loss-of-coolant accident. For plants that have 

amended their licensing basis to use an alternate source term, see Appendix I of Regulatory Guide 

1.183 (Ref. 7). 

10. Section 4.14 of ANSI N271-1976 states, “The piping between isolation barriers or piping, which 

forms part of isolation barriers, shall meet the requirements of 3.7 and applicable requirements for 

isolation barriers.” Piping between isolation barriers should meet the applicable requirements of 

Section 3.5 or Section 3.7 of ANSI N271-1976. 

D. IMPLEMENTATION 

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

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

back-fit in connection with its issuance. 

In some cases, applicants or licensees may propose an alternative or use a previously established 

acceptable alternative method for complying with specified portions of the NRC’s regulations. 

Otherwise, the design, testing, and maintenance requirements described in this guide will be used in 

evaluating acceptable design standards for plants. 
Rev. 1 of RG 1.141, Page 5 

REFERENCES 

1. 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” U.S. Nuclear 

Regulatory Commission, Washington, DC.1

2. ANSI N271-1976, “Containment Isolation Provisions for Fluid Systems,” American Nuclear 

Society, La Grange Park, IL, June 1976.2

3. U.S. Nuclear Regulatory Commission, “Standard Review Plan for the Review of Safety Analysis 

Reports for Nuclear Power Plants,” NUREG-0800.3

4. Regulatory Guide 1.33, “Quality Assurance Program Requirements (Operation),” U.S. Nuclear 

Regulatory Commission, Washington, DC.4

5. Regulatory Guide 1.7, “Control of Combustible Gas Concentrations in Containment following a 

Loss-of-Coolant Accident,” U.S. Nuclear Regulatory Commission, Washington, DC. 

6. Regulatory Guide 1.89, “Environmental Qualification of Certain Electric Equipment Important to 

Safety for Nuclear Power Plants,” U.S. Nuclear Regulatory Commission, Washington, DC. 

7. Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis 

Accidents at Nuclear Power Reactors,” U.S. Nuclear Regulatory Commission, Washington, DC. 

 

1

 All NRC regulations listed herein are available electronically through the Electronic Reading Room on the NRC=s public Web 

site, at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Copies are also available for inspection or copying for a fee from the 

NRC=s Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, 

Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail pdr.resource@nrc.gov. 

2 Copies of American National Standards (ANS) may be purchased from the American National Standards Institute (ANSI), 

1819 L Street, NW., 6th floor, Washington, DC 20036 [phone: (202) 293-8020)]. Purchase information is available through the 

ASCE Web site at http://webstore.ansi.org/ansidocstore/.

3

 All NUREG-series reports listed herein were published by the U.S. Nuclear Regulatory Commission. Most are available 

electronically through the Electronic Reading Room on the NRC=s public Web site, at 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/. Copies are also available for inspection or copying for a fee from the 

NRC=s Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, 

Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548; and e-mail pdr.resource@nrc.gov. 

4

 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission. Most are available 

electronically through the Electronic Reading Room on the NRC=s public Web site, at http://www.nrc.gov/ 

reading-rm/doc-collections/reg-guides/.