ML20094B853
| ML20094B853 | |
| Person / Time | |
|---|---|
| Issue date: | 11/24/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-07.1, NUREG-75-87, NUREG-75-87-7.1, SRP-07.01-01, SRP-7.01-1, NUDOCS 9511010317 | |
| Download: ML20094B853 (14) | |
Text
..
NUREG 75/087
- ps nog \\, U.S. NUCLEAR REOULATCRY CUMMl:SION p'
STANDARD REVIEW PLAN 4
0 OFFICE OF NUCLEAR REACTOR REGULATION v
SECTION
7.1 INTRODUCTION
REVIEW Resp 0NSIBILITIES Primary - Electrical instrumentation and Control Systems Branch (EICSB)
Secondary - Auxiliary and Power Conversion Systems Branch (APCSB)
ContainmentSystemsBranch(CSB)
Reactor Systems Branch (RSB) 1.
AREAS OF REVIEW Section 7.1 of the applicant's safety analysis report (SAR) contains information pertaining to safety-related instrumentation and control systems, their design bases, and the applicable acceptance criteria. EICSB reviews this information as detailed in III of this plan, and also determines the adequacy of the information presented with reference to the infomation requirements of the corresponding section of the Standard Format (Item 4.1 of Ref.1).
The secondary review branches (APCSB, CSB RSB) review the safety-related system tabula-tions for completeness, i.e., to verify that all safety-related systems within their respective areas of primary review responsibility have been identified. If systems other than those identified are deemed to be safety-related, this infomation is transmitted to EICSB.
This review plan also includes evaluation of the proposed technical specifications given in SAR Chapter 16 to assure that they are adequate with regard to safety system settings, limiting conditions for operation, and periodic surveillance testing of instrumentation and controls.
II. ACCEPTANCE CRITERIA The identification of safety-related systems is acceptable when it can be concluded that the integrated response of these systems assures the safety of the plant in normal operation, anticipated operational transients, and postulated accidents.
Table 7-1, " Acceptance Criteria for Controls." lists the criteria currently applicable to safety-related instrucentation and control systems (acceptance criteria for safety-related USNRC STANDARD REVIEW PLAN e,Yw.NNS.
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electric power systcm3 are listed in Table 8-1).
Conformance to these critoria does not necessarily establish the adequacy of the functional performance Gnd reliability of these systems. However, omission of any of the criteria will in most cases be an indication of system inadequacy. Therefore, the identification of the criteria applicable to safety-relat> instrumentation and control systems is acceptable if it includes all of the criteria listed in Table 7-1, and if the SAR contains a statement to the effect that these criteria are implemented, at the operating license (0L) stage, or will be implemented, at the construction permit (Cp) stage, in the design of these systems.
The fundamental bases for acceptance of the proposed technical specifications are that the limiting conditions for operation are such that sufficient equipment is required to be available for operation to meet the single failure criterion; that equipment outages that are permissible for a short period of time still leave available sufficient equip-ment to provide the protective function assuming no failures; and that the provisions of the technical specifications are compatible with the safety analyses.
III. REVIEW PROCEDURES Safety-related systems fall into three categories: basic safety systems, auxiliary supporting systems, and other systems important to safety.
Basic safety systems are those that directly perfonn a protective function. Examples are the reactor trip system, the emergency core cooling system, the containment isolation system, and the containment spray system. The reactor trip system provides reactor protection by fast insertion of negative reactivity (control rods) when plant conditions approach design safety limits. All the other systems listed are engineered safety features (ESF) systems; their function is to mitigate the consequences of postulated design basis accidents.
Auxiliary supporting systems are those that must function to enable operation of the basic safety systems. Component cooling systems, service water systems, ventilation systems, and electric power systems which serve ESF and reactor trip components are examples of auxiliary supporting systems. These systems must meet the same criteria as the basic safety systems they support.
Other systems important to safety are those systems which operate to reduce the proba-bility of occurrence of specific accidents, or to maintain the plant (including other safety systems) within the envelope of operating conditions postulated in the accident analyses as being required to assure full protection capability. Examples of this type of system are the cold loop startup control (interlocks) system, the accumulator tank isolation valve control (interlocks, position indication, alarms) system, and the plant status and alarm systems that provide the operator with the information necessary for initiating manual protective action. These systems are primarily instrumentation and control systems characterized by having a functional interface with the operator.
The same safety criteria apply. However, in application to this type of system, the criteria are usually further defined by regulatory guides snd in branch technical posi-tions of the EICSB.
7.1-2 11/24/75 e-e
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Tha EICSB review encompasses all of the electric power, instrumentation, and control
]
systems associated with all three categories of safety-rklated systems discribed above, with particular emphasis on the elements which constitute the protection system (as defined in IEEE Std 279-1971) and the Class IE electric systems (as defined in IEEE Std 308-1971). The safety-related electric power systems are covered in the standard review plans for Chapter 8 of the SAR. The standard review plans for SAR Chapter 7 are concerned only with the safety-related instrumentation and control systems.
The review of SAR Section 7.1 and applicable portions of the plant technical specifi-cations is perfomed as follows:
1.
E!CSB will establish that all safety-related systems are identified, and that this identification does not conflict with the more detailed information provided in other sections of the SAR, particularly in Chapters 6 and 8 and in subsequent sections of Chapter 7.
The definitions of safety-related systems presented above should be used at an aid in assessing the completeness of the identification. The secondary review tranches (APCSB, CSB, RSB) will confirm the identification of all safety-related systems within their respective areas of primary review responsi-bility. If systems other than those identified are deemed to be safety-related, this infomation shovld be transmitted to E!CSB. Particular care should be exer-cised to assure that all systems postulated in the accident analyses (Chapter 15) as being required for safety are identified as safety-related systems.
2.
EICSB verifies that other systems described in the SAR (particularly in Chapters 5, 6, 8, 9,10,11, and 15) but not identified by the applicant in Section 7.1 are not required for safety. The reviewer should obtain concurrence from the secondary review branches with regard to systems considered to be safety-related by E!CSB, but which have not been ider tified as such by the applicant. Written requests for 4
evaluation should be made to the secondary review branches when there are novel designs or significant differences of opinion.
3.
EICSB verifies that the safety-related systems are categorized by supplier, i.e.,
those designed and supplied by the nuclear steam system supplier and those designed or supplied by others.
4.
EICSB verifies that systems identical to those of reference plants that have recently received construction permits or operating licenses and those ; hat differ from the reference plants are so identified.
5.
E!CSB verifies that for those systems that are different from those of the reference plants, the differences are described and justified to the extent necessary for an evaluation of their safety significance.
6.
E!CSB confirms that the criteria identified as being applicable to the design of safety-related instrumentation and control systems include those criteria listed j
in Table 7-1.
This identification meets the applicable requirements of General 7.1-3 11/24/75
D: sign Criterion 1. " Quality Standards and R2 cords " of Appendix A of 10 CFR Part
- 50. General Design Criterion 1 also requires that, "Structurzs, systems and comp-onents important to safety shall be designed, fabricated, erected and tested to quality standards comensurate with the importance of the safety function to be performed.* Therefore, the SAR should include (1) a discussion regarding the applicability of each criterion listed, and (2) a statement to the effect that the criteria are implemented (OL) or will be implemented (CP) in the oesign of
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safety-related instrumentation and control systems.
7.
E!CSB verifies that technical design bases are provided (reference to other sections of the SAR is acceptable) for all the various functions of the protection system.
8.
Applicable portions of the proposed plant technical specifications (SAR Chapter 16) are reviewed by E!CSB and the secondary review branches to:
Confirm the suitability of the safety limits, limiting safety system settings, and a.
the limiting conditions for operation.
b.
Verify that the frequency and scope of periodic surveillance requirements are adequate.
For a CP review, it is only necessary to confirm that the applicant has identified those variables, conditions, or other items which have been determined to be probable subjects of the technical specifications (See 10 CFR 550.34(a)(5).). The applicant's justifica-tion for the selection of those items is evaluated with special attention to any that may significantly influence the final design. The specific provisions of the proposed technical specifications are not approved during the CP review. However, any specific provisions which are known to be unacceptable or which may influence acceptance of the preliminary design of the plant should be brought to the applicant's attention and, if approrpiate, included in that portion of the staff's safety evaluation report rertaining to the design of the affected systems.
For an operating license review, the proposed technical specifications are reviewed and evaluated in depth in accordance with the requirements of 10 CFR 550.36. For the EICSB areas of review, a check is made that the limiting conditions for operation (LCO) agree with the surveillance requirements, i.e., for each system or component that is the subject of a LCO, there must be corresponding surveillance requirements. Each system or component that performs a function for which credit is taken in the accident analyses should be the subject of an LCO. The limiting safety system settings should be in accordance with the values assumed in the accident analyses, including appropriate allowances for instrument error, drift, etc. If the acceptance of the design of a particular system is based upon required plant conditions or particular operating procedures, such requirements should be included in the final technical specifications and, if appropriate, noted in that portion of the staff's safety evaluation report pertaining to the design of the affected system.
7.1-4 11/24/75
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IV. EVALUATION FINDINGS EICSB verifies that sufficient infonnation has been providtd and that the review supports conclusions of the following type,'to be included in the staff's safety evaluation report:
"The applicant has identified the safety-related instrumentation and control systems and the applicable safety criteria and has documented his intent to design and implement these systems in accordance with the criteria. It is concluded that implementation of these systems in accordance with the criteria provides assurance that the plant will perform as designed in nonnal operation, anticipated operational
- transits, and postulated accident conditions, and meets the applicable requirements of General Design Criterion 1."
V.
REFERENCES 1.
Standard Review Plan Table 7-1, " Acceptance Criteria for Controls."*
s
- All references for this plan are included in Standard Review Plan Table 7-1.
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1 OFFICE OF NUCLEAR REACTOR REGULATION TABLE 7-1 ACCEPTANCE CRITERIA FOR CONTROLS Table 7-1 contains the acceptance criteria for the review plans of Chapter 7.
These acceptance criteria include the applicable general design criteria. IEEE standards, regulatory guides, and branch technical positions (BTP) of the Electrical. Instrumentation and Control Systems Branch (EICSB). The table was prepared by EICSB for use by its members in reviewing Chapter 7 and for use by secondary review branch reviewers.
The applicability of these criteria to specific sections of Chapter 7 is indicated by an X in the matrix listing of criteria and SAR sections. There is a corresponding table (8-1) at the end of Chapter 8 covering the acceptance criteria of safety-related power supplies. The BTP listed in Tables 7-1 and 8-1 are contained in Appendix 7-A to the Chapter 7 review plans.
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CRITERIA TITLE APPLICABILITY (SAR Section)
-REMARKS
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7.1 7.2 7.3 7.4 7.5 7.6 7.7 1.
10 CFR 550.34 Contents of Application: Technical Informattor X X
X X
.X X
X b.
10 CFR 550.36 Technical Specifications X
X X
X
-X X
~
c.
10 CFR 550.55a Codes and Standards X
.X X
X X
X X
l 1
2.
General Design Criteria (GDC),
I Appendix A to 10 CFR Part 50 1
a.
GDC 1 Quality Standards and Records X
X X
X X
X b.
GDC 2 Design Bases for Protection Against Natural
[
Phenomena X
X X
X X
X t
c.
GDC 3 Fire Protection X
X X
X X
X t
d.
GDC 4 Environmental and Missile Design Bases X
X X
X X
X I
e.
GDC 5 Sharing of Structures, Systems, and Components X
X X
X X
i.
[
f.
GDC 10 Reactor Design X
X X
X X
X
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g.
GDC 12 Suppression of Reactor Power Oscillations X
X X
X I
h.
GDC 13 Instrumentation and Control X-X X
X X
X X
1.
GDC 15 Reactor Coolant System Design X
X X
X t
j.
GDC 19 Control Room X
X X
X X
X X
\\
k.
GDC 20 Protection System Functions X
X X
X X
X i
1.
GDC 21 Protection System Reliability and Testability X
X X
X X
X e
m.
GDC 22 Protection System Independence X
X X
X X
X i
6 Table 7-1:
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TABLE 7-1 (CONTINUED)
APPLICABILITY (SAR Section)
CRITERIA TITLE 7.1 7.2 7.3 7.4 7.5 7.6I 7.7 REMARKS n.
GDC 23 Protection System Failure Modes X
X X
X X
X v
o.
GDC 24 Separation of Protection and Control Systems X
X X
X X
X X
p.
GDC 25 Protection System Requirements for Reactivity Control Maltunctions X
X X
i i
q.
GDC 26 Reactivity Control System Redundancy and X
X X
X X
Capability r.
GDC 27 Combined Reactivity Control Systems Capability X
X X
X X
s.
GDC 28 Reactivity Limits X
X X
X X
t.
GDC 29 Protection Against Anticipated Operational Occurrences X
X X
X X
X X
l u.
GDC 33 Reactor Coolant Makeup X
X X
X v.
GDC 34 Residual Heat Removal X
X X
X X
w.
GDC 35 Emergency Core Cooling X
X X
X X
x.
GDC 37 Tes&'ng of Onergency Core Cooling System X
X X
X X
l 5
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GDC 38 Containment Heat Removal X
X X
X I
z.
GDC 40 Testing of Containment. Heat Removal System X
X X
X aa.
6DC 41 Containment Atmosphere Cleanup X
X X
X bb.
GDC 43 Testing of Containment Atmosphere Cleanup X
X l
X X
Systems cc.
GDC 44 Cooling Water X
X X
X j
f dd.
GDC 46 Testing of Cooling Water System X
X X
X i
i X
X X
lContainmentDesignBasis r
ee.
Gnc 50 i
s to Table 7-1:
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J TABLE 7-1 (CONTINUED) b APPLICABILITY (SAR Section) g CRITERIA TITLE 7.1 7.2 7.3 7.4 73 7.6 7.7 REMARKS ff. GDC 54 Piping Systems Penetrating Containment X
X X
X l
gg. GDC 55 Reactor Coolant Pressure Boundary r
Penetrating Containment X
X X
X hh.
GDC 56 Primary Containment Isolation X
X X
X
- 11. GDC 57 Closed Systems Isolation Valves X
X X
X
[
3.
Institute of Electrical and Electronics Engineers (IEEE) t Standards:
a.
IEEE Std 279-1971 Criteria for Protection Systems for See 10 CFR 550.55a(h)
Nuclear Power Generating Stations X
X X
X X
X X
and Reg. Guide 1.62 b.
IEEE Std 308-1971 Criteria for Class IE Electric Systems for X
X X
X See Reg. Guide 1.32.
Nuclear Power Generating Stations c.
IEEE Std 317-1972 Electric Penetration Assemblies in Contain-See Reg. Guide 1.63.
ment Structures for Nuclear Power Generating Stations X
X X
X X
X X
i d.
!nsta11ation, Inspection and Testing Require-See Reg. Guide 1.30.
(ANSI N45.2.4-1972) mery.= er Instrumentation and Electric Equip-mer.t During the Construction of Nuclear Power Gen e ating Stations.
X X
X X
X X
X e.
IEEE Std 338-1971 Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems X
X X
X X
X i
f.
IEEE Std 344-1971 Guide for Seismic Qualification of Class I
[
Electrical Equipment for Nuclear Power l
Generating Stations X
X X
X X
X g.
IEEE Std 379-1972 Guide for the Application of the Single See Reg. Guide 1.53.
i (ANSI N41.2)
Failure Criterion to Nuclear Power Generating j
l Station Protection Sytems X
X X
X X
X X
3 h.
IEEE Std 384-1974 Criteria for Separation of Class IE Equipment X
X X
X (ANSI N41.14) and Circuits t
i Table 7-1:
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I TABLE 7-1 (CONTINUED)
APPLICABILITY (SARSection)
CRITERIA TITLE 7.1 7.2 7.3 7.4 7.5 7.6 7.7 REMARKS 4.
Regulatory Guides (RG) a.
RG 1.6 Independence Between Reduncant Standby (Onsite)
Power Sources and Between Their Distribution X
X X
Systems b.
RG 1.7 Control of Conbustible Gas Concentrations in Containment following a Loss of Coolant Accident X
X X
c.
RG 1.11 Instrument Lines Penetrating Primary Reactor Containment d.
RG 1.22 Periodic Testing of Protection System Actuation Functions X
X X
X X
X e.
RG 1.29 Seismic Design Classification X
X X
X X
X f.
RG 1.30 Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment X
X X
i X
X X
g.
RG 1.32 Use of IEEE Std 308-1971, " Criteria for Class IE Electric Systems for Nuclear l
Power Generating Stations" X
X X
X i
h.
RG 1.47 Bypassed and Inoperable Status Indication for Use in conjunction Nuclear Power Plant Safety Systems X
X X
X X
X with Position 3, i
?
1 RG 1.53 Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems X
X X
X X
X J.
RG 1.62 Manual Initiation of Protection Actions X
X X
X X
1 k.
RG 1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants X X
'X X
X X
X l.
RG 1.68 Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors X
X j
^
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TABLE 7-1 (CONTINUED)
R w
y, APPLICABILITY (SAR Section)
CRITERIA TITLE 7.1 7.2 7.3 7.4 7.5 7.6 7.7 REMARKS m.
RG 1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants. Rev. 2.
X X
X X
X X
X n.
RG 1.75 Physical Independence of Electric Systems X
X X
X o.
RG 1.78 Assumptions for Evaluating the Habitabilicy of a Nuclear Power Plant Control Room Du+1ng a Postulated Hazardous Chemical Release X
X p.
Qualification of Class IE Equipment for Nuclear Power Plants X
X X
X X~
X q.
RG 1.96 Design of Main Stear. Solation Valve Leakage Control Systems for Bolling Water Reactor Nuclear Power Plants X
X 5.
Branch Technical Positions (BTP) EICSB a.
BTP EICSB 1 Backfitting of the Protection and Emergency Power Systems of Nuclear Reactors X
X X
X X
b.
BTP EICSB 3 Isolation of Low Pressure Systems from the High Pressure Reactor Coolant System X
X X
c.
BTP EICSB 4 Requirements on Motor-Operated Valves in the ECCS Accumulator Lines X
X X
d.
BTP EICSB 5 Scram Breaker Test Requirements - Technical Specifications X
X e.
BTP EICSB 9 Definition tnd Use of " Channel-Calibration" -
Technical Specifications X
X X
X X
f.
BTP EICSB 10 Electrical and Mechanical Equipment Seismic.
Qualification Program X
X X
X X
g.
BTP EICSB 12 Protection System Trip Point Changes for Operation with Reactor Coolant Pumps Out of Sersice X
X X
Table 7-1: -6 i
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TABLE 7-1 (CONTINUED)
APPLICABILITY (SAR Section)
CRITERIA TITLE 7.1 7.2 7.3 7.4 7.5 7.6 7.7 REMARKS h '.
BTP EICSB 13 Design Criteria for Auxiliary Feedwater X
X Systems 4
1.
BTP EICSB 14 Spurious Withdrawals of Single Control Rods in Pressurized Water Reactors X
X j
j.
BTP EICSB 15 Reactor Coolant Pump Breaker Qualification X
X k.
BTP EICSB 16 Control Element Assembly (CEA) Interlocks in 7
X X
Combustion Engineering Reactors 1.
BTP EICSB 18 Application of the Single Failure Criteria to Manually-Controlled Electrically-Operated Valves X
X X
X e
i m.
BTP EICSB 19 Acceptability of Design Criteria for Hydrogen X
X X
Mixing and Drywell Vacuum Relief Systems n.
BTF EICSB 20 Design of Instrumentation and Controls Provided to Accomplish Changeover from Injection s
to Recriculation Mode X
X X
X l
o.
BTP EICSB 21 Guidance for Application of Reg. Guide 1.47 X
X X
X X
X p.
BTP EICSB 22 Guidance for Application of Reg. Guide 1.22 X
X X
X X
X i
q.
BTP EICSB 23 Qualification of Safety-Related Display Instrumentation for Post-Accident Condition Monitoring and Safe Shutdown X
X r.
BTP EICSB 24 Testing of Reactor Trip System and Engineered Safety Feature Actuation System Sensor Response Times X
X X
X X
s.
BTP EICSB 25 Guidance for the Interpretation of General Design Criterion 37 for Testing the Operability of the Emergency Core Cooling System as a Whole X
X X
{
Requirements for Reactor Protection System t.
BTP EICSB 26 Anticipatory Trips X
X U
u.
BTP EICSB 27 Design Criteria for Thermal Overland Protection for tiotors of tiotor-Operated Valves X
X X
X a
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