ML20083L826
| ML20083L826 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1984 |
| From: | Serkiz A Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| REF-GTECI-A-01, REF-GTECI-PI, TASK-A-01, TASK-A-1, TASK-OR NUREG-0800, NUREG-0800-06.3-R2, NUREG-800, NUREG-800-6.3-R2, SRP-06.03, SRP-6.03, NUDOCS 8404170375 | |
| Download: ML20083L826 (16) | |
Text
t NUREG-0800 (Formerly NUREG-75/087)
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U.S. NUCLEAR REGULATORY COMMISSION igj/li STANDARD REVIEW PLAN k;.'.u OFFICE OF NUCLEAR REACTOR REGULATION Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants 6.3 Section No.
Revision No. 2 Appendix No. N/A Revision No. N/A Branch Tech. Position _R_SB 6-1 Revision No. 1 Date issued April 1984 FILING INSTRUCTIONS PAGES TO BE REMOVED NEW PAGES TO BE INSERTED PAGE NUMBER DATE PAGE NUMBER DATE l
6.3-1 Rev. 1 July 1981 6.3-1 Rev. 2 April 1984 thru thru 6.3-12=
6.3-13 BTP RSB 6-1 to Rev. 1 July 1981 BTP RSB 6-1 to Rev. 1 July 1981 SRP Section SRP Section (No change except 6.3 6.3 for page numbers) 6.3-13 6.3-14 thru thru 6.3-14 6.3-15 8404170375 840430 PDR NUREQ N,
0800 R PDR The U.S. Nuclear Regulatory Commission's Standard Review Plan, NUREG-0800, prepared by the Office of Nuclear Reactor Regulation. is available for sale by the National Technical Information Service. Springfield, VA 22161.
NU REG-0800 (Formerly NUREG-75/087) mO.r N o.s.NoCLoo EGoLATOav CO-ssiON i'%i STANDARD REVIEW PLAN Qv OFFICE OF NoCLEAR REACTOR REGoLATION
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e eee 6.3 EMERGENCY CORE COOLING SYSTEM REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB)
Secondary - None I.
AREAS OF REVIEW The RSB reviews the information presented in the applicant's safety analysis report (SAR) regarding the emergency core cooling system (ECCS). The major elements of the review are:
1.
Design Bases The design bases for the ECCS are reviewed to assure that they satisfy g) applicable regulations, including the general design criteria and the
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amendments tc 10 CFR Part 50 regarding ECCS acceptance criteria issued by the Commission on December 28, 1973 (Ref. 1).
2.
Design The design of the ECCS is reviewed to determine that it is capable of per-forming all of the functions required by the design bases.
3.
Test Program l
The preoperational and initial startup test programs for the ECCS are reviewed by the Procedures and Systems Review Branch (PSRB) to determine if they are sufficient to confirm the performance capability of the ECCS.
RSB reviews the need for special design features to permit the performance of adequate test programs.
Rev. 2 - April 1984 USNRC STANDARD REVIEW PLAN Standard review plans are prepared for the guidance of the Office of Nuclear Reactor Regulation staff responsible for the review of applications to construct and operate nuclear power plants. These documents are made available to the public as part of the jm Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
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plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
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Not all sections of the Standard Format have a corresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa-tion and experience.
Comments and suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D.C. 20556.
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4.
Technical Specifications The proposed technical specifications are reviewed to assure that they are adequate in regard to limiting conditions of operation and periodic surveillance testing.
The ability of the ECCS to mitigate the consequences of a spectrum of loss-of-coolant accidents is reviewed by RSB under SRP Section 15.6.5.
In addition the RSB will coordinate with other branches evaluations that interface with the overall ECCS review as follows:
Auxiliary Systems Branch (ASB), as part of its primary review responsibility for SRP Sections 9.2.1, 9.2.2, 9.2.5, and 9.2.6, reviews those auxiliary systems essential for ECCS operation (service water system, component cooling system, ultimate heat sink, and condensate storage facility) and assesses the capability of these systems to perform all functions required by the ECCS.
The ASB will supply, on request, evaluations of portions of the power conversion systems (e.g., steam supply lines, steam generators, feedwater systems) which interface with the reactor coolant system in such a way as to influence the course of a loss-of-coolant accident (LOCA) for a particular plant.
The ASB also reviews the effects of pipe breaks This rev ew includes the effect of pipe i
outside containment on ECCS.
whip, jet impingement forces, and environmental conditions created as part of its primary review responsibility for SRP Section 3.6.1.
Instrumenta-tion and Control Systems Branch (ICSB), as part of its primary review responsibility for SRP Section 7.3, reviews the adequacy of ECCS-associated controls and instrumentation with regard to the features of automatic actuation, remote sensing and indication, and remote control. The Containment Systems Bran'ch (CSB) verifies that portions of the ECCS penetrating the containment barrier are designed with acceptable isolation features to maintain containment integrity for all operating conditions, including accidents, as part of its primary review responsibility for SRP Section 6.2.4.
The Power Systems Branch (PSB) as part of its primary review responsibility for SRP Sections 8.1, 8.2, 8.3.1, and 8.3.2, reviews the adequacy of the power supply for the ECCS. The Mechanical Engineering Branch (MEB), as part of its primary review responsibility for SRP Section 3.9.3, reviews the loading combinations (operational, LOCA, and seismic) and the associated stress limits.
In addition, the MEB, as part of its primary review responsibility for SRP Section 3.6.2, reviews the criteria used for postulating the effects of pipe breaks both inside and outside containment on ECCS.
This review includes criteria used for postulating the effects of pipe whip, jet impingement forces, and any related environmental conditions.
The ECCS is also reviewed by MEB to assure that system and components have the proper seismic and quality group classifications.
This aspect of the review is performed as part of its primary review responsibility for SRP Sectons 3.2.1 and 3.2.2.
The i
Structural and Geotechnical Enginearing Branch (SGEB) reviews the I
structures housing the ECCS for the proper seismic classification as part of its primary review responsibility for SRP Sections 3.8.1, 3.8.2, and 3.8.3.
The Materials Engineering Branch (MTEB), on a generic basis, reviews the thermal shock effect of water injected into the primary coolant system from the ECCS. The Procedures and Systems Review Branch (PSRB) reviews the proposed preoperational and initial startup test pro-grams to determine that they are consistent with the intent of Rcgulatory 6.3-2 Rev. 2 - April 1984
Guides 1.68 and 1.79 as part of its primary review responsibility for SRP 1
Section 14.2.
The PSRB also has primary review responsibility for Task Action Plan items l
II.K.1 (C.1.10) of NUREG-0694 (0Ls only) and I.C.6 of NUREG-0718 (cps only) regarding procedures to ensure that system operability status is known. The Radiological Assessment Branch (RAB) has primary review responsibility for SRP Sections 12.1 through 12.5 including Task Action t
Plan items II.B.2 of NUREG-0694 and NUREG-0718 which involve radiation and shielding design review to_take corrective actions to ensure adequate access to vital areas and protection of safety equipment (cps and OLs).
The review for Technical Specifications and Quality Assurance are coordinated and performed by the Standardization and Special Projects l
Branch and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 16.0 and 17.0, respectively.
For those areas of review identified above as being reviewed as part of the primary review responsibility of other branches, the acceptance criteria necessary for the review and their methods of application are contained in the referenced SRP section of the corresponding primary branch.
II. ACCEPTANCE CRITERIA The RSB acceptance criteria are based on meeting the relevant requirements of the following regulations:
s 1
A.
General Design Criterion 2 as it relates to the seismic design of struc-tures, systems, and components whose failure could cause an unacceptable reduction in the capability of the ECCS to perform its safety function.
Acceptability is based on meeting position C2 of Regulatory Guide _1.29.
j-B.'
General Design Criterion'4 as related to dynamic effects associated with flow instabilities and loads (e.g., water hammer).
C.
General Design Criterion 5 as it-relates to structures, systems, and com-
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ponents important to safety,shall not be shared among nuclear power units unless it can be demonstrated that sharing will not impair.their ability l
'to perform their safety function.'
D.
General Design Criterion 17 as it relates to the design'of the ECCS.having
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sufficient capacity and capability to assure that specified acceptable.
fuel design limits and the design conditions of the reactor coolant pres-sure boundary are not exceeded and that the core is cooled during antici-pated operational occurrences and accident conditions.
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E.
General Design Criterion 27 as itJrelates.to the' system design having the capability'to assure that under postulated accident conditions and with appropriate margin for stuck rods,. the capability to cool the core ~ is i
maintained.
O F.
General Design Criteria 35, 36,'and 37 as they relate to the ECCS being designed to provide an abundance of core cooling to transfer heat from the l-core at a rate so that fuel and clad damage will not-interfere with 6.3-3' Rev. 2.
April 1984 l
l continued effective core cooling, to permit appropriate periodic inspec-tion of important components, and to permit appropriate periodic pressure and functional testing.
G.
10 CFR Part 50, 650.46, and Appendix K to 10 CFR Part 50 as it relates to the ECCS being designed so that its cooling performance is in accordance with an acceptable evaluation model.
Specific acceptance criteria, Regulatory Guides,.and Task Action Plan items that provide information, recommendations, and guidance and in general describe a basis acceptable to the staff that may be used to implement the requirements of the Commission regulations identified above are as follows:
In regard to the ECCS acceptance criteria (Ref. 1), the five major performance criteria deal with:
1.
Peak cladding temperature.
2.
Maximum calculated cladding oxidation.
3.
Maximum hydrogen generation.
4.
Coolable core geometry.
5.
Long-term cooling.
These areas are reviewed as a part of the effort associated with the LOCA analysis (SRP Section 15.6.5).
However, the impact of various postulated single failures on the operability of the ECCS is evaluated under this SRP section.
The ECCS must meet the requirements of GDC 35 (Ref. 6).
The system must have alternate sources of electric power, as required by GDC 17 (Ref. 4), and must be able to withstand a single failure.
The ECCS should retain its capability to cool the core in the event of a failure of any single active component dur-ing the short term immediately following an accident, or a single active or passive failure during the long-term recirculation cooling phase following an accident.
The ECCS must be designed to permit periodic inservice inspection of important components, su-h as spray rings in the reactor pressure vessel, water injection nozzles, piping, pumps, and valves in accordance with the requirements of GDC 36 (Ref. 7).
The ECCS must be designed to permit testing of the operability of the system throughout the life of the plant, including the full operational sequence that brings the system into operation, as required by GDC 37 (Ref. 8).
The combined reactivity control system capability associated with ECCS must meet the requirements of GDC 27 (Ref. 5) and should conform to the recommenda-tion of Regulatory Guide 1.47 (Ref. 11). The primary mode of actuation for the ECCS must be automatic, and actuation must be initiated by signals of suitable diversity and redundance.
Provisions should also be made for manual actuation, monitoring, and control of the ECCS from the reactor control room.
The design of the ECCS should conform to the recommendations of Regulatory Guide 1.1 (Ref. 9).
Design features and operating procedures, designed to prevent damaging water hammer due to such mechanisms as voided discharge lines and water entrainment l
6.3-4 Rev. 2 - April 1984
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I kl' in steain lines shall be provided, in order to meet the requirements of General l ',1 Design Criterion 4 (Ref. 17).
j3 The design of those portions of the system which are not safety related, whose N'
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4 failures could have an adverse effect on the ECCS system, must be in accordance 4
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with GDC 2 (Ref. 2), and acceptance is based on meeting Position C2 of Regulatcry Guide 1.29 (Ref. 10).
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Interfaces between the ECCS and component or servi::e water systems must be such that operation of one does not interfere with, and provides proper support (where re4taired) for, the other.
In relation to these and other shared 4
systems, e.g., residual heat removal (RHR) and containment heat removal systems, the ECCS must conform to GDC 5 (Ref. 3).
J The requirements of the following Task Action Plan items must also be t
satisfied:
,s 6 1.
Task Action Plan Item II.B.8'of NUREG-0718 (Ref. 14) which involves I
description by the applicants of the degree to which the designs conform j
to the proposed interim rule 'on degraded core accidents (cps and OLs).
I 2.
Task Action Plan Item III.D.1.1 of NUREG-0694 and NUREG-0718 which involves primary coolant sources outside of containment (cps and OLs).
i 3.
Task Action Plan Item II.E.2.1 of NUREG-0737 which involves reliance on ECCS.
4.
Task Action Plan Item II.K.3(10) of NUREG-0737 and NUREG-0718 which l
involves final recommendations by B&O task force regarding applicant's proposal of use of anticipatory trips only at high power for selected plants.
5.
. Task Action Plan Item II.K.3(15) of NUREG-0737 and NUREG-0718 which involves isolation of PCI and RCIC for BWR. plants.
6.
. Task Action Plan Item II.K.3(18) of NUREG-0737 and NUREG-0718 involving ECCS outages for all' plants.
7.-
Task Action Plav Item II.K.3(21) of NUREG-0737 and NUREG-0718 which involves a study evaluatirg restart of'LPCS and LPCI after manual trip for BWR plants.
8.
-Task Action Plan Item II.K.3(39) of NUREG-0660 which' involves evaluation.
of effects of water slugs in piping caused by.HPI_and CF1 flows in B&W.
plants.
In addition to the above criteria, the acceptability of the ECCS may be based-on the degree of design similarity with previously approved plants.
h III.. REVIEW PROCEDURES The procedures below are used during the construction permit-(CP) review'to assure that the' design criteria and bases and the preliminary design as set
.6.3-5 Rev. 2 - April 1984
forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II of this SRP section.
For operating license (0L) reviews, the procedures are utilized to verify that the initial design criteria and bases have been appropriately implemented in the final design as set forth in the final safety analysis report.
The OL review also includes the proposed technical specifications to assure that they are adequate in regard to limiting conditions of operation and periodic surveillance testing.
Much of the review described below is generic in nature and is not performed for each plant.
That is, the RSB reviewer compares the ECCS design and param-eters to those of previously reviewed plants and then devotes the major portion of the review effort to those areas where the application is not identical to previously reviewed plants.
The following steps are taken by the RSB reviewer to determine that the acceptance criteria of subsection II have been met.
These steps should be adapted to CP or OL reviews as appropriate.
1.
The relationship of the system under review to other previously approved plants is established.
Systems or design features claimed to be identical or equivalent to those of previously approved plants are confirmed to be identical or equivalent.
2.
Piping diagrams are reviewed to evaluate the functional reliability of the system in the event of single failures.
That is, by referring to piping and instrumentation diagrams, the existence of the redundancy required by the criteria is confirmed.
3.
The significant design parameters (e.g., pump net positive suction head, pump head vs. flow, accumulator volume and pressure, water storaga volume, system flow rate and pressure, etc.) are examined for each component to confirm that these' parameters satisfy operating requirements and the recommendations of Regulatory Guide 1.1 (Ref. 9).
4.
The piping and instrunientation diagrams are checked in consultation with MEB to see that essential ECCS components are designated seismic Category I and Safety Class II (the cooling water side of heat exchangers can be Safety Class III).
5.
The ECCS design is reviewed to confirm that the system can function in postaccident environments, considering possible mechanical effects, missiles, and the pressure, temperature, moisture, radioactivity, and chemical conditions resulting from LOCA.
Protection against valve motor flooding should be confirmed by the RSB reviewer.
Regarding the effects of pressure, temperature, etc., the RSB reviewer should confirm that accident conditions are specified which provide the basis for proof tests for environmental qualification of ECCS components.
6.
The criteria, supporting analyses, plant design provisions, and operator actions that will be taken are reviewed to ensure that there will not be unacceptably high concentrations of boric acid in the core region (result-ing in precipitation of a solid phase) during the long-term cooling phase following a postulated LOCA.
6.3-6 Rev. 2 - April 1984
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7.
The ECCS design is reviewed to confirm that there are provisions for main-tenance of the long-term coolant recirculation and decay heat removal systems, e.g., pump or valve overhaul, in the post-LOCA environment (including consideration of radioactivity).
8.
The availability of an adequate source of water for the ECCS is confirmed, and the source volume, location, and susceptibility to failure (e.g.,
freezing) are evaluated. (RSB will request ASB review as required.) In l
PWRs, the piping from the water source to the ECCS safety injection pumps is evaluated for conformance with RSB 6-1 (Ref. 13).
9.
The ECCS flow paths are reviewed to determine the extent to which flow from the ECCS pumps is diverted as a backup feature to other safeguards equipment (e.g., RHR, containment spray).
The reviewer should confirm that the remaining portion of the flow provides abundant core cooling, despite the most severe single failure that affects ECCS flow.
10.
For a boiling water reactor (BWR), the reactor coolant automatic depressurization system is reviewed to confirm-the capability to satisfy LOCA pressure relief functions, including consideration of a single failure.
4 11.
The design of ECCS injection lines is reviewed to confirm that the isolation provisions at the interface with-the reactor coolant system are 4
adequate. The number and type of valves used to form the interface between low pressure portions of the ECCS and the reactor coolant system must provide adequate assurance that the ECCS will not be subjected to a pressure greater than its design pressure..This may be accomplished by any of the following provisions:
a.
One or more check valves in series with a normally closed motor-operated valve. The motor-operated valve is to be opened upon'
-receipt of a safety injection signal ~once the reactor coolant pressure has decreased below the-ECCS design pressure.
i b.
Three check' valves in series.
c.
Two check valves in series, provided that there are design provisions to permit periodic testing of the check ~ valves.for leaktightness and the~ testing is performed at least annually.
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- 12. The reviewer should identify those portions of.nonsafety-related systems-
_which could have an adverse'effect on ECCS and'should ensure that modi-fications are in place to correct these situations.-
13.
Motor-operated isolation valves in ECCS lines connecting the accumulators 1
- to the reactor' coolant. system in a pressurized water reactor (PWR) are.
a
. reviewed to ensure that adequate provisions are made against.. inadvertent isolation;-
- 14..The: capacity and' settings of relief valves provided forLthe ECCS to
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satisfy system overpressure-protection requirements are reviewed.
In
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'V particular, for-PWRs, the reviewer confirms that the accumulator relief ;
6.3-7 Rev. 2. Aprf1 - 1984
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valves have adequate capacity so that leakage from the reactor coolant system will not jeopardize the integrity of the accumulators.
15.
The ECCS is reviewed to evaluate the adequacy of design features that have been provided to prevent damaging water (steam) hammer due to such mechanisms as voided discharge lines, water entrainment in steam lines and steam bubble collapse.
For systems with a water supply above the discharge lines, voided lines are prevented by proper vent location and filling and venting procedures.
However, for the core spray and low pressure coolant injection systems of BWRs, the low elevation of the suppression pool will result in line voidage because of back leakage through pump discharge check valves and leaking valves in the full flow test line.
Proper vent location and filling and venting procedure are still needed.
In addition, a special keep-full system with appropriate alarms is needed to supply water to the discharge lines for any system which has a water source below the level of the highest pump discharge lines and at sufficiently high pressure to prevent voiding.
For the High Pressure Coolant Injection (HPCI) system of BWRs which uses a steam-driven turbine, typical design features for the steam supply line include (a) drain pots with testable drain pot level switches, (b) sloped lines, and (c) limitations on opening and closing sequences and seal-ins for manual operation of the isolation valves to prevent introducing water slugs into the line.
The turbine exhaust line features include sloped lines and vacuum breakers.
16.
The reviewer confirms that no component or feature of the ECCS in one reactor facility on a multiple plant site is shared with the ECCS in another facility, or that shared features clearly meet the requirements of GDC 5 (Ref. 3).
17.
The reviewer confirms that within an individual reactor facility, any com-ponents shared between the ECCS and other systems (e.g., coolant makeup systems, residual heat removal systems, containment cooling systems) satisfy engineered safeguard feature design requirements and that the ECCS function of the shared component is not diminished by the sharing.
18.
The reviewer confirms that ECCS components located exterior to the reactor containment are housed in a structure which, in the event of leakage from the ECCS, permits venting of releases through iodine filters designed in accordance with Regulatory Guide 1.52.
19.
The complete sequence of ECCS operation from accident occurrence through long-term core cooling is examined to see that a minimum of manual action is required and, where manual action is used, a sufficient time (greater than 20 minutes) is available for the operator to respond.
20.
The reviewer confirms that long-term cooling capacity is adequate in the event of failure of any single active or passive component of the INS.
If an intermediate heat transport system, such as the component cooling water system, is used to provide long-term cooling capability, the system must be designed and constructed to an appropriate group classification, must be seismic Category I, and must be capable of sustaining a single active or passive failure without loss of function.
6.3-8 Rev. 2 - April 1984
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21.
The RSB reviewer consults with the ICSB reviewer to:
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a.
Confirm that the power requirements of the ECCS, including the timing of electrical loads, are compatible with the design of onsite emergency power systems, both a-c and d-c.
b.
Confirm that there are sufficient instrumentation and controls avail-able to the reactor operator to provide adequate information in the control room to assist in assessing post-LOCA conditions, including the more significant parameters scch as coolant flow, coolant temperature, and containment pressure.
If ECCS flow is diverted as a backup to other safeguards systems, the reviewer confirms that instrumentation and controls are available to provide sufficient information in the control room to determine that adequate core cooling is being provided.
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c.
Confirm that automatic actuation and remote-manual valve controls are capable of performing the functions required, that suitable 4
interlocks are provided, which do not impair separation of power t
trains.or inhibit the required valve motions, and that instrumenta-tion and controls'have sufficient redundancy to satisfy the single failure criterion.
22.
Analyses are provided by the applicant in Chapter 15 of the SAR to assess the capability of the ECCS to meet-functional requirements.
These analyses are reviewed by the RSB, as described in SRP Section 15.6.5, to determine conformance to.the. acceptance criteria for ECCS.
However,-the following portions of the review of ECCS response in loss-of-coolant accidents are performed by the RSB reviewer under this SRP section:
a.
. The lower limit of break size for_ which ECCS operation is required is.
established; i.e., the maximum break size for which' normal reactor =
coolant makeup systems can maintain reactor. pressure and coolant level-is determined. The capability of'the ECCS to actuate-and-
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perform at this lower limit of break size is confirmed, I
. The reviewer confirms that the analyses take into account 1a variety b.
of potential locations for postulated pipe-breaks, including'ECCS:
- injection lines, c.
The reviewer confirms that 'the analyses take into account' a variety of single active failures. The reviewer should keep 'in mind that different single failures may be. limiting, depending on the particular break location and break size postulated.
d.
The ECCS component response-times (e.g., for valves,1 pumps, power supply) are reviewed ~to confirm that'they.are within the delay times-used-in.theLaccident analyses.
The ECCS design adequacy'for all modes of reactor operation-(e.g.,
e.
full poweri low power, hot standby, cold shutdown,~ partial 1 cop-
-isolation) is confirmed.
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. The proposed plant technical specifications _are reviewed to:
23.
6.3-9
- Rev. 2 - April 1984'
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a.
Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restrictions for periods when ECCS equipment is inoperable due to repairs and maintenance.
The means of indicating that safety systems have been bypassed or are inoperable should be in accordance with Regulatory Guide 1.47 (Ref. 11).
b.
Confirm that the limiting conditions of operation ensure that the specified operating parameters (minimum poison concentrations, minimum coolant reserve in storage, etc.) are within the bounds of the analyzed conditions.
c.
. Verify that the frequency and scope of periodic surveillance testing is adequate.
24.
The reviewer confirms that the design provides the capability for periodi-cally demonstrating that the system will operate properly when an accident signal is received.
That is, it should be rem 3nstrated by an applicant that pumps and valves operate on normal and vargency power and that water pressure and flow are as designed when the plant is operating (periodic system surveillance). When the plant is shut down for refueling, the system should be tested for delivery of coolant to the vessel.
25.
The RSB reviewer contacts his counterpart in PSRB to discuss any special I
test requirements and to confirm that the proposed preoperational test program for the ECCS is in conformance with the intent of Regulatory Guide 1.68 (Ref. 12).
26.
The RSB review evaluates the applicant responses to the following Task Action Plan items:
(a)
II.B.8 of NUREG-0718 (cps only)
(b) III.D.1.1 of NUREG-0737 and NUREG-0718 (cps and OLs)
(c)
II.E.2.1 of NUREG-0660 (d) II.K.3(10) of NUREG-0660 (e) II.K.3(15) of NUREG-0660 (f) II.K.3(18) of NUREG-0660 (g)
II.K.3(21) of NUREG-0660 (h) II.K.3(39) of NUREG-0660 IV.
EVALUATION FINDINGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of statements and conclusions which should be included in the staff's safety evaluation report.
(For completeness, this evaluation finding includes the RSB review effort described in SRP Section 15.6.5.)
The emergency core cooling system (ECCS) includes the piping, valves, pumps, heat exchangers, instrumentation, and controls used to transfer heat from the core following a loss-of-coolant. accident.
The scope of review of the ECCS for the _
plant included piping and instrumentation diagrams, equipment layout drawings, failure modes and effects analyses, and design specifications for essential components.
The review has included the applicant's proposed 6.3-10 Rev. 2 - April 1984
O design criteria and design bases for the ECCS and the manner in which the design conforms to these criteria and bases.
The staff concludes that the design of the Emergency Core Cooling System is acceptable and meets the requirements of General Design Criteria 2, 4, 5, 17, l
27, 35, 36, and 37.
This conclusion is based on the following:
(1) The applicant has met the requirements of GDC 2 with regard to the seismic design of nonsafety systems or portions thereof which could have an adverse effect on ECCS by meeting position C.2 of Regulatory Guide 1.29.
(2) The applicant has met the requirements of GDC 4 as related to dynamic effects associated with flow instabilities and loads (e.g., water hammer).
(3) The applicant has met the requirements of GDC 5 with respect to sharing of structures, systems, and components by demonstrating that such sharing does not significantly impair the ability of the ECCS to puform its safety function including, in the event of an accident to one unit, an orderly shutdown and cooldown of the remaining units.
(4) The applicant has met the requirements of GDC 17 with regard to providing sufficient capacity and capability to assure that (a) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (b) the core is cooled and vital functions are maintained-in the event of postulated accidents.
(5) The applicant has met the requirements of GDC 27 with regard to providing combined reactivity control system capability to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained and the applicant's design meets _the guidelines of Regulatory Guide 1.47.
(6) The applicant has met the requirements of GDC 35 to provide abundant cooling for ECCS by providing redundant safety grade systems that meet the recommendations of Regulatory Guide 1.1.
(7) The applicant has met the requirements of GDC 36 with respect to the design of ECCS to permit appropriate periodic inspection of important components of the system.
(8) The applicant has met the requirements of GDC 37 with respect to designing
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the ECCS to permit testing of the operability of the system throughout the life of the plant, including the full operational sequence that brings the system into operation.
(9) The applicant has provided an analysis'of the proposed ECCS relative to the acceptance criteria of 10 CFR Part 50,.550.46, and Appendix K to demonstrate that their ECCS designs for peak cladding temperature, maximum calculated cladding oxidation, maximum hydrogen genera _ tion, coolable core geometry, and long-term cooling are in accordance with the acceptable evaluation model.
In addition, the applicant has met the requirements of the following Task Action Plan items:
6.3-11 Rev. 2 - April 1984 I:
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(1) Meeting Task Action Plan item II.B.8 of NUREG-0718 (Ref. 14) which involves description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded core accidents (cps only).
(2) Meeting Task Action Plan item II.D.1.1 of NUREG-0737 (Ref. 15) and NUREG-0718 (Ref. 14) which involves primary coolant sources outside of containment (cps and Ols).
(3) Meeting Task Action Plan item II.E.2.1 of NUREG-0660 (Ref. 16) which involves reliance on ECCS.
(4) Meeting Task Action Plan item II.K.3(10) of NUREG-0660 which involves applicant's proposal to limit anticipatory trip to high power for selected plants.
(5) Meeting Task Action Plan item II.K.3(15) of NUREG-0660 wnich involves isolation of HPCI and RCIC for BWR plants.
(6) Meeting Task Action Plan item II.K.3(18) of NUREG-0660 which invc,1ves ECCS outages for all plants.
(7) Meeting Task Action Plan item II.K.3(21) of NUREG-0660 which involves restart of LPCS and LPCI for BWR plants.
(8) Meeting Task Action Plan item II.K.3(3a) of NUREG-0660 which involves evaluation of effects of water slugs in piping caused by HPI and CFT flows in B&W plants.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees cagarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides, NUREGs, BTP RSB 6-1 and implementation of acceptance criterion subsection II.B is as follows:
(a) Operating plants and OL applicants need not comply with the provisions of this revision.
(b) CP applicants will be required to comply with the provisions of this revision.
VI.
REFERENCES 1.
10 CFR Part 50, S50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled Nuclear Power Reactors," and Appendix K to 10 CFR Part 50, "ECCS Evaluation Models," issued by the Commission December 28, 1973; Federal Register, Vol. 39, No. 3, January 4, 1974.
6.3-12 Rev. 2 - April 1984
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2.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
3.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems, and Components."
4.
10 CFR Part 50, Appendix A, General Design Criterion 17, " Electric Power Systems."
5.
10 CFR Part 50, Appendix A, General Design Criterion 27, " Combined Reactivity Control System Capability."
6.
10 CFR Part 50, Appendix A, General Design Criterion 35, " Emergency Core Cooling."
7.
10 CFR Part 50, Appendix A, General Design Criterion 36, " Inspection of Emergency Core Cooling System."
8.
10 CFR Part 50, Appendix A, General Design Criterion 37, " Testing of Emergency Core Cooling System."
9.
Regulatory Guide 1.1, " Net Position Suction Head for Emergency Core Cool-ing and Containment Heat Removal System Pumps."
10.
Regulatory Guide 1.29, " Seismic Design Classification," Revision 1.
11.
Regulatory Guide 1.47, " Bypass and Inoperable Status Indication for Nuclear Power Plant Safety Systems."
12.
Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."
l 13.
Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors."
14.
Branch Technical Position RSB 6-1, " Piping From the RWST (or BWST) and l
Containment Sump (s) to the Safety Injection Pumps," attached to SRP Section 6.3.
15.
NUREG-0718, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing Licenses."
16.
NUREG-0737, " Clarification of TMI Action Plan Requirements."
17.
10 CFR Part 50, Appendix A, General Design Criterion 4, " Environmental and Missile Design Basis."
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L) 6.3-13 Rev. 2 - April 1984
O r
BRANCH TECHNICAL POSITION RSB 6-1 PIPING FROM THE RWST (OR BWST) AND CONTAINMENT SUMP (S)
TO THE SAFETY INJECTION PUMPS A.
Background
Current PWRs utilize the refueling water storage tank (RWST) or the borated water storage tank (BWST) as the sole source of water for the safety injection pumps during the first 20 to 40 minutes of any accident that trips a safety injection signal.
Since acceptable results of safety analyses of the accidents are based on the operation of a minimum number of these pumps, interruption of this water supply for even a short period of time could result in unacceptably high fuel and cladding temperatures if the safety injection pumps fail because of cavitation or overheating.
General Design Criteria 35 requires that the emergency core cooling system have suitable redundancy in components and features and suitable interconnections to assure the system safety function can be accomplished assuming a single failure.
The principal problem appears to be a definition of single failure.
A recent draft of ANSI N658, " Single Failure Criteria for PWR Fluid Systems,"
defines an active failure as:
(a) "An active failure is a malfunction, exceeding passive failures, of a com-ponent which relies on mechanical movement to complete its intended function upon demand."
(b) " Spurious action of a powered component originating within its actuation system shall be regarded as an active failure unless specific design features or operating restrictions preclude such spurious action."
This branch position on the availability of the RWST is based on the above criteria and the recognition that water supplied from the RWST system to the ECCS system is absolutely essential in the event of a LOCA.
B.
Branch Position 1.
The single active failure criterion defined in (a) and (b) above will be applied in evaluating the design of the piping systems that connect the safety injection pumps to the RWST (BWST) and the containment sumps.
2.
The piping systems including valves, shall be designed to satisfy therequirementslistedbelowwithouttheneedtodisconnectthe power to any valve.
3.
The valves and piping between the RWST (or BWST) and the safety injection pump 3 must be arranged so that no single failure will prevent the minimum flow to the core required to satisfy 10 CFR Part 50, $50.46.
6.3-14 Rev. 1 - July 1981
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4.
ThevalvesandpipingbetweentheRWST(orBWST)andsafetyinjec-tion pumps must be arranged so that no single active failure will O
result in damage to pumps such that the minimum flow requirements for long-term core and containment cooling after a LOCA are not satisfied.
5.
The valves and piping that connect the RWST (or BWST) and the contain-ment sumps (s) to the safety injection pumps must be arranged so as not to preclude automatic switchover from the injection mode of ECCS operation to recirculation cooling from the sump.
These piping systems must be arranged so that the differential pressure between the sump and the RWST (or BWST), even if there is a single active failure, will not result in a loss of core cooling or a path that permits release of radioactive material from the containment to the environment.
C.
Implementation I
1.
cps Under Review and Future CP Reviews The proposed position will be applied to all CP reviews for which an 4
SER was not published prior.to April 16, 1975.
It is expected that all of the events of the proposed position will be applied for such reviews.
Taking this position on cps would eliminate the need for various schemes such as-locking out power to valves located in the line between the various ECCS pumps and refueling water storage tank.
2.
OLs Under Review For operating licenses that are presently under review and OLs to be reviewed in the future that are not covered by item 1, the pro-posed position'will not be completely applied.
Specifically, locking out power to valves will be permitted.
For most plants it is expected that this will be sufficient to meet the single failure criteria, However, in other plants changes to the piping and valving arrange-i ments may be required to satisfy the single failure criteria.
3.
Plants Under Construction These plants will be handled as discussed in ites C.2.
It is expected, however, that we will discuss the proposed position with each of the applicable PWR vendors.
It will be obvious to the vendors which plants now under construction may have a problem. Then a generic review may be conducted for those plants that have a severe problem.
4.
Operating Plants All of the operating plants are being evaluated as an ongoing part of the current ECC review.
The review should be conducted as dis-cussed in ites C.2 to assure that these plants meet the essential
. parts of the proposed position.
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' O 6.3-15
-Rev. 1 - July 1981
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