ML20071P365
| ML20071P365 | |
| Person / Time | |
|---|---|
| Issue date: | 06/06/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0800, NUREG-0800-FC, NUREG-800, NUREG-800-FC, SRP-NUREG-0800-FC, SRP-NUREG-800-FC, NUDOCS 8306070437 | |
| Download: ML20071P365 (87) | |
Text
NUREG-0000 (Farm:rly NUREt!-75/087) pa atc w%
p g
u.S. NUCLEAR REGULATORY COMMISSION V
) STANDARD REVIEW PLAN
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OFFICE OF NUCLEAR REACTOR REGULATION eeee Proposed Revisions to Standard Review Plan PSRP - 5.4.6, Rev. 3 PSRP - 5.4.7, Rev. 3 l
PSRP - 6.3, Rev. 2 PSRP - 9.2.1, Rev. 3 PSRP - 9.2.2, Rev. 2 PSRP - 10.3, Rev. 3 PSRP - 10.4.7, Rev. 3 These proposed revisions to the Standard Review Plan and its supporting g
value/ impact statement and associated technical documentation have not received a g
complete staff review and approval and do not represent an official NRC staff position.
These proposed revisions to the Standard Review Plan incorporate the resolution of generic issue USI A-1, " Water Hammer." Public coments are being solicited on these proposed SRP revisions, the associated value/ impact analysis and the technical support document NUREG-0927 prior to a final review and decision by the Office of Nuclear Reactor Regulation as to whether these proposed revisions should be approved. Comments should be sent to the U. S. Nuclear Regulatory Commission, Washington, DC 20555, Attention:
K. Kniel. All comments received by July 18, 1983 will be considered, and all of the associated documents and comments considered will be made publicly available prior to a decision by the Director, Office of Nuclear Reactor Regulation, on whether to implement this revision.
Copies of each of these documents are available upon written request to the Division of Technical Information and Document Control, U. S. Nuclear Regulatory l
Commission, Washington, DC 20555.
FOR COMMENT 8306070437 830606 PDR NUREG 0000 R PDR 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 (n\\
applications to construct and operato nuclear power plants. These documents are made available to the public as part of the Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are not substitutes for regulatory guides or the Commission's regu:ations and compliance with them is not required. The standard review p!an sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
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 informe-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. 201ll16.
4
NU REG-0800 (Fcrmtriy NUREG-75/087) pa atogA u.S. NUCLEAR REGULATORY COMMISSION U @e m
%v$j STANDARD REV EW PLAN s /
OFFICE OF NUCLEAR REACTOR REGULATION e.e.e Proposed Revisicn Standard Review Plan PSRP-5.4.6, Rev. 3 This proposed revision of the Standard Review Plan and its supporting value/ impact statement and associated technical documentation have not received a cenplete staff review and approval and do not represent an
[
official NRC staff position.
The proposed revision to the Standard Review
! ])
Plan incorporates the resolution of generic issue USI A-1, " Water Hamer."
V Public comments are being solicited on the proposcd SRP section and the associated value/ impact analysis and technical support document !!UREG-0927,
" Evaluation of Water Hanner Experience ir. ?!uclear Power Plants" (including any implementation schedules) prior to a final review and decision by the Office of fluclear Reactor Regulation as to whether this proposed revision should be aproved.
Corr.ents should be sent to the Secretary of the Commission, U. S. Nuclear Regulatory Commission, Washington, D. C.
- 20555, Attention:
Docketing and Service Branch.
All coments received by July 18, 1983 will be considered, and all of the associated documents and comments considered will be made publicly available prior to a decision by the Director, Office of fiuclear Reactor Regulation, en whether to inplenent this revision. Copies of each of these documents are available upon written request to the Division of Technical Information and Document Control, U. S.
Nuclear Regulatory Comission, Washington, D. C.
20555.
1 l
l l
USNRC STANDARD REVIEW PLAN Star dard 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 Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
/
N
(
)
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The
, i
/
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 inf orma-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 20555.
NU REG-0800 (Fcrmtrly NUREG 75/087) pa "s our U.S. NUCLEAR REGULATORY COMMISSION e
O' @N.dSTANDARD REV EW PLAN
' //
OFFICE OF NUCLEAR REACTOR REGULATION e
..e Proposed Revision 3 to 5.4.6 REACTOR CORE ISOLATION COOLING SYSTEM (BWR)
REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB)
Secondary - None I.
AREAS OF REVIEW The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) is a safety system which serves as a standby source of cooling water to provide a limited decay heat removal capability whenever the main feedwater system is isolated from the reactor vessel. Abnormal events which could cause such a situation to arise include an inadvertent isolation of all main steam lines, loss of condenser vacuum, pressure regulator failures, loss of feedwater, and the loss of offsite power.
Each of these transients is analyzed in Chapter 15 of the
[]
applicant's safety analysis report (SAR).
For each of these events, the high (v) pressure part of the emergency core cooling system (ECCS) provides a backup function to the RCIC system. This review of the,RCIC is performed to assure conformance with the requirments of General Design Criteria 4_, 5, 29, 33, 34 and 54 The RCIC system consists of a steam-driven turbine-pump unit and associated valves and piping capable of delivering makeup water to the reactor vessel and supplying steam to and removing condensate from the RCIC steam turbine where applicable.
Fluid removed from the reactor vessel following a shutdown from power operation is normally made up by the feedwater system, supplemented by inleakage from the control rod drive system.
If the feedwater system is inoperable, the RCIC turbine-pump unit starts automatically or is started by the operator from the control room.
The water supply for the RCIC system comes from the condensate storage tank, with a secondary supply from the suppression pool.
Rev. 3 USNRC STANDARD REVIEW PLAN Star.dard 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 Commission's policy to inform the nucl ear industry and the general public of regulatory procedures and policies. Standard review
/,m plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The s
{j J
standard review plan sections are lieyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
\\
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. 20555.
Th: r: view of the RCIC system includes the system design bases, design criteria, description, and the points noted below.
The RSB is responsible for performing the technical review of the RCIC system in the following areas:
1.
The piping and instrumentation diagrams are reviewed to determine that the system is capable of performing its intended function and of being preoperationally and operationally tested.
2.
The degree of separation of the RCIC system from the high pressure core spray (HPCS) system, or high pressure coolant injection (HPCI) system is reviewed for protection against common mode failure of redundant systems.
3.
The process flow diagram is reviewed to confirm that the RCIC system design parameters are consistent with expected pressures, temperatures and flow rates.
4.
The complete sequence of operation is reviewed to determine that the system can function as intended and that the system is capable of manual operation.
5.
The system is reviewed for compliance with the applicable requirements of NUREG-0737 (Ref. 1).
In addition, the RSB will coordinate other branch evaluations that interface with the overall review of the system as follows:
Auxiliary Systems Branch (ASB) reviews the RCIC and HPCI (or HPCS) systems for protection against common mode failures from missiles as part of its primary review responsibility for Standard Review Plan (SRP) Sections 3.5.1.1 and 3.5.1.2.
Protection against flooding of RCIC and redundant, equipment is reviewed by ASB as part of its primary review responsibility for SRP Section 3.4.1.
Protection against damage from pipe whip and jet impingement is reviewed by the Mechanical Engi-neering Branch (MEB) as part of its primary review responsibility for SRP Sections 3.6.1 and 3.6.2.
The Licensing Guidance Branch (LGB) reviews the proposed technical soecifications as part of its primary review responsibility for SRP Section 16.0.
The Procedure and Test Review Branch (PTRB) reviews the proposed preoperational and critical startup test programs as part of its primary review responsibility for SRP Section 14.2.
The MEB reviews the RCIC system to assure that it has the proper seismic and quality group classification as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2.
The RCIC is to be enclosed in a seismic Category I structure or building.
The design adequacy of this structure or building is evaluated by the Structural Engineering Branch (SEB) as part of its primary review responsibility for SRP Sections 3.3, 3.4, 3.5, 3.7, and 3.8.
The Containment Systems Branch (CSB) reviews the RCIC system, as part of its primary review responsibility for SRP Sections 6.2.4 and 6.2.6 to confirm that the design is compatible with the containment system and can be isolated.
The Instrumentation and Control Systems Branch (ICSB), as part of its primary review responsibility for SRP Section 7.4, evaluates the adequacy of controls and instrumentation of the RCIC system with regard to the required features of automatic actuation, remote sensing and indication, and remote control.
The Power Systems Branch (PSB), as part of its primary review responsibility for SRP Section 8.3, evaluates the adequacy of emergency onsite power, sufficiency of battery capacity, and the use of d-c power only.
The MEB, as part of its primary 5.4.6-2 Rev. 3
review responsibility for SRP Section 3.9.3, ensures that the design and installation of the RCIC system meet applicable codes and are adequate for its proper functioning.
The Equipment Qualification Branch (EQB) reviews RCIC
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system equipment to determine that it is seismically and environmentally V
qualified for its intended use as part of its primary review responsibility for SRP Sections 3.10 and 3.11.
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 RSB acceptance criteria are based on meeting the relevant requirements of General Design Criteria 4, 5, 29, 33, 34 and E4.
Specific criteria to meet the requirements of the above GDCs are as follows:
A_.
General Design Criteria 4, as related to dynamic effects associated with flow instabilities and loads (e.g., water hammer).
B.
General Design Criterion 5 as it relates to structures, systems and components important to safety not being shared among nuclear power units unless it can be demonstrated that sharing will not impair its ability to perform its safety function.
C.
General Design Criterion 29 at it relates to the system being designed to have an extremely high probability of performing its safety function in
/Q the event of anticipated operational occurrences.
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l D.
General Design Criterion 33 as it relates to the system capability to provide reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary so the fuel design limits are not exceeded.
E.
General Design Criterion 34 as it relates to the system design being capable of removing fission product decay heat and other residual heat from the reactor core to preclude fuel damage or reactor coolant pressure boundary overpressurization.
F.
General Design Criterion 54 as it relates to piping systems penetrating primary containment being provided with leak detection and isolation capabilities.
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:
1.
The general objective of the review is to determine that the RCIC system, in conjunction with the HPCS (or HPCI) system, the safety / relief valves, and the suppression pool cooling mode of the residual heat removal system meets the requirements of General Design Criterion 34 (Ref. 2) by providing (3
the capability for decay heat removal to allow complete shutdown of the reactor under conditions requiring its use.
It must maintain the reactor 5.4.6-3 Rev. 3
water inventory above the top of the active fuel until the reactor is depressurized sufficiently to permit operation of the low pressure cool-ing systems.
The RCIC system, in conjunction with the HPCS (or HPCI) system, the safety / relief valves, and the suppression pool cooling mode of the RHR system must be capable of removing fission product decay heat and other residual heat from the reactor core following shutdown so as to preclude fuel damage or reactor coolant pressure boundary overpressuriza-tion.
Since RCIC in conjunction with HPCS (or HPCI) is used to provide makeup inventory in some modes of residual heat removal, these systems should jointly meet the guidelines of BTP RSB 5-1, attached to SRP Section 5.4.7.
2.
The RCIC system is also used to supply reactor coolant makeup for small leaks.
Accordingly, the systems must meet the requirements of General Design Criterion 33 (Ref. 4) in this regard.
3.
Historically, credit has been taken for RCIC system capability to mitigate the consequences of certain abnormal events; however, since the caoling function is redundant to the HPCI or HPCS system, the RCIC system itself is not required to meet the single failure criterion, but in conjunction with HPCS (or HPCI) must satisfy the single failure criterion in this regard.
In addition, the RCIC system is to perform its function without the availability of any a-c power per the requirements of General Design Criterion 34 (Ref. 2), and in conjunction with HPCS (or HPCI) must be designed to assure an extremely high probability of accomplishing its safety function as required by General Design Criterion 29 (Ref. 6).
4.
As a system which must respond to certain abnormal events, the RCIC system must be designed to seismic Category I standards (discussed in SRP Section 3.2.1) and must not be shared among nuclear power units except as permitted by General Design Criterion 5 (Ref. 7).
5.
The RCIC and HPCS (or HPCI) systems must be protected against natural phenomena, external or internal missiles, pipe whip, and jet impingement forces so that such events cannot fail both systems simultaneously.
Acceptance criteria for these are discussed in SRP Sections 3.3.1 through 3.6.2.
Acceptance criteria for RCIC instrumentation are described in SRP Section 7.4.
6.
The RCIC system must meet the requirements of General Design Criterion 54 (Ref. 8) with regard to leak detection and isolation provisions for lines passing through the primary containment.
Other containment isolation criteria for RCIC are described in SRP Sections 6.2.4 and 6.2.6.
7.
The RCIC system must meet the recommendations of Task Action Plan items II.K.1.22, II.K.3.13, II.K.3.15, II.K.3.22, II.K.3.24, and III.D.I.1 of NUREG-0737 (Ref. 1) and NUREG-0718 (Ref. 11) with regard to actions needed f or operation, system initiation setpoint and automatic restart capability, break detection provisions, automatic suction switchover to the suppression pool, adequacy of space cooling, and leakage minimization, respectively.
8.
If the RCIC system is used to control or mitigate the consequences of an accident, either by itself or as a backup to another system, it must meet the requirements of an engineered safety feature.
The RCIC system must 5.4.6-4 Rev. 3
(~')T meet the guidelines of Regulatory Guide 1.1 (Ref. 9) regarding net
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positive suction head.
9.
In order to meet the requirements of General Design criterion 4 (Ref. 12) design features and operating procedures, designed to prevent damaging water hammer due-to such mechanisms as voided discharge lines and entrainment in steam lines, shall be provided._
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 forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II.
For the operating license (0L) review, the procedures are used 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.
1.
Using the RCIC operating requirements specified in SAR Section 5.4.6 and Chapter 15, the reviewer confirms that the RCIC system can maintain coolant inventory in the reactor vessel to keep the core covered and assure cladding integrity.
Tnis determination is based on engineer)ng
/n) judgment and independent calculations (where deemed necessary), using V
information as specified in steps 2 and 3 below.
The reviewer consults with the CPB to assure that the decay heat loads used in the RCIC analyses are applicable and suitably conservative.
2.
Using the description given in Section 5.4.6 of the SAR, including component lists and performance specifications, the reviewer determines that the RCIC system piping and instrumentation are such as to allow the system to operate as intended.
This is accomplished by reviewing the piping and instrumentation diagrams to confirm that piping arrangements permit the required flow paths to be achieved and that sufficient process sensors are available to measure and transmit required information.
3.
Using the comparison tables of SAR Section 1.3, the RCIC system is compared to designs and capacities of such systems in similar plants to see that there are no unexplained departures from previously reviewed plants.
Where possible, comparisons should be made with actual performance data from similar systems in operating plants.
4.
The reviewer checks the piping and instrumentation diagrams and equipment layout drawings for the RCIC and HPCS (or HPCI) systems to see that the systems are physically separated and can function independently.
5.
The reviewer examines the system design in SAR Section 5.4.6 to verif3 that the capability for automatic switchover of suction from the conden-m
[V]
sate storage tank to the suppression pool has been provided per the requirements of item II.K.3.22 of NUREGs-0737 and 0718 (Ref. 1 and 11).
l The reviewer also judges whether adequate control and monitoring infor-mation is available to allow the operator to actuate the system manually 5.4.6-5 Rev. 3 i
or to realign the RCIC system manually within the time allowed (i.e.,
change the RCIC system suction from the condensate storage tank to the suppression pool, or to the steam condensing mode of the residual heat removal system).
6.
The reviewer contacts ICSB to confirm that automatic actuation and remote-muual valve controls are capable of performing the functions required and that sensor and monitoring provisions are adequate.
The instrumentation and controls of the RCIC system, in conjunction with the HPCS (or HFCI) system are to have sufficient redundancy to satisfy the single failure criterion.
7.
The reviewer contacts PSB to ascertain that the RCIC system operation is not dependent on a-c power sources, and that there is sufficient battery capability to permit operation of the RCIC for a period of two hours without the availability of a-c power.
8.
The reviewer checks with MEB to verify that essential RCIC system components are designated seismic Category I.
9.
The reviewer contacts PTRB to verify that the applicant's proposed preoperational and initial startup test programs are in compliance with Regulatory Guide 1.68 (Ref. 10).
At the OL stage, the reviewer confirms" with PTRB that sufficient information is provided by the applicant to identify the test objectives, methods of testing, and test acceptance criteria (see par. C.2.b of Regulatory Guide 1.68).
PTRB also verifies that the proposed test programs will provide reasonable assurance that the RCIC system will perform its safety function.
As an alternative to this detailed evaluation, the reviewer may compare the RCIC systen design to that of previously reviewed plants.
If the design is essentially identical and if the proposed test programs are essentially the same, the reviewer may conclude that the proposed test programs are adequate for the RCIC system.
If the RCIC system differs significantly from that of previously reviewed designs, the impact of the proposed changes on the required preoperational and initial startup testing programs are reviewed at the CP stage.
This effort should particularly evaluate the need for any special design features required to perform acceptable test programs.
10.
The LGB is contacted in regard to the proposed plant technical specifications to:
Confirm the suitability of the limiting conditions of operation, a.
including the proposed time limits and reactor operating restrictions for period > when system equipment is inoperable due to repairs and maintenance, b.
Verify that the frequency and scope of periodic surveillance testing is adequate.
11.
The reviewer confirms that the RCIC is housed in a structure whose design and design criteria have been reviewed by other branches (i.e., ASB, SEB, MEB) to assure that it provides adequate protection against wind, tornadoes, floods, and missiles, as appropriate.
O 5.4.6-6 Rev. 3
12.
Upon request from the primary reviewer, other branches will provide input
(
for the areas of review stated in subsection I.
The primary reviewer xs obtains and uses such input as required to assure that this review procedure is complete.
- 13. The reviewer checks the automatic and nanual actions necessary for oroper functioning of the RCIC system (in conjunction with the HPCS or HPCI, the safety relief valves and the suppression pool cooling mode of RHR) for completeness and practicality when used for residual heat removal per the requirements of item II.K.1.22 of NUREGs-0737 and 0178 (Ref. I and 11).
14 The reviewer checks the RCIC system break detection provisions to see that the systen is protected against spurious trip signals per the requirements of item II.K.3.15 of NUREGs-0737 and 0718 (Ref. I and 11).
- 15. The reviewer confirms, in conjunction with ASB as necessary, that the RCIC system can withstand a loss of offsite power to its support systems, incluaing space coolers, for at least two hours per the requirenent of item II.K.3.24 of NURECs-0737 and 0718 (Ref. 1 and 11).
- 16. The reviewer confirms per the requirements of item II.K.3.13 of NUPEGs-0737 and 0178 (Ref. 1 and 11) that analyses have been provided or referenced to determine the need to separate the RCIC and the HPCS (or HPCI) initia-tion levels. Based on these study results, the reviewer checks the RCIC dasign for appropriate provisions.
In addition, the reviewer checks to see that automatic restart capability is provided for RCIC.
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- 17. The reviewer checks (by calculation as necessary) to see that adequate net positive suction head is available for RCIC suction from all potential sources (i.e., condensate storage tank, suppression pool, or RUR steam condensing c.cde discharge).
- 18. The reviewer examines the RCIC in conjunction with the HPCS or HPCI, the safety / relief valves and the suppression pool cooling mode of RHR for conformance to the recommendations of BTP RSB 5-1 to SRP Section 5.4.7 regarding residual heat removal.
19.
The RCIC,,systera is reviewed to,, evaluate the adequacy of design features _
l that have be,en provided to prev,ent damaging water (s,te,am) har,rier due t_o such mechariisms as voided discharge lines, wat,er, entrainment and _steem hiubFle collapse,,If the normal water, supply is above the discharge'llnes, voideo lines are p'e'evented by proper vent location and filling and venting l
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procedures. Howev r, if the, n.ormal aYignment of,_the suct'ic'n valves i_s to the suppression pool, back leakage through the pump discharge check vaWs WITT7esult in line voidaie Proper vent loEa'ticr end filling,aEd~ venting procedures are still r:eeded.
In addition, a special keep-full system with~
appropriate"aTarms is needed to sipply water to the discharge lines at
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sufficientTy~ high pressure to 'p'revent voiding.
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The RCIC sy, sten uses a steam-driven turbine.
Typi thestpap,supplylineinclud,e[s'(i)drainpots,(b),calcesignfeaturesfor slopedlines,and[(d
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l l p) limitations on open,ing ar.d closing sequences and-seal-ins f,or manual opera-(
tion of the isolatior valves.
The turbine exhaust line features include V
sloped 1ines"a'n'd vacuum breakers.
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5.4.6-7 Rev. 3
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 evaluatien report:
The reactor core isolation cooling (RCIC) system includes the piping, valves, Dun'ps, turbines, instrunentation, and controls usec' to maintain water inventory in the reactor vessel whenever it is isolated from the main feedwater system.
Certain engineered safety features (HPCS or HPCI) provide a redundant backup for this function.
The scope of review of the RCIC system for the plant included piping and instrumentation diagrams, equipment layout Trawings, and functional specifications for essential components. The review has included the applicant's proposed design criteria and design bases for the RCIC systen, his analysis of the adequacy of the criteria and bases, and the conformance of the design to these criteria and bases.
The staff concludes that the reactor core isolation cooling system design is acceptable and meets the reouirements of General Design Criteria 4 5, 29, 33, 2
34 and 54. This conclusion is based cn the followire:
1.
The applicant has met the requirements of (cite Reg.) with respect to (state limits of review) by:
(Use one or more of the foilowing as applicable) a.
meeting the regulatory position in Pec;ulatory Guide b.
providing and meeting an alternative n.ethod to the regulatory position in Regulatory Guide
, that the staff has reviewed and fcurd to be acceptable, c.
r.iecting the regulatory position in BTP d.
The calculational methcd used by the applicant for (state) has been previcusly reviewed by the staff and found acceptable; the staff has revieweo the key parameters in this case and found them to be suitably ccnservative.
e.
The applicant has ret the requirements of (industry stenaard - number and title' that has been reviewed by ibe staff and detemined to be appropriate for this application.
2.
Repeat the above discussion for each CDC listed.
In addition, conforriance with General Desier Criterion 55, 56, and 57 regarding centairnent isolation is discussed in Section 6.2 of this report. Conformance with General Design Criterien 2 and 4 for pretection against natural phenonena, environn. ental hazards and potential missiles is discussed in Secticns 3.3 through 3.6 of this report.
The RCIC and HPCS (or HPCI) systems, in conjunction with the safety / relief valves and the suppression pool cooling mcde of the resicual heat renoval system, have been found capable of removing core decay heat following feed-water syster isolation and reactor shutdown so that sufficient coolant 5.4.6-8 Rev. 3
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ensure cladding integrity. This capability has been tf,ound to be available" even with a loss of offsite. power and with a single active failure.
V.
1 IMPLEMENTATION 5 l.
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y sThe following is intended to provide guidance to ap'plicants and licensees regarding the NRC staff's plans'for using this SRP section.
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Except in those cases in which the applicant pEcposes an acceptable alterra'tive method for complying with specified portions of the Commission's regulations, the method described herein.will be tised by the staff in its evaluation of conformance With Commission regulations.
Iniplenientation schedules for conformance to parts of the r:ethod discussed
.herein are contained in the referenced regulatory guides ana WUREGs.
t
_ Implementation of the Proposed Item II.A will, apply to CP applicants.
1.
VI. REFERENCES r
1.
NUREG-0737, "Clarifiatior of THI Action Plan Requirements," Ucvember 1980.
P.
10 CFR Part 50, Appendix A. General Design Criterion 34, "Resicual Heat Remova l.'" ~
[m) 3.
Branch Technical Position RSB 5-l',' " Design Requirements of the Residual
.V Heat Removal System," attached to SRP S'ection 5.4.7.
4 4.
10tCFR-Part'50, Appendix A, General Desigr. Criteriop 33, " Reactor Coolant Makeup."
s r
3 i
L 5.
Regulatory Guide 1.53, " Single Failure Criteri_or.."
l 6.
10 CFR Part 50,. Appendix A, General Design Criterion 29, '.' Protection Agairst Anticipated Operational Cccurrences."
i 7.
10 f.FR Part f0, Appendix A, General Design Criterion 5, "Sharin'g of
.[
Structures, Systeos, 'and Components."
4 8.
10 CFR Part-50, Appendix A', Genera.1 Design Criterion 54, " Piping Systems Peretrating Contai ment.",
9.
Regulatory Ciuide 1.1, " Net Posidve Suc' tion Head for Fraergency Core Cool.ing and Containrcent Heat Removal Systems."
- 10. Regulatory Guide 1.68, "I'nitial Test Prograris for Water-Cooled Reactor Power Plants "
- 11. NUREG20718g. " Licensing R64uirements for Pending Applications for Constructiun{ermitssndManufacturingLicense."
12_.
10 CFR Part 50. Appendix A, General.0esign Criterion'4, "Environmertal
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i ac: Missile besign Bases"._
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5.4.6-9 Rev. 3 '
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$;%jf&j STANDARD REV EW PLAN af OFFICE OF NUCLEAR REACTOR REGULATION
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o' Proposed Perisiw StendardiReview P1ar, PSFP-5.4.7, Rev. 3 I
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Inis proposed revisioq of the Standard; Review P3aa and its supporting value/ impact statetent and assqciated technical'documentction have net received a conglete staff review ar.d approval and aq ngt ippresent an
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official NRC' staff' position.
The proposed revision' to the. Mandard Review C/
Plan incorpatates the resolv: ion of generic i' sue USI A-1, "Uater Hamer."
s Public coments are being soliciteo en the proposed SRP sectidia and the associated value/inpact analysis and technical support document NUREG-0977,.
" Evaluation of Water Hamider Experience in Nuclesr Power Plant %" (including.
.any implementation schedules) prior.to e fina. review and. decision by the Office of Nuclear Reactc.r Regulatien as to whethdr this progsed revisi'tn should be aproved. Comments should be sent to the Secreta,ry of the Commission, U. S. Nuclear Regulatory Commission, Washington, D. C.
- 20555, Attention:
Dccketing and Service Branch. All comments ret >dved by July 18, 1983 will be considered, ar:d z.ll of the associited documents and cornents considere('will be made publicly available prior to a de cision by the Director, Office of Nuclear Reactor Regulation, on whether to inplenent this revision.
Copies of each of these documents are availcble upon written i
recoest to the Division of Technical Infornation and Docunent Contre.0, l'.
S.
Uuclear Regniatary Ccmission, Washington, D. C.
20555.
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Star.dard review plans are prepared for the guidance of tne Tfice of Nuclear Reactor Replation staff responsible for the review of pS applications to construct and operate nuclear power riante. These (ocuments h e made avm3acle to the public as part OL the f
j Commission's policy to infrarm the nuclear industry and the peneral pt.blic of regula cry proemiures and policies. Standard review I
f plans are not substitutes for regulatory guides or the Comraission's regulations arrt comNian. e with them is not required. The i (,/
standard review plan sectiors are keyed to the Standard Format and Ccotent of Safety Anatois Reports for Nuclear Power Plants.
I Not all sections of the Standard Format have a corresponding review p'en.
1 1
Published standard revie w plans will be revised periodically. as appropriats, to accorrimo jate comments and to reflect new ir4orme tion and experience.
Comments and suggestions for improvement will be conside.mt and shot.ld be sent tu 1.he u.S. Nuclear Regulatory commission, office of Nuclear Reactor Reputation, Washington, D.C. 20555.
N U REG-0800 (Formtrly NUREG-75/087)
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(/paatog[s U.S. NUCLEAR REGULATORY COMMISSION p
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N % p/ OFFICE OF NUCLEAR REACTOR RE e..e.
Proposed Revision 3 to 5.4.7 RESIDUAL HEAT REMOVAL (RHR) SYSTEM REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB)
Secondary - None I.
AREAS OF REVIEW The residual heat removal (RHR) system is used in conjunction with the main steam
^
and feedwater systems (main condenser), or the reactor core isolation cooling (RCIC) system in conjunction with the safety / relief valves in a boiling water reactor (BWR), or auxiliary feedwater sytem in conjunction with the atmospheric dump valves in a pressurized water reactor (PWR) to cool down the reactor coolant j
system following shutdown.
Parts of the RHR system also act to provide low pres-sure emergency core cooling and are reviewed as described in SRP Section 6.3.
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Some parts of the RHR system also provide containment heat removal capability and l (j are reviewed as described in SRP Section 6.2.2.
The review by RSB is to ensure that the design of the RHR system is in conformance with General Design Criteria 2 2
l 4, 5, 19, and 34.
l l
Both PWRs and BWRs have RHR systems which provide long-term cooling once the j
reactor coolant temperature has been decreased by the main condenser, RCIC, or auxiliary feedwater systems.
In both types of plants, the RHR is typically a low i
l pressure system which takes over the shutdown cooling function when the reactor coolant system (RCS) temperature is reduced to about 300 F.
Although the RHR system function is similar for the two types of plants, the system design are different.
The RHR system in PWRs takes water from the RCS hot legs, cools it, and pumps it back to the cold legs or core flooding tank nozzles.
The suction and discharge lines for the RHR pumps have appropriate valving to assure that the low pressure RHR system is always isolated from the RCS when the reactor coolant pressure is greater than the RHR system design pressure.
The heat Rev. 3 USNRC STANDARD REVIEW PLAN Star dard 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
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Commission's policy to inform the nuclear industry an't 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
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standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
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 inf orma-tion and experience.
Comments and suggestions for improvement will be considered and should be sent to the U.S Nuclear Regulatory Commission.
Of fice of Nuclear Reactor Regulation. Washington, D.C. 2055s.
removed in the heat exchang:rs is transported to tha ultimate h:at sink by the component cooling water or service water system.
In PWRs, the RHR system is also used to fill, drain, and remove heat from the refueling canal during refueling operations, to circulate coolant through the core during plant startup prior to RCS pump operation, and in some to provide an auxiliary pressurizer spray.
The RHR system in BWRs is typically composed of four subsystems.
The contain-ment heat removal and low pressure emergency core cooling subsystems are discussed in SRP Sections 6.2.2 and 6.3.
The shutdown cooling and steam condensing (via RCIC) subsystems are covered by this SRP section.
These subsystems make use of the same hardware, consisting of pumps, piping, heat exchangers, valves, monitors, and controls.
In the shutdown cooling mode, the BWR RHR system can also be used to supplement spent fuel pool cooling.
As in the PWR, the low pressure RHR piping is protected from high RCS pressure by isolation valves.
The steam condensing mode of RCIC operation in BWRs (when included in the plant design) provides an alternative to the main condenser or normal RCIC mode of operation during the initial cooldown.
Steam from the reactor is transferred to the RHR heat exchangers where it is condensed.
The condensate is piped to the suction side of the RCIC pump.
The RCIC pump returns the condensate to the reactor vessel. The heat removed in the heat exchangers is transported to the ultimate heat sink by the service water system.
Other means of removing decay heat in the event that the RHR system is inoper-able have been proposed for some BWRs.
These approaches use some of the piping that is used for the steam condensing mode of RCIC.
These approaches are also covered by this SRP section.
The reactor coolant temperatures and pressure must be decreased before the low pressure RHR system can be placed in operation; therefore, the review of the decay heat removal function must consider all conditions from shutdown at normal reactor operating pressure and temperature to the cold depressurized condition.
RSB reviews the requirements for reliability and capability of removing decay heat identified in NUREG-0660 (II.E.3.2 and II.E.3.3), NUREG-0718 (II.B.7), and NUREG-0737 (III.D.1.1).
With respect to the staff review for compliance with Branch Technical Position RSB 5-1 (Ref. 5), the Auxiliary Systems Branch (ASB),
Chemical Engineering Branch (CMEB), and RSB effort is divided as follows:
1.
For BWRs, the RSB reviews the processes and systems used in the cooldown of the reactor for the entire spectrum of potential reactor coolant system pressures and temperatures during decay heat removal.
2.
For PWRs, the RSB reviews the approach used to meet the functional require-ments of BTP RSB 5-1 with respect to cooldown to the conditions permitting operation of the RHR system.
Since an alternate approach to that normally used for cooldown may be specified, the reviewers identify all components and systems used.
The CMEB has primary review responsibility for the review of the pertinent portions of the CVCS (SRP Section 9.3.4).
The ASB, as part of its primary review responsibility for SRP Sections 10.3 and 10.4.9 reviews the atmospheric dump valves and the source for auxiliary feedwater, respectively, for conformance to BTP RSB 5-1.
The RSB reviews the pressurizer relief valve and ECCS, if used.
In addition, the R$8 reviews the tests and supporting analysis concerning mixing of borated water and cooldown under natural circulation as required in BTP RSB 5-1.
5.4.7-2 Rev. 3
Illi.
3.
For both PWRs and BWRs, the ASB reviews the component cooling or service
~
water systems that transfer decay heat from the RHR system to the ulti-mate heat sink as part of its primary review responsibility for SRP Sections 9.2.1 and 9.2.2.
4.
The RSB reviews the design and operating characteristics of the RHR system with respect to its shutdown and long-term cooling function.
Where the RHR system interfaces with other systems (e.g., RCIC system, component cooling water system) the effect of these systems on the RHR system is reviewed.
Overpressure protection provided by the valving between the RCS and RHR system is also reviewed.
In addition, the Reactor Systems Branch will coordinate evaluations of other branches that interface with the overall review of the RHR system as follows:
The Containment Systems Branch verifies that portions of the RHR system pene-trating the containment barrier are designed with acceptable isolation features to maintain containment integrity for all operating conditions including acci-dents as part of its primary review responsibility for SRP Section 6.2.4; The Structural Engineering Branch (SEB) determines the acceptability of the design analysis, procedures and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural phenomena such as safe shutdown earthquake (SSE), the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 thru 3.7.4, 3.8.4 and 3.8.5.
The Materials Engineering Branch (MTEB) verifies that inser-vice inspection requirements are met for system components as part of its primary review responsibility for SRP Section 6.6 and, upon request, verifies the compatibility of the materials of construction with service conditions as
_s l
part of its primary review responsibility for SRP Section 6.1.
The Mechanical Engineering Branch (MEB) determines that the components, piping and structures are designed and tested in accordance with applicable codes and standards as part of its primary review responsibility for SRP Sections 3.9.1 through 3.9.3.
The MEB also determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2 The effects of pipe breaks inside and outside of containment, such as pipe whip and jet impingement, are reviewed by MEB and ASB as part of their primary review responsibilities for SRP Sections 3.6.2 and 3.6.1, respectively.
The MEB also reviews adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6.
The Procedure and Test Review Branch (PTRB) reviews the proposed preoperational and startup test programs to confirm that they are in conformance with the intent of Regulatory Guide 1.68 as part of its primary review responsibility for SRP Section 14.2.
The PTRB also has primary review responsibility for Task Action Plan items II.K.1 (C.1.10) of NUREG-0737 (0Ls only) and I.C.6 of NUREG-0718 (cps only) regarding procedures to ensure that system operability status is known.
The ASB reviews flood protection as part of its primary review responsibility for SRP Section 3.4.1.
The ASB identifies the structures systems and components to be protected against externally generated missiles and reviews the adequacy of protection against such missiles as part of its primary review responsbility for SRP Section 3.5.1.4 and 3.5.2.
The ASB also reviews protection against internally generated missiles both inside and outside of containment as part of its primary review responsibility for SRP Sections 3.5.1.1 and 3.5.1.2.
o 5.4.7-3 Rev. 3
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Th3 Pow:r Systems Br:nch (PSB) id:ntifies th safety-related electrical loads and determines that power systems supplying motive or control power for the RHR system meet acceptable criteria and will perform these intended functions during all plant operating and accident conditions as part of its primary review responsibility for SRP Sections 8.1, 8.2, 8.3.1, and 8.3.2.
The Instru-mentation and Control Systems Branch (ICSB), as part of its primary review responsibility for SRP Sections 7.1 and 7.4 reviews the instrumentation and control systems for the RHR system to determine that it will perform its design function as required and conform to all applicable acceptance criteria.
The ICSB also reviews the provisions taken to meet GDC 19 with respect to equipment outside of the control room for hot and cold shutdown.
The Radio-logical Assessment Branch (RAB) has primary review responsibility for SRP Section 12.1 through 12.5 including Task Action Plan items II.B.2 of NUREG-0737 and NUREG-0718 which involve a radiation and shielding design review and corrective actions taken to ensure adequate access to vital areas and protec-tion of safety equipment (cps and OLs).
The review for Fire Protection, Technical Specifications, and Quality Asurance are coordinated and performed by the CMEB, Licensing Guidance Branch (LGB) and Quality Assurance Branch (QAB) as part of their primary review responsibility for SRP Sections 9.5.1, 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 Reactor Systems Branch acceptance criteria are based on meeting the require-ments of the following regulations:
A.
General Design Criterion 2 with respect to the seismic design of systems, structures and components whose failure could cause an unacceptable reduction in the capability of the residual heat removal system.
Accept-ability is based on meeting position C-2 of Regulatory Guide 1.29 or its equivalent.
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 which requires that any sharing among nuclear power units of structures, systems and components important to safety will not significantly impair their safety function.
D.
General Design Criterion 19 with respect to control room requirements for normal operations and shutdcwn, and; E.
General Design Criterion 34 which specifies requirements for a residual heat removal system.
Specific criteria necessary to meet the requirements of General Design Criteria 2, S 5, 19, and 34 are as follows:
1.
The system or systems are to satisfy the functional, isolation, pressure relief, pump protection and test requirements specified in Branch Technical Position RSB 5-1.
5.4.7-4 Rev. 3
2.
In order to meet the requirements of General Design Criterion 4 (Ref 11),
design features and operating procedures shall be provided to prevent p
damaging water hammer due to such mechanisms as voided discharge lines, I
water entrainment in steam lines and steam bubble collapse.
gd 3.
Iaterfaces between the RHR system and RCIC and component or service water systems should be designed so that operation of one does not interfere with, and provides proper support (where required)-for, the other.
In relation to these and other shared systems (e.g., emergency core cooling and containment heat removal systems), the RHR system must conform to GDC 5.
4.
The requirements for the reliability and capability of removing decay heat under the following Task Action Plan items must also be satisfied:
a.
Meeting Task Action Plan item II.E.3.2 of NUREG-0660 which involves systems reliability.
NRR will conduct a generic study to assess the capability and reliability of shutdown heat removal systems under various transients and degraded plant conditions including complete loss of all feedwater.
Deterministic and probabilistic methods will be used to identify design weaknesses and possible system modifica-tions that could be made to improve the capability and reliability of these systems under all shutdown conditions.
(cps and OLs).
Specific requirements will be based on the results of this study.
b.
Meeting Task Action Plan item II.E.3.3 of NUREG-0660 which involves a coordinated study of shutdown heat removal requirements.
An effort to evaluate shutdown heat removal requirements in a comprehen-
)
sive manner is required, thereby permitting a judgment of adequacy Q/
in terms of overall system requirements.
As part of this project, NRR will conduct a study to assess the desirability of and possible requirement for a diverse heat-removal path, such as feed and bleed, particularly if all secondary-side cooling is unavailable.
The NRC staff will work with the recently established ACRS Ad Hoc Subcommit-tee on this matter to develop a mutually acceptable overall study program.
(cps and OLs).
Specific requirements will be based on the results of this study.
c.
Meeting Task Action Plan item II.B.8 of NUREG-0718 (Ref. 7) 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) d.
Meeting Action Plan item III.D.1.1 of NUREG-0737 (Ref. 8) and l
NUREG-0718 (Ref. 7) which involves primary coolant sources outside of containment (cps and OLs).
5.
When the RHR system is used to control or mitigate the consequences of an accident, it must meet the design requirements of an engineered safety feature system.
This includes meeting the guidelines of Regulatory Guide 1.1 regarding net positive suction head.
III. REVIEW PROCEDURES (O/
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 5.4.7-5 Rev. 3 1
forth in the Preliminary Safety Analysis Report meet the acceptance criteria given in subsection II.
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.
As noted in subsections I and II, the RSB review for PWRs is limited to the low pressure - low temperature RHR system.
For BWRs, the review is to include all of the systems used to transfer residual heat from the reactor over the entire range of potential reactor coolant temperatures and pressures.
The following steps are to be applied by the reviewer for the appropriate systems, depending on whether a PWR or BWR is being reviewed.
These steps should be adapted to CP or OL reviews as appropriate.
1.
Using the description given in the applicant's Safety Analysis Report (SAR), including component lists and performance specifications, the reviewer determines that the system (s) piping and instrumentation are such to allow the system (s) to operate as intended, with or without offsite power and given any single active component failure.
This is accomplished by reviewing the piping and instrumentation diagrams (P& ids) to confirm that piping arrangements permit the required flow paths to be achieved and that sufficient process sensors are available to measure and transmit required information.
A failure modes and effects analysis (or similar system safety analysis) provided in the SAR is used to determine conformance to the single failure criterion.
2.
Using the comparison tables of SAR Section 1.3, the RHR system is compared to designs and capacities of such systems in similar plants to see that there are no unexplained departures from previously reviewed plants.
Where possible, comparisons should be made with actual performance data from similar systems in operating plants.
3.
From the system description and P& ids, the reviewer determines that the isolation requirements of Branch Technical Position RSB 5-1 (Ref. 5) are satisfied.
4.
The reviewer determines that the RHR system design has provisions to prevent damage to the RHR pumps in accordance with Branch Technical Position RSB 5-1 (Ref. 5).
The reviewer checks the isolation valves in l
the suction line for potential closure, NPSH requirements, pump runout, and potential loss of miniflow line during pump testing.
If operator action is required to protect the pumps, the reviewer evaluates the instrumentation required to alert the operator and the adequacy of the time frame for operator action.
L The RHR system 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 RHR systems of BWRS, the low elevation of the suppression pool will result in line voidage 5.4.7-6 Rev. 3
because of 'uack leakage through pump discharge check valves and leaking valves in the full flow test line.
Proper vent location and filling and y' 's venting procedures are still needed.
In addition, a special keep-full
(' ')
system with appropriate alarms is needed to supply water to the discharge lines at sufficiently high pressure to prevent voiding.
For RHR systems of BWRs which use the steam condensing mode of operation, the evaluation should include consideration of water hammer due to (a) water entrainment in the steam supply line during startup, (b) formation of steam bubbles in the RHR system heat exchangers resulting from leakage past valves in the steam supply line, and (c) water entrainment in the discharge line of the pressure relief valve used to prevent overpressurization of the system during operation in the steam condensing mode.
6.
Using the system process diagrams, P& ids, failure modes and effects analysis, and component performance specifications, the reviewer deter-mines that the system (s) has the capacity to bring the reactor to conditions permitting operation of the RHR system in a reasonable period of time, assuming a single failure of an active component with only either onsite or offsite electric power available.
For the purposes of this review, 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is considered a reasonable time period.
The ASB is responsible for the review of the initial cooldown phase for PWRs.
Therefore, this review effort is to be coordinated with that branch.
For the purposes of the review of both PWRs and BWRs, only the operation of safety grade equipment is to be assumed.
7.
The cooldown function is to be reviewed to determine if it can be per-formed from the control room assuming a single failure of an active
[
component, with only either onsite or offsite electric power available.
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Any operation required outside of the control room is to be justified by the applicant.
Like Item 5, the initial cooldown for PWRs is to be reviewed by ASB.
8.
By reviewing the system description and the P& ids, the reviewer confirms the RHR system satisfies the pressure relief requirements of Branch Technical Position RSB 5-1 (Ref. 5).
9.
By reviewing the piping arrangement and system description of the RHR system, the reviewer confirms that the RHR system meets the requirements of GDC 5 (Ref. 2) concerning shared systems.
10.
The RSB reviewer contacts the ASB reviewer in conjunction with his review of the RHR system heat sink and refueling system interaction to inter-change information and assure that the reviews are consistent with regard to the interfacing parameters.
For example, the ASB review determines the maximum service or component cooling water temperature.
The RSB reviewer then reviews the RHR system description tn determine that this maximum temperature has been allowed for in the RHR system design.
11.
The RSB reviewer contacts his counterpart in the ICSB to obtain any needed information from their review.
Specifically, ICSB confirms that automatic actuation and remote-manual valve controls are capable of performing the functions required, and that sensor and monitoring pro-
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visions are adequate. The instrumentation and controls of the RHR system are to have sufficient redundancy to satisfy the single failure criterion.
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5.4.7-7 Rev. 3
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12.
The RSB reviewer contacts his counterpart in CSB so that the information needed concerning their reviews will be interchanged.
13.
The RSB reviewer contacts his counterpart in PTRB to discuss any special test requirements and to confirm that the proposed preoperational test program for the RHR system is in conformance with the intent of Regulatory Guide 1.68.
14.
The proposed plant technical specifications are reviewed to:
a.
Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restric-tions for periods when system equipment is inoperable due to repairs and maintenance.
b.
Verify that the frequency and scope of periodic surveillance testing is adequate.
15.
The reviewer contacts the SEB reviewer to confirm that the systems employed to remove residual heat are housed in a structure whose design and design criteria provide adequate protection against wind, tornadoes, floods, and missiles, as appropriate.
16.
For PWRs, the reviewer confirms that the auxiliary feedwater supply satisfies the requirements of Branch Technical Position RSB 5-1.
17.
The RSB reviewer provides information to other branches in those areas where the RSB has a review responsibility that is not explicitly covered in steps 1-15 above.
These additional areas of review responsibility include:
a.
Identification of engineered safety features (ESF) and safe shutdown electrical loads, and verification that the minimum time intervals for the connection of th ESF to the standby power systems are satisfactory.
b.
Identification of vital auxiliary systems associated with the RHR system and determination of cooling load functional requirements and minimum time intervals.
c.
Identification of essential components associated with the main steam supply and the auxiliary feedwater system that are required to operate during and following shutdown.
l 18.
The RSB review evaluates the applicant responses to the following Task l
Action Plan items:
a.
II.E.3.2 of NUREG-0660 (cps and OLs) l b.
II.E.3.3 of NUREG-0660 (cps and OLs) c.
II.B.8 of NUREG-0718 (cps only) d.
II.D.l.1 of NUREG-0737 and NUREG-0718 (cps and OLs) l 5.4.7-8 Rev. 3
e IV.
EVALUATION FINDINGS n
The reviewer verifies that the SAR contains sufficient information and his
)
review supports the following kinds of statements and conclusions, which
'L/
should be included in the staff's Safety Evaluation:
For PWRs The residual heat removal function is accomplished in two phases:
the initial cooldown phase and the residual heat removal (RHR system) operation phase.
In the event of loss of offsite power, the initial phase of cooldown is accomplished by use of the auxiliary feedwater system and the atmospheric dump valves.
This equipment is used to reduce the reactor coolant system temperature and pressure to values that permit operation of the RHR system.
The review of the initial cooldown phase is discussed in Section of the SER.
The review of the RHR system operational phase is discussed below.
The residual heat removal (RHR) system removes core decay heat and provides long-term core cooling following the initial phase of reactor cooldown.
The scope of review of the RHR system for the plant included piping and instrumentation diagrams, equipment layout drawings, failure modes and effects analysis, and design performance specifications for essential components.
The review has included the applicant's proposed design criteria and design bases for the RHR system and his analysis of the adequacy of those criteria and bases and the conformance of the design to these criteria and bases.
/7 The staff concludes that the design of the Residual Heat Removal
(
)
System is acceptable and meets the requirements of General Design Criteria 2, 4 5, 19, and 34.
This conclusion is based on the 2
following:
(1) The applicant has met the General Design Criterion 2 with respect to position C-2 of Regulatory Guide 1.29 concerning the seismic design of systems, structures and components whose failure could cause an unacceptable reduction in the capability of the residual heat removal system.
l (2) The applicant has met the General Design Criterion 4 with respect to dynamic effects associated flow instabilities and loads (e.g., water hammer).
(3) The applicant has met the requirements of General Design l
Criterion 5 with respect to sharing of structure, systems and l
components by demonstrating that such sharing does not signifi-cantly impair the ability of the Residual Heat Removal System to perform it safety function including in the event of an accident to one unit, an orderly shutdown and cooldown of the remaining units.
l l
(4) The applicant has met General Design Criterion 19 with respect to the main control room requirements for normal operations and Cx shutdown and General Design Criterion 34 which specifies require-
- i,, )
ments for the residual heat removal system by meeting the regulatory position in Branch Technical Position RSB 5-1.
5.4.7-9 Rev. 3
In addition, the applicant has met the requirements of the following Task Action Plan Items:
(1) Task Action Plan item II.E.3.2 of NUREG-0660 (Ref. 10) as it relates to systems capability and reliability of shutdown heat removal systems under various transients.
(2) Task Action Plan item II.E.3.3 of NUREG-0660 (Ref. 10) as it relates to a coordinated study of shutdown heat removal requirements.
(3) Task Act on Plan item II.B.8 of NUREG-0718 (Ref. 7) as it relates to description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded core accidents (cps only).
(4) Task Action Plan item III.D.1.1 of NUREG-0737 (Ref. 8) and NUREG-0718 (Ref. 7) as they relate to primary coolant sources outside of containment (cps and OLs).
For BWRs The residual heat removal function is accomplished in two phases:
the initial cooldown phase and a low pressure-temperature operation phase.
In the event of loss of offsite electrical power, the initial cooldown phase is accomplished using the reactor core isolation cooling (RCIC) system and the safety / relief valves.
The low pressure-temperature mode of operation is usually accomplished by the residual heat removal (RHR) system.
However, certain single failures can render the RHR system inoperative.
In that event, two alternate systems that use components of the RCIC and RHR system are available to bring the reactor to cold shutdown conditions.
The scope of review of these systems for the plant included piping and instrumentation diagrams, equipment layout drawings, failure modes and effects analysis, and design performance specifica-tions for essential components.
The review has included the applicant's proposed design criteria and design bases for these systems and his analysis of the adequacy of those criteria and bases and of the conformance of the design to these criteria and bases.
The staff concludes that the design of the Residual Heat Removal System is acceptable and meets the requirements of General Design Criteria 2, 4 5, 19, and 34.
This conclusion is based on the following:
2 (1) The applicant has met General Design Criterion 2 with respect l
to position C-2 of Regulatory Guide 1.29 concerning the seismic design of systems, structures and components whose failure could cause an unacceptable reduction in the capability of the l
residual heat removal system.
l (2) The applicant has met the General Design Criterion 4 with respect to dynamic effects associated flow instabilities and loads (e.g., water I
l hammer).
5.4.7-10 Rev. 3
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. n).
(3) The applicant has met the requirements of General Design
(
Criterion 5 with respect to sharing of structures, systems, and V
components by demonstrating that such sharing does not signifi-cantly impair the ability of the Residual Heat Removal System to perform 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 General Design Criterion 19 with respect to the main control room requirements for normal operations and shutdown and General Design Criterion 34 which specifies require-ments for the residual heat removal system by meeting the regulatory position in Branch Technical Position RSB 5-1.
In addition, the applicant has met the requirements of the following Task Action Plan Items:
(1) Task Action Plan item II.E.3.2 of NUREG-0660 as it relates to systems capability and reliability of shutdown heat removal systems under various transients.
(2) Task Action Plan item II.E.3.3 of NUREG-0660 as it relates to a coordinated study of shutdown heat removal requirements.
(3) Task Action Plan item II.B.8 of NUREG-0718 (Ref. 7) as it relates to description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded
()
core accidents (cps only).
(4) Task Action Plan item III.D.1.1 of NUREG-0737 (Ref. 8) and NUREG-0718 (Ref. 7) as they relate to primary coolant sources outside of containment (cps and OLs).
In addition to the above criteria, the acceptability of the RHR system may be' based on the degree of design timilarity with previously approved plants. Deviations from these criteria from other types of RHR systems (e.g., systems that are designed to withstand reactor coolant system operating pressure or systems located entirely inside containmnt) will be considered on an individual basis.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alter-native 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 C
herein are contained in the referenced BTP RSB 5-1, regulatory guides, and I
v Inplementation of the propose _d_ Item II.b will apply to CP applicants.
5.4.7-11 Rev. 3
F VI.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems and Components."
3.
10 CFR Part 50, Appendix A. General Design Criterion 19, " Control Room."
4.
10 CFR Part 50, Appendix A, General Design Criterion 34, " Residual Heat Removal."
5.
Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," attached to SRP Section 5.4.7.
6.
Reglatory Guide 1.29, " Seismic Design Classification."
7.
NUREG-0718, " Licensing Requirements for Pending Applications for Construc-tion Permits and Manufacturing License."
8.
NUREG-0737, " Clarification of TMI Action Plan Requirements."
9.
Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Systems."
10.
NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident."
11.
10 CFR Part 50, Appendix A, General Design Criterion 4, " Environmental and Missile Design Bases" l
l 9
5.4.7-12 Rev. 3
1 1
BRANCH TECHNICAL POSITION RSB 5-1 DESIGN REQUIREMENTS OF THE RESIDUAL HEAT REMOVAL SYSTEM a
/
BACKGROUND v
GDC 19 states that, "A control room shall be provided from which actions can be taken to operate the nuclear power unit under normal conditions..."
Normal operating conditions including the shutting down of a reactor; therefore, since the residual heat removal (RHR) system is one of several systems involved in the normal shutdown of all reactors, this system must be operable from the control room.
GDC 34 states that " Suitable redundance.
.shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available), the system safety function can be accomplished, assuming a single failure."
In most current plant designs the RHR system has a lower design pressure than the reactor coolant system (RCS), is located outside of containment and is part of the emergency core cooling system (ECCS).
However, it is possible for the RHR system to have different design characteristics.
For example, the RHR system might have the same design pressure as the RCS, or be located inside of containment.
Plants which may have RHR systems that deviate from current designs will be reviewed on a case-by-case basis.
The functional, isolation, pressure relief, pump protection, and test requirements for the RHR system are
-~s included in this position.
I
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/
BRANCH POSITION A.
Functional Requirements The system (s) which can be used to take the reactor from normal operating conditions to cold shutdown
- shall satisfy the functional requirements listed below.
1.
The design shall be such that the reactor can be taken fram normal operating conditions to cold shutdown using only safety grade systems.
These systems shall satisfy General Design Criteria 1 through 5.
l 2.
The system (s) shall have su' table redundancy in components and features, and suitable interconnections, leak detection, and isolation l
capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.
AProcesses involved in cooldown are heat removal, depressurization, flow circulation, and reactivity control.
The cold shutdown condition, as described in the Standard Technical Specifications, refers to a sub-i (*;
critical reactor with a reactor coolant temperature no greater than 200 F
\\
)
for a PWR and 212 F for a BWR.
l 5.4.7-13 Rev. 3 l
L
3.
The system (s) shall be capable of being operated from the control room with either only onsite or only offsite power available.
In demonstrating that the system can perform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.
4.
The system (s) shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.
B.
RHR System Isolation Requirements The RHR system shall satisfy the isolation requirements listed below.
1.
The following shall be provided in the suction side of the RHR system to isolate it from the RCS.
(a) Isolation shall be provided by at least two power-operated valves in series.
The valve positions shall be indicated in the control room.
(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure.
Failure of a power supply shall not cause any valve to change position.
(c) The valves shall have independent diverse interlocks to protect against one or both valves being open during an RCS increase above the design pressure of the RHR system.
2.
One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:
(a) The valves, position indicators, and interlocks described in item 1(a) thru 1(c) above, (b) One or more check valves in series with a normally closed power-operated valve.
The power-operated valve position shall be indicated in the control room.
If the RHR system discharge line is used for an ECCS function, the power-operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressere has decreased below the ECCS design pressure.
i l
(c) Three check valves in series, or i
(d) Two check valves in series, provided that there are desion provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.
O 5.4.7-14 Rev. 3 l
e:
C.
Pressure Relief Requirements The RHR system shall satisfy the pressure relief requirements listed below.
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1.
To protect the RHR system against accidental overpressurization when it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code.
The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system.
For example, during shutdown cooling in a PWR with no steam bubble in the pres-surizer, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be considered in selection of the design bases.
2.
Fluid discharged through the RHR system pressure relief valves must be collected and contained such that a stuck open relief valve will not:
(a) Result in flooding of any safety-related equipment.
(b) Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.
(c) Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment.
O) 3.
If interlocks are provided to automatically close the isolation valves when the RCS pressure exceeds the RHR system design pressure, adequate relief capacity shall be provided during the time period while the valves are closing.
D.
Pump Protection Requirements The design and operating procedures of any RHR system shall have provisions to prevent damage to the RHR system due to overheating, cavitation or loss of adequate pump suction fluid.
E.
Test Requirements The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in the RHR mode.
Testability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22.
The preoperational and initial startup test program shall be in conformance with Regulatory Guide 1.68.
The programs for PWRs shall include tests with supporting analysis to (a) confirm that adequate mixing of borated water added prior to or during ceoldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing, and (b) confirm that the cooldown under natural circulation conditions can be achieved within the limits specified in the emergency operating procedures.
Comparison with performance of previously tested plants of similar design may 1
be substituted for these tests.
1 LJ 5.4.7-15 Rev. 3
o F.
Operational Procedures The operational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33.
For pressurized water reactors, the operational procedures shall include specific procedures and information required for cooldown under natural circulation conditions.
G.
Auxiliary Feedwater Supply The seismic Category I water supply for the auxiliary feedwater system for a PWR shall have sufficient inventory to permit operation at hot shutdown for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by cooldown to the conditions permitting operation of the RHR system.
The inventory needed for cooldown shall be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure.
H.
Implementation For the purposes of implementing the requirements for plant heat removal capabilitity for compliance with this position, plants are divided into the following three classes:
Class 1 Full compliance with this position for all plants (custom or standard) for which CP or PDA applications are docketed on or after January 1, 1978.
See Table 1 for possible solutions for full compliance.
Class 2 Partial implementation of this position for all plants (custom or standard) for which CP or PDA applications are doc.keted before January 1, 1978, and for which an OL issuance is expected on or after January 1, 1979.
See Table 1 for recommended implementation for Class 2 plants.
Class 3 The extent to which the implementation guidance in Table 1 will be backfitted for all operating reactors and all other plants (custom or standard) for which issuance of the OL is expected before January 1, 1979, will be based on the combined I&E and 00R review of related plant features for operating reactors.
O 5.4.7-16 Rev. 3
^
U U
V e
4 TABLE 1.
POSSIBLE SOLUTION FOR FULL COMPLIANCE WITH BTP RSB 5-1 AND RECOMMENDED IMPLEMENTATION FOR CLASS 2 PLANTS Dssign Requirements Process and [5ystem Possible Solution for Recommended Implementation for of BTP RSB 5-1 or Component]
Full Compliance Class 2 Plants (see Note 1)
I.
Functional Requirement for Long-term cooling [RHR drop Provide double drop line (or valves Compliance will not.be required if Taking to Cold Shutdown line]
in parallel) to prevent single valve it can be shown that correction for failure from stopping RHR cooling single failure by manual actions
- a. Capability Using Only Safety function. (Note: This requirement inside or outside of containment or Grade Systems in conjunction with meeting effects return to hot standby until manual of single failure for long-term actions (or repairs) are found to
- b. Capability with either only cooling and isolation requirements be acceptable for the individual onsite or only offsite power involve increased number of plant.
and with single failure independent power supplies and (limited action outside CR to possibly more than four valves),
meet SF)
- c. Reasonable time for cooldown assuming most limiting SF and w
,a only offsite or only onsite y
power.
Heat removal and RCS circulation Provide safety grade dump valves, Compliance required.
during cooldown to cold shutdown operators, and power supply, etc. So (Note: Need SG cooling to main-that manual action should not be tain RCS circulation even after required after SSE except to meet RHR in operation when under single failure, natural circulation [ steam dump valves].)
Depressurization (Pressurizer Provide upgrading and additional Compliance will not be required if auxiliary spray or power-valves to ensure operation of aux-a) dependence on manual actions operated relief valves).
iliary pressurizer spray using only inside containment af ter SSE or safety grade subsystem meeting single single failure or b) remaining at failure. Possible alternative may hot standby until manual actions
=
involve using pressurizer power-or repairs are complete are found operated relief valves which have to be acceptable for the individual been upgraded. Meet SSE and single plant.
failure without manual operation inside containment.
TABLE 1.
POSSIBLE SOLUTION FOR FULL COMPLIANCE WITH BTP RSB 5-1 AND RECOMMENDED IMPLEMENTATION FOR CLASS 2 PLANTS Design Requirements Process and [Systen Possible Solution for Recommended Implementation for of BTP RSB 5-1 or Component]
Full Compliance Class 2 Plants (see Note 1)
Boration for cold shutdown Provide procedure and upgrading where Same as above.
necessary such that boration to cold shutdown concentration meets the requirements of I.
Solution could range from (1) upgrading and adding valves to have both letdown and charg-ing paths sa'ety grade and meet single failure to (2) use of backup procedures involving less cost. For example, bor-ation without letdown may be acceptable and eliminate need for upgrading let-down path. Use of ECCS for injection of borated water may also be accept-able. Need surveillance of boron P
concentration (boronometer and/or sampling). Limited operator action y
inside or outside of containment g
if justified.
II. RHR Isolation RHR System Comply with one of allowable Compliance required. (Plants arrangements given.
normally meet the requirement under existing SRP Section 5.4.7).
III. RHR Pressure Relief Collect and contain relief RHR System Determine piping, etc., needed to Compliance will not be required, discharge meet requirement to provide in if it is shown that adequate design.
alternate methods of disposing of discharge are available.
E' s
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O O
O
-g G
G L) e TABLE 1.
POSSIBLE SOLUTION FOR FULL COMPLIANCE WITH BTP RSB 5-1 AND RECOPMENDED IMPLEMENTATION FOR CLASS 2 PLANTS Design Requirements Process and (System Possible Solution for Recommended Implementation for of BTP RSB 5-1 or Component]
Full Compliance Class 2 Plants (see Note 1)
V. Test Requirement Meet R.G. 1.68.
For PWRs, Run tests confirming analysis to Compliance required.
test plus analysis for cooldown meet requirement.
under natural circulation to confirm adequate mixing and cooldown within limits specified in E0P.
VI. Operational Procedure Meet R.G. 1.33.
For PWRs, Develop procedures and information Compliance required.
include specific procedures and from tests and analysis, information for cooldown under P
natural circulation.
VII. Auxiliary Feedwater Supply Seismic Category I supply for Emergency Feedwater Supply From tests and analysis obtain Compliance will not be required, auxiliary FW for at least four conservative estimate of auxiliary if it is shown that an adequate hours at hot shutdown plus FW supply to meet requirement and alternate seismic Category I cooldown to RHR cut-in based provide seismic Category I supply.
source is available, on longest time for only onsite or only offsite power and assumed single failure.
Note 1: The implementation for Class 2 plants.does not result in a major impact while providing additional capability to go to cold shutdown. The major impact results from the requirement for safety grade steam dump valves.
E' kb
NU REG-0800 (Formuly NUREG-75/087) pa asog 8
i U.S. NUCLEAR REGULATORY COMMISSION nU Qj$s) STANDARD REV EW PLAN
%, v /
OFFICE OF NUCLEAR REACTOR HEGULATION e e...
Proposed Revision i
Standard Review Plan PSRP-6.3, Rev. 2 This proposed revision of the Standard Review Plan and its supporting value/inpact statement and associated technical documentation have not received a complete staff revicu and approval and do not represent an O
official NRC staff position.
The proposed revision to the Standard Review
("/
Plan incorporates the resolution of generic issue USI A-1, " Water Hamer."
Public coments are being solicited on the proposed SRP section and the associated value/inpact analysis and technical support document HUREG-0927,
" Evaluation of Water Hanner Experience in Nuclear Power Plants" (including any implementation schedules) prior to a final review and decision by the Office of Nuclear Reactor Regulation as to whether this proposed revision should be aproved.
Coments should be sent to the Secretary of the Connission, U. S. Nuclear Regulatory Comission, Washington, D. C.
- 20555, Attention:
Docketing and Service Branch. All connents received by July 18, 1983 will be considered, and all of the associated docunents and comments considered will be maoe publicly availabic prior to a decision by the Director, Office of Nuclear Reactor Regulation, on whether to inplement this revision.
Copies of each of these documents are available upon written reouest to the Division of Technical Infornation and Document Control, U. S.
Nuclear Regulatory Comission, Washington, D. C.
20555.
USNRC STANDARD REVIEW PLAN Star dard 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
[% \\
Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
(
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The l
L/
standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
1 Not all sections of the Standard Format have a corresponding review p!an.
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. 20555
NU REG-0800 (Form:rly NUREG-75/087)
!e ascgI%
U.S. NUCLEAR REGULATORY COMMISSION O iMi STANDARD REV EW PLAN d
8 OFFICE OF NUCLEAR REACTOR REGULATION N.i.9 %. o e..
Proposed Revision 2 to 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 appli-cable regulations, including the general design criteria and the amendments to 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 f
The preoperational and initial startup test programs for the ECCS are reviewed by the Procedures and Test Review Branch (PTRB) 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.
l 4.
Technical Specifications The proposed technical specifications are reviewed to assure that they are adequate in regard to limiting conditions of operation and periodic surveil-lance testing.
Rev. 2 USNRC STANDARD REVIEW PLAN Staridard 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 CN Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
(
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The bj standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
s Not all sections of the Standard Format have a correspondmg review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa-tion and experience.
l l
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. 20555.
The ability cf th: ECCS to citigate thm cons:qu:nces of a sp:ctrum 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 aactor 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 outside containment on ECCS.
This review includes the effect of pipe whip, jet impingement forces, and environmental conditions created as part of its primary review responsibility for SRP Section 3.6.1.
Instrumentation and Control Systems Branch (ICSB), as part of its primary review responsi-bility 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 Branch (CSB) verifies that portions of the ECCS penetrating the containment barrier are designed with acceptable isolation features to maintain contain-ment 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 load-ing 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 Structural Engineering Branch (SEB) reviews the 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 Test Review Branch (PTRB) reviews the proposed preoperational and initial startup test pro-grams to determine that they are consistent with the intent of Regulatory Guides 1.68 and 1.79 as part of its primary review responsibility for SRP Section 14.2.
The PTRB also has primary review responsibility for Task Action Plan items 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 Plan items II.B.2 of NUREG-0694 and NUREG-0718 which involve radiation and shielding design 6.3-2 Rev. 2
I review to take corrective actions to ensure adequate access to vital areas and protection of safety equipment (cps and Ols).
The review for Technical Oc Specifications and Quality Assurance are coordinated and performed by the (j
Licensing Guidance 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:
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 af the ECCS to perform its safety function.
Acceptability is based on meeting position C2 of Regulatory Guide 1.29.
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-ponents important to safety shall not be shared among nuclear power units unless it can be demonstrated that sharing will not impair their ability to perform their safety function.
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D.
General Design Criterion 17 as it relates to the design of the ECCS having 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.
E.
General Design Criterion 27 as it relates 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 maintained.
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 core at a rate so that fuel and clad damage will not interfere with 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, 950.46, 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:
6.3-3 Rev. 2 m
- ---~ - --
-m
=-
3-+----v-
- w,--w-
- - - - = * - - - * -
In regard to th: ECCS ccc ptanca crit:ria (R3f. 1), th fiva tajor p::rformanca 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 nhase following an accident.
The ECCS must be designed to permit periodic inservice inspection of important components, such 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 in steam lines shall be provided, in order to meet the requirements of General Design Criterion 4 (Ref. 17).
The design of those portions of the system which are not safety related, whose failures could have an adverse effect on the ECCS system, must be in accordance with GDC 2 (Ref. 2), and acceptance is based on meeting Position C2 of Regulatory Guide 1.29 (Ref. 10).
Interfaces between the ECCS and component or service water systems must be such that operation of one does not interfere with, and provides proper support (where required) for, the other.
In relation to these and other shared systems, e.g.,
residual heat removal (RHR) and containment heat removal systems, the ECCS must conform to GDC 5 (Ref. 3).
6.3-4 Rev. 2 b
The requirements of the following Task Action Plan items must also be satisfied:
1.
Task Action Plan Item II.B.8 of NUREG-0718 (Ref. 14) which involves descrip-
/' sx
(
)
tion by the applicants of the degree to which the designs conform to the N- '
proposed interim rule on degraded core accidents (cps and OLs).
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).
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 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 HPCI and RCIC for BWR plants.
6.
Taks Action Plan Item II.K.3(18) of NUREG-0737 and NUREG-0718 involving ECCS outages for all plants.
7.
Task Action Plan Item II.K.3(21) of NUREG-0737 and NUREG-0718 which involves a study evaluating 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 CFT flows.in B&W
'y 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.
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 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 por-tion 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 N
to determine that the acceptance criteria of subsection II have been met.
These
)
steps should be adapted to CP or OL reviews as appropriate.
sv 6.3-5 Rev. 2
1.
Tha relationship of the systea und:r review to oth r prsviously approv d 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 storage volume, system flow rate and pressure, etc.) are examined for each component to confirm that these parameters satisfy operating requirements and the recom-mendations of Regulatory Guide 1.1 (Ref. 9).
4.
The piping and instrumentation 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 he 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.
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 con-sideration 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 ACSB review as required.)
In 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.
l 10.
For a boiling water reactor (BWR), the reactor coolant automatic depressuri-zation system is reviewed to confirm the capability to satisfy LOCA pressure relief functions, including consideration of a single failure.
6.3-6 Rev. 2
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The design of ECCS injection lines is reviewed to' confirm that the isolation 11.
spr'ovisions at the interface with the reactor coolant system are adequate.
('~'y
'The number and type of valves used to' form the i,nterface between low
'j pressure portions of the ECCS and the reactor coolant system must provide x
~~
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 '
val.ve. 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.
b.
TQree 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 is4 performed at least annually.
- 12. The reviewer shoula 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 to the reactor coolant system in a pressurized water reactor (PWR) are reviewed to ensure that adequate provisions are made against inadvertent
' isolation.
(n) 14.
The capacity and settings of relief valves provided for the ECCS to satisfy
\\s,,/
system overpressure protection requirements are reviewed.
In particular, for PWRs, che reviewer confirms that the accumulator relief valves have adequate capacity so that leakage from the reactor coolant system will nbt jeopardize the integrity of the accumulators.
15.
The ECCS is reviewed to evaluate the adenuacy 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 ste6m bubble collapse.
For syste.ns with a water supply above the discharge lines, voided lines are prevented by proper vent-location-and filling and venting proce-dures.
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
' Tilling 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 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, (b) sloped lines, and (c) limitations on opening
^
and closing sequences and seal-ins for manual operation of the isolation valves.
The turbine exhaust line features include sloped lines and vacuum s
breakers.
x_ -)
l' 6.3-7 Rev. 2
16.
Th; r!. viewer confirms th:t no compon:nt or fcature of the ECCS in on:
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 contirms that within an individual reactor facility, any ccm-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 Tor the operator to respond.
l 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 ECCS.
If an intermediate heat transport system, such as the component cooling water system, is used to provide long-term cooling capability, the systcm 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.
21.
The RSB reviewer consults with the ICSB reviewer to:
Confirm that the power requirements of the ECCS, including the tim-a.
ing 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 2 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 such as coolant flow, coolant tempera-ture, 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 suf ficient information in the control room to determine that adequate core cocling is being provided.
Confirm that automatic actuation and remote-manual valve controls c.
are capable of performing the functions required, that suitable interlocks are provided, which do not impair separation of power 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 t
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 6.3-8 Rev. 2
following portions of the review of ECCS response in loss-of-coolant accidents are performed by the RSB reviewer under this SRP section:
- /'~'s t
I The lower limit of break size for which ECCS operation is required a.
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 perform at this lower limit of break size is confirmed.
b.
The reviewer confirms that the analyses take into account a variety 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 tnat different single failures may be limiting, depending on the particular
)
break location and break size postulated.
1 d.
The ECCS component response times (e.g., for valves, pumps, power supply) are reviewed to confirm that they are within the delay times used in the accident analyses.
e.
The ECCS design adequacy for all modes of reactor operation (e.g.,
full power, low power, hot standby, cold shutdown, partial loop isolation) is confirmed.
u
.s 23.
The proposed plant technical specifications are reviewed to:
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/
a.
Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restric-s tions 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 w
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 demonstrated by an applicant that pumps and valves operate on normal and emergency 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 re. sewer contacts his counterpart in PTRB to discuss any special test requirements and to confirm that the proposed preoperational test
/'~'N program for the ECCS is in conformance with the intent of Regulatory
(
)
Guide 1.68 (Ref. 12).
- s_s y
/
6.3-9 Rev. 2
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, beat 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 com-ponents.
The review has included the applicant's proposed 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, 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 Reg"latory 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 perform 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 6.3-10 Rev. 2
~
N U REG-0800 (Formsrly NUREG-75/087)
+*faarog%
f
) U.S. NUCLEAR REGULATORY COMM J
@\\ Sff'OFFICE OF NUCLEAR REACTOR REGULATION e..ee Proposed Pevisinn Standard Review Plac i
PSRP-9.2.1, Rev. 3 This proposed revision of the Standard Leview Plan and its suppciting value/inpact statenent and associeted technical docunud etion have r,ct received a cer.plete staff review at d approval ard do not represent an
[
of ficial fiRC staff position.
The proposed revision to the St6ndard Review Plan incorporates the resolution of generic issue USI A-1, "liater Harrer."
Public corrents are being solicited on the proposed SRP Lection dvd the associated value/inpact analysis ard technical support docunent fiUREG-0927,
" Evaluation of Water Har.ver Experience in fluclear Power Plants" (incit. ding any inplementation schedules) prior to a fint1 review and decision by the Office of fluclear Reactor Reguletion as to whether this proposed revision should be aproved.
Connents should be sent to the Secretary of ibe Connission, U. S. fluclear Regulatory Connission, Washington, D. C.
- 20555, Attention:
Docketing and Service Brantb.
All connents receiv(c by July 18, 1983 will be considered, and all of the asscciated decurents and connents considered will be nade publicly avcilable prior to a decision by the Director, Office of fluclear Reactor Regulation, en whether te inplenent this revision.
Copics of each of these docunents are availabit t.pon writter request to the Division of Technical Inf ornation and Docunent Centrol, U. S.
flucleer Regulatory Commissien, Washington, D. C.
P0555.
i l
L USNRC STANDARD REVIEW PLAN Star.dard 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 h
Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
)
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.
Not all sections of the standard Format have a corresponding review plan.
,Nblished standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new inf orma-tion Od 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. 20555.
NU REG-0800 (Formsrly NUREG 75/087) p# 8IG(/g p
s U.S. NUCLEAR REGULATORY COMMISSION
+
o U
Rhf<f)STANDARD REV EW PLAN N '-
OFFICE OF NUCLEAR REACTOR REGULATION ee..e Proposed Revision 3 to 9.2.1 STATION SERVICE WATER SYSTEM REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB)
Secondary - None I.
AREAS OF REVIEW The service water system (SWS) provides essential cooling to safety-related equip-ment and may also provide cooling to nonsafety-related auxiliary components that are used for normal plant operation.
The ASB reviews the system from the service water pump intake to the points of cooling water discharge to assure conformance with the requirements of General Design Criteria 2, 4, 5, 44, 45, and 46.
The ultimate heat sink (reviewed under SRP Section 9.2.5) provides the intake source of water to the SWS for long-term cooling of station features required for plant shutdown and also any special equipment required to prevent or mitigate the con-sequences of postulated accidents and as such is an interface system to thu SWS.
/'
N The SWS pump performance characteristics will be compared to the high and low
(")
water levels of the ultimate heat sink to assure that pumping capability can be provided for extended periods of operation following postulated events.
1.
The ASB reviews the characteristics of the SWS components (pumps, heat exchangers, pipes, valves) with respect to their functional performance as affected by adverse operational (i.e., water hammer) and environmental occurrences including cold weather protection, by abnormal operational requirements, and by accident conditions such as a loss-of-coolant accident (LOCA) with the loss of offsite power.
Since the SWS normally has require-ments that relate to cooling functions during normal plant operation as well l
as for safety functions, the review will include an evaluation of the capa-bility of the system to perform these multiple functions.
2.
The ASB also reviews the design of the SWS with respect to:
a.
The capability for detection, control, and isolation of system leakage including the capability for detection and control of radioactive leak-l age into and out of the system and prevention of accidental releases to l
the environment.
Rev. 3 -
USNRC STANDARD REVIEW PLAN star.dard review plans are prepared for the guidance of the office of Nuclear Reactor Regulation staff responsible for the review uf gx applications to construct and operate nuclear power plants. These documents are made available to the public as part of the e
i Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
(
)
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The Nj standard review plan sections are keyed to the standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
[
Not all sections of the Standard Format have a curresponding review plan.
Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new inf orma-tion and experience.
j Comments and suggestions for improvement will be considered and should be sent to the U S. Nuclear Regulatory Commission, j
office of Nuclear Reactor Regulation. Washington, D.C. 20555.
l
b.
Measures to preclude long-term corrosion and organic fouling that would tend to degrade system performance.
c.
Provisions for system and component operational testing, including the instrumentation and control features that determine and verify that the system is operating in a correct mode (i.e., valve position, pressure and temperature indication).
d.
The effects of the failure of nonseismic Category I equipment, struc-tures or components of safety-related portions of the SWS are taken into account in the design.
3.
The ASB reviews the SWS capability to flood the reactor containment should this be required in a post-accident recovery situation.
4.
The ASB reviews the system to determine that a malfunction, a failure of a component, or the loss of a cooling source will not reduce the safety-related functional performance capabilities of the system.
Specifically, ASB performs the following reviews under the SRP sections indicated:
Review for flood protection is performed under SRP Section 3.4.1.
a.
b.
Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1.
c.
Review of the structures, systems and components to be protected against externally generated missiles is performed under SRP Sec-tion 3.5.2.
d.
Review of high and moderate energy pipe breaks is performed under SRP Section 3.6.1.
In addition, the ASB will coordinate other branches evaluations that interface with the overall review of the system as follows:
The Reactor Systems Branch (RSB) identifies essential components associated with the reactor coolant system and the emergency core cooling systems that are required for operation during normal operations or accident conditions.
The R$8 establishes accident cooling load functional requirements and minimum time intervals.
The RSB performs these reviews as part of its primary review responsibility for SRP Sections 5.4.7, 5.4.8, 6.0 and 15.0.
The Structural Engineering Branch (SEB) determines the acceptability of the design analyses, procedures, and criteria used to estab-lish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural phenomena such as the safe shutdown earthquake (SSE), probable maximum flood (PMF), and tornado missiles as part as its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5.
The Mechanical Engineer-ing Branch (MEB) determines that the components, piping and structures are designed in accordance with applicable codes and standards as part of its primary review responsibility for SRP Sections 3.9.1 through 3.9.3.
The MEB also determines the acceptability of the seismic and quality group classifica-tions for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2.
The MEB also reviews the adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6.
The Materials Engineering Branch (MTEB) verifies that inservice inspection requirements are met for system components as part of its primary review responsibility for SRP Section 6.6 and, upon 9.2.1-2 Rev. 3
+
o request, verifies the compatibility of the materials of construction with service conditions.
The Instrumentation and Control Systems Brar.ch (ICSB) and p
Power Systems Branch (PSB) will evaluate the system controls, instrumentation, Q
and power sources with respect to capabilities, capacity, and reliability for supplying power during normal and emergency conditions to safety-related pbmps, valves and other components as part of their primary review responsibility for SRP Sections 7.1 and 8.1, respectively.
The reviews for Fire Protection, Technical Specifications and Quality Assurance are coordinated and performed by the Chemical Engineering Branch, Licensing Guidance Branch and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0, and 17.0, respectively.
For those areas of review identified above as being the responsibility of other branches, the acceptance criteria and their methods of application are contained in the SRP sections identified as the primary review responsibility of those branches.
II.
ACCEPTANCE CRITERIA Acceptability of the design of the service water system, as described in the applicant's safety analysis report (SAR), including related sections of Chap-ters 2 and 3 of the SAR is based on specific general design criteria and regu-latory guides.
Listed below are specific criteria as they relate to the SWS.
The design of the service water system is acceptable if the integrated system design is in accordance with the following criteria:
1.
General Design Criterion 2, as related to structures. housing the system (Vp) and the system itself being capable of withstanding the effects of earth-quakes.
Acceptance is based on meeting the guidance of Regulatory Guide 1.29, Position C.1 for safety-related portions and Position C.2 for nonsafety related portions.
2.
General Design Criterion 4, as related to dynamic effects associated with flow instabilities and loads (e.g., water hammer) during normal plant operation as well as during upset or accident conditions.
3.
General Design Criterion 5, as related to the capability of shared systems and components important to safety being capable of performing required safety functions.
4.
General Design Criterion 44, as related to transferring heat from struc-tures systems and components important to safety, to an ultimate heat sink.
Acceptance is based on the following:
a.
The capability to transfer heat loads from safety-related structures, systems, and components to a heat sink under both normal operating and accident conditions.
b.
Component redundancy so that the safety function can be performed assuming a single active component failure coincident with the loss of offsite power.
,/n c.
The capability to isolate components, oubsystems, or piping if required so that the system safety function will not be compromised.
9.2.1-3 Rev. 3
=
d.
Meeting task action plan item II.K.1-C.l.22 of NUREG-0694 for boiling water reactors regarding automatic and manual actions necessary when the main feedwater system is not operable.
Meeting task action plan item II.K.1.22 of NUREG-0718 for B&W plants e.
regarding automatic and manual actions for proper functioning of the auxiliary heat removal systems when the main feedwater system is not operable.
5.
General Design Criterion 45, as related to design provisions to permit inservice inspection of safety-related components and equipment.
6.
General Design Criterion 46, as related to design provisions to permit operational functional testing of safety-related systems and components.
III. REVIEW PROCEDURES The procedures set forth below are used during the construction permit (CP) application review to determine that the design criteria and bases and the preliminary design as set forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II.
For review of operating license (0L) applications, the review procedures and acceptance criteria are utilized to verify that the initial design criteria and bases have been apptopriately implemented in the final design as set forth in the final safety analysis report.
Upon request from the primary reviewer, the coordinating review branches will provide input for the areas of review stated in subsection I.
The primary reviewer obtains and uses such input as required to assure that this review procedure is complete.
As a result of the various SWS designs provided, there will be variations in system requirements.
For the purpose of this SRP section, a typical system is assumed which has fully redundant systems, with each of the systems having an identical essential (safety features) portion and an identical non-essential portion (used for normal operation).
For cases where there are variatiuns from the typical arrangement, the reviewer will adjust the review procedures given below.
However, the system design will be required to meet the acceptance criteria given in subsection II.
Also, the reviewer will need to refer to SRP sections for other systems that would interface with the SWS, depending upon the nature and conditions of the ultimate heat sink cooling water (e.g., salt water).
l 1.
The SAR is reviewed to determine that the system description and piping and instrumentation diagrams (P& ids) show the SWS equipment that is used for normal operation, and the minimum system heat transfer and flow requirements for normal plant operation. The system performance require-ments will also be reviewed to determine that it describes component allowable operational degradation (e.g., pump leakage) and describes the procedures that will be followed to detect and correct these conditions when they become excessive.
2.
The reviewer, using the results of failure modes and effects analyses as appropriate, comparisons with previously approved systems, or independent calculations, determines that the system is capable of sustaining the loss of any active component and meeting minimum system requirements 9.2.1-4 Rev. 3
o (cooling load and flow) for the degraded conditions.
The system P& ids, layout drawings, and component descriptions and characteristics are then
/'
}
reviewed for the following points:
\\
/
a.
Essential portions of the SWS are correctly identified and are isol-able from the non-essential portions of the system.
The P& ids are reviewed to verify that they clearly indicate the physical division between each portion and indicate the required classification changes.
System drawings are also reviewed to see that they show the means for accomplishing isolation and the system description is reviewed to identify minimum performance requirements for the isolation valves.
The drawings and descriptions are reviewed to verify that automati-cally operated isolation valves separate non-essential portions and components from the essential portions.
b.
Essential portions of the SWS, including the isolation valves separat-ing essential and non-essential portions, are classified Quality Group C and seismic Category I.
Components and system descriptions in the SAR that identify mechanical and performance characteristics are reviewed to verify that the above seismic and safety classifica-tions have been included, and that the P& ids indicate any points of change in piping quality group classification.
Design provisions have been made that permit appropriate inservice c.
inspection and functional testing of system components important to safety.
It will be acceptable if the SAR information delineates a testing and inspection program and if the system drawings show the necessary test recirculation loops around pumps or isolation valves p) that would be required by this program.
(v d.
The review of seismic design is performed by SEB and the review for seismic and quality group classification is performed by MEB as indi-cated in subsection I of this SRP section.
3.
The reviewer determines that the safety function of the system will be maintained, as required, in the event of adverse environmental phenomena such as earthquakes, tornadoes, hurricanes, and floods, or in the event of certain pipe breaks or loss of offsite power. The reviewer uses engi-neering judgment, the results of a failure mode and effects analyses, and the results of reviews performed under other SRP sections to verify the following:
a.
The failure of portions of the system or of other systems not designti to seismic Category I and located close to essential por-tions of the system, or of non-seismic Category I structures that house, support, or are close to essential portions of the SWS, will not preclude operation of the essential portions of the SWS.
Refer-ence to SAR Chapter 2 describing site features and the general arrangement and layout drawings will be necessary as well as the SAR tabulation of seismic design classifications for structures and systems.
Statements in the SAR that verify that the above conditions are met are acceptable.
(CP) m
/
i b.
The essential portions of the SWS are protected from the effects of
(_ /
floods, hurricanes, tornadoes, and internally or externally generated 9.2.1-5 Rev. 3
s missiles.
Flood protection and missile protection criteria are dis-cussed and evaluated in detail under the Section 3 series of the SRP.
The reviewer will utilize the procedures identified in these SRP sec-tions to assure that the analyses presented are valid.
A statement to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected, or that compo-nents of the system will be located in individual cubicles or rooms that will withstand the effects of both flooding and missiles is acceptable.
The location and the design of the system, structures, and pump rooms (cubicles) are reviewed to determine that the degree of protection provided is adequate.
The SWS pumps will have sufficient available net positive suction c.
head (NPSH) at the pump suction locations, considering low water levels.
Reference to SRP Section 2.4, which indicates the lowest probable water level of the heat sink, and to drawings indicating the elevation of service water pump impellers will be necessary.
An independent calculation verifying the applicant's conclusion will be necessary for acceptance.
d.
Provisions are made in the system to detect and control leakage of radioactive contamination into and out of the system.
It will be acceptable if the system P& ids show radiation monitors located on the system discharge and at components susceptible to leakage, and these components can be isolated by one automatic and one manual valve in series.
The essential portions of the system are protected from the effects e.
of high and moderate energy line breaks.
Layout drawings are reviewed to assure that no high or moderate energy piping systems are close to essential portions of the SWS, or that protection from the effects of failure will be provided.
The means of providing such protection will be given in Section 3.6 of the SAR and the procedures for reviewing this information are given in the correspond-ing SRP sections.
f.
Essential components and subsystems necessary for safe shutdown can function as required in the event of loss of offsite power.
The system design will be acceptable if the SWS meets minimum system requirements as stated in the SAR assuming a concurrent failure of a single active component, including a single failure of an auxiliary electric power source.
The SAR is reviewed to determine that for each SWS component or subsystem affected by the loss of offsite power, system flow and heat transfer capability meet or exceed mini-mum requirements.
The results of failure modes and effects analyses are considered in assuring that the system meets these requirements.
This will be an acceptable verification of system functional reliability.
g.
Provisions are made for protection of the essential service water supply from potential failures or malfunctions caused by freezing, icing, and other adverse environmental conditions.
Statements in the SAR that would indicate that safety grade heating sources will be used for this purpose, considering the equipment necessary for safe shutdown, will be acceptable.
9.2.1-6 Rev. 3
/7 4.
The descriptive information, P& ids, SWS drawings, and failure modes and Q
effects analyses in the SAR are reviewed to assure that essential portions of the system can function following design basis accidents assuming a concurrent single active component failure. The reviewer evaluates the failure mode and effects analysis presented in the SAR to assure function of required components, traces the availability of these components on system drawings, and checks that the SAR contains verification that mini-mum system flow and heat transfer requirements are met for each accident situation for the required time spans._ For each case the design will be acceptable if minimum system requirements are met.
5.
The SAR is reviewed to assure that the applicant has described all the automatic and manual actions necessary for proper functioning of the service water system when the main feedwater system is not operable. The design will be acceptable in this regard if sufficient detail is presented to provide reasonable assurance that the requirements of items II.K.1.22 of NUREG-0718 and II.K.1-C.1.22 of NUREG-0694 are property implemented.
~6.
The SAR is reviewed to assure that the applicant has committed to address the potential for wa'ter hammer in open loop systems and wilt provide for venting and filling of such systems, and operating procedures for avoidance of water hammer.
IV.
EVALUATION FINDINGS n
The reviewer determines that sufficient information has been provided and his
(
)
review supports conclusions of the following type, to be included in the staff's
(./
safety evaluation report:
The service water system (SWS) includes all components and piping from the SWS pump intake to the points of cooling water discharge.
Portions of the SWS that are necessary for safe shutdown accident prevention, or accident mitigation are designed to seismic Category I, Quality Group C requirements. Based on the review of the applicant's proposed design criteria, design bases and safety classification for the service water l
system regarding the requirements for continuous cooling of safety-related components necessary for a safe plant shutdown, the staff concludes that the design of the service water system is acceptable and meets the require-ments of General Design Criteria 2, 4, 5, 44, 45, and 46.
This conclusion is based on the following:
1.
The applicant has met the requirements of General Design Criterion 2 with respect to safety-related portions of the system being capable of withstanding the effects of earthquakes. Acceptance is based on meeting Regulatory Guide 1.29 position C.1 for the safety-related I
portions and position C.2 for the nonsafety-related portions.
2.
The applicant has met the requirements of GDC 4 with respect to dynamic effects associated with flow instabilities (i.e., water I
hammer loads) with respect to impairment of the required service l
water systems during normal plant opera _tions, and under upset or O
accident conditions. Acceptance is based on the following:
4 l
9.2.1-7 Rev. 3 1
r a.
Vents shall be provided for venting of components and piping at high points in liquid filled, but normally idle piping (or systems) where voiding can occur.
These vents should be designed for ease of operational testing on a periodic basis.
b.
Consideration will be given to voiding which can occur following pump shutdown, or during standby.
If the system design is such that voiding could occur, means should be provided for a slow system fill upon pump start for avoidance of water hammer.
c.
Operating and maintenance procedures will be reviewed by the applicant to assure that sufficient measures have been taken for avoiding water hammer (e.g., rapid fill due to pump start,
-pjriodic fill and vent checks, avoidance of sudden valve movement, or realignment).
3.
The applicant has met the requirements of General Design Criterion 5 with respect to sharing of structures, systems and components by demonstrating that such sharing does not significantly impair the ability of the service water system to perform its safety function, including in the event of an accident in one unit, an orderly shut-down and cooldown of the remaining units.
4.
The applicant has met the requirements of General Design Criterion 44 with respect to cooling water by providing a system to transfer heat from structures, systems and components important to safety to an ultimate heat sink.
The applicant has demonstrated that the service water system can transfer the combined heat load of these structures, systems, and components under normal operating and acci-dent conditions assuming loss of offsite power and a single failure and that portions of the system can be isolated so that the safety function of the system will not be compromised.
The applicant has also met task action plan items II.K.1-C.1.22 of NUREG-0694 and II.K.1.22 of NUREG 0718 in meeting General Design Criterion 4.
5.
The applicant has met the requirements of General Design Criterion 45 with respect to inspection of cooling water systems by providing a service water system design which permits inservice inspection of safety-related components and equipment.
6.
The applicant has met the requirements of General Design Criterion 45 with respect to testing of cooling water systems by providing a service water system design which permits operational functional testing of the system and its components.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
Except in those cases in which the applicant proposes an acceptable alterna-tive method for complying with specified portions of the Commission's Regula-tions, the method described herein will be used by the staff in its evaluation of conformance with Commission Regulations.
9.2.1-8 Rev. 3
p Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced Regulatory Guide and NUREGs.
Implementation of Item II.2 is as follows:
a.
Operating plants and OL applicants need not address these revisions.
b.
CP applicants will be required to comply with the_se proposed revisions.
VI. REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Struc-tures, Systems, and Components."
3.
10 CFR Part 50, Appendix A, General Design Criterion 44, " Cooling Water."
4.
10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cooling Water System."
5.
10 CFR Part 50, Appendix A, General Design Criterion 46, " Testing of Cool-ing Water Systems."
6.
Regulatory Guide 1.29, " Seismic Design Classification."
7.
NUREG-0694, "TMI-Related Requirements for New Operating Licenses."
8.
NUREG-0718, " Proposed Licensing Requirements for Pending CP's and Manu-facturing License."
9.
10 CFR Part 50, Appendix A, General D,esign Criterion 4, "_ Environmental and Missile Design Bases."
i 9.2.1-9 Rev. 3 1
NU R EG-0800
+
(Formsrly NUREG.75/087)
!e ncgIe U.S. NUCLEAR REGULATORY COMMISSION O iW } STANDARD REV EW PLAN 8
OFFICE OF NUCLEAR REACTOR REGULATION o
e...e Proposed Revision Standard Review Plar-PSPP-9.2.2, Rev. 2 This proposed revision of the Standarc Review Plan and its supporting value/inpact statement and associated technical documentation have not received a complete staff review and approval and do not represent an official URC staff position.
The proposed revision to the Standard Review
.)
Plan incorporates the resolution of generic issue USI A-1, " Water Hamer."
v Public conments are being solicited on the proposed SPP section and the associated value/inpact analysis and technical support docunent huREG-G927,
" Evaluation of Water Hanmer Experience in Nuclear Power Plants" (including any implementation schedules) prior to a final review and decision by the Office of Nuclear Reactor Regulation as to whether this proposed revision should be aproved. Coments should be sent to the Secretary of the Comission, U. S. Nuclear Regulatory Commission, Washington, D. C.
- 20555, Attention:
Docketing and Service Branch. All coments received by July 18, 1983 wili be considered, and all of the associated documents and conments considered will be made publicly available prior to a decision by the Director, Office of Nuclear Reactor Regulation, on whether to inpleuent this revision.
Copies of each of these documents are available upon written request to the Division of Technical Infornation and Docunent Control, U. S.
Huclear Regulatory Comission, Washington, D. C.
20555.
USNRC STANDARD REVIEW PLAN Star dard 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 fm Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
/
)
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.
\\/
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 inf orma-tion and experience.
Comments and suggestions for imorovement will be considered and should be sent to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington. D.C. 20555.
L
NUREG-0800 (Form rly NUREG-75/087)
/paascg\\
U.S. NUCLEAR REGULATORY COMMISS;ON L+'?#n OFFICE OF NUCLEAR REACTOR REG
. i C')
gN p e...e Proposed Revision 2 to 9.2.2 REACTOR AUXILIARY COOLING WATER SYSTEMS REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB)
Secondary - None I.
AREAS OF REVIEW The ASB reviews reactor auxiliary cooling water systems (CWS) that are required for safe shutdown during normal, operational transient, and accident conditions and for mitigating the consequences of an accident or preventing the occurrence of an accident.
These include closed loop auxiliary cooling water systemt for reactor system components, reactor shutdown equipment, ventilation equipment, and components of the emergency core cooling system (ECCS).
The review of these systems includes components of the system, valves and piping, and points of connection or interfaces with other systems.
Emphasis is placed on the CWS for safety-related components such as ECCS equipment, ventilation equip-p) ment, and reactor shutdown equipment.
The ASB reviews reactor auxiliary cooling (V
water systems to ensure conformance with the requirements of General Design Criteria 2, 4, 5, 44, 45, and 46.
1.
The ASB reviews the capability of the auxiliary cooling systems to provide adequate cooling water to safety-related ECCS components and reactor auxi-liary equipment for all pianned operating conditions.
The review includes the following points:
a.
The functional performance requirements of the system including the ability to withstand adverse operational (i.e. water hammer) and environmental occurrences, operability requirements for normal operation, and requirements for operation during and subsequent to postulated accidents.
b.
Multiple performance functions (if required) assigned to the system and the necessity of each function for emergency core cooling and safe shutdown.
Rev. 2 USNRC STANDARD REVIEW PLAN Star dard 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 Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review gg j
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The i
\\ j) standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
\\
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. 20555.
c.
Th3 cap:bility of the systa surge tank to perform its intended function.
d.
The capability of the system to provide adequate cooling water during all operating conditions.
e.
The sizing of the system for core cooling and decay heat loads and the associated design margin.
2.
Other system aspects that are reviewed include:
a.
The effects of non-seismic Category I component failures on the seismic Category I portion of the system.
b.
The provisions for detection, collection, and control of system leak-age and the means provided to detect leakage of activity from one system to another and preclude its release to the environment.
The requirements for operational testing and inservice inspection of c.
the system.
d.
The capability of the system to provide adequate cooling to the seals and bearings of all reactor coolant pumps.
e.
Instrumentation and control features necessary to accomplish design functions, including isolation of components to deal with leakage or malfunctions and actuation requirements for redundant equipment.
f.
A simplifieri reliability analyses using event-tree and fault-tree logic techniques.
3.
ASB also performs the following reviews under the SRP sections indicated:
a.
Review of flood protection is performed under SRP Section 3.4.1, b.
Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1, c.
Review of the protection of structures, systems and components against the effects of externally generated missiles is performed under SRP Sections 3.5.1.4 and 3.5.2, and d.
Review of high and moderate energy pipe breaks is performed under SRP Section 3.6.1.
In addition, the ASB will coordinate other branches evaluations that interface with the overall review of the system asi follows.
The Reactor Systems Branch (RSB) will identify engineered safety feature components associated with the reactor coolant system and the emergency core cooling systems that are required for operation during normal operations, transients, and accident conditions.
RSB will establish cooling load functional requirements and minimum time inter-vals associated with safety-related components.
The RSB performs these reviews as part of its primary review responsibility for SRP Sections 5.4.7, 5.4.8, 6.0, and 15.0.
The Structural Engineering Branch (SEB) will determine the accepta-bility of the design analyses, procedures, and criteria used to establish the ability of Category I structures that house the system and supporting systems 9.2.2-2 Rev. 2
a to withstand the effects of natural phenomena such as the safe shutdown earth-quake (SSE), the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1, (V) 3.7.4, 3.8.4 and 3.8.5.
The Mechanical Engineering Branch (MEB) determines that the components, piping and structures are designed in accordance with applicable codes and standards as part of its primary review responsibility for SRP Sec-tions 3.9.1 and 3.9.3.
The MEB also determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRE Sections 3.2.1 and 3.2.2.
The MEB also reviews the adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6.
The Material Engineering Branch (MTEB) verifies that inservice inspection requirements are met for system components as part of its primary review responsibility for SRP Section 6.6 and, upon request, verifies the compatibility of the materials of construction with service conditions.
The Instrumentation and Control Systems Branch (ICSB) and Power Systems Branch (PSB) will determine the adequacy of the design, installa-tion, inspection, and testing of all essential electrical components, system controls, and instrumentation required for proper operation as part of their primary review responsibilities for SRP Sections 7.1 and 8.1, respectively.
The review for Fire Protection, Technical Specifications, and Quality Assurance are coordinated and performed by the Chemical Engineering Branch (CMEB), Licens-ing Guidance Branch (LGB) and Quality Assurance Branch (QAB) as part of their primary review responsibility for SRP Sections 9.5.1, 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 neces-sary for the review and their methods of application are contained in the referenced SRP section of the corresponding primary branch.
[
i y/
II.
ACCEPTANCE CRITERIA Acceptability of the designs of cooling water systems as described in the appli-cant's Safety Analysis Report (SAR), including related sections of Chapters 2 and 3 of the SAR, is based on specific general design criteria and regulatory guides, and on independent calculations and staff judgments with respect to system functions and component selection.
The design of a CWS is acceptable if the integrated system design is in accordance with the following requirements and recommendations:
1.
General Design Criterion 2, as related to structures housing the system and the system itself being capable of withstanding the effects of earth-quakes.
Acceptance is based on meeting the guidance of Regulatory Guide 1.29, Position C.1 for safety-related portions and Position C.2 for non-safety-related portions.
2.
General Design Criterion 4, as related to dynamic effects associated with flod instabilities and attendant loads (i.e., water hammer) during normal plant operation as well as during upset or accident conditions.
3.
General Design Criterion 5, as related to shared systems and components important to safety being capable of performing required safety functions.
4.
General Design Criterion 44, as its relates to:
,m i
\\
\\%'/
9.2.2-3 Rev. 2
s a.
Th2 capability to transfer h at loads from safety related structures, systems, and components to a heat sink under both normal operating and accident conditions.
b.
Component redundancy so that safety functions can be performed assum-ing a single active component failure coincident with the loss of off-site power.
c.
The capability to isolate components, systems, or piping, if required, so that the system safety function will not be compromised.
d.
Task Action Plan items II.K.2.16 and II.K.3.25 of NUREGs-0718 and 0737 as they related to loss of cooling water to reactor coolant pump (RCP) seals.
e.
A single failure in the CWS does not result in fuel damage or reac-tor coolant leakage in excess of normal coolant-makeup capability.
Single failure includes but is not limited to operator error, spurious activation of a valve operator, and loss of a cooling water pump.
A moderate-energy leakage crack or an accident that is initiated from a failure in the CWS piping does not result in excessive fuel damage or reactor coolant leakage in excess of normal coolant-makeup capa-bility.
A single r * ;ve failure is considered when evaluating the consequences of tr 3 accident.
Moderate leakage cracks are determined in accordance with the guidelines of Branch Technical Position ASB 3-1, "Prctection Against Postulated Failures in Fluid Systems Outside Containment."
It has been demonstrated by testing that the reactor coolant pumps will withstand a complete loss of cooling water for 20 minutes, and instrumentation in accordance with IEEE 279 that alarms in the control room is provided to detect a loss of cooling water to ensure a period of 20 minutes is available so that the operator would have sufficient time to initiate manual protection of the plant.
Alternatively, if it is not demonstrated by the necessary pump testing that the reactor coolant pumps will operate for 20 minutes without operator corrective action:
1.
Instrumentation in accordance with IEEE 279 is provided consist-ent with the criteria for the protection system to initiate auto-matic protection of the plant upon loss of cooling water to a pump.
For this case, the component cooling water supply to the seal and bearing of the pump may be designed to nonseismic Cate-gory I requirements and Quality Group D, or 2.
The component cooling water supply to each pump is designed to be capable of withstanding a single active failure or a moderate-energy line crack as defined in Branch Technical Position ASB 3-1 and to seismic Category I, Quality Group C, and ASME Section III Class 3 requirements.
4.
General Design Criterion 45, as related to the design provisions to permit inservice inspection of safety-related components and equipment.
9.2.2-4 Rev. 2
5.
General Design Criterion 46, as related to the design provisions to permit operational functional testing of safety-related systems or components to p!
ensure:
rU a.
Structural integrity and system leak tightness, b.
Operability and adequate performance of active system components.
c.
Capability of the integrated system to perform required functions during normal, shutdown, and accident situations.
III. REVIEW PROCEDURES The procedures set forth below are used during the construction permit (CP) application review to determine that the design criteria and bases and the pre-liminary design as set.forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II of this SRP section.
For the review of operating license (0L) applications, the review procedures and accept-ance criteria given in subsection II will be used 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.
One of the main objectives in the review of a CWS is to determine its function with regard to safety.
Some cooling systems are designed as safety-related systems in their entirety, others have only portions of the system that are safety-related, and others are classified as nonsafety-related because they do not perform any safety function.
To determine the safety category of a CWS, the ASB will evaluate its necessity for achieving safe reactor shutdown condi-I' 3 tions or for accident prevention or accident mitigation functions.
The safety functions to be performed by these systems in all designs are essentially the same, however, the method used varies from plant to plant depending upon the individual designer.
Upon request from the primary reviewer, the coordinating review branches will provide input for the areas of review stated in subsection I of this SRP section.
The primary reviewer obtains and uses such input as required to ensure that this review procedure is complete.
In view of the various designs provided, the procedures set forth below are for a typical CWS designed entirely as a safety-related system.
Any variance of the review procedures to take account of a proposed unique design will be such as to ensure that the system meets the criteria of subsection II.
The reviewer will select and emphasize material from this SRP section, as may be appro-l priate for a particular case.
1.
The information provided in the SAR pertaining to the design bases and design criteria, and the system description section are reviewed to verify that the equipment used and the minimum system heat transfer and l
flow requirements for normal plant operations are identified.
A review l
of the system piping and instrumentation diagrams (P& ids) will show which components of the system are used to:
l a.
Remove heat from the reactor primary coolant system necessary to O
achieve a safe reactor shutdown.
9.2.2-5 Rev. 2 I
e-
~.
b.
Provida essential cooling for containment components or systems such as the sprays, ventilation coolers, or sump equipment.
c.
Provide cooling for decay heat removal equipment.
d.
Provide cooling for emergency core cooling pump bearings or other emergency core cooling equipment necessary to prevent or mitigate the consequences of an accident.
2.
The system performance requirements section is reviewed to determine that it describes allowable component operational degradation (e.g., pump leak-age) and describes the procedures that will be followed to detect and correct these conditions when degradation becomes excessive.
3.
The reviewer, using the results of failure-modes and -effects analyses, determines that the system is capable of sustaining the loss of any active component and, on the basis of previously approved systems or independent calculations, that the minimum system requirements (cooling load and flow) are met for these failure conditions.
The system P& ids, layout drawings, and component descriptions and characteristics are then reviewed for the following points:
a.
Essential portions of the CWS are correctly identified ani are isol-able from the nonessential portions of the system.
The P& ids are reviewed to verify that they clearly indicate the physical division between each portion and indicate required classification changes.
System drawings are reviewed to see that they show the means for accomplishing isolation and the SAR description is reviewed to identify minimum performance of the isolation valves.
The drawings and description are reviewed to verify that automatically operated isolation valves separate nonessential portions and components from the essential portions.
b.
Essential portions of the CWS, including the isolation valves separating seismic Category I portions from the nonseismic portions, are Quality Group C and seismic Category I.
System design bases and criteria, and the component classification tables are reviewed to verify that the heat exchangers, pumps, valves, and piping of essen-tial portions of the system will be designed to seismic Category I requirements in accordance with the applicable criteria.
The review of seismic design is performed by SEB and the review for seismic and quality group classification is performed by MEB as indicated in sub-section I of this SRP section, c.
The system is designed to provide water makeup as necessary.
Cooling water systems that are closed loop systems are reviewed to ensure that the surge tanks have sufficient capacity to accommodate expected leakage from the system for eeven days or that a seismic source of makeup can be made available within a time frame consistent with the surge tank capacity (time zero starts at low level alarm).
The surge tank and connecting piping are reviewed to ensure that makeup water can be supplied to either header in a split header system.
Redundant surge tanks (one to each header) or a divided surge tank design are acceptable to ensure that in the event of a header rupture, the loss of the entire contents of the surge tank will not occur.
9.2.2-6 Rev. 2
J' 4
d.
The system is designed for removal of heat loads during normal opera-tion and of emergency core cooling heat loads during accident condi-tions, with appropriate design margins to ensure adequate operation.
A comparative analysis is made of the system flow rates, heat levels, maximum temperature, and heat removal capabilities with similar designs previously found acceptable. To verify performance characteristics of the system, an independent analysis may be made.
Design provisions are made that permit appropriate inservice inspec-e.
tion and functional testing of system components important to safety.
The applicant should ensure that the SAR information delineates a testing and inspection program and the system drawings show the necessary test recirculation loops around pumps or isolation valves necessary for this program.
f.
Essential portions of the system are protected from the effects of high-energy and moderate-energy line breaks.
The system description and layout drawings will be reviewed to ensure that no high-or moderate-energy piping systems are close to essential portions of the CWS, or that protection from the effects of failure will be pro-vided. The means of providing such protection will be given in Section 3.6 of the SAR, and the procedures for reviewing this infor-mation are given in the corresponding SRP sections.
g.
Essential components and subsystems (i.e., those necessary for safe shutdown) can function as required in the event of a loss of offsite power and instrument air systems.
The system design will be accept-able in this regard if the essential portions of the CWS meet mini-g' mum system requirements as stated in the SAR assuming a concurrent tV) failure of a single active component, including a single failure of any auxiliary electric power source.
The SAR is reviewed to deter-mine that for each CWS component or subsystem affected by the loss of offsite' power or instrument air systems, system flow and heat transfer capability exceed minimum requirements.
The results of failure-modes and -effects analyses are considered in ensuring that the system meets these requirements.
This will be an acceptable verification of system functional reliability.
The effects of loss of cooling water to RCP seals as a result of loss of power will be reviewed as indicated in Task Action Plan items II.K.2.16 and II.K.3.25 of NUREGs-0718 and 0737.
4.
The system design information and drawings are analyzed to ensure that the following features will be incorporated.
a.
A leakage detection system is provided to detect component or system leakage.
An adequate means for implementing this criterion is to provide sumps or drains with adequate capacity and appropriate alarms in the immediate area of the system.
b.
Components and headers of the system are designed to provide indi-vidual isolation capabilities to ensure system function, control system leakage, and allow system maintenance.
OU 9.2.2-7 Rev. 2
c.
Design provisions are made to ensure the capability to detect leakage of radioactivity or chemical contamination from one system to another.
Radioactivity monitors and conductivity monitors should be located in the system component discharge lines to detect leakage.
An alterna-tive means is to prevent leakage from occurring by operating the system at higher pressure to ensure that leakage is in the preferred direction.
d.
The system is designed to provide cooling to the reactor coolant pump seals and hearings during normal plant operating conditions, antici-pated transients, and following postulated accidents.
Instrumenta-tion in accordance with IEEE 279 with alarms in the control room should be provided to detect a loss of cooling water in order to ensure that a period of 20 minutes is available to the operator to initiate manual protection of the plant, if necessary.
It has been demonstrated by testing that the reactor coolant pumps could poten-tially operate with loss of cooling water for 20 minutes without the need for operator action.
As an alternative to pump testing, the reviewer verifies that:
(1)
Instrumentation in accordance with IEEE 279 is provided consist-ent with the criteria for the protection system to initiate automatic protection of the plant upon loss of water to a pump.
For this case, the component cooling water supply to the seal and bearing of the pump may be designed to nonseismic Category I requirements and Quality Group D, or (2) The component cooling water supply to each pump is designed to be capable of withstanding a single active failure or a moderate-energy line crack as defined in Branch Technical Position ASB 3-1 and to seismic Category I, Quality Group C, and ASME Section III, Class 3 requirenents.
S.
The reviewer verifies that the system has been designed so that system functions will be maintained as required in the event of adverse environ-mental phenomena such as earthquakes, tornadoes, hurricanes, and floods.
The reviewer evaluates the system using engineering judgment and the results of failure-modes and -effects analyses to determine the following:
a.
The failure of portions of the system or of other systems not designed to seismic Category I standards and located close to essential portions of the system, or of non-seismic Category I structures that house, support, or are close to essential portions of the CWS, will not pre-clude essential functions.
The review will identify these nonseismic category components or piping and ensure that appropriate criteria are incorporated to provide isolation capabilities in the event of failure.
Reference to SAR Chapter 2, describing site features, and the general arrangement and layout drawings will be necessary as well as the SAR tabulation of seismic design classifications for structures and systems.
b.
The essential portions of the CWS are protected from the effects of floods, hurricanes, tornadoes, and internally-or externally generated missiles.
Flood protection and missile protection criteria are 9.2.2-8 Rev. 2
r discussed and evaluated in. detail under the SRP sections for Chapter 3
,m
(
\\
of the SAR. The reviewer will use the procedures identified in these
(,/
SRP sections to ensure that the analyses presented are valid. A state-ment to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected or that compo-nents of the system will be located in individual cubicles or rooms that will withstand the effects of both flooding and missiles is acceptable. The location and design of the system, structures, and pump rooms (cubicles) are reviewed to determine that the degree of protection provided is adequate.
6.
The descriptive information, P& ids, CWS drawings, and failure-modes and
-effects analyses in the SAR are reviewed to ensure that essential por-tions of the system will function following design basis accidents assuming a concurrent single, active component failure. The reviewer evaluates the information presented in the SAR to determine the ability of required components to. function, traces the availability of these com-ponents on system drawings, and checks that the SAR information contains verification that minimum system flow and heat transfer requirements are met for each accident situation for the required time spans.
For each case, the design will be acceptable if minimum system requirements are met.
7.
The SAR is reviewed to assure that the applicant has committed to address the potential for water hammer in the auxiliary cooling water systems and will provide means for prevention, or avoidance, such as O
venting and filling capability and operating procedures for avoicfa~nce
)
of water ha'mmer.
~
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and his review supports conclusions of the following type to be included in the staff's safety evaluation report:
The reactor auxiliary cooling water systems include pumps, heat exchangers, valves and piping, expansion tanks, makeup piping, and the points of connection or interfaces with other systems.
Portions of the reactor auxiliary cooling water systems that are necessary for safe shutdown, acci-dent prevention or accident mitigation are designed to seismic Category I and Quality Group C requirements. Based on the review of the applicant's proposed design criteria, design bases, and safety classification for the reactor auxiliary cooling water systems with regard to the requirements for providing adequate cooling water for the safety-related ECCS compo-l nents and reactor auxiliary equipment for all conditions of plant opera-tion, the staff concludes that the design of the reactor auxiliary cool-ing water systems is acceptable and meets the requirements of General Design Criteria 2, 4, 5, 44, 45, and 46. This conclusion is based on the following:
l 1.
The applicant has met the requirements of Ceneral Design Criterion 2 with respect to safety-related portions of the systems being capable l(Ov) meeting Regulatory Guide 1.29, Position C.1 for the safety-related of withstanding the effects of earthquakes. Acceptance is based on portions and position C.2 for the nonsafety-related portions.
9.2.2-9 Rev. 2 I
l
~2.
The applicant has met the requirements of GDC 4 with respect to dynamic efTects associated with' flow instabilities'and attendant
~
loads (i.e., water hammer) with respect to impairment of the
~
required functions of auxiliary coofing systems during normal plant operations,"and under upset or ' accident condit, ions. Acceptance will be based on_the following commitments by the applicant:
a.
Vents shall be provided for venting components and piping at high points in liquid filfid systems which is n'ormally idle and in which voids could occur. These vents should le located for ease of operation and t'es_ ting on a periodfc' basis.
b.
Operating and maintenance procedures shall be reviewed by the applicant to assure that adequate measures are taken to avoid _ water haEImer due to voided fine conditions.
3.
The applicant has met the requirements of General Design Criterion 5 with respect to sharing of structures, systems and components by demonstrating that such sharing does not significantly impair the ability of the reactor auxiliary cooling water systems to perform their safety function, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.
4.
The applicant has met the requirements of General Design Criterion 44 with respect to cooling water by providing a system to transfer heat from structures, systems and components important to safety to an ultimate heat sink. The applicant has demonstrated that the reactor auxiliary cooling water systems can transfer the combined heat load of these structures, systems and components under normal operating and accident conditions assuming loss of offsite power and a single failure, and that portions of the system can be isolated so that the safety function of the system will not be compromised.
5.
The applicant has met the requirements of General Design Criterion 45 with respect to inspection of cooling water systems by providing reactor auxiliary cooling water systems design features which permit inservice inspection of safety-related components and equipment.
6.
The applicant has met the requirements of General Design Criterion 46 with respect to testing of cooling water systems by providing reactor auxiliary cooling water systems design features which permit opera-tional functional testing of the system and its components.
7.
Also in nieeting the requirements of General Design Criterion 44, the applicant has demonstrated that the system can withstand a loss of power without damage to RCP seals in accordance with items II.K.2.16 and II.K.3.25 of NUREGs-0718 and 0737.
V.
IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section.
O 9.2.2-10 Rev. 2
u Except in those cases in which the applicant proposes an acceptable alternative y]
[
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 con-formance with Commission Regulations.
Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced Regulatory Guide and NUREGs.
Implementation,of Item II.2_is as follows:
a_.
Operating plants and OL applicants need not address these revisions _.
b_.
CP applicants will be required to comply with these propos_ed revisions.
VI.
REFERENCES 1.
General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena," of Appendix A to 10 CFR Part 50.
2.
General Design Criterion 5, " Sharing of Structures, Systems, and Compo-nents," of Appendix A to 10 CFR Part 50.
3.
General Design Criterion 44, " Cooling Water," of Appendix A to 10 CFR Part 50.
(
4.
General Design Criterion 45, " Inspection of Cooling Water System," of Appendix A to 10 CFR Part 50.
5.
General Design Criterion 46, " Testing of Cooling Water System," of Appendix A to 10 CFR Part 50.
6.
Regulatory Guide 1.29, " Seismic Design Classification."
l 7.
NUREG-0718 " Proposed Licensing Requirements for Pending Applications for Construction Pernits and Manufacturing License."
8.
NUREG-0737 " Clarification of TMI Action Plan Requirements."
9_.
General Design Criterion 4, "Envir_onmental and Missile Design Basis."
q 9.2.2-11 Rev. 2
NU R EG-0800 (Formsrly NUREG-75/087)
/pa atc I U.S. NUCLEAR REGULATORY COMMISSION w
O (N Ndi$p)s STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION e...e Preposed Revision Standard Review Plan FSRP-10.3, Rev. 3 This proposed revision of the Standard Review Plan and it:, supporting value/ impact statenent and associated technical documentatien have not received a conplete staff review anc approval and cc rot represer.t an official NRC staff position.
The proposed revision to the Standard Revia, s
)
Plan incorporates the resolution of generic issue USI A-1, " Water Hamer."
(v' Public coments are being solicited on the proposed SRP section and the associated value/inpact analysis and technical support docunent NUREG-0927,
" Evaluation of Water Hamer Experience in Nuclear Power Plants" (incleding any implenentation schedules) prior to a final review and decision by the Office of Nuclear Reactor Regulation as to whether this proposed revision should be aproved. Comnents should be sent to the Secretary of the Comission, U. S. Nuclear Rcoulatory Comission, Washington, D. C.
- 20555, Attention:
Docketing and Service Branch.
All connents received t'y July 18
, 1983 will be considered, and all of the associated docunents anc coments considered will be nade publicly avaiictle prior to a decision by the Director, Office of Nuclear Reactor Regulation, on whether to inplenent this revision.
Copies of each of these docunents are available upon written reauest to the Division of Technical Information and Document Control, U. S.
Nuclear Regulatory Comission, Washington, D. C.
20555.
USNRC STANDARD REVIEW PLAN Star dard review plans are prepared for the guidance of the Office of Nuclear Heactor Regulation staf f responsible for the review of i
I applications to construct and operate nuclear power plants. These documents are made available to the public as part of the j {[ ] {
Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review plans are nat substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. Th:
)
[ \\
/
standard review plan sections are keyed to the Standard Format and Content of Safety Analvsis Reports for Nuclear Power Plants.
l V
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 inf orma-tion and experience.
Comments and suggestions for emprovement will be considered and should be sent to the U.S. Nuclear Regulatory Commission, i
Office of Nuclear Reactor Regulation. Washington, D.C. 20555.
i
N U REG-0800 (Form:;rly NUREG 75/087) pe nog %
U.S. NUCLEAR REGULATORY COMMISSION O
(yg/OFFICE OF NUCLEAR REACTOR REG
%; w <
e eeee Proposed Revision 3 to 10.3 MAIN STEAM SUPPLY SYSTEM REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB)
Power Systems Branch (PSB)
Secondary - None I.
AREAS OF REVIEW The main steam supply system (MSSS) for both boiling water reactor (BWR) and pressurized water reactor (PWR) plants transports steam from the nuclear steam supply system to the power conversion system and various safety-related or non-safety-related auxiliaries.
Portions of the MSSS may be used as a part of the heat sink to remove heat from the reactor facility during certain operations and may also be used to supply steam to drive engineered safety feature pumps.
The MSSS may also include provisions for secondary system pressure relief in PWR plants.
ry The MSSS for the BWR direct cycle plant extends from the outermost containment isolation salves up to and including the turbine stop valves, and includes con-nected piping of 2-1/2 inches nominal diameter and larger up to and including the first valve that is either normally closed or is capable of automatic closure during all modes of reactor operation.
The MSSS for the PWR indirect cycle plant extends from the connections to the secondary sides of the steam generators up to and including the turbine stop valves, and includes the containment isolation valves, safety and relief valves, connected piping of 2-1/2 inches nominal diameter and larger up to and including the first valve that is either normally closed or capable of automatic closure during all modes of operation and the steam line to the auxiliary feedwater pump turbine.
The ASB is responsible for the review of the MSSS from the containment up to and including the outermost isolation valve.
The PSB is responsible for the review of the remainder of the MSSS.
(The turbine stop valve review is included in SRP Section 10.2.)
The PSB also determines the adequacy of the design, installation, inspection, and testing of the electrical power supplies for essential components required for proper operation of the MSSS.
The design of the MSSS must be in accordance with General Design Criteria 2, 4, 5, and 34.
Rev. 3 USNRC STANDARD REVIEW PLAN Star.dard review plans are prepared for the guidance of the office of Nuclear Reactor Regulation staff responsible for the review )f apphcations to construct and operate nuclear power plants. These documents are made available to the public as part of the j#\\
Commission's policy to inform the nuclear industry and the general public of regulatory,* ocedures and policies. standard review j
plans are not substitutes for regulatory guides or the Commission 3 regulations and cornpliance with them is not required. The
&(j standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
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 ard 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. 20555.
r 1.
Th ASB and PSB review the MSSS to determine which, ir any, portions of the, system are essential for safe shutdown of the reactor or for preveating or' mitigating the consequences of accidents.
The system is reviewed to ve'rify that; a.
A single malfunction or failure of an active component would not preclude safety-related portions of the system from functioning as required during normal operations, adverse environmental occurrences,
.and accident conditions, including loss of offsite power.
b.
Appropriate quality group and seismic design classification are met forsafety-relatedportions;ofthesystem.
Failures of nonseismic dategory I equipment or structures, or pipe c.
cracks or breaks in high-and moderate-energy piping will not preclude essential functions of safety-related portions of the system, d.
The system is capable of peiforming multiple functions such as trans-porting steam to the power conversion system, providing heat sink capacity or pressure relief cepability, or supplying steam to drive safety system pumps (e.g., turuine-driven auxiliary feedwater pumps),
as may be specified for a particular design.
e.
The design of the MSSS includes the capability to operate ti:e atmo-spheric dump valves remotely from the control room following a safe shutdown earthquake coincident with the loe, of, offsite power so that
- a. cold shutdown can be achieved with depe7dence upon safety grade comnonents only.
f.
The system design capability can withstand adverse dynamic loads, such as steam hammer and relief valve fluid discharge loads 2.
The ASB reviews the MSSS with regard to measures provided to limit blow-down of the system in the event of a steam line break.
3.
The ASB and PSB also review the design of the MSSS with respect to the following:
a.
The. functional capability of the system to transport steam from the nuclear steam supply system as required during air operating conditions.
b.
The capability to detect and control system leakage, and to isolate portions of the system in case of excessive leakage or component malfunctions.
c.
The capability to preclude accidental releases to the environment.
d.
Provisions for functional testing for safety-related portions of the system.
4.
ASB alco performs the following reviews under the SRP sections indicated:
a.
Review for flood protection is performed under SRP Section 3.4.2.
10.3-2 Rev. 3
3 b.
Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1.
A l
c.
Review of the structures, systems, and components to b9 protected-(d against externally generated missiles is performed under SRP Section 3.5.2.
d.
Review of high-and moderate-energy pipe breaks is performed under SRP Section 3.6.1.
In the review of the main steam supply system, the ASB and PSB will coordinate other branches' evaluations that interface with the overall review of the system as follows:
The Reactor Systems Branch (RSB) identifies essential components associated with the portion of the MSSS inside the primary containment that are required for normal operations and accident conditions, establishes shutdown cooling load requirements versus time, and verifies the design transient used in establishing the flow capacity and setpoint(s) of steam generator relief and safety valves as part of its primary review responsibility for SRP Section 5.2.
The Structural Engineering Branch (SEB) determines the accept-ability of the design analyses, procedures, and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural phenomena such as the safe shutdown earthquake (SSE),'the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5.
The Equipment Qualification Brar.cV(EQB) reviews the seismic and environmental qualification of components under SRP Sections 3.10 and 3.11.
The Mechanical Engineering Branch (MEB) determines that the components, piping, and supports are designed in accordance with appli-
,i
,[/
cable codes and standards as part of its primary review responsibility for SRF'
?
(
Sections 3.9.1 throdgh 3.9.3.
The MEB determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2.
The MEB also reviews the adequacy of the inservice testing program of the system valves as part-of its primary, review responsibility for SRP Section 3.9.6.
The Materials
- Engineering Branch (MTEB) verifies, upon request, the compatibility of the, materials of construction with service conditions.
The Instrumentation and Control Systms Branch (ICSB) reviews portions of the MSSS with respect to the adequacy of design, installation, inspection, and testing of essential components necessary for instrumentation 'and control functions as part of its primary review responsibility for SRP Sed Ms '7.1, 7.4, 7.5, and 7.7.
The Procedures and Test Review Branch determine-t 4 acceptability of the preoperational and startup tests as part of its o % < review responsibility for SRP Section 14.0.
The reviews for fi e, et.. inn, technical specifications, and quality assurance are coordinated ar.m jed
-d by the Chemical Engineering Branch, Licensing Guidance Branch, ar.d Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0, and 17.0, respectively.
4 For those areas of review identified above as being part of the primary review responsiblity of other branches, the acceptance criteria necessary for the review and their methods of application are contained in the referenced SRP sections of the corresponding primary branchU O)
\\x, 10.3-3 Rev. 3
II.
ACCEPTANCE CRITERIA Acceptability of the design of the MSSS, as described in the applicant's safety analysis report (SAR), is based on specific general design criteria and regulatory guides.
The design of the MSSS is acceptable if the integrated design of the system is in accordance with the following criteria:
1.
General Design Criterion 2, as related to safety-related portions of the system being capable of withstanding the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, and floods, and the positions of x
the following:
a.
Regulatory Guide 1.29, as related to the seismic design classification of system components, Positions C.1.a, C.1.e, C.1.f, C.2, and C.3.
b.
Regulatory Guide 1.117, as related to the protection of structures, systems, and components important to safety from the effects of tornado missiles, Appendix Positions 2 and 4.
2.
General Design Criterion 4, with respect to safety related portions of the system being capaole of withstanding the effects of external missiles and internally generated missiles, pipe whip, and jet impingement forces associated with pipe breaks, and the position of Regulatory Guide 1.115 as related to the protection of structures, systems, and components impor-tant to safety from the effects of turbine missiles, Position C.1.
The system design should adequately consider steam hammer and relief valve discharge loads to assure that system safety functions can be achieved and should assure that operating and maintenance procedures include adequate precautions to avoid steam hammer and relief valve discharge loads.
3.
General Design Criterion 5, as related to the capability of shared systems and components important to safety to perform required safety functions.
4.
General Design Criterion 34, as related to the system function of transfer-ring residual and sensible heat from the reactor system in indirect cycle plants, and the following:
a.
The positions in Branch Technical Position RSB 5-1 as related to the design requirements for residual heat removal.
b.
Issue Number 1 of NUREG-0138 as related to credit being taken for all valves downstream of the main steam isolation valves (MSIV) to limit blowdown of a second steam generator in the event of a steam line break upstream of the MSIV.
III.
REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to deternine that the design criteria and bases and the preliminary design as set forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II of this SRP section.
For review of operating license (0L) 10.3-4 Rev. 3
lO applications, the procedures are used to verify that the initial design criteria V) and bases have been appropriately implemented in the final design as set forth in the final safety analysis report.
The procedures for OL applications include a determination that the content and intent of the technical specifications prepared by the applicant are in agreement with the requirements for system testing, minimum performance, and surveillance, developed as a result of the LGB review, as indicated in subsection I of this SRP section.
The primary reviewers, will coordinate this review with the other brancnes' areas of review as stated in subsection I of this SRP section. The primary reviewers obtain and use such input as required to assure that this review procedure is complete.
The review procedures below are written for typical MSSSs for both direct and indirect cycle plants.
The reviewer will select and emphasize material from this SRP section, as may be appropriate for a particular case.
1.
There are significant differences in the design of the MSSS for an indirect cycle (PWR) plant as compared to that for a direct cycle (BWR) plant.
Further, different portions of the MSSS are safety-related in different plant designs, although the safety functions of the system are much the same in all PWR plants, and also in all BWR plants. The first step in the review of the MSSS, then, is to determine which portions are designed to perform a safety function.
For this purpose, the system is evaluated (m) to determine the components and subsystems necessary for achieving safe
(/
reactor shutdown in all conditions or for performing accident prevention or mitigation functions.
2.
The reviewer determines that essential (safety-related) portions of the MSSS are correctly identified and are isolable to the extent required'from nonessential portions of the system. The system description and piping and instrumentation diagrams (P& ids) are reviewed to verify that they clearly indicate the physical division between each portion. System arrangement drawings are reviewed to identify the means provided for accomplishing system isolation.
3.
The SEB reviews the seismic design bases and MEB reviews the quality and seismic classification as indicated in subsection I of this SRP section.
The SAR is reviewed by ASB and PSB to verify that essential portions of the MSSS are designed to Quality Group B and/or seismic Category I require-ments, and to verify that the design classifications specified meet the acceptance criteria specified in subsection II of this SRP section.
In general:
a.
The main steam lines from the steam generators to the containment isolation valves in PWR plants are classified seismic Category I and Quality Group B.
b.
The main steam lines in BWR plants extending from the outermost con-
,(O tainment isolatfor valve and connected piping up to and including
)
the first valve that is either normally closed or capable of automatic closure during all modes of normal reactor operations but not including 10.3-5 Rev. 3
the turbine stop and bypass valves are classified seismic Category I and a quality group classification in accordance with BTP RSB 3-1.
Alternatively, for BWRs containing a shutoff valve (in addition to the two containment isolation valves) in the MSSS, seismic Category I and a quality group classification in accordance with BTP RSB 3-2 should be applied to that portion of the MSSS extending from the outermost containment isolation valves up to and including the shutoff valve.
4.
The SAR is reviewed to assure that design provisions have been made to permit appropriate functional testing of system components important to safety.
It is acceptable if the SAR delineates a testing and inspection program and the system drawings show any test recirculation loops or special connections around isolation valves that would be required by this program.
5.
The system description, safety evaluation, component table, and P& ids are reviewed to verify that the system has been designed to:
a.
Provide the necessary quantity of steam to any turbine-driven safety system pumps.
The reviewer verifies that the design is capable of providing the required steam flow to the turbine so that an adequate supply of water can be pumped.
(OL) b.
Assure safe plant operation by including appropriate design margins for pressure relief capacity and setpoints for the secondary system, and for removal of decay heat during various accident conditions, as may be applicable in a particular case.
The review is done on a case-by-case basis, and system acceptability is based on a comparison of system flow rates, heat loads, maximum temperatures, and heat removal capabilities to those of similarly designed systems for previously reviewed plants.
For PWRs the design is reviewed to verify system capability for controlled cooldown to about 350 F to allow actuation of RHR system.
Provide leakage detection means for steam leakage from the system in c.
the event of a steam line break.
Temperature or pressure sensors are acceptable means for initiating signals to close the main steam line isolation valves and/or turbine stop valves to limit the release of steam during a steam line break accident.
d.
Assure that in the event of a postulated break in a main steam line in a PWR plant, the design will preclude the blowdown of more than one steam generator, assuming a concurrent single active component failure.
In this regard, all main steam shut-off valves downstream of the MSIVs, the turbine stop valves, and the control valves are considered to be functional.
The reviewer should verify that the main steam isolation valves, shut-off valves in connecting piping, turbine stop valves, and bypass valves can close against maximum steam flow.
The reviewer verifies that the SAR provides a tabulation and descriptive text of all flow paths that branch off the main steam lines between the MSIVs and the turbine stop valves.
The descript.ive information shall include the following for each flow path:
(1) System identification 10.3-6 Rev. 3
-s p
(2) Maximum steam flow in pounds per hour tVl (3) Type of shut off valve (s)
(4). Size of valve (s)
(5) Quality of the valve (s)
(6) Design code of the valve (s)
(7) Closure time of the valve (s)
(8) Actuation mechanism of the valve (s) (i.e., solenoid operated, motor operated, air operated diaphragm valve, etc.)
(9) Motive or power source for the valve actuating mechanism.
e.
In the event of a main steam line break, termination of steam flow from all systems identified in d, above, except those that can be used for mitigation of the accident, is required to bring the reactor to a safe cold shutdown.
For these systems the reviewer verifies that the SAR describes what design features-have been incorporated to assure closure of the steam shut-off valve (s) and what operator actions, if any, are required.
If the systems that can be used for mitigation of the accident are not available, or the decision is made to use
,- m other means to shut down the reactor, the reviewer verifies that the
)
SAR decribes how these systems are secured to assure positive steam
(./
shut-off and what operator actions, if any, are required.
f.
Assure that in the event of a postulated safe shutdown earthquake in a PWR plant, the design includes the capability to operate atmospheric dump valves remotely from the control room so that cold shutdown can be achieved using only safety grade components, assuming a concurrent loss of offsite power (refer to Branch Technical Position RSB 5-1 attached to SRP Section 5.4.7).
6.
The reviewer verifies that the system is designed so that essential functions will be maintained, as required, in the event of adverse environmental phenomena, certain pipe breaks, or loss of offsite power.
The reviewer uses engineering judgment and the results of failure modes and effect_ analyses to determine that:
a.
Failure of nonseismic Category I portions of the MSSS or of other systems located close to essential portions of the system, or of nonseismic Catagory I structures that house, support, or are close to essential portions of the MSSS, do not preclude operation of the essential portions of the MSSS.
Reference to SAR sections describing site features and the general arrangement and layout drawings will be necessary, as well as the SAR tabulation of seismic design classi-fications for structures and systems.
Statements in the SAR that confirm that the above conditions are met are acceptable.
t(,,)
b.
Essential portions of the MSSS are protected from the effects of floods, hurricanes, tornadoes, and internally and externally generated missiles.
Flood protection and missile protection 10.3-7 Rev. 3
criteria are evaluated under the SRP Section 3 series.
The loca-tions and the design of the system and structures are reviewed to determine that the degree of protection provided is adequate.
A statement to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected, or that components of the system will be located in individual cubicles or rooms that will withstand the effects of winds, flooding, and tornado missiles is acceptable.
Essential portions of the MSSS are protected from the effects of high c.
and moderate energy line breaks and cracks, including pipe whip, jet forces, and environmental effects.
The means of providing such protec-tion will be given in Section 3.6 of the SAR and procedures for reviewing this information are given in SRP Section 3.6.
d.
Essential components and subsystems necessary for safe shutdown can function as required in the event of loss of offsite power.
The SAR is reviewed to verify that for each MSSS component or subsystem affected by a loss of offsite power, the system functional capability meets or exceeds minimum design requirements.
Statements in the SAR and results of failure modes and effects analyses are considered in assuring that the system meets these requirements.
This is an accept-able verification of system functional reliability, 7.
The descriptive information, P& ids, MSSS drawings, and failure modes and effects analyses in the SAR are reviewed to assure that essential portions of the system will function following design basis accidents assuming a concurrent single active component failure.
The reviewer evaluates the analyses presented in the SAR to assure function of required components, traces the availability of these components on system drawings, and checks that the SAR contains verification that minimum requirements are met for each accident situation for the required time spans.
For each case the design is acceptable if minimum system requirements are met.
8_.
The SAR is reviewed to assure that the applicant has committed to address the potential for steam hammer and relief valve discharge loads, and will take adequate procedures action to avoid such occurrence.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and his review supports conclusions of the following type, to be included in the staff's safety evaluation report:
The main steam supply system (MSSS) includes all components and pip-ing from the outermost containment isolation valves (for BWRs) [from the steam generator connection (for PWRs)] up to and including the turbine stop valves.
The essential portions of the MSSS are designed to quality Group B [for PWRs, from the steam generator to the contain-ment isolation valves, and connected piping up to and including the first valve that is normally closed] [for BWRs, from the outermost containment isolation valves and connecting piping up to and includ-ing the first valve that is either normally closed or capable of 10.3-8 Rev. 3
s t
+
(O automatic closure during all modes of normal reactor operation, but
)
not including the turbine stop and bypass valves].
Those portions of-the MSSS necessary to mitigate the consequences of an accident such as a steam line break are designed to the quality standards commensurate with the importance to its safety function, and are designed to the following standards:
The scope of review of the MSSS for the plant included layout drawings, piping and instrumentation diagrams, and descriptive information for the system.
The basis for acceptance of the MSSS in our review was conformance of the applicant's design criteria and bases to the Commission's regula-tions as set forth in the General Design Criteria (GDC) of Appendix A to 10 CFR Part 50.
The staff concludes that the plant design is accept-able and meets the requirements of GDC 2, 4, 5, and 34.
This conclusion is based on the following:
1.
The applicant has met the requirements of GDC 2, " Design Bases for Protection Against Natural Phenomena," with respect to the ability of structures housing the safety-related portion of the system and the safety-related portions of the system being capable of withstanding the effects of natural phenomena such as earth-quakes, tornadoes, hurricanes, and floods and GDC 4 " Environmental and Missile Design Bases" with respect to structures housing the safety-related portions of the system and the safety related portions of the system being capable of withstanding the effects p) of external missiles, and internally generated missiles, pipe (U
whip and jet impingement forces associated with pipe breaks.
The essential portions of the MSSS (as identified in the above discussion) are designed Seismic Category I and housed in a Seismic Category I structure which provides protection from the effects of tornadoes, tornado missiles, turbine missiles, and floods.
This meets the positions of Regulatory Guide 1.29, " Seismic Design Classification," Position C.1.a, C.1.e, C.2 and C.3 or C.1.f, C.2 and C.3; Regulatory Guide 1.115, " Protection Against Low Trajectory Turbine Missiles," Position C.1; and Regulatory Guide 1.117, " Tornado Design Classification," Appendix Positions 2 and 4.
In addition, the system design capabilities should include the capability to accommodate steam hammer dynamic loads resulting from rapid closure of systems valves (including turbine bypass g stop valves), and safety / relief valve operation without compmmising required safety functions.
Operating and mainte-nance procedures are to be reviewed by the applicant to alert plant personnel to the potential for such occurrences and means to j
minimize such occurrences.
This commitment should be stated in the applicants' SAR.
2.
The applicant has met the requirements of GDC 5, " Sharing of Structures, Systems, and Components with Respect to the Capability
('n}
of Shared Systems and Components," important to safety to perform v'
10.3-9 Rev. 3
~
required safety functions. We have reviewed the interconnections from the MSSS of each unit to The interconnections are designed so that the capability to mitigate the consequences of an accident in either unit and achieve safe shutdown in that unit is retained without reducing the capability of the other unit to achieve safe shutdown.
or Each unit of the plant has its own MSSS with no interconnections between the safety-related and/or nonsafety-related portions.
3.
The applicant has met the requirements of GDC 34, " Residual Feat Removal," with respect to the system function of transferring residuel and sensible heat from the reactor system in PWR plants.
The f1SSS is capable of providing heat sink capacity and pressure relief capability and supplying steam to the steam driver safety-related pumps necessary for safe shutdown. The liSSS is also designed to include the capability to operate the atmospheric pump valves remotely from the control room following a safe shutdown earthquake coincident with the loss of offsite power so that a cold shutdown can be achieved with dependence upon safety-grade components only.
This meets the positions in Branch Technical Position RSB 5-1,
" Design Requirements of Residual Heat Removal System," and in Issue 1 of NUREG-0138.
V.
Il1PLEliEf4TATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans fcr using this SRP section.
Except in there cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Conmission's regulaticns, the method described herein will be used by the staff in its evaluation of con-formance with Commission regulations.
Implemenation schedules for confornance to parts of the method discussed herein are contained in the referenced regulatory guides and NUREG.
Implementation of Item II.2 is as follows:
Operating pl_an_ts and OL applicants. need not address th,ese revisicns.
a.
b.
CP applicants,will be required to comply with these prop _osed revisions.
O 10.3-10 Rev. 3
.o VI.
REFERENCES 73(V.
i 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 4, Environmental
-and Missile Design Bases."
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 34, " Residual Heat Removal."
5.
Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants."
6.
Regulatory Guide 1.29, " Seismic Design Classification."
7.
Regulatory Guide 1.115, " Protection Against Low-Trajectory Turbine Missiles."
8.
Regulatory Guide 1.117, " Tornado Design Classification."
9.
Branch Technical Positions ASB 3-1, " Protection Against Postulated Piping Failures in Fluid Systems Outside Containment," attached to SRP Section 3.6.1,
[O Branch Technical Position MEB 3-1, " Postulated Break and Leakage Locations
()
in Fluid System Piping Outside Containment," attached to SRP Section 3.6.2.
10.
Branch Technical Position RSB 3-1, " Classification of Main Steam Components Other than the Reactor Coolant Pressure Boundary for BWR Plants," attached to SRP Section 3.2.2.
11.
Branch Technical Position RSB 3-2, " Classification of BWR/6 Main Steam and Feedwater Components Other Than the Reactor Coolant Pressure Boundary,"
attached to SRP Section 3.2.2.
- 12. Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Rerraval System," attached to SRP Section 5.4.7.
l 13.
NUREG-0138, " Staff Discussion of Fifteen Technical Issues Listed in Attach-ment to November 3,1976, memorandum from Director NRR to NRR Staff."
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10.3-11 Pev. 3 1
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NU R EG-0800 (Form rly NUREG-75/087) a arc
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U.S. NUCLEAR REGULATORY COMMISSION nU Q<if i STANDARD REV EW PLAN k
..e 8
OFFICE OF NUCLEAR REACTOR REGULATION 0
- e Proposed Revision Standard Review Piar PSRP-10.4.7, Rev. 3 This proposed revision of the Standard Review Plan and its supr.orting value/inpact statement and associated technical docunentation have not received a complete staff review and approval and do not represent an
/
official f4RC staff position.
The proposed revision to the Standard Review
(",)/
Plan incorporates the resolution of generic issue USI A-1, " Water Hamer."
Public coments are being soliciteo on the proposed SRP section cod the associated value/ impact ana!ysis and technical suppert document flVREG-0927,
" Evaluation of Water Harner Experience in Nuclear Power Plants" (including any inplenentation schedules) prior to a final review and decision by the Office of Nuclear Reactor Regulaticn as to whether this proposed revisien should be aproved.
Comments should be sent to the Secretary of the i
Commission, U. S. Nuclear Regulatory Comission, Washington, D. C.
- 20555, Attention:
Docketing and Service Branch. All corrents received by July 18, 1983 will be considered, and all of the associated docunents and comments considered will be made publicly evailable prior to a decision by the Director, Office of Huclear Reactor Regulation, on whether to inplenent
(
this revision.
Copies of each of these docunents are available upon written i
reouest to the Division of Technical Information and Docunent Control, U. S.
l Huclear Regulatory Comnission, Washington, D. C.
20555.
l l
l l
USNRC STANDARD REVIEW PLAN Stir.dard review plans are prepared for the guidance of the office of Nuclear Reactor Regulation staff responsible for the review of fx applications to construct and operate nuclear power plants. These documents are made available to the public as part of the l
Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review l ((
/
plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The v'
standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the Standard Format have a corresponding review plan.
i Published standard review plans will be revised periodically, as appropriate, to accommodate comments and to reflect new informa-I 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. 20555.
~
NU REG-0800 (Form:rly NUREG-75/087)
@ MGg (3
f h
U.S. NUCLEAR REGULATORY COMMISSION U
(M i STANDARD REV EW PLAN N"
8 OFFICE OF NUCLEAR REACTOR REGULATION 0
- aee*
Proposed Revision 3 to 10.4.7 CONDENSATE AND FEEDWATER SYSTEM REVIEW RESPONSIBILITIES Primary - Auxiliary Systems Branch (ASB)
Secondary - None I.
AREAS OF REVIEW The condensate and feedwater system (CFS) provides feedwater at the required tempera-ture, pressure, and flow rate to the reactor for boiling water reactor (BWR) plants and to the steam generators for pressurized water reactor (PWR) plants.
Condensate is pumped from the main condenser hotwell by the condensate pumps, passes through the low pressura feedwater heaters to the feedwater pumps, and then is pumped through the high pressure feedwater heaters to the nuclear steam supply system.
n hD)
ASB reviews the CFS from the condenser outlet to the connection with the nuclear steam supply system and to the heater drain system to assure conformance to General Design Criteria 2, 4, 5, 44, 45 and 46.
For indirect cycle plants, there are also interfaces with the secondary water makeup system and the auxiliary feedwater system.
The CFS is used for normal shutdown.
The only part of the CFS classified as safety-
- related, i.e.,
required for safe shutdown or in the event of postulated accidents, is the feedwater piping from the steam generators for PWRs and from the nuclear steam supply system for BWRs, up to and including the outermost containment isola-tion valve.
1.
The ASB reviews the characteristics of the CFS with respect to the capability to supply adequate feedwater to the nuclear steam supply system as required for normal operation and shutdown.
2.
The ASB review determines that an acceptable design has been established for:
a.
The interfaces of the CFS with the auxiliary feedwater system (PWR), the reactor core isolation cooling system (BWR), and the condensate cleanup system with regard to functional design requirements and seismic design classification.
Rev. 3 USNRC STANDARD REVIEW PLAN Star.dard review plans are prepared for the guidance of the office of Nuclear Reactor Regulation staff responsible for the review of
/
i applications to construct and operate nuclear power plants. These documents are made available to the public as part of the
{
J Commission's policy to inform the nuclear industry and the general public of regulatory procedures and policies. Standard review
"'y plans are not substitutes for regulatory guides or the Commission's regulations and compliance with them is not required. The s
standard review plan sections are keyed to the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants.
Not all sections of the standard Format have a corresponding review pian.
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. Washmgton, D.C. 20555.
b.
The feedwater system (PWR), including the auxiliary feedwater system piping entering the steam generator, with regard to possible fluid flow instabilities (e.g., water hammer) during normal plant operation as well as during upset or accident conditions.
c.
The detection of major system leaks that could affect the functional performance of safety-related equipment.
3.
ASB also performs the following reviews under the SRP sections indicated:
(a) Review for flood protection is performed under SRP Section 3.4.1, (b) Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1, (c) Review of the structures, systems, and components to be protected against externally generated missiles is performed under SRP Section 3.5.2, and (d) Review of high-and moderate-energy pipe breaks is performed under SRP Section 3.6.1.
The ASB will coordinate evaluations performed by other branches that interface with the overall evaluation of the system as follows:
The Reactor Systems Branch (RSB) determines that transients resulting from feedwater flow control malfunctions will not violate the primary system pres-sure boundary integrity criterion as part of its primary review responsibility for SRP Sections 15.1.1 through 15.1.4, and that the loss of normal feedwater flow will not violate the fuel damage criterion or the system pressure boundary integrity criterion as part of its primary review responsibility for SRP Section 15.2.7.
The Power Systems Branch (PSB) evaluates the system power sources with respect to their capability to perform safety-related functions during normal, transient, and accident conditions as part of its primary review responsibility for SRP Section 8.3.1.
The Structural Engineering Branch (SEB) determines the accepta-bility of the design analyses, procedures, and criteria used to establish the ability of seismic Category I structures housing the system ar.d supporting systems to withstand the effects of natural phenomena such as the safe shutdown earthquake (SSE), the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5.
The Mechanical Engineering Branch (MEB) determines that the components, piping and structures are designed in accordance with applicable codes and standards as part of its primary review responsibility for SRP Sections 3.9.1 through 3.9.3.
The MEB determines the acceptability of the seismic and quality group classifications for system components as part of its primary review responsibility for SRP Sections 3.2.1 and 3.2.2.
The MEB also reviews the adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6.
Upon request, the MEB determines the acceptability of design analyses, procedures, and criteria used to establish the adequacy of devices or restraints as they may relate to significant water hammers in system piping and the MEB reviews test programs of components that may be affected by water hammers.
The Materials Engineering Branch (MTEB) verifies that inservice inspection requirements are 10.4.7-2 Rev. 3
.s i
[]
net for system components as part of its primary review responsibility for SRP
("/
Section 6.6 and, upon request, verifies the compatibility of the materials of construction with service conditions. The review for Fire Protection, Technical Specifications, and Quality Assurance are coordinated and performed by the
' Chemical Engineering Branch, Licensing Guidance Branch, and Quality Assurance Branch as part of their primary review responsibility for SRP Sections 9.5.1, 16.0,'and 17.0, respectively. The Equipment Qualification Branch (EQB) reviews the seismic qualification of Category I instrumentation and electrical equipment and the environmental qualification of mechanical and electrical equipment as part of its primary review responsibility for SRP Sections 3.10 and 3.11, respectively. Upon request, the Instrument and Control Systems Branch (ICSB)-
will review the instrumentation and controls associated with the feedwater control system (BWR) or steam generator level control system (PWR).
For those areas of review identified above as being 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 sections of the corresponding primary branches.
II. ACCEPTANCE CRITERIA Acceptability of the condensate and feedwater system, as described in the applicant's safety analysis report (SAR), is based on the specific requirements i
of General Design Criteria and the positions of regulatory guides.
Listed below are the specific criteria as they relate to the CFS.
/%
/CJ 1.
General Design Criterion 2, as related to the system being capable of with-standing the effects of earthquakes. Acceptance is based on meeting the guidance of Regulatory Guide 1.29, Position C.1 for safety-related portions, and Position C.2 for nonsafety-related portions.
2.
General Design Criterion 4, as related to the dynamic effects associated with possible fluid flow instabilities (e.g., water hammers) during normal plant operation as well as during upset or accident conditions (pWRs enly). Acceptance is based on meeting the guidance contained in the attached Branch Technical Position ASB 10-2 on reducing the potential for water hammers in steam generators with-tep-feedving-designs.
3.
General Design Criterion 5, as related to the capability of shared systems and coc.gonents important to safety to perform required safety functions.
4.
General Design Criterion 44, as it relates to:
a.
The capability to transfer heat loads from the reactor system to a heat sink under both normal operating and accident conditions, b.
Redundancy of components so that under accident conditions the safety function can be performed assuming a single active component failure.
(This may be coincident with the loss of offsite power for certain events.)
c.
The capability to isolate components, subsystems, or piping if required v
so that the system safety function will be maintained.
10.4.7-3 Rev. 3
5.
General Design Criterion 45, as related to design provisions to permit periodic inservice inspection of system components and equipment.
6.
General Design Criterion 46, as related to design provisions to pemit appropriate functional testing of the system and components to assure structural integrity and leak-tightness, operability and performance of active components, and capability of the integrated system to function as intended during normal, shutdown, and accident conditions.
III. REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to determine that the design criteria and bases and the preliminary design as set forth in the preliminary safety analysis report meet the acceptance criteria given in subsection II of this SRP section.
For the review of operating license (0L) applications, the procedures are used 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 primary reviewer will coordinate this review with the areas of review of interfacing branches as stated in subsection I of this SRP section.
The primary reviewer obtains and uses such inputs as required to assure that this review procedure is complete.
The reviewer will select and emphasize material from this SRP section as may be appropriate for a particular case.
The SAR is reviewed to determine that the system description and diagrams delineate the function of the condensate and feedwater system under normal and abnormal conditions. The reviewer verifies the following:
1.
The system has been designed to function as required for all modes of operation.
The results of failure modes and effects analyses presented in the SAR, if any, are used in making this determination.
2.
The system piping is designed to preclude hydraulic instabilities from occurring in the piping for all modes of operation. As appropriate, the reviewer evaluates the results of model tests and analyses that are relied on to verify that water hammer will not occur, or proposed tests of the installed system that are intended to verify design adequacy.
Steam generators that use top feed designs are reviewed in accordance with Branch Technical Position ASB 10-2.
The feedwater control valve design and controller shall be verified to be compa,tible with s'y's' tem (s,), imposed oper'ating conditions fe'.'g., control functions required, range of control and pressure drop characteristics,
~
~
valve stroke, trim, etc. ).
Te'st, data or operatin'g' e,xperience data 's'h'a'll be used where available.
In, addition, the applicant has committed to reviewJ ant operating a_nd maintenance procedures to assure that i
precautions for avoidance of steam / water, hammer and water hammer _
occurrences,h_ ave been provided.
3.
The outermost containment isolation valves and all downstream piping to the nuclear steam supply system are designed in accordance with seismic Category I requirements. The review for seismic design is performed by 10.4.7-4 Rev. 3 l
l
s%
SEB and the review for seismic and quality group classification is per-(/
formed by MEB as indicated in subsection I of this SRP section.
4.
The CFS design is such that the plant can be safely shut down using the auxiliary feedwater system or the reactor core isolation cooling system, if required.
5.
The CFS design, or other plant systems, provide the capability to detect and control leakage from the system.
6.
The reviewer verifies that the essential portion of the system has been designed so that system function will be maintained as required in the event of adverse environmental phenomena or loss of offsite power. The review for protection against natural phenomena is performed in the Chapter 3 SRP sections. The reviewer evaluates the system, using engineer-ing judgment and the results of failure modes and effects analyses, to determine that the failure of nonessential portions of the system or of other systems not designed to seismic Category I standards and located close to essential portions of the system, or of nonseismic Category I structures that house, support, or are close to essential portions of the CFS, will not preclude operation of the essential portions of the CFS.
IV.
EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and his
/7 review supports conclusions of the following type, to be included in the staff's
(
)
safety evaluation report:
The condensate and feedwater system includes all components and equipment from the condenser outlet to the connection with the nuclear steam supply system and to the heater drain system [ secondary water makeup system, and auxiliary feedwater system interfaces. (PWRs only)].
Based on the review of the applicant's proposed design criteria, the l
design bases, and safety classification for the safety-related por-tions of the condensate and feedwater system and the requirements for system performance for all conditions of plant operation, the staff concludes that the design of the condensate and feedwater system and supporting systens is in conformance with the Commission regula-tions as set forth in General Design Criterion 2, 4, 5, 44, 45 and
- 46. This conclusion is based on the following:
1.
The applicant has met the requirements of General Design Criterion 2 with respect to safety-related portions of the system being capable of withstanding the effects of earthquakes by meeting Regulatory Guide 1.29 Position C.1 for the safety-related portions and Position C.2 for the nonsafety-related portions.
2.
The applicant has met the requirements of General Design Criterion 4 with respect to the dynamic effects associated with possible fluid flow instabilities (e.g.,.iater hammers) by having ry the feedwater system designed in accordance with the guidance
(
)
contained in Branch Technical Position ASB 10-2 and thereby il eliminating or reducing the possibility of water hammers in steam generators (PWRs only).
10.4.7-5 Rev. 3
That the applicant has adequately addressed feedwater control valve and con' troller, f unctions wiTh respec,t to water hammer
~
potential and the app'licant has committed to review operating and maintenance procedures"to assume that prec'autions taken will _
minimize, or avold, water h'ahimers.
~~
~
3.
The applicant has met the requirements of General Design Criterion 5 with respect to the capability of shared systems and components important to safety to perform required safety functions. We have reviewed the interconnections of the CFS between each unit. The interconnections are designed so that the capability to mitigate the consequences of an accident in either unit and achieve safe shutdown in that unit is retained without reducing the capability of the other unit to achieve safe shutdown.
4 The applicant has met the requirements of General Design Criterion 44 with respect to cooling water by providing a redundant and isolable system capable of transferring heat loads frcri the reactor system to a heat sink under both normal opera-ting and accident conditions.
The applicant has demonstrated that the condensate and feedwater systen can provide sufficient cooling water to transfer the heat load of the reactor system under normal operating conditions and accident conditions assuming loss of offsite power and a single failure and that portions of the system can be isolated so that the safety function of the system will not be compromised.
5.
The applicant has met the requirements of General Design Criterion 45 with respect to inspection of cooling water systems by providing a feedwater system design that permits inservice inspection of safety-related components and equipment.
6.
The applicant has met the requirements of General Design Criterion 45 with respect to testing of cooling water systems by providing a feedwater system design that permits operational functional testing of the safety-related portion of the system and its components.
The staff concludes that the design of the CFS conforms to all applicable GDCs and positions of the regulatory guide cited and is, therefore, acceptable.
V.
IMPLEMENTATION The following is intended to provide guidance to all applicants and licensees regarding 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 guide.
10.4.7-6 Rev. 3
[/i Implementation of Item II.2 is as follows:
\\
a.
Operating plants and OL applicants need not address these revisions.
b.
CP applicants will be required to comply with these proposed revisions, c.
It should be noted that steam generators in operating plants and NT0L's where a SER has been issued, have been backfitted to comply with the rev,ised BTP ASB 10-2.
VI. REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena."
2.
10 CFR Part 50, Appendix A, General Design Criterion 4, " Environmental and Missile Design Bases."
3.
10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures Systems and Components."
VI. REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for (f 's)
Protection Against Natural Phenomena."
w./
2.
10 CFR Part 50, Apperdix A, General Design Criterion 5, " Sharing of Structures, Systems, and Components."
3.
10 CFR Part 50, Appendix A, General Design Criterion 44, " Cooling Water."
4.
10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cooling Water System."
5.
10 CFR Part 50, Appendix A, General Design Criterion 46, " Testing of Cooling Water System."
6.
Regulatory Guide 1.29, " Seismic Design Classification."
7.
Branch Technical Position ASB 10-2, " Design Guidelines for Water Hammer in Steam Generators with Top Feedring Designs."
(v 10.4.7-7 Rev. 3
BRANCH TECHNICAL POSITION ASB 10-2 DESIGN GUIDELINES FOR AVOIDING WATER HAMMERS IN STEAM GENERATORS BACKGROUND Plant operational experience has shown that top-feed steam generators containing feedwater spargers with bottom drain holes incur steam condensation induced water hammers.
This type of water hammer has frequently occurred after the feedwater sparger was uncovered (due to some plant transient) and cold auxiliary feedwater flow was subsequently initiated.
The initiation of the auxiliary feedwater flow into the steam generator produces a water slug in the sparger or feedwater piping, which is then accelerated by the unbalanced pressures produced by the condensation of a steam pocket in the line.
The resultant impulse could be of a sufficient magnitude to cause damage to the steam generator internal components and feedwater systems piping.
The most damaging of such water hammer incidents occurred at Indian Point No. 2 in 1973, where the water hammer loads resulted in rupture of an 18-inch feedwater pipe and damage to the containment inner liner.
The repeated occurrence of such water hammers and potential severity such flow instabilities resulted in the NRC in engaging Creare Inc. in 1976 to evaluate causes and effects, aid to develop recommendations for avoidance of top feed steam generator water hammer, and design methods minimize associated dynamic loads.
The underlying causes of water hammer in top-feed steam generators were extensively studied by Creare, Inc. who reported findings and recommended design modifications to minimize or preclude such water hammar occurrence in NUREG-0291 (1977).
These recommendations called for:
(a) use of J-tubes on the feedring to minimize loss of water when uncovered, (b) early initiation of auxiliary feedwater to keep piping and feedring full of water, (c) short horizontal FW pipe lengths at the SG nozzle to reduce magnitude of slug formation and impact, (d) limit FW recovery flow rates to less than 150 gpm/SG to minimize steam-water entrainment and subsequent formation of a water slug.
The design and operational modifications were implemented by plants experiencing SG water hammer and appear to have essentially eliminated SGWH.
NUREG-0918 details plant specific modifications which were made.
In addition, experience sustains maintaining preoperational tests to verify the absence of SGWH.
More recently, Westinghouse and Combustion Engineering have introduced steam generators of the preheat type, wherein the majority of feedwater enters the steam generator at the bottom through a preheater section.
The potential for condensation-induced water hammer in preheat steam generators was studied by BNL and reported in NUREG/CR-1606, "An O
10.4.7-8 Rev. 3 i
)
Q]
[
Evaluation of Condensation-Induced Water Hammer in Preheat Steam Generators,"
June 1980.
This report, citing the lack of definitive experimental and analytical results, recommended full scale verification tests to demonstrate the absence of damaging water hammer in preheat steam generators and connecting feedwater piping (i.e., preoperational tests).
B&W steam generators, which are a "once through" flow design, have generally not reported water hammer occurrence.
However, in May 1982, several B&W plants (following inservice inspection) reported damaged internal auxiliary feedwater headers and-support structures.
The cause was attributed to steam pocket collapse.
The internal auxiliary feedring design concept is similar to CE & W top feedring concepts which have experienced water hammer before corrective design measures were implemented.
For these B&W plants, the OTSG's are being modified to return-to the previous design using auxiliary feedwater injection manifolds which are external to the steam generator.
The staff believes that SGWH evidence and studies performed to date warrant the establishment of design guidelines for steam generators and the associated piping.
Guidelines have been developed that may be used to reduce the probability of a damaging steam condensation induced water hammer, particularly for the Westinghouse and Combustion Engineeering PWR designs which use top-feed steam generators.
BRANCH TECHNICAL POSITION In CP and OL application reviews, the staff requires the applicant to provide
(
the following design capability and verification:
LJ Top-Feed Steam Generator Designs To eliminate or reduce possible water hammer in the feedwater system:
a.
Prevent or delay water draining from the feedring following a drop in steam generator water level by means such as J-Tubes.
b.
Minimize the volume of feedwater piping external to the steam generator which could pocket steam using the shortest possible (less than seven feet) horizontal run of inlet piping to the steam generator feedring.
c.
Perform tests acceptable to NRC to verify that unacceptable feedwater hammer will not occur using the plant operating procedures for nromal and emergency restoration of steam generator water level following loss of normal feedwater and possible draining of the feedring.
Provide the procedures for these tests for approval before conducting the tests and submit the results from such tests.
Preheat Steam Generator Designs a.
Minimize the horizontal lengths of feedwater piping between the steam generator and the vertical run of piping by providing downward turning elbows immediately upstream of the main and auxiliary feedwater nozzles.
10.4.7-9 Rev. 3 t
b.
Provide a check valve upstream of the auxiliary feedwater connection to the top feedwater line.
c.
Maintain the top feedwater line full at all times.
d.
Perform tests acceptable to NRC to verify that unacceptable feedwater hammer will not occur using plant operating procedures for normal and emergency restoration of steam generator water level following loss of normal feedwater.
Also perform a water hammer test at *% of power by using feedwater through the auxiliary feedwater (top) nozzle at the lowest feedwater temperature that the plant standard operating procedure (50P) allows and then switching the feedwater at that temperature from the auxiliary feedwater nozzle to the main feedwater (bottom) nozzle by following the SOP, and submit the results of such tests.
Once Through Steam Generator (OTSG) Designs Provide auxiliary feedwater to the steam generator through an externally a.
mounted supply header.
b.
Perform tests acceptable to NRC to verify that unacceptable feedwater hammer will not occur using the plant operating procedures for normal and emergency restoration of steam generator water level following loss of normal feedwater.
Provide the procedures for these tests for approval I
before conducting the tests, and submit the results of such tests.
REFERENCES (1) Block, J. A. et.al., "An Evaluation of PWR Steam Generator Water Hammer," NUREG-0291, June 1977.
(2) Chapman, R.
L., et.al., " Compilation of Data Concerning Known and Suspected Water Hammer Events in Nuclear Power Plants," NUREG/CR-2059, May 1982.
(3) Anderson, N. and Han, J.
T., " Prevention and Mitigation of Steam Generator Water Hammer Events in PWR Plants," NUREG-0918, December 1982.
^The power level at which feedwater flow is transferred from the auxiliary feedwater nozzle to the main feedwater nozzle.
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