ML19221B105
| ML19221B105 | |
| Person / Time | |
|---|---|
| Issue date: | 03/31/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-75-087, NUREG-75-087-05.4.7, NUREG-75-87, NUREG-75-87-5.4.7, SRP-05.04.07, SRP-5.04.07, NUDOCS 7907120433 | |
| Download: ML19221B105 (16) | |
Text
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STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTICN 5.4.7 RESIDUAL HEAT REMOVAL (RHR) SYSTEM REVIEW RESPCNSIBILITIES Primary - Reactor Systens Branch (RSB)
Secondary - Auxiliary Systems Branch (ASB)
Core Performance Branch (CPB)
Containment Systems Branch (CSB)
Instrumentation and Control Systems 3 ranch (ICSB)
Structural Engineering Branch (SEB)
I.
AREAS OF REVIEW The residual heat removal (RHR) system is used in conjunction witn the main steam and feedwater systems (main condenser), or the reactor &
,lation cooling (RCIC) system in conjunction with the safety / relief valves in a bu sing water reactor (BWR), or auxiliary feedwater system in conjunction with the atmospheric dump valves in a pres-surized water rec tor (PWR) to cool down the reactor coclant system following shutdown.
Parts of the RHR system also act to provide low pressure emergency core cooling and are reviewed as described in SRP Section 6.3 Some parts of the RHR system also provide containment heat removal capability and 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, 4, 5, 34, 44, 45, 46, 55, 56 and 57.
Both PWRs and BWRs have RhR systems which provide long-tern cooling once the reactor coolant temperature has been decreased by the main condans r, RCIC, or auxiliary feed-water systems. In both types of plants, the RHR is typically a low 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 designs 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 sucticn 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 removed in the heat exchangers is transported to the ultimate heat sink by the componelt cooling water or service water system.
In PWRs, the RHR system is also used to fill, drain, and remcVe 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 l
spray.
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" *" I 7907120 63
The RHR system in BWRs is typically composed of four subsystems. The containment 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) sub-systems 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 condensir.g 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 inoperable have been proposed f er 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 l
and temperature to the cold depressurized condition. This effort is divided between the RSB and the ASB as follows 1.
For BWRs, the R58 reviews the processes and systems used in the cooldown of the reactor for the entire spectrum of po*ential reactor coolant system pressures and temperatures during decay heat removal.
2.
For PWRs, the R58 reviews the approach used to meet the functional requirements of BTP RSB 5-1 with respect to cooldown to the conditions permitting operation of j
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 ASB has primary review responsibility for the review of the pertinent portions of the CVCS (SRP 9.3.4), the atmospheric dump valves (SRP 10.3), and the source for auxiliary f eedwater (SRP 10.4.9) for conformance to BTP RSB 5-1.
The RSB reviews the pressurizer relief valve and ECCS, it used.
In addition, the RSB reviews the tests and supporting analysis concerning mixing of borated water and cooldown under natural circulation as required in BTP RSB 5-1, 3.
Fcr both PWRs and BWRs, the ASB reviews the component cooling or service water systems that transfer decay heat from the RHR system to the ultimate heat sink as described in SRP Sections 9.2.1 and 9.2.2.
Rev. 1 5.4.7-2 j
4.
The RSB reviews the design and operating characteristics of the RhR systera with respect to its shutocan and long-term cooling function. Wnere the RHR system 9
interfaces with other systers (e.g., RCIC system, component cnoling water system) the effect of these systems on the RHR system is reviewed.
Overpressure protec-tion provided by the vwlving < tween the RCS and RHR system is also reviewed.
The proposed RHR system preoperational and initial sta +ue te3L programs are reviewed and the proposed technical specifications are evaluated in &qard to limiting condi-tions of operation and periodic surveillance testing.
The RHR r.ystem is reviewed to assure tnat it has the proper seismic and qua'ity group classifications. This aspect of the review is performed as a portion of the effort described in SRP Sections 3.2.1 and 3.2.c.
The RHR system ic to be enclosed in a stincture having the proper seismic classification. The review is done as a part of the effort described in SRP Section 3.2.2.
The RHR system shall be enclosed in a seismic Category I structure or building. The design adequacy of this structure or building is evaluated by SEB as described in SRP Secticns 3.3, 3.4, 3.5, 3.7 and 3.8.
The C58, as described in SRP Section 6.2.4, reviews the design of the RHR system to see that it is compatible with the function of the containment and that adequate isola. ion capabilities are provided.
The ICSB, as described in SRP Section 1.4, reviews motor operated valve controls, interlocks, sensors for interlocks, position indicators, and power sources.
ICSB determines that the interlocks on motor-operated valves used as barriers between the high and low pressure RHR piping are suitably indepandent and diverse and that trip signals close the valves when the pressure is too high.
The MEB, as described in SRP Section 3.9.3, reviews the design and installation of the RHR system to see that applicable code requirements are met.
The MIEB reviews the materials and inservice inspection program for the RHR system, as described in SRP Sections 6.1.1 and 6.6.
The CPR reviews the core decay energy output on which the design is based to see that it is applicable and suitably conservative.
Ihe MEB, as described in SRP Section 3.6.2, reviews the effects of pipe breaks both in and outside containment to confirm that requirements for protection of the reactor shutdown systems are met.
The ASB, as described in SRP Section 3.6.1, reviews the plant oesign fo protection against piping failures outside containment to assure that 9
could be safely shutdown in the event of c'Ih failures.
such failures would not cause loss of needed safety related systems and that the plant h
Rev. I 5.4.7-3
II.
ACCEPfANCE CRITERIA The general objective of the review is to determine that the system or systems employed to remove residual heat meets the requirements of General Design Criterion (GDC) 34 (Ref 4) reprding residual heat removal, and GDC 19 (Ref. 16) regarting operabilitv from the cor. trol r.
9.
As noted to subsection i, reactor ;oo:d'wn covers a w.de range of potential reactor coolant temperatures and pressuree A portion this ranc,e is reviewed by the ASB for PWRs.
This SRP section deals specifically with the areas of review perf or med by the Rs3.
The acceptance criteria are discussed in the following paragraphs The system or systems must be capable of performing the function of transferring heat from the reactor to the environment using only safety grade systems. The system (s) and the system (s) enclosure must be in conformance with GDC 2 (Ref. 1) and GDC 4 (Ref. 2),
and should conform to the rec'_mmendations of Regulatory Guide 1.26 (Ref. 17),
Regulatory Guide 1.29 (Ref. 12), Regulatory Guide 1.46 (Ref. 13), and the staff positions on protection against piping failures outside containment (Ref. 15).
The system (s) are to satisfy the functional, isolation, pressure relief, pump protec-tion, and test requirements specified in Branch Technical Position RSB 5-1 (Ref. 11).
Interfaces between the RHR system and RCIC and component or service water systems should be designed so that operation of one dces not interfere with, and provides proper support (where required) for, the other.
In relation to these and other shared syst e s (e.g., emergency core cooling and coritainment heat removal systems), the RPR system mu t conform to GDC 5 (Ref. 3).
Component cooling and service water systems removing heat from the RHR beat exchangers must conform to GDC 44, 45 and 46 (Refs. 5, 6 and 7).
Contairment isolation provisions for the RHR system must conform to GDC 55, 56 aad 57 (Refs. 8, 9 and 10).
In addition to the above criteria, the acceptability of the RHR syste ay be based on the degree of design similarity with previously approved plants. Dew!ations from these criteria fron other types of RHR systems (e.g., systems that are designed to withstand reactor coolant system operating pressure or systems located entirely inside contain-ment) will be considered on an individual basis.
For those areas of review identified in subsection I of this SRP section as being the responsibility of other branches, the acceptance criteria and their methods of applica-tion are contained in the SRP sections corresponding to those branchos.
III. REVIEW PROCEDURES lhe procedures below are used during the construction permit (CP) review to assure that the design criteria and bases and the pr eliminary design as set forth in the Prelim-inary Safety Analysis Report meet the acceptance criteria given in subsection II.
147 159 Rev. 1 5.4.7-4
For operating license (OL) reviews, toe procedures are utilized to verify that the initial design criteria and tases have been appropriately implemented in the final 9
desiga a, 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 testir.g.
As noted in subsections I and II, the RSB review for PWRs is limited to the low pressure low temperature RHR system. For RWRs, the review is to incloce all of the systems used to transfer residual heat from the reactor over the entire range of poten-tial 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 t;e 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 as 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 instrumenta-tion 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 conform-ance to the single fai'ure 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, compari-sons 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. 11) 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. 11).
The reviewer checks the isolation valves in 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.
5.
Using the system process diagrams, P& ids, failure modes and effects analysis, and component performance specifications, the reviewer determines that the system (s) has the capacity to bring the reactor to cold shutdown conditions in a reasonable period of time, assuming a single failure of an active component with only either 147 M0 S.4.7-5 Rev. 1
onsite or offsite electric power available. f or the purposes of the review, 36 nours is considered a reasonable time period. The reviewer consults with the CPB to confirm that the proper core decay energy output was assumed for the analy-sis.
The ASB is responsible for the review of the initiel 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 safely grade equipment is to be assumed.
6.
The cooldown function is to be reviewed to determine if it can be performed from the control room assuming a single failure of an active component, with only either onsite or offsite electric power available. 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.
7.
By reviewing the system description and the P& ids, the reviewer confirms the RHR system satisfies the pressure relief requirements of Branch Technical Position R5B S-1 (Ref. 11).
8.
The reviewer checks the P&lDs to see that essential components of the systems employed to remove residual heat are designated seismic Category I and Safety Class II (the cocling water side of heat exchangers can be Safety Class III).
Based on st:tements made in SAR Section 5.4.7 or on the reviews made by other branches, the R5B reviewer confirms that the RHR system meets the requirements of GDC 2 (Ref. 1) and GDC 4 (Ref. 2), and conforms to the recommendations of Regula-tory Guides 1.29 (Ref. 12) and 1.46 (Ref. 13) and the staff positions on piping failures outside containment.
9.
By reviewing the pioing arrengement and system description of the RHR system, the reviewer confirms that the RHR system meets the requirements of GDC 5 (Ref. 3) concerning shared systems.
10.
The R5B reviewer contacts ihe ASB reviewer in conjunction with his review of the RHR system heat sink and refueling system interaction to interchange information and assure that the reviews are consistent in regard to the interfacing para-meters.
For example, the ASB review determines the maximum service or component cooling water temperature. The R58 review:r then reviews the RHR system descrip-tion to determine that this maximum temperature has been allowed for in the RHR system design, 11.
The RSB reviewer contacts his counterpart in the IC58 to obtaia any needed infor-mation from their review.
Specifically, ICSB confirms that automatic actuation and remote manual valve :.ontrols are capable of perf orming the f unctions required, and that sensor and monitoring provisions are adequate. The instrumentation and controls of the RHR system are to have sufficier.t redundancy to satisfy the single failure criterion.
heV. I b.4.7-6
\\47 \\6\\
12.
The RSB reviewer contacts his counterpart in CSB so that the information needed concerning their reviews will be interchanged.
13.
The applicant's proposed initial startup test programs are reviewed to determine that they are consistent with the requirements of Branch Technical Position RSB 5-1 (Ref 11) and with the intent of Regulatory Guide 1.68 (Ref. 14).
At the OL stage, the RSB reviewer assures that sufficient information is provided by the applicant to identify the test objectives, methods of testing, and test acceptance criteria (see paragraph C.l.b of Regulatory Guide 1.68).
The RSB reviewer evaluates the proposed initial test programs to determine if they provide reasonable assurance that the RHR system will perform its safety function.
As an alternative to this detailed evaluation, the reviewer may compare the RHR system 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 RHR system.
If the RHR system differs significantly from that of previously reviewed designs, the impact of the proposed changes on the required 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 14.
The prcposed plant technical specifications are reviewed to:
a.
Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restrictions 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 desigr 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 (Ref. 11).
17.
The RSB reviewer provides information to other branches in those areas where the RSB has a review responsibility that is not (xplici tly covered in steps 1-16 above. These additional areas of review responsibility include:
a.
Identification of engineered safety features (ESF) and saft shutdowa electri-cal loads, and verification that the minimum time intervals for the connec-tion of the ESF to the standby power system', are satisfactory.
5.4.7-7 Rev. I
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 asseriated with the main steam supply and the auxiliary feedwater system that are required to operate during and following shutdown.
18.
Upon request from the primary reviewer, other secondary 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 assure that this review procedure is complete.
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.
For PhRs "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 useJ to reduce the reactcr coolant system temperature and pressure to values that permit operation of the RhR system.
The review of the initial cool-down phase is discussed in Section of the SER.
The review cf the RHR system operational phase is discussed below.
The residual heat removal (RHR) system removes core decay heat and provioes long-term core cooling following the initial phase of reactor cooldown. The sccpe of review of the RHR system for the plant included piping and instrumentation diagrams, equipment layout draw-ings, failure modes and effects analysis, and design performance specificatiors 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.
"The drawings, component descriptions, design criteria, and supporting analyses associated with the RHR system have been reviewed and have been found to ccnform to Commission regulations, regulatcry guides, and staff technical positions. The RHR system has been found to conform ta General Design Criteria 2, 4, 5, 19, 34, 44, 45, 46, 55, 56 and 57, to Regulatery Guides 1.26, 1.29, 1.46 and 1.68, and to Branch fechnical Position RSB 5-1.
The System was found capable of performing its shutdown cooling functiona frcm the control room with only onsite electrical power or offsite power available, assumirg the most restrictive single active component failure. It was also found that two independent and redundant barriers are always b 4 7-0 Pev. 1 147 163
in place between the reactor coolant systems (RCS) and RHR system whenever the RCS pressure is higher than the RHR system design pressure.
"The staff concludes that the design of the resviual heat removal system conforms to the Commission's regulations, and to applicable *egulations, guides and staff positions, and is acceptable."
For EWRs "The residual heat removal function is accomplished in two phases: the ;nitial 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 saf ety/ relief valves.
The low pressuae-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 compo-nents of the RCIC and RHR system are available to bring the reactor to cold shut-down conditions.
"The scope of review of these systems for the plant included piping a1d instrumentation diagrams, equipment layout drawings, failure mode and effects analysis, ar.d dasign performance specifications for essential components The review has included the applicant's proposed design criteria and desic,n 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 drawings, component descriptions, design criteria, and supporting analyses associated with the systems employed to remove residual heat from the reactor have been reviewed and have been found to conform to Commission regulations, regulatory guides, and staff technical positions. These systems have been found to conform to General Design Criteria 2, 4, 5, 19, 34, 44, 45, 46, 55, 56 and 57, to Regula-tory Guides 1.26, 1.29, 1.46 and 1.68, and to Branch Technical Position RSB 5-1.
The system was found capable of performing its shutdown cooling functions from the control room with only onsite electrical power or offsite power available, assum-ing the most restrictive single active component failure. It was also found that two independent and redundant barriers are always in place between the RCS and RHR system whenever the RCS pressure is higher than the RHR system design pressure.
"The staff concludes that the des *gn of the systems employed to remove residual heat from the reactor conform to the Commission's regulations and to applicable regulatory guides and staff technical positions, and is acceptable."
V.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 2, " Design Bases for Protec-tion / gainst Natural Phenomena."
h 5.4.7-9 Rev. 1
2.
10 CFR Part 50, Appendix A, General Design Criterion 4, " Environmental and Missile Design Bases."
3.
10 CFR Part 50, Aopendix 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.
10 CF R Part 50, Appendix A, cn.a1 Design Criterion 44, " Cooling Water."
6.
10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cooling Water System."
7.
10 CFR Part 50, Appendix A, General Design Criterica 46, " Testing of Cooling Water System."
8.
10 CFR Part 50, Appendix A, General Design Criterion 55, " Reactor Coolant Pressure Boundary Penetrating Cor*ainment."
9.
10 CFR Part 50, Appendix A, General Design Criterion 56, " Primary Containment Isolation."
10.
10 CFR Part 50, Appendix A, General Design Criterion 57, " Closed Syst Isolation Valves."
11.
Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System," attached to SRP Section 5.4.7.
12.
Regulatory Guide 1.29, " Seismic Design Classification."
13.
Regulatory Guide 1.46, "Protecticn Against Pipe Whip Inside Cont ainment."
14.
Regulatory Guide 1.68, " Initial Test Programs for Water-Cooled Reactor Power Plants."
15.
Branch Technical Positions ASB 3-1, " Protection Against Postulated Piping Failures in fluid Systems Outside Containment,' attached to SRP Section 3.6.1, and MEB 3-1,
" Postulated Break and Leakage Locations in Fluid System Pipino Outside Containment," attached to SRP Section 3.6.2.
16.
10 CFR Part 50, Appendix A, General Design Criterion 19, " Control Room."
17.
Regulatory Guide 1.26, " Quality Group Classifications and Standards for Water, 5 team, and Radioactive-Waste-Containing Corrponents of Nuclear Power Plants."
O 147 165 Rev. 1 5.4.7-10
O BRANCH TECHNICAL PCSIf!CN RSB 5-1 Of SIGN RE CllTREME NT S OF THF RISIOUAL HEAT REMOVAL SYSTEM BACKGROUND GDC 19 states that, "A control room shall t>e provided f rom which actions can be taken to operate the nuclear power unit onder normal conditions.
Normal operating conditions include the shutting dosn of a reactor; therefore, since the residual heal removal (RHR) system is one of seseral 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 a oure that for onsite electrical power system operation (assumin; offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available), the system safety function can be acccmplished, assuming a single failure."
.'n 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 cure cooling system (ECCS).
However, it is possible for the RHR system to have different design characteristics. For example, the PHR system might have the same design pressure as the RCS, or be located inside of containment. Plants which may have PH9 systems that deviate from curre..c 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 included in this position.
BRANCH FOSITION A.
Functional Requirements The system (s) which can be used to take the reactor from normal operating condi-tions to cold shutdown
- shall satisfy the functional requiremants listed below.
1.
The design shall be such that the reactor can be taken from normal operatin s
conditions to cold shutdown
- Using only safety grade systems. These systems shall satisfy General Design Criteria 1 through 5.
IProcesses involved in cooldown are heat removal, depressurization, flow circulation, and reactivity ccntrol. The cold shutdown condition, as described in the Standard Technical Specifications, refers to a subcritical reactor with a reactor coolant temperature no greater than 200 F fo' a PWR and 212 F for a BWR.
147 166 5.4.7-11 Rev. 1
2.
The system (s) shall have suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilitics 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 tingle failure.
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 suitabiy 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.
E.
RHR System Isolation Requirements The RHR system shall satisfy the isolation requirements listed below.
1.
The following shall be provided 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 belnw the RHR system desira pressure. Failure of a power supply shall not cause any valve to 4
change position.
(c) The valves shall have indeoendent diverse interlocks to pratect against ene or both valves being open during an RCS increase above the desic.
pressure of the RHR systen..
2.
One of the following shall be provided on the discharge side of the RHR systen to isolate it from the RCS:
(a) The valves, positian indicators, and interlocks described in item 1(a) - (c),
(b) One or more check valves in series with a normally closed power-operated valve.
Tie power-operated valve position shall be indicated in the control room.
If the RHR,ystem discharge line is used for an ECCS function, the power-operated valve is to be opened upon receipt of a safety injection signal once tre reactor coolant pressure has decreased below the ECCS design pressore.
\\A7 \\6i Re /. 1 5.4.7-12
(c) Three check valves in series, er (d) Two cneck valves in series, provided that there are design provisicas to permit periodic testing of the check valves for leak tightness and the testing is perfore.ed at least annually.
C.
Pressure Relief Requirements The RHR system shall s3tisfy the pressure relief requirements listed below.
1.
To pr)tect 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 wh?n selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a PWR with ro steam bubble in the nrc su
,ar, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be ccasiderJd in selection of he design bases.
2.
Fluid discharged through the RHR system pressure relief valves must be col-lected anc.ontained such that a stuck open relief valve will not:
a.
Result in flooding of any safety-related equipment, b.
Reduce the car ability of the ECCS below thct needed to mitigate the consequences if a pcstulated LOCA.
c.
Result in a non " latable situation in which the water provided to the RCS to maintsir, thi core in a safe conoition is discharged outside of tha containment.
3.
If interlocks are provided to automatically close the isolation valves when the RCS pressure exceeds the RHR system design reessure, adequate relief capacity shall be orovided during the time perica while the valves are closing.
D.
Pump Protection Requirements The design and operating procedures of any RHR system shall have provisions tu prevelt damage to the RHR system pumps due to overheating, cavitation or loss of adequate pump suction fluid E.
Test Requir0ments The isolation valve operability and interlock circuits must be designed so as to pernit on line testing when operating in the RHR mode.
Testability shall meet the requirements of IEEE Stanoard 338 and Regulatory Guide 1.22.
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5.4.7-13 Pev. 1
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 cooldown can be achieved under natural circulation conditions and permit estimatico 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 be substituted for these tests.
F.
Operational Procedures The operational procedures for beinging 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.
O.
Auxiliary Feedwates Supply The seismic Category I water supply for the auxiliary feedwater system for a P4R 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 />, ollowed by cooldown to tha conditions permitting operation of the RHR nventory needed for cooldown shall be based on the longest couldown systea.
The i 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 capa-bility 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 compiiance.
Class 2 Partial implementat.cn of this position for all plants (custom or standard) for which CP or PDA 1ppl1 Cations are docketed 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 ; mi
.-nt s ' ion guidance in Table I will be backfitted for all oper s and all other plants (custom e
or standard) for wh ths vL is expected before t W combined I&E and DOR review January 1, 1979,s
.1 of related plant fea uu tor operating ructors.
Pev. 1 5.i.7-14 g
If> t f 1 Possible (51uticns for full Cr. pliance with EIP t t 5-1 ard Recorer
- d c lev ntatice for Class 2 riants Desinn Fecairerents F ro-:e's an t t i vs t' Fcssible Solut,on for RecorrerN Ir rlenint3ticn for of Bir csR F-1 or r
r,< nont'--
full -<
x~
pliece r.13 s,
e' plants (see ' ate 11
!. Fsoctien 31 Pm 2ir erent fer Lana-terr ccolirq /pp; er, l iro '
d Prey tde doutle drop line (or valves curelicce will n:t s;irej if Taking to Cold Snatdcwn ir ;3rallel) to prevent sirqle valve it can be sPcwn th u cticn fcr
- a. Capcility Using Only Safety f3ilJre fruc stm ping H cooling single f ailure by ranai ntions Grade System
~
?"rcticn.
(Note:, inis requirm ent ins 11 c r cutside et ccnt mert er s
in conjunction with neetinq effects return to N t starc y until N 'u31
- b. Capability with eitrer cnly c' single failure for lona-tern actions (cr re; airs) are fe e d s'
ensite or enl:, offsite power c oo l i n,1 and isolation res; ire ~ents te acce table fcr the individ;31 ar.d with single failure in,olve inc reased nur er of (licit. Etion catside Ck to giant.
ind ;eernt,c er supplim and ;!assibly reet Si )
ccre tun fn;r valwes.)
- c. Re.isenable tire for ccoldown assu~ino
>t liniting SF and only offsite or only onsite pos r Feat re". oval anj PE circulation f rc vide safety-grade d e n valves, Compliance required.
durino coold^wn to cold shutdown.
c; <. ra t ors, a n] power supply, etc. so (Note: Need SG coolino t o riaint.iin that manu.sl action should not be reauired RC'. circulation esen after - RS in atter SSE except to rieet single f ailure, coeration when under natural circula-tion) / steam darp valves Depressurization /Frassurizer f revide upgradinq and a-Jitional Cc"pliance will nct be rovirei if auxiliary spray or power-c;.erated valves to ensure op ration of da -
a) dependerce cn manual actions relief valves iliary pressurizer spray using only insicc conta inrer.t af ter SSE cr
~
saf ety-grade subsysten roeting single single failure cr b) re aininq at f ailure, rossibic alternative ray insolve het standbv until ranual acticcs using pressurizer rower-cperated relief or repairs are corrlete are salves which have t en uograded. Meet fcund to be acceptable for the SSE and sinqle failur e without manual individual slant.
cLeration inside cc-tainrent.
Foration for cold shatdown /CW 5 Previde procedure and unnrading where Same as above.
and boron sampling /
necessary such th.it t cration to cold shutdcwn concentration reets the requirements of I. Solution could rance fron (1) uoqradina and adding
~?
ulves to have t:oth letdown and N
cnarqira catbs safety crade and reet sinqle failure to (2) use of backup procedures involvinq less cost. For exr ple, baration without letdewn ray s
be aueptable and eliminate need for uparadinq letoown path, Use of ECCS for C
injection of t, ora ted wa t er may al so be acceptable, eed surseillance of baron cercentration (bororometer and/or samplirg Limited operator action inside or outside of containnent if justified.
5.4.7-13
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