ML20065S171

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Forwards Addl Info Re Leak Detection Instrumentation Sys, in Response to NRC 820409 Request.Info Will Be Incorporated Into Future PSAR Amend
ML20065S171
Person / Time
Site: Clinch River
Issue date: 10/26/1982
From: Longenecker J
ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To: Check P
Office of Nuclear Reactor Regulation
References
HQ:S:82:115, NUDOCS 8211010121
Download: ML20065S171 (39)
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Department of Energy Washington. D.C. 20545 Docket No. 50-537 HQ:S:82:ll5 OCT 2 01982 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Check:

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

Reference:

Letter, P. S. Check to J. R. Longenecker, "CRBRP Request for Additional Information," dated April 9,1982 This letter formally responds to your request for additional infor-mation contained in the reference letter.

Enclosed are responses to Questions CS ~421.36 and 42; which will be incorporated into a future PSAR amendment.

Sincerely, J n R. Longen ker Acting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy 2 Enclosures cc: Service List Standard Distribution Licensing Distribution 8211010121 821026 Nf PDR ADOCK 05000537 A PDR

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- ~ ~ " rese ~~O'ot-vopu tt,~2rau Ouestion CS421.36 Provide a more detailed discussion of the CRBR Leak Detection system and how it meets the provisions contained in the Light Water Reactor Regulatory Guide 1.45. The discussion should include detection methods, detector sensitivity, detector response time, signal correlations and calibration, seismic qualification, testabiiIty, and the provisions for technical spectfications.

Resoonser PSAR Section 7.5.5.1.1 has been revised to provide a more detailed discussion of the CRBRP Leak Detection Instrumentation System. A comparison to the provisions of Regulatory Guide 1.45 is contained in Section 5.3 of WARD-O-185,

" Integrity of the Primary and Intermediate Heat Transport System Piping in Containment", (Ref erence 2 of PSAR, Section 1.6).

Technical Specifications will be developed at the FSAR stage. The Technical Specification will require that the plant will be placed in either the hot shutdown or refueling condition if there is a confirmed leak in either the primary or Intermediate heat transport system.

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QCS421.36-1 Amend. 72 l Oct. 1982

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7.5.5.1.1 Design Bases and Design Criteria For the Liquid Metal-To-Gas Leak Detection System The design bases of the Liquid Metal-to-Gas Leak Detection System arises f rom the need to protect plant equipment, considerations of maintenance and plant availability, and the corrosion ef fects of sodium compounds on stalnless steel s at high temperatures.

Considering the signif icance of corrosion with respect to piping integrity, it is appropriate that the design criteria assure that the Liquid Metal-to-Gas Leak Detection System provide reliable detection f or the Primary and Intermediate Heat Transport in-Containment Systems in a small fraction of the nominal time to penetrate the pipe by local corrosion. The ef fects of corrosion on the CRBRP RiTS piping have been thoroughly assessed in WARD-0-185

" Integrity of the Primary and Intermediate Heat Transport System Piping in-Containment," Ref erence 1.6 of the PS AR. In summary, leaks of 100 gm/hr may cause local corrosgon in 3600 hrs and general corrosgon in 18,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at temperatures near 1000 F. At temperatures less than 700 F, the corrosion rate becomes extremely slow. The Leak Detection System will detect leaks of 400 gm/hr in pipes and components operating at temperatures greater than 700 F in less than 250 hrs.

Design Criteria have been established to guarantee reliable plant operation with pipe temperatures greater than 700 F. These include:

1. The PHTS and in-containment IHTS shall be monitored for leaks by

, diverse methods each capable of providing the required time response.

2. Capability shall be provided to procure a filter sample f or laboratory analysis to provide a highly reliable confirmation method. Filter samples should be analyzed a minimum of once every 1000 hrs.
3. The Liquid Metal-to-Gas Leak Detection System must operate af ter an l operating basis earthquake (OBE).
4. The leak detection system shall be equipped with provisions to readily permit testing f or operabil ity and cal ibration during plant operation.
5. A reliable sel f-monitoring provision shall be provided to detect component f ail ure.

7.5-18a Amend. 72 Oct. 1982

6. Upon loss of ability to fulfill the specified time response, the plant will be placed in a hot shutdown condition.
7. The system shall be quallfled to operate in its environment.

Additional Design Criteria of the Liquid Metal-To-Gas Leak Detection System required to protect plant and capital investment, limit maintenance. and protect plant availability are outlined below:

1. The Liquid Metal-To-Gas Leak Detection System shall detect and locate l iquid metal-to-gas l eaks tgroughout the pl ant between the temperatures of 375 to 1000 F as required to f ul fill continuous monitoring requirements of Appendix G, "CRBRP Plan For Inservice and Preservice inspections."
2. The Leak Detection System shall be able to identify the general location of the leak.

This wstem is not needed for initiation of plant shutdown, for removal of decay heat or for reduction of of f-site radiation exposure to acceptable level s; theref ore, it is classified as a non-safety system. The safety related instrumentation provided to accommodate lIquid metal leaks is described in Section 7.5.3.1.1. The passive engineered safety features provided to mitigate the ef fects of liquid metal leaks are described in Section 3.8.

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1 7.5-18b Amend. 72 l Oct. 1982

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7.5.5.1.1.1 Design Descrfotion General Detection equipment is provided to monitor the primary and intermediate sodium coolant boundaries to identify comparatively small leaks when they occur.

The leak detection methods selected f or the following . installations are:

1. Particulate monitors (radiation detectors), Sodium lonization Detectors (aerosol detectors), and chemical analysis f or atmosphere monitoring in selected cells.
2. Plugging filter aerosol detectors (PFADs) for Main Heat Transfer System piping and guard vessels, major components, and for Inerted cell atmosphere monitoring.
3. Contact detectors in the space between the bellows and the stem packing of the bellows sealed sodium valves.
4. Cable detectors in guard vessels and under major IIquid metal component s.

Of the types of leak detection devices that comprise the Leak Detection System, only sodium aerosol leak detection devices show a dif ference in their response when operated in an air atmosphere as opposed to an inert atmosphore.

The time f or a detector to respond to a leak in air is generally shorter than in an inert atmosphere. The electrical sensing types such as cable and contact detectors show no dif ference in response due to operating atmospheres.

However, the potential for higher moisture content in air can result in greater inhibition to sodium flow when the leak is very small.

Considerations which materially af fect detection times include: sodium leak

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rate, sodium temperature, and cell size. Test data (See Reference 5) conf inn that sodium leaks of 100 gm per hour in an air or inert ahno' sphere can be l detected by aerosol detection over the operating temperature ranges, within j the detection time periods identified in Figure 5.1.1 of WARD-D-0185,

" integrity of Primary and Intermediate Heat Transport System Piping in Conta i nment", (Ref erence 2, PSAR Section 1.6). Larger leaks (on the order of kg/ min) will be readily detected by two or more systems in minutes.

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7.5-19 Amend. 72 Oct. 1982

e The electrical sensing types of detectors (cables and contact) respond with an alarm when IIquid metal causes an electrical short between the electrode and l Its protective sheath. The Sodium lonization Detectors (SIDs) pgido an

{ alarm when the aerosol concentration reaches a level of about 10 gm/cc.

The PFADs, which are integrating devices, respond with an alarm when the dif ferential pressure across a filter has increased by 2 inches of water. The time for this response is relaty to aerosol concentration as shown on Figure 7.5-7. For example, at 1 x 10' gm/cc, the time response is appro'ximately 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />. Both SIDs and PFADs have filters which are chemically examined for sodium on a monthigbasis so that leaks which result in aerosol concentrations lower than 1 x 10' gm/cc w il l al so b detoded. A leak m u m ng in a concentration of approximately 2 x 10'y3 gm/cc is detectable by chemical examination of the filter pads. The sodium aerosol concentration resulting from a 100-gm/h leak in inerted CRBRP celis ranging in volume f rom 15,000 to 115,000 ft 3 is shown on Figure 7.5-8. In the operating temperature range of 700-1000 F, the leak deiection criteria are easily met with either SIDs or PFADs. In addition, during reactor operation, the radiation particulate monitoring systemgili detect leaks resulting in aerosol concentration of approximately 10' gm/cc in those cells containing primary sodium.

The aerosol detectors are connected to the PDH&DS so that the rate at which the signal is changing.can be checked af ter a leak alarm is obtained. A rapid increase in PFAD dif ferential pressure (less than I hour from normal reading to alarm) accompanied by leak alarms f rom other detectors in the same area would Indicate a large leak (greater than I gpm). Conversely, a leak signal that took 10 to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or more to reach the alarm level would Indicate a smal I (100-1000 gm/h) leak. The SIDs are cal Ibrated so that aerosol concentration can be related to the signal level. Instruments are set to alarm at specif ic aerosol concentrations. The I Iquid metal-to-gas leak I detection system is designed to f unction af ter an OBE. The radiation

particulate monitoring system is designed to f unction af ter an SSE. All leak detection equipment wilI be tested periodically to demonstrate operability.

The increase in cell atmosphere temperature and pressure in the event leaks Iarger than 20 kg/ min as detected by temperature and pressure sensors can provide an additional source of leak detection.

The abil ity to detect small leaks ( 100 gm/hr) by several methods in hours I plus the ability to detect Iarge ieaks (>kg/ min) in minutes w11I provide a highly rellable leak detection system that provides the operator information to enable shutdown to repair def ects without extensive time for cleanup operations.

Af ter a sodium or NaK leak has occurred, the Liquid Metal-to-Gas Leak Detection System equipment impacted by the leak will be either replaced or cleaned (pneumatic system rinsed with alcohol) to remove sodium leak residue products. The system will then be acceptance tested and calibrated in accordance with the preoperational test specification criteria utilized prior to inital plant startup.

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l l 7.5-20 Amend. 72 j Oct. 1982

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Table 7.5-3 gives the primary and back-up methods of leak detection f or the principal sodium systems and components in the plant. The methods shown in the table are related to the three sizes of leaks defined in Section 7.5. 5 .1. 2. The principal methods of leak dctection are described below.

Aerosol Monitoring Aerosol monitoring will be perf ormed by measuring the pressure drop across a membrane filter with a constant flow of gas sampled f rom the annular space between major piping and its Insulation, from the space within guard vessels, and from cells containing IIquid metal systems. Another cell aerosol monitoring method uses a sodium ionization detector. Liquid Metal aerosols or vapor are ionized by a hot filanent and the ion current is measured.

Increases in the ion current Indicate a leak.

Based upon the experimental results, these methods provide f or detection of leaks of 100 gm/hr and less, with a response time depending on temperature and the vol ume being monitored.

The major f unction of this Instrumentation will be t'o provide Indication of the presence of small leaks which do not present a significant contamination hazard, but which might result in undesirable long-term corrosion.

Contact Detectors (Scark-Plug)

Contact detectors consist of a stainless-steel-sheathed, mineral oxide-

, insulated, twe-wire probe with the sensing end open and the wire ends exposed.

Contact detectors are installed, for example, on bellows sealed valves with the sensing end between the bellows and the mechanical backup seal. A leak is detected by the reduction in circuit electrical resistance caused by sodium contacting the wire ends.

Cable Detectors Cable detectors consist of stainless-steel-sheathed, mineral-oxide-Insulated, cable with holes penetrating the sheath to permit leaked liquid metal to come in contact with the conductors. Cabl e detectors w il l be pl aced, f or exampl e, in the bottom of guard vessels and below large tanks.

Other Detection Methods Pressure and temperature measurements available in the Inerted cells (Section 9.5.1.5) will provide immediate Indication of the presence of large leaks over the 20 kg/ min size. In the case of systems containing radioactive sodlum, the detection of airborne radioactivity arising f rom Na-24 or Na-22 in the aerosols will be perf ormed by particulate radiation monitoring equipment (Section 11.4.2) which proviggs a sensitive detection method f or aerosol concentrations as low as 10 gm/cc.

Chemical analysis provides positive '-

concentrations of approximately 10 )petection capabil Ity for aerosolgm/cc, depending period.

7.5 -21 Amend. 72 Oct. 1982 m .

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Liquid Sodium Level Sensors in the reactor, the EVST, the IHTS expansion tank,

and sodium storage tanks will provide Indications of large leaks. Smoke detectors (Fire Protection System) will detect combustion products originating from sodium leaks in air (See Section 9.13.2). .

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I 7.5-21a t

Amend. 72 l Oct. 1982

Indication in Control Room An audible group alarm is sounded in the control room upon indication of a leak or certain f ailures of contact, cable, or aerosol channels. The channel number producing the alarm and the location of the region covered by this channel are displayed on an annunciator on a local panel. This information will identify the leak as occurring in a specific major comonent or series of pipe sections, or specific bellow-sealed valve, or the cell containing the l eaking system. The leak detection system uses the Plant Data Handling System f or channel failure monitoring, data and trend logging; the sampl Ing time Interval will nominally be approximately 30 seconds.

No automatic isolation f unctions or reactor scram are initiated by the Liquid Metal-To-Gas Leak Detetton System. Isolation or shutdown of a system showing a leak will be performed manually, following verification of the leak and review of the operating conditions.

7.5.5.1.2 Deslan Analysis The Liquid Metal-to-Gas Leak Detection System will meet the appropriate requirements of CR3R Design Criterion 30, " Inspection and Surveillance of Reactor Coolant Boundary and Criterion 33, " Inspection and Surveillance of Reactor Cool ant Boundary. Criterion 30 requires that means be provided for detecting and Identifying the location of the source of reactor coolant leakage from the reactor coolant boundary to the extent necessary to assure that timely discovery and correction of leaks which could lead to accidents whose consequences could exceed the limits prescribed for protection of the

, health and saf ety of the publ Ic. Criterion 33 requires that means be provided f or detecting intermediate coolant leakage from the Intermediate coolant boun.t.y. In order to denonstrate how the Intent of the criteria will be satisfied, the Instrumentation requirements met by this system f or three dif ferent ranges of leaks are discussed. These ranges have been selected to analyze situations which cover the complete range of leak detection instruments. Section 15.6 discusses the consequences of leaks f or the health and saf ety of the publ ic.

Laroe Leaks

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This category covers f ailures up to those resulting in a leak of 30 gpm or 100 kg/ min. A significant physical characteristic of leaks of this size is that they would result in pressure and temperature changes in the primary cells if l the leak occurs in PHTS pipe sections. This feature sets the lower boundary l of the leak at about 20 kg/ min; this being an estimte of the amount of sodium l which would result in measurable changes In cell pressure and temperature. If the leak occurs in a guard vessel, continuity detectors will provide detection l

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! 7.5-22 Amend. 72

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of these large leaks. Leaks of this msgnitude would be detected in five minutes or less for the primary and intermediate heat transport system. The operator would then be able to Initiate and complete plant shutdown within ten minutes after the start of the leak.

The pressure and temperature measurements available in the inerted cells will, in conjunction with the aerosol detectors, continuity detectors and radiation monitors, provide the response required for proper operator action in case of leaks of this magnitude.

Intermediate Leaks intermediate leaks were defined as those leaks which would not result in significant changes in cell pressures and temperatures but where the extent of the resulting contamination and plant maintenance makes plant shutdown desirable. The range of leak rates covered extends from the lower limit of the large leaks previously considered down to a leak of 100 gm/hr. The detection times f or the wide range of leaks in this group would vary from a few minutes to several hours depending on the rate of leakage. Based upon experimental results, it is concluded that several systems would detect a leak of this magnitude in several hours at least and possibly in minutes.

Instrumentation capable of detecting leaks of this magnitude include radiation monitors, continuity detectors, and the dif ferent types of aerosol detectors.

l Small Leaks Small leaks at or below 100 gm/hr were defined as those events resulting in releases of sodium which do not pose a contamination or maintenance problem but might result in undesirable long-term corrosion (see Section 5.3.3). The methods for detecting leaks of this range are aerosol detectors and radiation monitors in the case of the primary system.

In the course of test programs, aerosol concentrations produced by leaks of down to 5 gm/hr were found to be within the detection capability of both a Sodium lonization Detector and a Plugging Filter Aerosol Detector in test l chambers. The test results show that leaks of this s.lze can be detected in l

the range of one hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by annull monitors depending upon the sodium l temperature and gas environment. It is deduced from the test results that very small leak (<1 gm/hr) will be detected by annuli monitors in several days.

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l 7.5-23 Amend. 72 l Oct. 1982

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't i 1 Tests during 1975 and 1976 showed that under environmental conditions typical of LMFBR operation, small leaks from typical piping configurations can be detected by both Sodium lonization and Plugging Filter Aerosol Detectors.

Continuity (cable or contact) detectors did not reliably detect small pipe i

leaks under these conditions. Testing in 1978 verified the performance of aerosol detectors using prototypic CRBRP cell atmosphere recirculation as well

as pipe / Insulation design.

It is deduced from the test results that the sodium vapor / aerosol systems will, in conjunction with existing radiation monitoring technology, provide adequate indication of the smallest sizes of leaks of Interest.

i Sodlum Leaks Into an Air Atmosohere Test results indicate that the methods applicable to sodium leaks in inerted cells will also operate when applied in an air atmosphere. The additional use of smoke detectors and the accessibility of piping located in an air atmosphere to visual inspection assist in the selection of an ef fective sodium-to-air leak detection system.

7.5.5.2 Intermediate to Primarv Heat Transoort Svstem Leak Detection 7.5.5.2.1 Design Descriotion The IHTS pressure is maintained at least 10 psi higher than the Primary Heat Transport System at the IHX to prevent radioactive primary sodium from entering the IHTS in the event of a tube leak. Maintaining a positive pressure differential across the IF0( is a limiting condition for operation of the plant (Chapter 16 - Technical Specifications). This provides assurance that a zero or negative dif ferential will not exist during any extended Interval. A loss of this pressure or a reversai of it is not expected to occur except during accident conditions. Such an occurance would necessitate an orderly plant shutdown to correct the problem. Sinco a reverse

, dif ferential cannot occur for a significant interval. the potential leakage of primary sodium into the intermediate system, through an lHX tube leak, is small.

Leakage of primary sodium into the IHTS, should it occur. will be detected by radiation monitors provided on the IHTS piping within the SGB. The radiation monitor system will provide an Indication of the radiation level and will provide alarms f or conditions of excessive radiation inficative of Ingress of primary sodium. Since the only activity expected in the ! HTS is a low level of tritium, the radiation monitors will be very sensitive to the presence of significant amounts of radioactive primary sodium in the intermediate system.

For accidents which involve a loss of IHTS boundary integrity the radiological ef fects have been evaluated. The results of these evaluations are presented in Sections 15.3.2.3, 15.3.3.3 and 16.6.1.5.

7.5-24 Amend. 70 Aug. 1982

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References to Section 7.5

1. Ford, J. A., "A Recent Eval uation of Foreign and Domestic Wastage Data f rom Sodium Water Reaction Investigation", APDA CTS-73-05, January,1973.
2. Morejon, J. A., " Sodium-to-Gas Leak Detection Mockup Tests",

N707-TR-520-004, September 17, 1975. (Atomics International)

3. Greene, D. A., J. A. Gudahl and J. C. Hunsicker, " Experimental Investigation of Steam Generator Materials by Sodium-Water Reactions, Vol ume 1, GEAP-14094, January 1976.
4. Gudahl, J. A. and P. M. Magee, "Microleak Wastage Test Results",

GEFR-00352, March 1978.

5. Matl in, E., Witherspoon, J. E., Johnson, J. L., "L iquid Metal-to-Gas Leak Detection Instruments".

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Figure 7.5-7 Liquid Metal / Gas Leak Detection System Response Time Vs. Sodlun Aerosol Concentration (Inerted Cells)

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Figure 7.5-8 Liquid Metal / Gas Leak Detection System Predicted Aerosol Concentration for 100 g/HR Leak in N 2(1% 0 2) Atmosphere 10*7 , ,

PREDICTED AEROSOL CONCENTRATION FOR 100 g/hr LEAK IN N 2 (1% O2) ATMOSPHERE TEST CELL-RANGE OF

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CELL GROUP VOLUME (ft ) i 7.5-55 Anend. 72 Oct. 1982

. nam Question 421.42 Section 7.1.2 and 7.2.2 of Chapter 7 of the PSAR reference the use of IEEE standards. Other sections in Chapter 7 make reference to Section 7.1.2 but do not identify specific IEEE standards which were implemented in the system design. Justify why Section 7.3 through 7.7 of the PSAR do not provide enought inf ormation to determine whether the IEEE standards are implemented in the design.

Resoonse:

Chapter 7 has been revised to add specific identification of IEEE standards when appropriate as described below. Compliance with IEEE standards f or non-saf ety related systems is not required and theref ore use of IEEE standards f or those systems is not discussed.

Section 7.2 - This sectioa is amended to clarify the use of IEEE standards.

Section 7.3 - This section is amended to clarify the use of IEEE standards.

Section 7.4 - This section is amended to clarify the use of IEEE standards.

Section 7.5.1 - The Wide Range and Power Range Flux Monitors discussed in this section are safety related, the IEEE standards of Table 7.1-3 are appi led to the designs.

, Section 7.5.2 - Addresses the types of f unctions and the sensors used in the plant and does not specifically identify these instruments as saf ety rel ated or not. Tabl e 7.5-1 identifies the variables which are saf ety related as does Section 7.2. Paragraph 7.5.2.2 states that the instruments which are a part of the Protection system comply with the requirements of Section 7.1.2 and 7.2.2 which encompasses the IEEE standards l isted in Tabl e 7.1-3.

P Section 7.5.3 - The sodium level probes discussed in this section are 1E.

The remaining instrumentation is non-1E. Secti on 7.5.3.2 states that the sodium level probes are part of the Reactor Shutdown system and will comply with PPS Design Requirements (Sections 7.1.2 and 7.2.2). The probes, theref ore, wil l comply with IEEE standards identified in these sections as eppl icable to PPS.

Section 7.5.4 - The Failed Fuel System is not saf ety related.

Section 7.5.5 - The leak detection systems discussed in this section are not saf ety rel ated.

QCS421.42-1 Amend. 72 Oct. 1982 t

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Section 7.5.6 - SWRPRS instrumentation and control has two f unctions. One is to initiate a reactor trip, the other is to isolate the af fected loop. The reactor trip f unction is part of the Pl ant Protection system and as stated in 7.5.6.2 compi les w Ith Sections 7.1.2 and 7.2.2. Isolation of the af fected loop is not safety related since it does not compranise the abil Ity to remove decay heat fran the unaf fected Ioops.

Sections 7.5.7, 7.5.8 and 7.5.9 -

The Instruments discussed in these sections are safety related, the IEEE standards of Table 7.1-3 are appl ied to the designs.

Sections 7.6.1, 7.6.2, 7.6.4 and 7.6.6 -

These Sections have been revised to incorporate appl icable IEEE Standards.

Section 7.6.5 - The SGB Flooding Protection System is saf ety related and section 7.6.5 is amended to clarify the use of IEEE standards.

Sections 7.7 and 7.8 -

No IEEE standards are appiled in these sections since the systems described therein are non safety related systems.

Section 7.9 -

This section has been amended to ciarify the use of IEEE standards.

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QCS421.42-2 Amend. 72 Oct. 1982

TMB LE 7.1 -3 LIST OF IEEE STANDARDS APPLICABLE TO SAFETY RELATED INSTRUENTATION AND CONTROL SYSTEMS lEEE-279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations IEEE-308-1974 Criteria for Class IE Power Systems for Nuclear Power Generating Stations IEEE-317-1976 Electric Penetration Assemblies in Containment Structures f or Nuclear Power Generating Stations I EEE-323-1974 Qualifying Class IE Electric Equipment f or Nuclear Power Generating Stations IEEE-323-A-1975 Supplement to the Foreword of IEEE 323-1974 IEEE-336-1971 IEEE Standard: I nstal l ation, inspection, and Testing Requirements for Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations IEEE-338-1977 Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems IEEE-344-1975 IEEE Std. 344-1975, IEEE Recommen' d ed Practices f or Seismic Qual ification of Cl ass 1 Equipment f or Nuclear Power Generating Stations IEEE-352-1975 General Principles f or Reliabil Ity Analysis of Nuclear Power Generating Station Protection Systems IEEE-379-1972 lEEE Trial-Use Guide f or the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems IEEE-383-1974 Standard f or Type Test of Class 1E Electric Cables, Field Spl ices, and Connections f or Nuclear Power Generating Station.

IEEE-384-1974 IEEE Trial Use Standard Criteria for Separation of Class IE Equipment and Circuits IEEE-420-1973 Trial-Use Guide for Class IE Control Switchboards f or Nuclear Power Generating Stations IEEE-494-1974 IEEE Standard Method f or identif ication of Documents Related to Class 1E Equipment and Systems f or Nuclear Power ..

Generating Station 7.1 -9 Amend. 72 Oct. 1982 e%9 nu ., __ - . - -

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o Environmental Changes All electrical equipment is subject to performance degradation due to major changes in the operating environment. Where practical, PPS equipment is designed to minimize the of fects of environmental changes; If not, the performance at the environmental extremes is used in the analysis.

Measures have been taken to assure that the RSS electronics are l capable of performing according to their essential performance requirements under variations of temperature. The range of temperature environment specified for all the electronic equipment considered here is greater than is expected to occur during normal or abnormal conditions. Electronics do not f all catastrophically when these Iimits are exceeded even though this is the assumed f ailure mode. The detailed design of the circuit boards, board mounting and racks includes f ree ventilation to minimize hot spots. Ventilation is a result of natural convection ale fIow.

The RSS is designed to operate under or be protected f rom a wider range of reiative humidity than that produced by normal or postuiated l accident conditions.

Vibration and shock are potential causes of f ailure in electronic

.. components. Design measures, including the prudent location of equipment, minimize the vibration and shock experienced by RSS electronics. The equipment is qualified to shock and vibration specif Ications which exceed al I normal and of f-normal occurrences.

The RSS comparators and protective logic are designed to operate over a power source voltage range of 108 to 132 VAC and a power ' source f requency range of 57 to 63 HZ. The maximum variation of the source voltage is expected to be 110%. More extreme variations in the power source may result in the af fected channel comparator or logic train outputting a trip signal. In addition, testing and monitoring of RSS i equipment is used, where appropriate, to warn of impending equipment l degradation. Therefore, it is not expected that changes in the enviro.vnent wilI cause total felIure of an instrument channel or Iogic train, much less the simultaneous f ailure of all instrument channels or logic trains.

The majority of the RSS electronics is located in the control bullding, and is not subjected to a radioactive environment. Any PPS l

l equipment located in the radioactive areas (such as the head access l area) w!!I be designed to withstand the ievel of activity to which it will be subjected, if its f unction is required.

l 7.2-12 Amend. 72 Oct. 1982 l

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o Tornado The RSS is protected from the ef fects of the design basis tornado by locating the equipment within tornado hardened structures.

o Local Fires ,

Al l RSS eq ui pment, incl uding sensors, actuators, signal conditioning l equipment, wiring, scram breakers, and cabinets housing this equipment is redundant and separated. These characteristics make any credible fire of no consequence to the safety of the plant. The separ,ation of the redundant components increases the time required f or f ire to cause extensive damage and also allows time for the fire to be brought to the attention of the operator such that corrective action may be .

Initiated. Fire protection systems are also provided as discussed in Section 9.13.

o Local Exolosions and Missiles All RSS equipment essential for reactor trip is redundant. Physical l separation (distance or mechanical barriers) and electrical isol ation exists between redundant components. This physical separation of redundant components minimized the possibility of a local explosion or missile danaging more than one redundant component. The remaining redundant components are still capable of perf orming the required protective f unctions.

o Earthauakes All RSS equipment, incl uding sensors, actuators, signal conditioning l equipment, wiring, scram breakers and structures (e.g., cabinets) housing such equipment, is classed as Seismic Category 1. As such, all RSS equipment is designed to remain f unctional under CBE and SSE conditions. The characteristics of the OBE and SSE used f or the l ovaluation of the RSS are found in Section 3.7.

[

7.2.2 Analvsis The Reactor Shutdown System meets the saf ety related channel perf ormance and rel labil Ity requirements of the NRC General Design Criteria, IEEE Standard l 279-1971, appiIcable NRC Regulatory Guides and other appropriate criteria and standards.

The RSS Logic is designed to conf orm to the IEEE Standards !Isted in Table 7.2-4.

General Functional Reaufrement The Plant Protection System is designed to automatically initiate appropriate protective action to prevent unacceptable plant or component damage or the release or spread of radioactive materials.

7.2-13 Amend. 72 Oct. 1982

- '- ___ _ - ~ . - -- - . ....... -

l TABLE 7.2-4  ;

LIST OF IEEE STANDARDS APPLICABLE TO l

THE REACTOR SHUTDOWN SYSTEM LOGIC (1) l l

IEEE 279-1971 lEEE Standard: Criteria for Protection Systems f or Nuclear Power Generating Stations IEEE 308-1974 Criteria for Class 1E Power Systems f or Nuclear Power Generating Stations IEEE 317-1976 Electric Penetration Assemblies in Containment Structures f or Nuclear Power Generating Stations IEEE 323-1974 IEEE Trial-Use Standard: General Guide for Qualifying Class 1E Electric Equipment f or Nuclear Power Generating Stations IEEE 323-A-1975 Supplement to the Foreward of IEEE 323-1974 IEEE 336-1971 IEEE Standard: Installation, inspection and Testing Requirements f or Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations IEEE 338-1977 IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems I EEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Seismic Qualification of Class 1 Equipment for Nuclear Power Generating Stations

(

IEEE 352-1975 IEEE Guide f or General Principles f or Rel iabil ity Analysis of Nuclear Power Generating Station Protection Systems IEEE 379-1972 IEEE Trial-Use Guide f or the Applicaton of the Single Failure Criterion to Nuclear Power Generating Station Protection Systems IEEE 384-1974 IEEE Trial Use Standard Criteria for Separation of Class IE Equipment and Circuits lEEE 494-1974 IEEE Standard Method f or Identification of Documents Related l to Class IE Equipment and Systems f or Nuclear Power l Generating Station (1) lEEE Standards applicable to the Instrumentation and monitoring systems are l isted in Section 7.5. ,

l 7.2-23 a Amend. 72 Oct. 1982

Head Access Area Radiation The Head Access Area Radiation Subsystem initiates closure of the containment isolation valves in the event of large radiation releases in the head access area. Three radiation sensors are located in the head access area to provide early initiation and closure of the isolation valves to assure that releases from design basis events do not exceed the guidel Ine val ues of 10CFR100, 7.3.1.2.2 Essential Performance Reautrements To implement the required isolation function within the specified limits, the CIS must meet the f unctional requirements specified below:

The closure time requirement f or the inlet and exhaust isolation valves is 4 seconds with a three second or less detection time in the heating and ventilating system. A 10 second transport time f rom sensing point to the val ve exists (see Section 15.1.1). The 3 seconds includes sensor time response, comparator and logic time del ays.

The CIS is designed to meet these requirements f or the environmental conditions described in Section 7.2.1.

7.3.2 Analvsts The design of the CIS provides the necessary design features to meet the f unctional and perf ormance requirements as described below. The CIS logic is designed to conform to the IEEE Standards listed in Table 7.3-2.

7.3.2.1 Functional Performance The analyses in Sections 15.5 and 15.6 shows the results of the postulated f ault conditions. These analyses assumed a closed containment where the events occurred wIth the containment hatch closed. For the l imiting event, primary drain tank fire during maintenance, scoping analyses have been

perf ormed to determine the required closure time of the containment isolation val ves. For the primary drain tank fire, closure within 20 minutes is adeq uate. Further, analyses to determine the required closure time under postulated accident conditions have been performed and are discussed in i

Section 15.1.1. These analyses are used to determine the available design i margin. The results of this assumed condition do not exceed the guideline i

val ues of 10CFR100 if the main exhausi and inlet valves are closed within 4

, seconds assuming the normal air transport time from the detector to the valve is 10 seconds or more, a 14,000 Cfm normal ventilation rate.

Since the automatic Containment isolation System is designed to isolate within the above time response requirements, all of the design basis conditions are terminated within the necessary limits for the present design concept.

7.3.2.2 Deslan Features The CIS Instrunentation, controls and actuators are designed to meet the requirements of IEEE-279-1971. The analyses of compliance with these are summarized bel ow.

7.3-3 Amend. 72 Oct. 1982 L as-sees - .- ----

--.-.-...; 9 ._- m g.,ropw v , T ypy - - - - - - -

TABLE 7.3-2 LIST OF IEEE STANDARDS APPLICABLE TO THE CONTAINENT ISOLATION SYSTEM LOGIC IEEE 279-1971 IEEE Standard: Criteria f or Protection Systems 'f or Nuclear Power Generating Stations IEEE 308-1974 Criteria for Class IE Power Systems for Nuclear Power Generating Stations IEEE 317-1976 Electric Penetration Assemblies in Containment Structures f or Nuclear Power Generating Stations IEEE 323-1974 IEEE Trial-Use Standard: General Guide for Qualifying Class IE Electric Equipment f or Nuclear Power Generating Stations IEEE 323-A-1975 Supplement to the Foreward of IEEE 323-1974 IEEE 336-1971 IEEE Standard: I nstal l ation, Inspection and Testing Requirements for instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations lEEE 338-1977 IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE 344-1975 IEEE Standard 344-1975, IEEE Recommended Practices for Seismic Qualification of Class 1 Equipment for Nuclear Power Generating Staiions IEEE 352-1975 IEEE Guide f or Gene < al Principles f or Reliability Analysis of Nuclear Power Ger.erating Station Protection Systems IEEE 379-1972 IEEE Trial-Use Guide f or the Applicaton of the Single l Failure Criterion to Nuclear Power Generating Station

! Protection Systems IEEE 384-1974 IEEE Trial Use Standard Criteria for Separation of Class 1E Equipment and Circuits IEEE 494-1974 IEEE Standard Method f or Identification of Documents Related to Class 1E Equipment and Systems f or Nuclear Power Generating Station 7.3-5a Amend. 72 Oct. 1982

7.4 INSTRUMENTATION AND CONTROL SYSTEMS REOUIRED FOR SAFE SHUTDOWN The Instrunentation and Control Systems necessary for safe shutdown are those associated with monitoring of core criticality, decay heat removal (SGAHRS portion), outlet steam isolation, and control room habitabil ity.

Monitoring of core criticality is ef fected by the Flux Monitoring System

( Sect i on 7.5.1 ) . The control room habitability is covered in Chapter 6.

Thus, this section treats the control and instrumentation needs for decay heat removal by the Steam Generator Auxiliary Heat Removal System (SGAHRS) and outlet steam isolation by the Outlet Steam isolation System (OSIS); control and Instrumentation for Direct Heat Renovel Service (DHRS) is discussed in Section 7.6.

7.4.1 Steam Generator Auxillarv Heat Removal Instrumentation and Control

. SYG1RM 7.4.1.1 Design Descriotion 7.4.1.1.1 Function The SGAHRS (fluid system and mechanical components as described in Section 5.6.1, and electrical components as described below) provides the heat removal path and heat sink f or the nuclear steam supply system following upset, emergency, or f aulted events which render the normal heat sink unavailable.

The SGAHRS Instrumentation and Control System in conjunction wIth the PPS detects the need f or, Initiates, and controis the alternate heat removal path when the normal heat sink is unavailable. The SGAHRS nominal control setpoints shown in Table 7.4-2 are discussed in the following subsections.

The SGAHRS Instrumentation and Control System is designed to the IEEE Standards l isted in Tabl e 7.4-3.

7.4.1.1.2 Eculoment Design The mechanical system for which the SGAHRS IAC is provided is briefly described bel ow, l

When actuated, the SGAHRS draws water from a Protected Water Storagc Tank and pumps it to each steam drum. Two supply lines are provided f or each steam l dr um. One i Ine is suppl ied by two hal f-sized, motor-driven f eedwater pumps i

while the other is supplied by a f ull-sized, turbine-driven pump. Each supply l line provides a flow control valve and an isolation valve at the inlet to each steam drum. The isolation valves are provided to isolate the auxiliary l feedwater system from the steam generator system during power operation and to provide leak isolation during SGAHRS operation.

l l In addition, a Protected Air Cooled Condenser (PACC) supplied with each steam i drum is placed into operation. This system rejects heat to the atmosphere via j convection. Saturated steam is supplied to the condenser f rom the steam drum I

. 7.4-1 Amend. 72 l Oct. 1982 i

( .__

s 7.4.1.2 Design Analysis To provide a high degree of assurance that the SGAHRS will operate when necessary, and in time to provide adequate decay heat removal, the power for the system is taken from energy sources of high reliability which are readily avail abl e. As a safety related system, the instrumentation and controls critical to SGAHRS operation are subject to the safety criteria identified in Section 7.1.2.

Redundant monitoring and control equipment will be provided to ensure that a single f ailure will not impair the capability of the SGAHRS Instrumentation and Control System to perform its intended safety function. The system will be designed f or f all saf e operation and control equipment where practical and w il l, in the event of a f ailure, assume a f ailed position consistent with its Intended saf ety function.

Because there are three redundant decay heat removal loops, the instrumentation and controls associated with each Individual loop (e.g.,

auxiliary feedwater flow and air cooled condenser control systems) do not independently meet singl e f ail ure criteria. However, when taken collectively as a system, they provide the single f ail ure capabil Ity required.

7.4.2 Outlet steam isolation Instrumentation and control System 7.4.2.1 Design Descriotion 7.4.2.1.1 Function The Outlet Steam isolation Subsystem (OSIS) provides isolation of steam system pipe breaks. Steam system isolation is a necessary function f or safe shutdown in those pipe break conditions af fecting the three steam supply systems and is' provided if needed on a per loop basis. By definition, this zone of protection will include the high pressure steam supply system downstream from the Individual loop check valves.

The OSIS Controls are designed to the IEEE Standards listed in Table 7.4-3.

1 7.4-6 Amend. 72 Oct. 1982 c ,_ i ne n , , ., . __. , . . . _ -

~ ~ ~ ' ' ' ~ ~ ~ ' ' ' ~

-- '-i'ep #rtcs 1t>uu> taDTH4 ^ ^ ^ ^ ~

TM3LE 7.4-3 LIST OF lEEE STANDARDS APPLICABLE TO SGAHRS AND OSIS INSTRUENTATION AND CONTROL SYSTEMS lEEE-279-1971 lEEE Standard: Criteria for Protection Systems'f or Nuclear Power Generating Stations IEEE-308-1974 Criteria for Class 1E Power Systems f or Nuclear Power Generating Stations IEEE-323-1974 IEEE Trial-Use Standard: General Guide for Qualifying Class 1E Electric Equipment for Nuclear Power Generating Stations IEEE-323-A-1975 Supplement to the Foreword of IEEE 323-1974 IEEE-336-1971 lEEE Standard: Installation, inspection, and Testing Requirements f or Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations I EEE-338-1977 Criteria f or the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Seismic Qualification of Class 1 Equipment for Nuclear Power Generating Stations I EEE-352-1975 General Principles f or Rel labil ity Analysis of Nuclear Power Generating Station Protection Systems l IEEE-379-1972 IEEE Trial-Use Guide f or the Appl Ication of the l Singl e-Fall ure Criterion to Nuclear Power Generating i

Station Protection Systems IEE-382-1980 IEEE Standard f or Qualification of Safety-Related Valve Actuators IEEE-384-1974 lEEE Trial Use Standard Criteria f or Separation of Class 1E Equipment and Circuits l

IEEE-494-1974 IEEE Standard Method f or Identif ication of Documents Related to Class IE Equipment and Systems f or Nuclear Power Generating Station 7.4-10d Amend. 72 Oct. 1982

Page 6 (82-0454) [8,07] #43 7.4.4.2 Design Analvsts The Remote Shutdown System provides the RSMP from which an operator can assess the progress of the plant shutdown and command the local operation of the plant systems (primarily SGMRS) to ef fect the shutdown. It should be noted that the PACC subsystem of SGMRS is autmatically initiated by all reactor trips, and it remains in operation for the duration of the plant shutdown or as long as the reactor generates significant decay heat.

)

The Remote Shutdown System imposes no special requirements on the plant systems, but takes advantage of the following system design features:

o The ability to operate in both local and remote modes with isolation frm and annunciation in the Control Room when operating in the local mode.

o The redundancy diversity, separation, Isolation and reliability of the saf oty grade systems.

o The design and locatic,a of safety grade systems equipment that minimize the probabil Ity and ef feet of f f res and explosions on the ability of the systems to perf orm their saf ety f unction.

o The redundant safety grade SGMRS provides the capability to achieve and maintain hot shutdown and, if desired, to cool the piant to and maintain the pl ant at ref ueling conditions, o When transferring SGMRS to the local mode, the operator manually starts g

SGMRS. Once started, SGMRS automatically controls those parameters used to rmove decay heat.

The RSMP is a non-Class 1E Seismic Class lll assembly and therefore, is not subject to the separation requirments of IEEE 384-1974, or to the seismic qualification requirments of IEEE 344-1974, or to any of the other IEEE Standards l isted in Tabl e 7.1-3.

t N

l 7.4-8f Amend. 72 Oct. 1982 ll L1 __ _

~ ~ ^

'Page'4 (82-0454) [8,07] #43 ~

7.4.4.1.4 Eautoment Deslan The RSMP is the only piece of equipment provided by the Remote Shutdown Sy stem. It will be a vertical sided, non-Class IE cabinet assembly containing l meters and a phone Jack panel. The meters will receive buf fered signals from I the initiating systems and, thus, do not require transfer switches to isolate l them f rom the Control Room. The phone Jack panel will permit the operator at 4 the RSMP to communicate with the five NSSS or Nuclear Island buildings by means of any of the three MCJ circuits provided in each of the bulldings. -In addition, communications among the bulldings can be established through the phone Jack panel on the RSMP.  !

The Indications provided on the RSMP are as follows: ,

1 o For each primary heat transport system ioop, 1 - Purnp outlet sodium temperature indicaton (3 total) 1 - Reactor iniet sodium temperature Indication (3 total) 1 - Sodium pump shaft speed Indication (3 total) o For each intermediate heat transport system loop, l

1 - lHX outlet sodium temperature indication (3 total)

I - lHX inlet sodium temperature indication (3 total) l 1 - Sodium pump shaf t speed Indication (3 total) o For each superheated steam loop, 1 - Temperature Indication (3 total) 1 - Steam flow indication (3 total)

One reactor vessel sodium level meter (long probe) o o For each Diesel Generator (3 total)

[ ~ '

1 - Wattmeter 1 - Frequency meter 1 - Varmeter 1 - Voltmeter wIth phase selector switch s 1 - Ammeter with phase selector switch in addition to the foregoing Indications, other indications used during remote shutdown operations that are not on the RSMP will be available as follows:

o SGAHRS Controls and indicators used for the operation of each SGAHRS division are located on the three seperate SGAHRS panels in cells 272A, B, and C. Each SGAHRS division is separate and redundant from the other divisions. See l the response to Question CS421.04 for additional Information about SGAHRS division assignments.

7.4-8d Amend. 72 Oct. 1982 l

- - , _ . , . . . ~ , . = -

-n

~ ~ ~ ~ ~ ~ ~

~"--bge ~~T"rezi oos1Td,IJ f F41 E. Operating Basin Overflow F. EPSW Makeup Pump Discharge Pressure G. Emergency Cool ing Tower Basin Level H. EPSW Flow to Emergency Chillers

l. EPSW Temperature at the Discharge of Emergency Chillers J. EPSW Flow from Diesel Generators Heat Exchangers K. EPSW Temperature at the Discharge of Diesel Generators Heat Exchangers L. Dif f. Pressure Across Emergency Chillers M. Transfer of Controlling Capabilities from Control Room to Local Panels N. Pump and Fan Status Process variables identif ied above with ' A' and 'H' are designated as accident monitoring variables to assess plant and environs conditions during and f ollowing an accident. Refer to Section 7.5.11 of PSAR for detailed requirements on Accident Monitoring.

7.6.1.3.3 Inouts to PDH&DS The following process variables are provided as inputs to Plant Data Handling

& Displ ay System (Non-Saf ety System):

A. EPSW Discharge Temperature B. Emergency Cool ing Tower Basin Level C. EPSW Temperature at the Discharge of Emergency Chiller D. EPSW Temperature at the Discharge of Diesel Generator Heat Exchanger E. EPSW Fl ow to Emergency Chil ler inoperable status of EPSW Pumps; Makeup Pumps; and Cooling Tower Fans is also monitored through inoperable Status Monitoring System.

7.6.1.1.3.4 Deslan Analvsis l

EPSW System is designed to operate automatically. The system is operated only during emergency conditions. EPSW System components are cascaded to operate in sequence. Starting of EPSW Pumps will operate EPSW Makeup Pumps and Cool ing Tower Fans. System will not operate when the EPSW Pump pit level is l low or when electrical f ault exists.

l The design of the EPSW System is in conformance with the following IEEE standards l isted in Tabl e 7.6-2.

I 7.6-2a Amend. 72 Oct. 1982

w 7.6.2.2.3 Inouts to PDHADS The following process variables are provided as inputs to Plant Data Handling

& Displ ey System (Non-Safety System):

A. EW Temperature at the inlet of Emergency Chiller B. EW Temperature at the Discharge of Emergency Chiller C. EW FIow fran Emergency Chiller D. E W Chiller Trip Status E. EW Containment Isolation Valves Status

~

F. Secondary Coolant Expansion Tank DT-J Leakage 7.6.2.2.4 Design Analysis Em System is designed to operate automatically. The system is operated only during emergency condition. EW System components are cascaded to operate in seq uence. Low flow of.NW to EW Ioop signal w11I align EW isolation valves and operate EW Pumps, Emergency Plant Service Wa+er Loops, and EW Chillers.

System will not operate when the EPSW flow through chiller is not established or when electrical fault exists.

The design of the EW System is in conformance wIth the IEEE Standards iIsted In Tabl e 7.6-2.

7.6-2e Amend. 72 Oct. 1982 on i nu - . _ . _ . -

~

Yage~-' Y wz % W g g n y -- -

3. Unit cooler or HVAC unit supply air temperature high or air temperature entering cooling coil low.
4. Smoke, ammonia, chl orine, fl uorine or radiation present In Control Room main or remote air intake.
5. Control switch in the local mode (Control Room only alarm only).

C. Typically, process variables are provided as inputs to the Plant Data Handling & Display System as follows:

1. Control Roan and computer room humidity.
2. Containment dif ferential pressure.
3. Annul us dif ferential pressure.
4. RSB confinement differential pressure (four different cells).
5. Control Room dif ferential pressure.
6. Air temperature entering and leaving each filter unit.
7. Air temperature entering and leaving each HVAC unit.
8. Cell temperature of each area being serviced by a unit cooler or HV AC unit.
9. Inoperable or bypass status of components.

D. The following process variables are classified as Accident Monitoring variables and are used to assess plant and environs conditions during and following an accident:

1. Annul us to abnosphere dif ferential pressure.
2. RSB conf inement to atmosphere dif ferential pressure.
3. HVAC units discharge air temperature.
4. Filter units adsorbent f ilter leaving air temperature.
5. HVAC and f ilter units air flow low.
6. Damper and valve position Indication.
7. Fan operation status Indication.

7.6.4.3 Design Analvsis ,,

The HV AC Instrumentation and Control System is designed to perf orm the f unctions described in Section 7.6.4 while meeting the criteria listed in Section 7.6.4.1. All HVAC l&C circuits shall meet the requirements of Section 7.1 with the exception of alarm circuits and inputs to the PDH&DS which are 7.6-8 Amend. 72 Oct. 1982 L

_ _ ._., 9._ nom w ,7og ,

1 i

I i

Non-Cl ass 1E circuits. The design of the HVAC Instrumentation and Contrci  !

system is in conformance with the IEEE Standards and listed in Table 7.6-2.

Ref er to PS AR Section 7.1.2 for conf ormance to appl icable IEEE Standards.

7.6.5 Steam Generator Building (SGB) Floodino Protection Subsystem 7.6.5.1 Design Basis The SGB Flooding Protection Subsystem is provided to prevent flooding of SGAHRS equipment resulting f rom postulated SGS water / steam line ruptures, thereby assuring the availabil ity of SGAHRS for reactor decay heat removal following water / steam line rupture events.

The SGB Flooding Protection Subsystem is designed to the IEEE Standards listed in Tabl e 7.6-3.

7.6.5.2 Design Reauirements The SGB Flooding Protection Subsystem is designed to perf orm the following f unctions:

a) Detect the presence of large steam / water piping ruptures (see Section 15.3.3.1) by temperature and moisture sensors in each cell.

b) Detect water level flooding conditions in each cell by water level sensing el ements.

c) Provide the signals to Initiate the alarms and activate the equipment which provides the SGB flooding protection.

7.6.5.3 Design Descriotion 7.6.5.3.1 Instrumentation Instrumentation provided f or this subsystem consists of Class 1E temperature, and moisture transducers. In addition, non-Class 1E level transducers are prov ided. The transducers and associated control logic are located in the SGB cel ls containing main f eedwater or recircul ation piping. Three independent moisture and temperature measurements in each cell are utilized f or Identifying a major water / steam line rupture. Water level measurements in each cell confirm a flooding condition and are annunciated in the main control room.

7.6.5.3.2 Controls Each heat removal loop Isolates the main feedwater supply upon detection of a major pipe rupture. The start-up and main f eedwater control valves close upon activation by a two-out-of-three logic using measurements of moisture and temperature in each cell. The main f eedwater Isolation valve is Independently closed upon activation by a two-out-of-three logic using the same three moisture and temperature measuremonts from each cell. Separation and isolation is maintained between the control valve and isolation valve activation l ogic.

7.6-9 Amend. 72 Oct. 1982

m.

Small water / steam leaks are identified in each SGB cell by measuring water

, level. Manual corrective control of flooding is Initiated by the operator upon annunciation in the main control room.

l l

r l

I 7.6-9a Amend. 72 i

Oct. 1982 on , nc , _ .. - . ~ . . . . . .- -

. . m y , um mm, y Inoperable status of subsystem f ans (MA, E, EA, EB) and Isolation valves (two per subsystem) is also monitored through inoperable Status Monitoring System.

7.6.6.2.4 Desion AnalvsIs Ref er to PSAR Section 7.1.2 for conf ormance to appl icable IEEE Standards.

EC system is designed to operate automatically. The system and its saf ety-related subsystems are operated during normal as well as emergency conditions.

The RGC system components are cascaded to operate in ' sequence. Starting a fan wIlI open associated supply and return gas Isolation valves. A subsystem w11I not operate when high watar vapor or cooler high water level or electrical fault exists.

As discussed in Section 9.16 each subsystem of RGCs supplles cooling to redundant components, so no additional redundancy is provided in its components and instrumentation.

The systems are designed for f all safe operation and control equipment will assume a f ailed position consistent with its Intended safety function.

The coolant supply to safety-related subsystems MA, 2, EA, EB is provided by Emergency Chilled Water System. The f an motors for these subsystems are provided wIth AC power from Class IE power sources to continue operating during loss of of fsite power, except for the booster f an of the subsystem EB which is not required to operate during loss of power condition. Subsystems MA and EA and the EM pumps cooled by these two subsystems are served by Class IE power supply Division 1. Also, subsystems MA and EA are served by Emergency Chil led Water Loop "A". Subsystems 2 and EB are served by Emergency Chilled Water Loop "B", and Class 1E power supply Division 2. The EM pumps cooled by subsystems E and EB are also connected to Class 1E power supply Division 2. Autcmatic Isolation valves are designed as f all open valves so as to be in their safety position upon loss of power.

Fan and Isolation Valve controf switches are Iocated in the Iocal panels as well as in the back panels for subsystems MA, E, EA and EB, except for booster f an. Thus, in case of control room evacuation the f ans and valves can be controlled from outside the control rocrns, using local panels.

i The design of the Recirculating Gas System is in conformance with the IEEE Standards Iisted in Table 7.6-2.

I l

l 7.6-16 Amend. 72 Oct. 1982

- " "-~ - ~ ' - ~ ^ - -

- r:yu ~ Y taz-tvo4 r T O ,1 3 gau-- - - - -

TM LE 7.6-2 LIST OF IEEE STANDARDS APPLICABLE TO EERGENCY PLANT SERV ICE WATER, EERGENCY OilLLED WATER, HV AC, AND RECIRCULATING GAS INSTRUENTATION AND CONTROL SYSTEM a) IEEE Standard 279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations b) IEEE Standard 308-1974 Criteria for Class 1E Power Systems f or Nuclear Power Generating Stations c) lEEE Standard 323-1974 Qual Ifying Class 1E Electrical Equipment f or Nuclear Power Generating Stations d) IEEE Standard 338-1977 Criteria for Periodic Testing of Nuclear Power Generating Station Saf ety Systems e) IEEE Standard 379-1972 IEEE Trial-Use Guide for the Appl Icabil Ity of the Singl e-Fail ure Criterion to Nuclear Power Generating Station Protection Systems f) lEEE Standard 383-1974 Standard f or Type Test of Class 1E Electric Cabies, Field Splices and Connections for Nuclear Power Generating Stations g) IEEE Standard 384-1974 lEEE Trial-Use Standard Criteria for Separation of Class IE Equipment and Circuits.

l l

l 7.6-18a Amend. 72 Oct. 1982

W TABLE 7.6-3 LIST OF IEEE STANDARDS APPLICABLE TO SGB FLOODING PROTECTION SUBSYSTEM IEEE-279-1971 IEEE Standard: Criteria f or Protection Systems .f or Nuclear Power Generating Stations IEEE-323-1974 IEEE Trial-Use Standard: General Guide f or Qualifying Class 1E Electric Equipment f or Nuclear Power Generating Stations IEEE-323-A-1975 Supplement to the Foreword of IEEE-323-1974 IEEE-336-1971 lEEE Standard: Installation, inspection, and Testing, Requirements f or Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations IEEE-338-1971 IEEE Trial-Use Critaria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Seismic Qualification of Class 1 Equipment for Nuclear Power Generating Stations IEEE-352-1972 IEEE Trial-Use Guide: General Principles f or Rel iabil ity Analysis of Nuclear Power Generating Station Protection Systems IEEE-379-1972 lEEE Trial-Use Guide f or the Application of the Single-Fallure Criteric to Nuclear Power Generating Station Protection Systems IEEE-384-1974 IEEE Trial-Use Standard Criteria f or Separation of Class 1E Equipment and Circuits IEEE-494-1974 lEEE Standard Method for identification of Documents Related to Class 1E Equipment and Systems f or haclear Power Generating Station l

l l

1 l

7.6-18b Amend. 72

, Oct. 1982

i. _ . _ . _ -._ _ _

=-- ' mm -- ~~- -- - -

f :-- ny.-ino6[Wg 3,-) g yg 3 .

Since the Main Control Panel includes safety related equipment, the sections including this equipment are designed to Seismic Category I and qualifled in ,'

accordance seith IEEE Std. 323 and IEEE Std. 344. Structures, wiring, wireways, and connectors are designed and installed to ensure that safety i related equipment on the control panel remains operational during and af ter l the SSE. The Main Control Panel is constructed of heavy gauge steel within 4

appropriate supports to provide the requisite stif fness. -

l I

Within the bounderles of the Main Control Panel Sections, modules are arranged according to control functions. Fire' retardant wire is used. Moduler train wiring is formed into wire bundles and carried to meta 1 wire ways (gutters).

Gutters are run into metal vertical wireways (rlsers). The risers are the interf ace between external wire trays feeding the panel and Main Control Panel wiring. Risers are arranged to maintain the seperated routing of the external wire trays. (See Figures 7.9-3 and 7.9-4).

1 Mutually redundant safety train wiring is routed so as to maintain separation in accordance with the criteria of IEEE Std. 384. A minimum of six inches air ,

separation is maintained between wires associated with dif ferent trains.

Where such air separation is not available, mechanical barriers are provided in Ileu of air space.

The Main Control Panel protection system circuits are designed and selected to  ;

ensure that system performance requirements are met and channel integrity and independence are maintained as required by IEEE Std. 279. Power division

separation and isolation are maintained in accordance with the requirements of
IEEE Std. 308.

7.9.3 Local control Stations Local control panels are provided for systems and components which do not require f ull time operator attendance and are not used on a continuous basis.

In these cases, however, appropriate alarms are activated in the Control Room to alert the operator of an equipment mal function or approach to an of f-normal condition.

7.9.4 bommunications Commtnications are provided between the Control Room and alI operating or l manr.e'd areas of the pl ant. In addition to publIc address and interpt ant l communications and the private automatic exchange (used f or in-plant and

'x! external communications) a sound powered maintenance emunication Jacking t system is provided. Redundant and separate methods of c 19unication between the co6 trol room and other TVA generating plants is alto provided.

L 7.9.5 Des!on Evaluatton Foi l ow Ing the Three Mi l e I si and acci dent, a I arge task f orce was f ormed f or the purpose of performing a thorough review of the CRBRP Control Room design.

This overalI revlew was divided into three parts; a planning phase, a revlew phase, and assessment and Implementation phase. Fol low ing the task f orce ef fort, NJREG-0700 was issued. NUREG-0700 is similar in intent to the CRBRP I

Control Roan design eval uation.

7.9-6 Amend. 72 Oct. 1982

7.9.5.1 Plannino Phase in the planning phase the objectives and scope of the task f orce were identified, and criteria were establ ished f or personnel selection. A charter was developed which contained the scope and objectives, and personnel selection was accompi ished.

The task f orce charter required a review of the Control Room design and the '

operating procedure outlines to ensure that the systems designs, the integration of the systems, and the man-machine interf aces properly supported saf e operations of the plant during both normal and abnormal condif f ons. A task analysis was established f or observing the operator conducting various duties. Specif ic items included in the review are:

1. Overall Control Room and Individual panel designs and features, and their interf ace with the operator.
2. System and overall plant operating procedure outlines.
3. Administrative approaches f or plant operations.
4. Recommendations from other Key System Review Task Forces.*
5. Recommendations made by NRC and other parties as a result of the Three Mile Island occurrence.
6. Computer util ization by the operators.
7. Operator training requirements.
8. Remote shutdown capabilities and safety system status indication in the Control Room.

Criteria were established for personnel selection of those to participate on the task f orce. Nuclear experience was considered necessary in the areas of design, analysis, operations, testing, maintenance, and training. Personnel whose background included sodium plants and light water plants were selected.

Licensed and qualified operators were consider ed mandatory. Personnel with human f actors education and experience both inside and outside the nuclear industry were incl uded.

Human f actors considerations were emphasized in the planning phase. Previous i Control Room design ef f orts had attempted to optimize the man-machine interface. However, a major objective of the Control Room Task Force was to re-eval uate th is interf ace. Prior to the evaluation ef fort a seminar was i hel d, under the direction of three leading human f actors personnel, to teach i

the Task Force discipl ined methods f or considering human f actors. Based on this training and f urther assistance from human engineers, check lists were l prepared to eval uate the man-machine Interf ace.

  • See Ref erence 7.9-1 I

7.9 -6 a Amend. 72 Oct. 1982 on.<nca ___

~

.o s ut-wu,inogry su ~~~ -

u ap 7.9.5.2 Review Phase in the review phate extensive analysis of plant events were conducted.

Functional analyses were made of the operator in his response to automatic equipment actions, manual actions which had to be perf ormed in the Control Room, and manual actions required by operators external to the Control Room.

More than 200 walk-throughs of plant events were conducted. -

The Control Roan design and operating instructions were thoroughly reviewed in four areas:

1. Proper identification of systems to be operated from the Main Control Roan.
2. Proper staf fing of the Control Room.
3. Proper overall layout of the Control Room to enhance the man-machine interf aces and support the integrated cperation of plant's systems.
4. Proper layout and design of Individual Control Room panels, Instr uments, Indicators, and control s to enhance the man-machine Interf ace and support the integrated operations of the plant's sy stems.

A f ull scale mockup of the Control Room was used. The events chosen to be evaluated were caref ully selected so they would umbrella all of the operations that are either expected to occur or might be postulated to occur over the l if e of CRBRP. The of f-normal events include plant responses to single and multiple f ailures.

The methodology of perf orming this review consisted of using three groups of people; simul ators, operators, and eval uators.

The Simulators analyzed the events which were to be evaluated prior to the walk-throughs and then, during the walk-through evaluations, simulated the control panel indicator s. Some of these events had previously been analyzed via computer while other events required additional computer runs to enable mocking up the panel as it would appear to the operator. The control panels were mocked up by the Simulators to represent the changing plant conditions and the inf ormation flow into the Control Room during the event. This made the walk-through as realistic as possible.

The Operators pl ayed the part of the Control Room operators and carried out the steps of the procedure being evaluated. They touched each switch they were required to operate, and observed each Indicator which was part of the particul ar event.

Tne Eval uators included a Human Factors Engineer and a Systems Engineer.

Their function was to fill out the Operating Sequence Diagram and the eval uation sheets f or each procedure and event reviewed.

7.9-6b Amend. 72 Oct. 1982

- - - - - - - - ~ ~ - - - - - " 2 rage vTez-ruertu,/yy49 As problems or concerns were encountered, recommendations were made. These were, in some cases, of a broad nature and reflected the need for reconsideration of decisions made in the four most important evaluation areas described above. Other problems and concerns related to specific detalis of the Control Room design or the procedure outiines.

7.9.5.3 Assessment and imolementation Phase

  • The evaluation and implementation of the recommendations started with a check of the consistency of all of the recommendations by the task force. Small models of the overalI Control Room and Main Control Panel were made assuming alI recommendations were incorporated into the design. The recommendations were modifled based on the smalI model to provide a coordinated and consistent set of f Inal recommendations. Senior Project Management reviewed the final set of recommendations and issued them to the Project iIne organization for assessment and implementation. The cognizant design engineers have two choices. They can either accept the recommendation if it is valid, and include it into the plant design via normal proceduros, or reject the recommendation and provide adequate justification if the recommendation is i nval Id. Each assessment is reviewed and approved by senior project management.

1 7.9.5.4 Conclusions The Control Room Task Force Design Review is documented in f urther detail in Ref erence 7.9-1. In September 1981, NUREG-0700 entitled "Guidel ines f or Control Room Design Review" was issued. A comparison between these two documents leads to the conclusion that although NUREG-0700 was issued af ter the Control Room Task Force Review, the intent of the NRC in promulgating NUREG-0700 is similar to the Project's intent in performing the Control Room Task Force Review, and the Intent of NUREG-0700 is bel leved met by CRBRP.

i 7.9-6c Amend. 72 Oct. 1982 1

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