ML20050A558

Amend 67 to PSAR

ML20050A558
Person / Time
Site:	Clinch River
Issue date:	03/31/1982
From:	ENERGY, DEPT. OF
To:
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ML20050A556	List: ML20050A555 ML20050A558
References
NUDOCS 8204010421
Download: ML20050A558 (100)

Text

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PAGE REPLACEMENT GUIDE FOR

. AMENDMENT 67 Ci. INCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT i

(DOCKET NO. 50-537) 4 Transmitted herein is Amendment 67 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537, Amendment 67 consists of new and replacement pages for the PSAR text and Responses to new NRC Questions.

Vertical lines on the right hand side of the page are used to identify question response information and lines ontthe'le'ft hand 4

side are used to identify new or changed design information.

The following attached sheets list Amendment 67 pages and

instructions for their incorporation into the Preliminary Safety Analysis Report.

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I 8204010421 820331-PDR ADOCK 05000537

.B PDR

, L - ,- , - . - , . - , . - . - . - . . - , , . . - . . . . -

J AVEN0 MENT 67 i

PAGE RCPLACEMENT GUIDE REMOVE THESE PAGES INSERT THESE PAGES Chapter 2 2.3-3 thru 6 2.3-3 thru 6 f Chapter 3 '

3.1-43, 44 3.1-43, 44

3.5-Sa, 6 3.5-Sa, 6 l Chapter 4 i 4.2-548, 548a, 548b, 549 4.2-548, 548a, 548b, 549 Chapter 5 5.4-9, 9a 5.4-9, 9a 5.4_-38, 39 5.4-38, 39 O 5.6-3,.4, 4a 5.6-30, 31 5.6-3, 4, 4a 5.6-30, 31 Chapter 7 7-ix, x 7-ix, x

'7.2-2, 2a, 3 thru 10 7.2-2, 2a, 3 thru 10

7.2-17 thru 23 7.2-17 thru 23 7.2-34 thru 37 7.2-34 thru 37 l 7.5-33b, 33c, 33d 7.5-33b, 33c j 7.5-42 7.5-42.

[ 7.5-48 thru 53 7.5-48 thru S3 j 7.8-1, 2 7.8-1 thru 4

7.9-3 thru 6 7.9-3 thru 6, 6a thru 6d

. Chapter 9 9.5-3, 4 9.5-3, 4 9.7-19, 20 9.7-19, 20 9.7-34 9.7-34 Chapter 11 11.1-3, 4 11.1-3, 4 A

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REMOVE THESE PAGES INSERT THESE PAGES Chapter 12. I i

l 12.2-11, 12, 13 12.2-11, 12, 13 ,

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j Chapter 15

! 15.1-94, 95 15.1-94, 95

! 15.1-98, 99 15.1-98, 99 1

APPENDIX-A A-205, 205a, 206 A-205, 205a, 206 i

! APPENDIX G t- i f G-iv, y G-iv, v j

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4 s/ AMENDMENT 67 QUESTION / RESPONSE SUPPLEMENT This Question / Response Supplement contains an Amendment 67 tab sheet to be inserted following Qi page Amendment 66, March 1982.

Page Qi Amendment 67 is to be inserted following the Amendment 67 tab sheet.

As a result of our recent restart of Licensing Activities and resultant new NRC Questions, a new tab entitled " RESPONSES TO NRC QUESTIONS SINCE FALL OF 1981" was issued in Amendment 66. All new NRC Question / Responses issued for incorporation in the PSAR will be

! inserted behind this new tab in numerical order. New NRC Question /

Responses provided in this amendment are listed below. The parenthesis beside each Question / Response shown indicates the number i of pages associated with each Question / Response:

Q471.1-1 (1)

Q471.2-1(2)

.Q471.3-1 (1)

Q471.4-1(3)

Q471.5-1 2)

Q471.6-1 1)

O' Q471.8-1 1)

Q760.1-1 (1)

Q760.2-1 (1)

Q760-3-1(1)

Q760.4-1(1)

Q760.5-1(1) i J

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f- g Hall is not - f requent but it does occur. On an index of potential hall damage 1

( / to residential property, (calculated for each area formed by one degree of j latitude and one degree of longtitude,) the Site is in a region of low potential loss due to hall (Reference 11). Maximum values of the Index occur in northwest Kansas where the index is 50. The index in eastern Tennessee is about 5. Therefore, on a geographical basis, the Site is situated in a region where hall is not a significant factor.

2.3.1.2.4 Tornados The Site is located in an area inf requently af fected by tornadoes (Reference 12 and 13). For the purpose of comparison, Tennessee ranked 25th among all states in the number of tornadoes f rom 1955 to 1967 (Ref erence 14). Dividing along the 86th Meridian, the western half of the state has reported observing three times as many tornadoes as were observed in the eastern half, which includes the Site (Reference 13). The Oak Ridge-Clinch River area has one of the lowest probabilities of tornado occurrence in the entire State (References 14 and 15).

Tornado frequencies calculated by Thom (Reference 16) for each one-degree square of latitude and longitude, for the period 1953 to 1962, show the Site to be situated in a one-degree square with an annual frequency of 0.5.

According to Thom's methodology, the probability that a tornado will strike any point in a ggrticular one-degree square, such as the Site, is calculated to be 3.63 x 10 per year. The recurrence interval is one divided by the probability, which is once in 2760 years. Raw count data on tornado occurrences for those counties near the Site, for the 57-year period from 1916

("}

x_, to 1972, are presented in Figure 2.3-2 (Ref erences 12 and 17) . Roane County is only one of several counties which are represented by the one-degree square used for the calculation of the tornado probability. Roane County itself has not recorded a tornado in the 57-year period f rom 1916 to 1972.

2.3.1.2.5 31 tong Winds and HurricaDRS The following table is based on Thom's report (Reference 18) on his analysis of data on f astest-mile wind speeds at 30 feet above ground level, for indicated recurrence intervals, for eastern Tennessee.

/N 2.3-3 Amend. 65 Feb. 1982

CALCULATED FASTEST MILE VS. RF 'JRRENCE INTERVAL (Ref erence 18)

EASTERN TENNESSEE Recurrence Interval (vears) Fastest MIIe (mph) 10 64 25 73 50 76 100 89 The peak gust recorded at the Oak Ridge City Of fice during a 17-year period was about 59 miles per hour (Reference 4). The f astest mile reported for the Knoxville Airport for a 31-year period was 73 miles per hour (Ref erence 3). A 33-year record at Chattanooga, Tennessee, shows a fastest-mile of 82 miles per hour (Ref erence 19).

Hurricanes are in the post hurricane stage with diminished winds by the time they reach the Site area. In the past 70 years, the remnants of nine hurricanes, classified as devastating when crossing the coastline of the U.S.,

have crossed Tennessee (Ref erence 20). Flooding in association with the remnants of a hurricane has occurred. (Reference 20). Visible damage associated with tropical storms has been reported about once in 25 years in eastern Tennessee (Reference 5).

2.3.1.2.6 High Air Pollution Potential According to a study by Holzworth (Ref erence 21), high air pollution potential can be expected to occur about 5 to 10 days enraally. Holzworth's results are based upon the f requency of occurrence of calculated mixing heights combined with concurrent calculated wind speeds. However, since CRBRP releases can be expected to be ground level, the f requency of stable atmospheric conditions in combination with low wind speeds should be more reflective of dispersion conditions at or near the Site. For the year of record f rom the onsite CRBRP permanent tower, stable conditions (Pasquill stability classes E, F, and G) with wind speeds of about 5 miles per hour or less were reported for about half of the hours.

l 2.3.2 Local Meteorology The CRRRP Site meteorological facilities, the Oak Ridge Area Station X-10, the Oak Ridge City Of fice and the Knoxville Airport (the latter three being the '

closest NOAA weather stailons to the Site) have been used as the primary sources of local meteorological data (Ref erences 1, 2, 3, and 4), with a few exceptions noted in the following discussion. Climatological statistics for these stations are belleved to be representative of the Site area.

Supplementary climatological data were obtained f rom TVA on relative humidities and fog frequencies (Reference 22).

O 2.3-4 Amend. 67 March 1982

2.3.2.1 Normal and Extreme Values of.Meteoroloalcal Parameters 2.3.2.1.1 Temoerature ~ '

Monthly and annual climatological temperature data for Ares Station X-10 and the annual mean temperature data and extremes of temperature for the Oak Ridge City Of fice and Knoxville vicinity, for comparison purposes, are presented in Table 2.3-4. It is apparent by inspection of these data that the three sites are quite similar with respect to temperature except for the extreme low of

-16 degrees F recorded in the Knoxville vicinity. This record low is a part of a much longer observation period, spanning 100 years, in which to observe extremes.

Based on these data one would expect local _ temperatures to range between abou~t l -15 degrees Fahrenheit and 105 degrees Fahrenheit. Temperatures above 90 degrees Fahrenheit should be moderately common from June to early September.

Freezing temperatures should be common from December to February, and have occurred !a al l months f rom October through May.

2.3.2.1.2 Winds Data from the permanent meteorological tower have been used to characterize U wind conditions at the Site. The period of record is February 17, 1977 through February 16, 1978. These data have been used to construct the joint frequency distributions of wind speed and wind direction by stability classes, presented in Tables 2.3-5 through 2.3-20.

From an examination of available data collected at or neer the Site, this one-year summary of on-site wird data appears reasonably representative of average conditions in an average year. The CRBRP meteorological data were used f or characterizing dispersion conditions because they are site specific; the measuring heights conform to NRC Regulatory Guide 1.23; TVA has first-hand knowledge of the instrumentation and the quality control and quality assurance i procedures applied toward meeting regulatory guide spectilcations; and the I CRBRP wind instruments have lower threshold wind speeds.

An analysis of the one-year summary of on-site wind data shows an average annual wind speed of 3.5 mph at the 33-foot level and 5.6 mph at the 200-foot level. The wind was most f requent from the west-northwest at 33 feet and f rom the west-southwest at 200 feet. An analysis of the Oak Ridge Area'Stailon X-10 data, where the wind sensor is mounted at a height of 102 feet, shows an average annual wind speed of 4.9 mph and a revailing wind direction of south to southwest (Ref erence 1). The Oak Ridge City Of fice shows a prevailing wind f rom the southwest with a mean speed of 4.4 mph (Reference 4), which is -

consistent with the other. wind data discussed above. Knoxville Airport data show that the prevailing wind is from the northeast, and the mean hourly speed fl 1s 7.3 mph. (Reference 3). A summary of these data are provided in Table 2.3-21.

O 2.3-5 Amend. 67 March 1982

/

2.3.2.1.3 Humidity Table 2.3-22 provides monthly and annual average relative humidities from i measurements at the 10-meter level at CRBRP. (February 17, 1977 - February 16, 1978). Direct comparisont with respective averages for a 13-year period (1 % 1-1973) at the Knoxvil le Airport (Table 2.3-23), suggests that the influence of the local CRBRP environment has resulted in higher relative humidity values. During four years (1970-1973) of hourly measurunents at the Bull Run Steam plant, (at the one-meter level), relative humidities reached I I-' 100 percent about 2 percent of the time and were at least 95 percent about 7 percent of the time; the temperature was less than 50 F concurrently with a relative hun,ldity of at least 90 percent, about 4 percent of the time.

(Saturation specific humidity at 50 F and 1,000 millibars is about 8 grams per kilogram). The Bull Run Steam plant, similar to the CRBRP site, is located along the Clinch River, about 15 direct miles away.

2 . 3 . 2 .1. 4 Precioltation Average annual precipitation was 51.52 inches at the Oak Ridge Area Station X- 10 based upon 21 years of record (Reference 1) . As indicated by table 2.3-24, on the average, winter was the wettest season with about 30 percent of the annual pr ecipitation. February and March had the highest monthly average of about 5.4 inches. October averaged the driest (2.82 inches). Maximum observed monthly rainfall and 24-hour precipitation (12.84 and 7.75 inches, respectively), both occurred in September. Monthly onsite precipitation of the period February 17, 1977 through February 16, 1978 is presented in Table 2.3-25.

Snow and ice pellet data for the Oak Ridge City Of fice are summarized in Table 2.3-26 (Reference 4). Data listed in the table show that the annual snowfall averaged about 10 inches. Maximum annual snowfall in the 26 year period was 41.4 inches, more than four times the annual mean. Snowf alls of more than six inches in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, were reported at least one for each month from November through March (Reference 4).

2.3.2.1.5 Egg The incidence of heavy fog (1/4 mile or less visibility) varies greatly around Tennessee (Reference 5). Typical annual values include 31 days at Knoxville

~

(Reference 3), 34 days at Oak Ridge City Office (Reference 4), and 36 days at Chattanooga (Pof erence 19) . Five months of the year had an average fog f requancy of three days or more at each of the three stations. At both the Oak Ridge City Of fice and Knoxville, October had the highest fog incidence, with an average of 8 and 5 occurrences, respectively (Table 2.3-27) .

Supplementary fog data (Table 2.3-28) were recorded at two sites along Melton Hill Lake, upstream from the CRBRP site, for the period January 1964 to October 1970. The sites are at Bull Run Creek (about 15 miles nor theast of the CRBRP Site) and Melton Hill Dam (about 4.5 miles east of the CRBRP site).

O 2.3-6 Amend. 65 Feb. 1982

Criterion 24 - Reactivity Control System Redundancy and Caoability O Two independent reactivity control systcms of dif forent design prir,ciples shal l bo provided. One of the systctns shal l use control rcds, pref erably including a positivo means f or Inserting the rods, and shall be capable of rollably controlling reactivity changes to assure ihat under conditions of normal operation, including anticipated operational occurrences, and with appropriate margln f or mal f unct!ons such as stuck rods, specifled acceptable f uel cosign l imits are not exceeded. The second reactivliy control system shall be capablo of rollably controlling the rate of reactivity changes l resulting f rce planned, normal power changes (including xenon burnup) to assure acceptable f uel design limits are not exceeded. One systcm shal I be capable of holding the reactcr core subcritical fcr any coolant tmperature lower than the hot shutdown condition.

Resoonso Two independent diverso reactivity control systcms are provided; namely, the primary and secondary shutdown systctns. The primary control systm is designed to meet the f uel burnup and load follow requirments fcr each cycle as well as to compensato for criticality and ref ueling uncertainties. The primary system will have suf ficient wcrth at any time in the reactcr cycle, assuming the f ailure of any single active component (i.e., a stuck rod) to shutdown the reactcr f rom any cperation condition and to maintain subcriticality over the f ull rango of coolant temperatures expected during shutdown. Allowance will also be made f cr the maximum reactivliy fauli associated with any anticipated occurrence.

loss of electrical power to a rotcr-roller not mechanism mounted on theScram actuation is obtained by roactcr closure head. The scr m release, located in the mechanism, causes coupled drivelino and control rod (movablo pin bundle) Insertion into the f ueled core region. Positive insertion is assured by a scre spring assist to sspplanent gravity accelerated Insertion.

The secondary control systm, using rods of significantly dif ferent design principles, will have suf ficient worth at any time in the reacter cycle, assuming the f ailure of any single active component (i.e., e stuck rod), to shutdown the reactcr frce any operating condition to the hot shu'cdown temperature of the coolant.

Allowance w il l also be made f or the maximum reactivity f ault associated with any anticipated occurrence. This reactivliy f ault alIowance is ineluded In the requiroments on both controf systems. The maximum reactivity f ault is postulated to occur upon the accidental uncontrolled cr withdrawal ItIcal conf Iguratt of the highest worth control rod in the reactor in any on. Secondary systcm diversity rolattve to the primary systcm features.

is provided by utilizing mechanical components of dif ferent design Scram actuation is initiated by Icss of eiectrical power to solenoids located in the mechanism mounted on the reactcr closure head. The solonolds vent pressure in a piston mounted in the mechanism. Loss of pressure actuates a scrm latch located in the control assembly and causes the control rod to scrm. A hydraulic scram assist suppiments gravity for positivo insertion.

O 3 .1 -43 Amend. 67 March 1982 s___... . . -

-u Criterion 25 - COMBINED __ REACTIVITY CONTROL SYSTEMS CAPABILIT_Y The reactivity control systems shall be designed to have a com-bined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriat, margin for stuck rods, the capability to cool the core is maintained.

Response; The CRBRP reactivity control system redundancy and capability, as identified in the response to Criterion 24 assures sufficient reactivity worth and insertion to shutdown the reactor under postulated events without loss of capability to cool the core. The primary and secondary systems individually provide this total reactivity control capability. In addition, tne requirement is established for both the primary and secondary systems to terminate transient events without loss of capability to cool the core.

Speed of response requirements for the primary system are established to limit fuel damage to limits of an operational incident, minor incident and major incident for Anticipated Faults, Unlikely Faults and Extremely Unlikely Faults, respectively. These limits are satisfied with provision for a stuck rod in the primary system. A major incident damage limit corresponds to preventing loss of coolable geometry while operational and minor incident limits allow less fuel damage than a major incident.

The secondary system, with provision for a stuck rod, has speed of response requirements which limit Anticipated and Unlikely Faults to minor incident and major incident damage limits, respectively. The combined primary and secondary system speed of response requirements together with total shutdown worth requirements assure that Criterion 25 is satisfied with margins for stuck rods.

32 Amend. 32 3.1-44 Dec. M

3.5.2.1.3 O Inictmediate Heat Transgort. SyMem (IHTSLPumo and Yor.11caLDrhn_AssemblyMiss!Ics Since the IHTS pumps are Idontical to the PHTS pumps the mechanisms for f ailure and subsequent missilo generation are the same. The higher IHTS pressures (during normal and accident conditions) will be an inconsequential contributor to missile generation.

3.5.2.1.4 EQ1m11Dg_ Component. Missile Selection from the Steam Donctator.lyM er 3 DSL Missilo generation f rom SGS compononts is postulated to occur as a result of recirculation pump overspood. The pump overspood condition would result from the sudden depressurization caused by a pipo break olther upstream or downstroom of the recirculation pump.

The recirculation pump and motor have boon designed such that es complete units the pump or motor will not becomo missiles. Analyses show that iho pump Impoller f alls bef oro the bolts that secure the motor and pump' to their supports f all .

Should the impoller fall, there is insufficlont energy to cause a failure of the support bolts. A twenty-five pound Impoller segment was chosen as a hypothetical missilo and the resultant analysis Indicated that this missile has insuf ficient energy to exit through the pump case or piping. (Sco Table 3.5-3).

( I The equipment within the cell and cell boundarios surrounding the recirculation pumps, which must perform a safety function have been ovaluated and shown not jeopardized by the of fects of a postulated missilo.

O 3.5-5a Amend. 65 Feb. 1982

3.5.2.1.5 Rotating Com enent Missile from Steam Generator Auxillarv Heat Renoval System (SGAHRS1 The auxillary feedwater pumps and pump drives are not postulated to genercte missiles, since they are not expected to overspeed. The motor drives are constant-speed synchronous motors. The drive turbine overspeed protection incorporates both an electronic high speed stop at 105% rated speed and a mechanical overspeed trip at 120% rated speed.

3.5.2.2 frfssurized comoonent Failure Missiles Pressurized component missiles result f rom the sudden release of stored strain energy (in the case of nuts, bolts and studs) or of confined fluid energy (in the case of equipment or vessels that contain highly pressurized fluids).

Structures, systems or portion of systems, and components such as valve bonnets, relief valve parts, hardware retaining bolts or Instrument wells are reviewed to determine their potential for generating missiles.

3.5.2.2.1 Missile Selection From Intermediate Heat Transport System (IHTS)

The IHTS is composed of three independent and physically separated low pressure (approximately within a pressure range frcen 98 psi to 210 psi) coolant loops. Since the pressure is low, the energy state of the conteined fluid is correspondingly low and no potential sources of high-energy missiles have been identifled.

3.5.2.2.2 Missile Selection From Steam Generator Auxillarv Heat Removal System (SGAHRS)

The selection of potential missiles and their characteristics are provided in Tabl e 3.5-2.

This listing of missiles represent an envelope in that the most energetic of the potential missiles in a given cell or area was chosen.

3.5.2.2.3 Missile Selection from the Steam Generator Svstem (SGS1 Potential missiles associated with the SGS are missiles generated by the recirculation pump as a result of pump overspeed, PACC Inlet isolation valve operator, and water dump valve operator. The recirculation pump postulated missiles are discussed in Section 3.5.2.1.

The remaining SGS equipment was excluded f rom consideration based on the criteria either that it could generate only limited impact forces or that the equipment design was sufficiently conservative such that fragmentation failure was not deemed credible. For example, bolting and drain line Isolation valves were eliminated as missiles because their calculated resultant kinetic energies are small. For large actuated valves, an adequate design margin is l required for the valve bonnet retainers. Valve snubbers are used to minimize seismic loads on the retainers. Thus, operators for such remaining actuated valves are not considered credible missiles.

O 3.5-6 Amend. 67 March 1982

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l l Figure 4.2-48L Decreaw in Yield Strength of Type 316 SS as a Function of Decrease in C+N Concentration

1. 85 2 i Reference 183) l 4.2-548a Amend. 66 Mar. 1932

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O conditions, Edition up to and load for the combinod including during the transient will bo included.

sot. Winter 1974 addenda, applicable to emergency The offects of cyclic pressure loading 5.4.2.3.2 Expansion _Innh An expansion tank is provided in each IHTS loop to accommodate the sodium volume chango in excess of that eccommodated in the IHTS pump tank, due to the thermal expansion associated with normal and off normal conditions. The pump tank and evpansion tank cover gas volumes are connected by a gas service line.

The expansion teak sodium lino is connected to the main lHTS cold leg piping just upstream of the IHTS pump suction nozzle.

5.4.2.3.3 1 HTS Ploing_anLSuppod The IHTS piping conducts sodium in a continuous loop to transport reactcr heat to the Steam Generation System. An isometric drawing of the IHTS piping and compononts 'n a typical steam generator cell is shown in Figure 5.4-2.

Each IHTS piping run is provided with the necessary elbows, teos and reducers to provido adequate loops for thermal expansion. While each loop contains iho necessary appondago piping and associated fittings, there are no valves in the main sodium flow piping. Each loop has similar components, although the loop piping differs in length and configuration because of the differences in distanco between the IHX units and steam generator modules.

O The hot leg piping is 24" OD x 1/2" wall Type 316H stainless stcol and extends from the IHX outlet nozzle through the reactor containment penetration to the superheater inlet nozzle. From the superheater outlet nozzle, the piping consic+c of two parallel runs of 18" OD x 9/16" wall 2-1/4 Cr - 1 Mo ferrific si col and extends to each of the two evaporator inlet nozzles. The cold leg piping commences at the evaporator outlet nozzles and consists of two parallel runs of 18" OD x 1/2" wall Type 304H stainless steel pipes. The linos are joined together ihrough expanders at a 24" x 24" x 24" Type 304 H mixing too and continuo as a single run of 24" OD .x 1/2" wall Typo 304H stair.less steel pipe which is connected to the 36" diameter IHTS coolant pump suction through a seven foot long diffuser. The cold leg continues as a 24" OD x 1/2" wall Type 304 H stainless steel pipo from the pump dischargo through the reactor containment ponotration and completes the loop at the IHX inlet nozzle. The Typo 316H and Typo 304H stainless steel pipos are joined to the 2-1/4 Cr - 1 Mo ferritic steel superheater and ovaporators, respectively, through Alloy 800H itonsition spool plocos. Transition spool pieces (2-1/4 Cr-1Mo/A800H/

316H) are used in the Instrumentation and steam generation vont lines.

The auxillary IHTS piping includes appendago piping for instrumentation, system high point vents and low point drains, fill linos, sodium dump lines and serulco connections for sodium purification.

O 5.4-9 Amend. 65 Feb. 1982

The lHTS piping except for the dump lines, will be designed in accordance with Section ill, Ciass I of the ASME Boller and Pressure Vesse Code. The dump / fill and drain lines downstream of the first dump valve and the gas equalization line downstream of the first valve and the rupture disc assembly wIII be designed in accordance wIth SectIon (lI, Class 2. The matrolal used in piping is In accordance with the applicable code and will be of all welded construction. Transition spool plecos (2-1/4 Cr-1Mo/A800H/316H) are utilized to join lines to iho dump tank.

All l HTS piping and components, except the dump / fill and drain piping downstroom of the first dump valve in series, the gas equalization line downstream of the first valve in series and the gas equalization bypass line downstream of the rupture disc assembly are required to be Saf ety Class 2 as a minimum and theref ore, ASME Boller and Pressure Vessel Code, Section ill, Class 2. as a minimum; however, they are being optionally upgraded and will be designed, constructed and code stamped in accordance with ASME B&PV Code, Section Ill, Class I. The dump /f!Il and drain piping downstream of the first valvo in series, the gas equalization line downstream of the first valve in series, and the gas equalization by-pass lIne downstream of the rupture disc assembly are required to be Saf ety Class 3 as a minimum, and therefore, ASME B&PV Code. Section 111, Class 3 as a minimum; however, they are being optionally upgraded and will be designed, constructed and code stamped in accordance with ASME B&PV Code, Section ill, Class 2.

O O

5.4-9a Amend. 67 March 1982

600 60 ll!

3 550 -

40 g HEOutRED NPSH - 963 RPM 8 8

C 500 - -

20 E z-E z

450 - -

0 -

90 t

400 -

80 PUMP EFFICIENCY - 963 RPM 2

w 350 -

70 E

- E U

z 5 3M I

60 9 HEAD CAPACITY - 963 RPM i

.J w

h l R

250 -

50 200 -

- 40 150 -

6 ,

O z

b 3

100 _

0 4 >-

POWER @ SPECIFIC GRAVITY = 0.87 d

W 963 RPM h s

w -

2E e

u 4

l l l l l l l k O O 12 16 20 24 28 32 36 40 SODIUM FLOW, THOUSANDS OF GALLONS PER MINUTE 82-136-02 Figure 5.4 3. INTERMEDIATE PUMP CHARACTERISTICS 5.4-38

?mnd. 67 liar. 1982

l O,

43 l Figure 5.4-4 has been intentionally deleted.

(

O i

i i

O 5.4-39 Amend. 43 Jan.1978 1

1

o Material Specifications A list of material specifications for the SGAHRS vessels, piping, pumps, and valves is given in Table 5.6-3. Corresponding weld materials specifications are listed in Table 5.6-4.

5.6.1.1.5 Leak Detection Requirements Leak detection requirements for the SGAHRS are as follows:

a. Excessive leakage of high pressure and temperature water from the steam generator system into the SGAHRS will be detectable

b. Excessive leakage of low pressure water will be detectable The methods used for leak detection are described in Section 5.6.1.2.5.

5.6.1.1.6 Instrumentation Requirements Functional requirements of the SGAHRS instrumentation are to monitor the following parameters and to warn the plant operator of any abnormal or dangerous conditions in the following parameters:

1. Protected water storage tank level, pressure, and temperature

2. Auxiliary feedwater pump inlet pressure and temperature V 3. Auxiliary feedwater pump discharge temperature and pressure

4. Auxiliary feedwater flow and temperature

5. Position of all isolation and control valves 58 l 6. Drive turbine steam supply and discharge pressures

7. Operating status of protected air cooled condenser

8. Operating status of all motors

9. Startup of air-cooled condenser

10. Startup of auxiliary feedwater pumps.

17 The plant protection system instrumentation and control equipment associated with the active components which must operate to insure that SGAHRS performs its safety function are described in Section 7.4.

5.5-3 Amend. 58 t

x s) Nov. 1980

5.6.1.2 Design Descriotion 5.6.1.2.1 Resign Methods and Procedures 5.6.1.2.1.1 Identification of Active and Passive Components which inhibit Leaks The equipment of the SG# IRS is shown schematically in Figure 5.1-5. Valves and pumps within the SG#1RS are classified as active or inactive, and their operating mode is given in Tables 5.6-5 and 6.

In the event of a pipe break in the auxiliary feedwater portion of SGAHRS, (ontinued heat rmoval capability will be assured by the multiple loop feature of the SGNIRS and heat transport system.

If a pipe break occurs in any portion of a steam generator loop, this will result in a reactor shutdown and an AFW initiation. Autcmatic isolation of the AFW supply to the af fected loop will occur within approximately 2 minutes when the steam drum pressure f alls below 200 psig. Operatcr action as a backup is available.

If after an AFW initiating event, a pipe break were to occur in the auxillary feodwater piping between the steam drum and the isolation valves immediately downstream of the control valves, the flow in the effective loop will increase until limited by the control valve (at approximately 110% of rated flow). A fIow Iimit alarm in the controf room wIII alert the operatcr to the f act thai corrective action is necessary. Following the control valve flow limit alarm, the operator verifles a leak from Information provided by the following instrumentation:

a) Saf ety-related steam drum level and pressure indication are provided or, each loop to assist in making a break determination. An inability to recover level cr maintain pressure on any steam drum with a corresponding flow limiting alarm on AFW provides a break Indication, b) The Steam Generator Building (SGB) Flooding Protection Subsystem annunciates abnormal SGB temperature, humidity, and sump level in the control room to alert the operator to pipe breaks that could compromise SG#tRS operation (see Section 7.6.5).

c) The plant trip signal: high or low steam to main feedwater flow ratio or Ict steam drum level . A trip of this type will direct the operator's attention to the steam / water-side of the plant.

Operator etion in the control room will close two AFW supply isolation valves to isolate the defective loop. In addition to the above, autcmatic isolation will occur when the AFW flow rmains above 150% for 5 sec. (Indicating a flow limiter failure). Due to the flow limiting capalblity of the control valves, the leakage flow will be minimized and proper flow to the two remaining steam drums will continue even though one loop has suf fered a pipe break.

O 5.6-4 Amend. 67 March 1982 m

r3 If a leak were to occur in the AFW system upstream of the AFW isolation valves y! during normal plant operation (AFW in standby), it would be detected by a change in Indicated water level in the PWST.

The PWST is sized to retain suf ficient water for at least 30 days of SGAHRS operation while allowing for 10 minutes to isolate the AFWS f rom a large break. A large break creates restricted flow in both AFW control valves to the af fected loop (154 lb/sec combined flow). Analysis of SGAHRS water use, following a pipe break, determines the required usable water volume to complete a 30-day mission (sco Table 5.6-9). This evaluation uses the specified operator action time of 10 minutes which shows that water retnalns in the PWST af ter 30 days. The water volume remaining in the initially full PWST and steam drums af ter normal 30-day SGAHRS operation is actually suf f!cient to allow 22 minutes prior to loop Isolation by operator action following a large break. Smaller break sizes with less than control valve limited flow would allow longer operator action times.

PWST water supply for at least 30 days is asscred with an operator design response time of 10 minutes to Isolate the loop with a pipe break. Operator action within 10 minutes is f acilitated by control room Information noted above and the control room operation to isolate the af fected loop. Exceeding 22 minutes for operator initiated Isolation may reduce the PWST water supply to the SGAHRS to less than 30 days. Prior to depletion of the PWST supply, an alternate water supply is likely to be available.

The piping runs from the AFWP to the AFW isolation valves and from the turbine OV drive steam supply isolation valve to the turbine drive are low-pressure and low-temperature lines during normal plant operation. Both lines are subjectcd to high-pressure during AFW operating periods and the turbine supply line is also subjected to high-temperature conditions during the time the turbine is operating. However, the SGAHRS operating time is anticipated to be less than 2% of the plant operating time since the auxiliary feedwater portion of SGAHRS will not be utilized unless either the normal heat rejection system (main condenser) or feedwater supply system has been lost. Therefore, no piping breaks will be postulated for those piping runs.

A pipe break between the pump suction isolation valve and the protected water storage tank is not considered credible because of the low temperature (200 F) and the low pressure (15 psia) operating conditions. If such an event were to occur, the alternate water supply could be brought into service by the opc.ator. The alternate supply is provided by a separate header connecting the feodwater pumps to the 250,000 gal. condensate storage tank.

5.6.1.2.1.2 Design of Active Pumos and Valves in order to assure the functional performance of active components of the SGAHRS, the active ASME Section 111 Class 2 or 3 pumps and valves will be designed and tested in accordance with Reference 12, PSAR Section 1.6.

. \.)

5.6-4a Amend. 67 March 1982

TABLE 5.6-2 CL ASSIFICATICf; 0F SCAH25 C0!'.PO';EilTS QUAL IlY SAFETY ,  !;ATIO!;AL QUAll1Y ASSURANCF -

COMP 0'iEllT CLASS CODES STAl:DARDS* AS.'E 3rotected SC-2 ASME I!I/2 Group B ii A-4000

'la t e r Stoi c.';e Tank (PWST)

PWST SC-2 ASME 111/2 Group B f;A-40C0 Pipind PWST SC-2 ASME 111/2 Group B I,'A- 4 000 Valves

_j ______

Protected SC-3 l ASME III/3 Group C iA-4000 Air Cooled Ccndenser.

(PACC)

"ACC Pi pir.g SC-3 ASME 111/3 Group C iia-4000

- - - - - - - - --- .s - - - - . ... . . _ . . . . . _

Auxiliary SC-3 I

ASME III/3 Group C  !?A-4000 Feeduater Systeu (AFS)

Pipi>g AFS Pumps SC-3 ASME III/3 Group C t.A-4000 AFS Valves SC-3 ASME 111/3 Group C flA-4000 g

• NR'C Regulatory Guide 1.26 " Quality Group Classifications and Standards'"

March 23, 1973.

Amend. 17 5.6-30 Apr. 1976

TABLE 5.6-3 SGAHRS E0UlPMENT LIST AND MATERI AL SPECIFICATIONS ASME DESIGN DES IGN SECTION lll TEMP PRESSURE SGAHRS COMPONENT CODE CLASS MATERIAL" ( F) (PSIG)_

Air Cooled Condenser Bundio 3 CS 650 2200 Air Cooled Condenser Fan, Motor, Louvers - --

100 ----

Auxiliary Feedwater Pump 3 CS 200 2200 Pump Motor Drive - --

104 ----

Pump Turbino Drive Downstream of Admission Valuos -

CS 600 1250 Upstream of Admission Values -

CS 650 2200 Water Storage Tank 2 CS 200 15 SGAHRS Piping:

PWST to First Isolation 2 CS 200 15 Valve First isolation Valve to AFW Pumps 3 CS 200 100 AFW Pumps to AFW Headers 3 CS 200 2200 AFW Headers 3 CS 200 2200 AFW Headers to Electrically Operated Isolation Valvo 3 CS 200 2200 AFW Pump Test Loop to and between isolation Valvos 3 CS 650 2200 AFW Pump Test Loop From isolation Valves to PWST Fill Lino 3 CS 200 100 Isolation Valve to Main FW Lino 3 CS 650 2200 Superheater inlet Line to PACC 3 CS 650 2200 PACC to Evaporator Recirc Lino 3 CS 650 2200 AFW Pump Recirc to Orifico 3 CS 200 2200 l Orl f ico to PWST-Rocirc 3 CS 200 250 Superheater Veni Line (Upstream of Valve) 3 1 1/4 Cr-1/2 Mo 935 1900 Steam Drum Vent Line (Upstream of Volvo) 3 CS 650 2200 Superhoater Vent Line (Down-stream of Valve) 3 1 1/4 Cr-1/2 Mo 850 250 Steam Drum Vent Lino (Down-stream of Valvo) 3 CS 400 250 Steam Supply Line to Drivo Turbino 3 CS 650 2200

• CS - Carbon Steel -

5.6-31 Amend. 67 March 1982

LIST OF TABLES TABLE N0. PAGE 7.1-1 Safety Related Instrumentation and 7.1-7 Control Systems 7.1-2 List of Regulatory Guides Applicable 7.1-8 to Safety Related Instrumentation and Control Systems 7.1-3 List of IEEE Standards Applicable to 7.1-9 Safety Related Instrumentation and Control Systems 7.1-4 List of RDT Standards Applicable to 7.1-10 Safety Related Instrumentation and Control Systems 7.1-5 Deleted 49 7.1-6 Safety Related Electrical Instrumentation 7.1-13 23 and Control Equipment 7.2-1 Plant Protection System Protective 7.2-18 Functions 7.2-2 PPS Design Basis Fault Events 7.2-19 7.2-3 Essential Performance Requirements for 7.2-23 PPS Instrumentation 7.2-4 thru 40 24 Deleted 7.3-1 Containment Isolation System Design Basis 7.3-5 7.4-1 Sequence of Decay Heat Removal Events 7.4-9 7.5-1 Instrumentation System Functions and 7.5-34 Summary 7.5-2 Reactor and Vessel Instrumentation 7.5-39 44 7.5-3 Surmiation of Sodium / Gas Leak Detection 7.5-40 Methods 34 49  ?.5-4 Post Accident Monitoring 7.5-42 7.6-1 Uce of Refueling Interlocks 7.6-4 7.9-1 Control Room Arrangements 7.9-8 Amend. 50 7 - ". June 1979

LIST OF FIGURES FIG. NO. EDGE 7.2-1 Reactor Shutdown System 7.2-24 7.2-2 HTS Coolant Pump Shutdown 7.2-25 7.2-2A Typical Primary PPS instrument Channel 7.2-26 Logic Diagram 7.2-2AA RSS Bypass Function Block Diagram 7.2-27 7.2-28 Primary PPS Logic Diagram 7.2-28 7.2-2C 'ypical Secondary PPS Instrument Channel 7.2-29

',ogic Diagram 7.2-2D Secondary PPS Logic Diagram 7.2-30 7.2-3 Typical Primary Subsystem 7.2-31 7.2-4 Typical Secondary Subsystem 7.2-32 7.2-5 Functional Block Diagrams of the Flux-Delayed 7.2-33 Fl ux, High Fl ux, Fl ux-Pressure, and Reactor Vessel Level Protective Subsystems 7.2-6 Functional Block Diagrams of HTS Pump Frequency 7.2-34 and Pump Speed Mismatch Protective Subsystems 7.2-7 Functional Block Diagrams of the IHX Primary 7.2-35 Outlet Temperature and Steam to Feedwater Flow Mismatch Protective Subsystems 7.2-8 Functional Block Diagrams of the Flux-Total 7.2-36 Flow, Startup Nuclear, Modified Nuclear Rate, and Primary to Intermediate Flow Rate Protective Subsystems 7.2-9 Functional Block Diagrams of the Steam 7.2-37 Drum Level and HTS Pump Voltage I

Protective Subsystems 7.2-10 Functional Block Diagrams of the Evaporator 7.2-38 Outlet Sodium Temperature and Sodium Water Reaction Protective Subsystems 7.3-1 Containment Isolatton System Block Diagram 7.3-6 O

7-x Amend. 67 March 1982 1

g

() Figure 7.2-2B is a logic diagram of the Primary RSS logic trains.

The outputs from the comparators and 2/3 functions are inputs to a 1 out of 24 general coincidence arrangement. The output of the 1/241s an input to a 1 out of 2 with the manual trip function to actuate the scram breakers. The scram breakers are arranged in a 2 of 3. When 2 or more logic trains actuate the associated scram breakers, power to the control rods is open circuited and the control rods are released for insertion to shutdown position with spring assisted scram force. Open circuiting the control rod power initiates Heat Transport System shutdown.

In the Secondary RSS, the sensed variables are signal conditioned and compared to specified limits by equipment which is different from the Primary RSS equipment. The secondary logic is configured in general rather than local coincidence to provide additional protection against comon mode failure. Each instrument channel comparator outputs its trip or reset signal to a 1 of 16 logic module. The 3 redundant secondary instrument channels from each subsystem feed 3 redundant logic trains, which are coupled to the secondary scram actuators. Figure 7.2-ED is a logic diagram for the Secondary RSS logic.

The Secondary RSS consists of 16 protective subsystems and monitors I a set of parameters diverse from the Primary RSS as shown in Table 7.2-1.

However, since a measure of nuclear flux is necessary in both the Primary and Secondary RSS, nuclear flux is sensed with compensated ionization chambers in the primary while fission chambers are used in the secondary. The Primary g'! RSS monitors primary and intermediate pump speed while the Secondary RSS

V monitors primary and intermediate coolant flow. Similarly, the steam flow to feedwater flow ratio is used in the Primary RSS while the steam drum level is sensed for the Secondary RSS.

Figure 7.2-2C is a typical Secondary RSS instrument channel logic diagram. Each protective subsystem has 3 redundant sensors to monitor a physical parameter. The output signal from each sensor is conditioned for transmission to the trip comparator located in the control room. Redundant ,

instrument channels are used. When necessary, calculational units are placed in front of the comparators to derive additional variables. The output of the comparators are input to redundant logic trains in a. general coin-cidence arrangement.

Bypass of secondary comparators is implemented in the same fashion as in the primary system except that different equipment is used to provide the permissive comparator function.

57 Figure 7.2-2D is a logic diagram of the Secondary RSS logic trains.

The outputs from the instrument channels are input to a 1/16 general coinci-dence arrangement. The 1/16 output controls the solenoid power sources through isolated outputs. Isolated outputs are also provided to initiate Heat Transport System shutdown. A trip latch-in function is provided to assure that once initiated, the scram will go to completion. The remaining 43 redundant logic trains provide the other two signals for the 2/3 function.

,/-

1 Amend. 57 7.2-2 Nov. 1980

Figure 7.2-2 shows the RSS Interf ace with 1rhe Heat Transport System (HTS) pump breaker control. Two HTS pump breakers are connected in series for each HTS pump.

Each HTS breakers receives input from the Primary RSS and Secondary RSS pump trip logic. Upon receipt of a reactor trip signal from either Primary or Secondary RSS, the HTS pump breakers open to rmove power from the primary and intermediate pumps.

Provisions are made to allow testing of the HTS breaker actuation function during reactor operation. A test breaker is used to bypass the main HTS breaker during a test cond tion. Test signals are then Inserted through the Primary or Secondary RSS pump trip logic to open the main HTS breaker.

Mechanical Interlocks are provided on the bypass breakers to prevent more than one main HTS breaker in any loop from being bypassed at a time. Control interlocks are provided which make the breaker test inputs inef fective unless the bypass breakers are properly Installed. Main HTS breaker and test breaker position status is supplied as part of the RSS status display on the main control panel.

The RSS subsystems do not directly require the reactor operator or control system to impiment a protective action. However, manual control devices to manually initiate each protective f unction are included in the design of the Plant Protection System.

Where signals are extracted f rce the Plant Protectic,n System, buf fers are provided. ~Inese buf fers are designed to meet the requirments of I EEE-279-1971, RDT Standards C16-1T, Dec.1969 and C16-3T, Dec.1971. The buf fers prevent the of fects of f ailures on the non-PPS side from af fecting the perf ormance of the PPS eq11pment. The buffers aro considered part of the PPS and meet all PPS criteria.

System TestabIIItv Both Reactor Shutdown Systems are designed to provide on-line testing capability. For the Primary RSS, overlapping testing is used. The sensors are checked by comparison with redundant sensor outputs and related measurements. Each Instrument channel includes provisions for Insertion of a signal on the sensor side of the signal conditioning electronics and test points to measure the performance at the comparator (or calculational unit)

Input. Where disconnection of the sensor is unavoidable for test purposes, the comparator is tripped when disconnected. The Instrument channel electronics including trip comparators and bypass permissive comparators are tested for ability to change value to beyond the trip point and provide a trip input to the logic. The comparators and logic are tested by the PPS Monitor.

A set of pulsed signals are inserted from the monitor into the comparators associated with one subsystem and the logic output is checked by the Monitor to assure that logic trip occurs for the correct combinations of comparator trips. The logic and scram breakers are tested by manually tripping one logic train and observing that the corresponding breakers trip. HTS breakers are tested by maintaining power to the pump through a bypass circuit breaker and manually inserting a test signal to the pump trip logic.

O 7.2-2a Amend. 57 Nov. 1980

f^)

>V For the Secondary RSS, insertion of the test signal into a channel causes the entire train (comparator, logic, and scram solenoid valves) to trip. Testing of the pump breaker trip is identical to that for the Primary RSS.

System Instrumentation The Instrumentation used by the RSS to detect the occurrence of of f-normal plant conditions includes:

o Neutron Flux The Primary RSS uses three compensated Ionization chamber power range nuclear sensors evenly spaced around the reactor vessel. The Secondary RSS uses three fission chamber wide range nuclear sensors evenly spaced around the reactor vessel. See Section 7.5.2 for detector details.

o Reactor Inlet Plenum Pressure The Primary RSS uses six pressure detectors, two per HTS primary loop, located as close as practical to the reactor vessel inlet plenum in the elevated primary cold leg piping. Each set of two detectors comprises an instrument channel. The outputs of the two detectors in each loop are auctioneered. The resultant output signal is provided to the comparator. See Section 7.5.2 for detector details.

p) o Sodium Pumo Soeed .

The Primary RSS uses threo redundant tachometers per primary and intermediate HTS pump to measure pump speed. See Section 7.5.2 for detector detaiis.

o Sodlum Flow The Secondary RSS uses six permanent magnet flowmeters to measure HTS sodium flows. One flowmeter is located in each of the primary and intermediate cold legs. Each flowmeter provides three redundant measurements of loop flow. See Section 7.5.2 for detector details.

o Reactor Vessel Sodlum Level The Primary RSS uses four sodlum level detectors evenly spaced within the reactor vessel. Threo of these detectors provide redundant active signals to the RRS. The fourth detector is used as a spare. See Section 7.5.3 for detector details.

o Underfreauencv Relav The Primary RSS uses three underf requency relays, one per coolant loop pump bus. The underf requency relays are located on the HTS pump buses.

b 7.2-3 Amend. 67 March 1982

o Stonm Ffsw Tho Primary RSS usos throo redundant steam mass flow signals por loop.

Each steam supply loop has ono venturi fic,wmeter, three dif ferential pressure censors, throo temperaturo sensors, and three pressure sonsors located between the suporheater oxit and the main steam hoodor. Throo rodundant steam mass flow signals are generated by pressuro and temperaturo componsailon of tho venturl flowmoter analog signal. Soo Section 7.5.2 for detector details.

o Ecedwater Flow One venturl flowmoter, three differential pressuro sensors, and throo temperaturo sonsors por steam generator loop supply the Primary RSS with three redundant temperature-compensated foodwater mass flow si gnal s. Soo Section 7.5.2 f or detector detail s.

o ,1HX Primary Outiet Sodium Temperaturo The Primary RSS usos three redundant thermocouples, mounted in throo thormowells, por loop to measuro the sodlum temporaturo in the primary cold leg. Soo Section 7.5.2 for detector detail s.

o Steam DCum_Laval The Secondary RS uses throo rodundant reference column levol sonsors to deformino the water level in each steam drum. Tho lovel sensor is density componsated. Soo Section 7.5.2 for detector details.

o Evaporator Outiot Sodlum Temocratura The Secondary RSS uses throo redundant thermocouples por loop, mounted in thormowolls. Those thermocouples are provided to mon' tor the sodlum temperatures in each Intermodlate cold log. Soo Section 7.5.2 for detector dotalls, o Undervoltage Relay The Secondary RSS usos throo undorvoltago relays, ono por coolant loop pump bus. The undervoltago relays are located on the HTS pump buses.

o Sodlum-Water Reaction The Secondary RSS usos throo rodundant pressuro sensors located in the reaction products vont lino Immt ,toly downstream from each rupturo disk to detect if tno rupture disks have blown. Soo Section 7.2.1 for details.

The configuration of the instrumentation in the protectivo subsystems is described in Section 7.2.1.2.

O 7.2-4 Amend. 67 March 1982

lPrImarv55utdownSystemLoalc The Primary RSS Logic is implemented using Integrated circuits to minimize the scram delay time. Other advantagos include minimizing power consumption and l spaco required and maximizing testability, maintainability and rollabilIty.

The Primary RSS Logic is arranged as shown in Figure 7.2-3.

l In each logic train, twenty-four 2/3 coincidence logic circuits feed a 1/24 module, whose output is coupled to the final actuation logic and rod actuators by a transistorized power ampl if ier. When only one comparator of any or all protective f unctions is tripped, the logic signal output rcrnains positive (reset). When any two comparators of a protective f unction trip and provido a negative logic signal to the protective logic, the output of the corresponding 2/3 modulo also trips to a negative logic signal. This negative logic signal in turn trips the 1/24 logic module which outputs a negative logic signal to the final actuation logic and rcrnovos power from the scram breaker undervoltage coll.

Light emitting diodos and phototransistors are utilized to provido complete electrical isolation at strategic points through the Primary RSS logic. There is no electrical connection between the comparator output and protective logic input. Consequently, an internal electrical fault in a single Instrument channel or comparatcr cannot propagate to the other channels, protective f unctions, or logic trains of the protectivo system. Each logic train is electrically isolated f rom the other so that protective action can bo initiated regardless of any Internal electrical f ault in a single logic train.

The equipment needed to impicrnent the 24 protective subsystems of the Primary RSS includos the sensors, signal transmitters and ampiIf lers or equivalent, calculational units, comparators, logic isolators, 2/3 logic modules,1/24 logic modutos, logic drivers, final scram actuation circuitry and breakers, buf f ers, permissivos and bypasses. A three section equipment cabinet is used to house the equipment for each of the three Instrument channels including the calculational units, comparators, power supplles and buf fers. A two section equipment cabinet is used to house the equipment for each of the throo logic trains and single equipment cabinets house signal conditioning equipment for each channol. This arrangement of equipment within cabinets provides the nocessary mechanical separation of redundant equipment.

O 7.2-5 Amend. 57 Nov. 1980

Secondary Reactor Shutdown System Loalc The Secondary RSS Logic consists of the 16 protective subsystms arranged in a general coincidence conf Iguration, as shown in Figuro 7.2-4. In this arrangement, the outputs of instrument channel A comparatcrs are directly coupled to a 1/16 logic circuit modulo in logic train A, as are the outputs of instrument channel B with logic train B and the outputs of instrument channel C with logic train C.

When the sensed parameter in an Instrument channel exceeds its sotpoint (trips), the comparator outputs a zero (trip) signal to the 1/16 logic module, which in turn outputs a zero (trip) signal to the scr m latch and scr m solonoid val ves (sco Figuro 7.2-2D). The 1/16 logic modulo output voltage changes to zero regardless of the output of the other comparatcrs. The output l of the 1/16 logic modules are combined in a 2/3 coincidence by the 3 solenoid valvos located within the Secondary RSS rod. Electrical isolation of the logic output to the solenoid drivers and Heat Transport Systm shutdown logic (HTS pump breakers) is shown in Figure 7.2-2D. Redundant isolated outputs are provided f rom each Secondary RSS logic train to the Secondary RSS pump trip logic where they are combined in a 2/3 logic. Trip signals are provided to the HTS pump breakers when 2/3 of the redundant Secondary RSS channels are in a tripped condition.

The equipment of the Secondary RSS includes sensors, signal conditioning equipment (transmitters), calculational units, comparators,1/16 logic modules, solenold drivers, secondary final actuation IogIc and actuatcrs, buf fors, permissives and bypasses. The equipment is designed using hardware which is diverse f rom that used in the Primary RSS. Since each Instrument channel is uniquely associated with a logic train, a four section equipment cabinot houses each of the Instrument channel comparators, logic trains and solonoid drivers. Single equipment cabinets are used to house signal conditioning equipment f or each channel. This arraagement of equipment within separate, compiotely metaliIcally enclosed cabinets providos the necessary mechanical separation between redundant equipment.

Channel Outout Monitoring Channel output monitoring is included to provide the operatcrs with early Indication of anomalous instrumentation perf ormance. This equipment is not sof oty rel ated. If the output of one channel dif fers from either of the redundant channels by more than a preset amouni, the channel output monitcring circuitry alarms this condition.

7.2.1.2 Design Basis Information The RSS Initiates and carries to completion Reactcr, Heat Transport and Balanco of Plant Shutdown if any of the of f-ncrmal plant conditions listed in Tabl o 7.2-2 occur. The table also shows the frequency classification of the postulated f ault, and the first Primary and Secondary RSS subsystms which act to terminate the f ault. As detailed in Chapter 15, the RSS design described below providos the perf ormance necessary to appropriately limit the results of the postuiated events. Tablo 7.2-1 shows the Primary and Secondary RSS subsystems which use the instrumentation described previously to determine the of f-normal conditions and trip the plant.

7.2-6 Amend. 59 Dec. 1980 m _

7.2.1.2.1 Primarv Reactor Shutdown Svstem Subsystems Hlah Flux Tho High Flux Protectivo Subsystem (Figure 7.2-5) Initiates trip for positive reactivity insertions at or near f ulI power. This subsystem assures that sustained operation will not occur with the f uel near incipient conterline mol ti ng. As shown on the figure, the subsystem comparos the compensated Ion chamber output signal with a flxod setpoint and initiates trip when the signal excoods the setpoint. Analysis of the perf ormanco of this subsystem is based on worst caso time response of the instrument of 50 milliseconds and worst case trip point of 115% of full power. This subsystem is never bypassed.

Flux-Dolaved Flux The Flux-00 layed Flux Protectivo Subsystems (Figure 7.2-5) initiate a trip for rapid sustelnod reactivity disturbances, either positive or negativo, which occur anywhoro In the load rango. Two subsystems are provided; one for positivo flux rates and one for negative flux rates. These subsystms provent undostrod thermal transients caused by rapid changes in power with flow hold corstant. As shown on the figure, the flux signal is compared with a signal proportional to the nominal load level as measured by pump speed and with the output of a long lag circuli whoso input is flux. To initiate trip f or increasing reactivity disturbancos, the flux signal is a negative input to the comparator. For decreasing reactivity disturbancos, the flux signal is positivo and the output of the lag circuit is negative. The operation of this subsystem is such that the trip point is dependent on the Initial condition, O. rate of power chango, and magnitudo of power changes.

power, thoro is a threshold magnitude of power change to cause a trip. A step For a given Initial chango of smaller magnitude than the threshold value will not cause a trip.

Power changes greater 'than the throshold value will initiate a trip with a lower total power change than a slower ramp rato. The trip equation constants are adjusted to provido the necessary protection for the range of normal power conditions without significantly impairing the plant operations. Worst case val ues of the constants, instrument response times and repeatabilItles are used in analyzirg the perf crmance of the subsystem. This subsystem is never bypassod. The negative flux rate subsystm must be bypassed for plant startup. Nuclear flux is usod as a permissivo signal, if nuclear flux is loss than 20% of util power flux a bypass can be manually instated. The bypass is automatically removed when power is increased above the permissive l ev el .

Flux- Pressure The Flux-Pressuro* Protectivo Subsystem (Figure 7.2-5) Initiates trip for positivo reactivity excursions or reductions in primary flow over the load rango. Two prossure sonsors are used for each redundant channel of the sy stem. This arrangement assures' appropriate rodundancy whilo providing of foctivo plant operational characteristics since pressure sensor replacement

• Formorly ref erred to as Fl ux-4 Pressure subsystm.

O 7.2-7 Amend. 67 March 1982

cannot be carried out on-lino. The use of the blSh auctioneer autcmatically accommodates f all t.ro of a sensor. All six pressure sensor outputs are comparod in the channel output monitoring circuitry to provide early indication of anomalous perf ormanco to the operating personnel. The subsystem perf ormanco is a f unction of initial operating level and is analyzed using worst caso values f or the Instrumentation and electronics including response time for the pressure instrumentation of 150 milliseconds. This subsystem must bo bypassed for plant startup. Nuclear flux is used as a permissive signal. If nuclear flux is less than 10% of full power flux a bypass can be manually instated. The bypass is autmatically rmoved when power is increased above the permissive level.

BTS Pumo Frecuencv The HTS Pump Frequency Subsystem (Figure 7.2-6) providos protection for loss of pumping power f cr two or throo HTS loops. Three underfrequency relays, one on each bus, aro used as redundan1 channels. If two of throo reli.ys are trippod, roactcr trip onsues. A timo delay !s used to allow the plant to continuo through mcmentary power outages.

Primarv-Intermediate Pumo Soeed Mismatch The Primary-intermediate Pump Speed Mismatch Subsystms (Figure 7.2-6) initiate trip f cr imbalances in heat rmoval capability between the primary cnd intermediate circuits within a heat transport loop. Three subsystems are incl uded, one f cr each HTS loop. As shown in the f Igure, the primary and intermediato spood signals are normalized and subtracted. The absolute val ue of this dif fcrence is compared with a flxod bias and a lIncar ratio of the primary speed to determine trip initiation. The actual trip point is dependent on initial conditions. Worst caso values are used f cr analysis incl uding a 20 mil lisecond tachomotor time constant. Those subsystems must be bypassed to start the plant. The permissive signal used is the nuclear flux, if the nuclear flux is less than 10% full power flux, the subsystm can be bypassed manually. For two loop operation, provisions are made to bypass the f unction associated with the shutdown loop. A manual bypass is Instated under administrativo control by changing the hardware conf Iguation. Two loop bypassos are also under permissivo control. Nuclear flux must be less than 10% of full power flux at the time of instating and the primary HTS pump speed in the shutdown loop must be loss than 15% of the speed producing ncrninal full flow or the two loop bypass is autcznatically removed.

Reactor Vossol Lovel Tho Roactcr Vessel Level subystm (Figuro 7.2-5) prevents reactcr oporation unless the sodium level in the reactcr vessel is at least 6 inches above the supprosscr plato. The output of the level sensor is compared with a f ixed sotpoint to determino the need for a reactor trip. Wcrst case values are used in the analysis of the perf crmance of this subsystm including a sensor time constant of 0.5 second. This subsystcm is never bypassed.

O 7.2-8 Amend. 67 March 1982

Steam-Feed Flow Mismatch n 57l The Steam-Feed Flow Mismatch subsystem (Figure 7.2-7) initiates reac-( j tor trip to prevent continued operation with large imbalances between the V steam and feedwater flow for each HTS loop. One of these subsystems is

. included in each HTS loop. These subsystems protect the steam generators and drums against unacceptable thermal transients. As shown in the figure, each subsystem compares the steam flow and feedwater flow, both of which are mul-tiplied by appropriate constants, in two individual comparators. If the difference between the two values exceeds the setpoint in either of the com-parators, a trip is initiated. Increasing steam flow and decreasing feed-water flow fault events are sensed by the first comparator. The second com-parator senses decreasing steam flow and increasing feedwater flow fault events. Analysis of this function is based upon worst case parameter values.

This subsystem must be bypassed for plant startup. A permissive is included which allows manual bypass of this subsystem for nuclear power less than 10%.

Two loop bypass provisions are also included for the shutdown loop. Two loop bypasses are also under permissive control. Nuclear flux must. be less than 10% of full pcwer flux at the time of instating and the primary HTS pump speed in the shutdown loop must be less than 15% of the speed producing 57 nominal full flow or the two loop bypass is automatically removed.

IHX Primary Outlet Temperature The IHX Primary Outlet Temperature subsystem (Figure 7.2-7) compares the sodium temperature in the primary cold leg of each IHX to a fixed set point. A reactor trip is initiated if the sodium temperature exceeds this set point. These subsystems assure that temperature increases in an inter- ,

A mediate loop sodium resulting from steam side fault events or intermediate C) flow reductions do not increase the reactor coolant temperature. There is one IHX primary outlet temperature subsystem per HTS loop. These subsystems are never bypassed.

Secondary Reactor Shutdown System Subsystems 57l 7.2.1.2.2 Modified Nuclear Rate The Modified huclear Rate subystems (Figure 7.2-8) initiate trip for J rapid sustained reactivity disturbances which occur in the load range. Two sub-systems are provided. One for positive flux rates and one for negative flux rates. These subsystems prevent undesired thennal transients caused by rapid changes in power with flow held constant. The reactor trip is based on flux rate measurements from the fission counters. A pennissive is included which allows manual bypass of the negative rate subsystem for nuclear power 57 less than 10%. The positive ~ rate subsystem is never bypassed.

Flux-Total Flow The Flux-Total Flow Subsystem '(Figure 7.2-8) provides protection against increasing and decreasing flow and power events over the 40 to 100%

load range. The primary flows of the three HTS loops are summed and multi-plied by an appropriate gain. -A nuclear power signal obtained from the fis-sion counters is subtracted in the comparator from the total flow value and

.fm this difference is compared to a fixed set point. If the difference exceeds

-(v j the set point, then a reactor trip is initiated. Analysis of this subsystem is. based on worst case parameter values, including a 500 msec. time delay for

-the~ flow detectors. This subsystem is never bypassed.

Amend. 57 7.2-9 Nov. 1980

j Startup Nuclear The Startup Nuclear Subsystem (Figure 7.2-8) obtains a wide range log channel measurement of nuclear power from the fission counters and compares It to a fixed-set point, if nuclear power is greater than the set point, a reactor trip is initiated.

A perrnissive module is provided which allows manual bypass of this subsystem upon the verification of the operation of the wide range linear channel. This subsystem provides protection against positive reactivity disturbances occurring during startup.

Erlmarv to intermediate Flow Ratio The Primary to intermediate Flow Ratio subsystems (Figure 7.2-8) protect against an Imbalance in the heat removal capability of the primary and intermediate loops. The heat rocoval capability of a particular loop is determined by measurement of the sodium flow within the loop. The Secondary RSS includes two of these subsystems, Primary Flow High and Primary Flow Low.

In the Primary Flow High subsystem, the output of the high primary flow auctioneer is compared to the summation of the outputs frcm the low Intermediate flow euctioneer and a signal proportional to the total primary f l ow . When the high primary flow auctioneer signal exceeds the low Intermediate flow auctioneer signal by an amount proportional to the total primary fIow, a reactor trip Is initiated.

Similarly in the Primary Flow Low subsystem, a comparison is made between low primary flow od high intermediate flow. When the high intermediate flow auctioneer signal exceeds the low primary flow auctioneer signal by an amount proportional to the total primary flow, a reactor trip is initiated. These subsystems are manually bypassed during plant startup.

used is based on reactor power. If reactor power is lessThe permissive signal than 10%, the subsystems can be manually bypassed.

Steam Drum Level The Steam Drum Level Subsystems (Ff' ).'

i and compare it to two Individual ice , t ,a!ats. d) measure steam drum water level A reactor tripp is initiated whenever the drum water level w;mses below this fixed setpoint.

There are three of these subsystems, one per HTS loop. Analysis of these subsystems are based upon worst case parameter values. For two loop operation, a manual bypass is instated under administrative control by changing the hardware conf iguation. Two loop bypasses are also under permissive control.

Nuclear flux must be less than 10% of full power flux at the time of. Instating and the primary flow in the shutdown icoa must be less than 15% of f ull flow or the two loop bypass is automatically removed.

HTS Pumo Voltage The HTS Pump Voltage Subsystem (Figure 7.2-9) provides protection for loss of pumping power for two or three HTS loops. Three undervoltage relays, one on each HTS pump bus, are used as redundant channels. If two of the three redundant channels are tripped, reactor trip ensues. A time delay is included to allow the plant to continue operation through momentary power outages.

(

7.2-10

{ Amend. 67 1

March 1982

O e Scram Actuator Logic

, o Heat Transport System (HTS) Shutdown Logic e Control Rod Drive Mechanism (CRDM) Power Train 57 e PPS Isolation Buffer Figures 7.2-3 and 7.2-4 provide assistance in locating the above 571 system level components within the overall RSS.

The probability of occurrence of each failure mode is listed in the tables of Appendix C, Supplement 1, in the Probability Column.

57l 41 The effects of each potential failure mode have also been categorized in the tables in the Criticality Column. Even though the failure of an individual element may result in the inability to initiate channel trip, the pro-vision of redundant independent instrument channels and logic trains assures that single random failures cannot cause loss of either che Primary or 57l Secondary RSS thereby meeting the design requirements of RDT C16-1T and IEEE 279-1971. The high reliability of components, redundant configuration, provision for on-line monitoring and on-line periodic testi'ig further assure that random failures will not accumulate to the point that trip

57) initiation by either Primary or Secondary RSS is prevented. All O..

failure effects are therefore categorized as not causing any degradation 41 or failure of'a system safety function. The majority of the identified failure modes can be eliminated from consideration based on their low probability of occurrence and the insignificance of their criticality. They are included in the FMEA, however, to document their consideration.

O 7.2-17 Amend. 57 Nov. 1980

TABLE 7.2-1 PLANT PROTECTION SYSTEM PROTECTIVE FUNCTIONS Primarv Reactor Shutdown System o Flux-Delayed Flux (Positive and Negative Flux Rate) o Flux- Pressure o High Flux l o Pump Speed Mismatch (Primary to Intermediate) o HTS Pump Frequency o Reactor Vessel Level o Steam-Feedwater FIow Mismatch l

o IHX Primary Outlet Tanperature '

Secondarv Reactor Shutdown Svstem o Modified Nuclear Rate (Positive and Negative Flux Rate) o FIux-Total FIow o Startup Flux o Flow Mismatch (Primary to Intermediate) o Steam Drum Level o Evaporator Outlet Sodium Temperature o HTS Pump Voltage o Sodium Water Reaction O

7.2-18 Amend. 67 March 1982

TABLE 7.2-2

_ PPS DESIGN BASIS FAULT EVENTS Primary Reactor Secondarv Reactor Fault Events Shutdown Svstem Shutdown Svstem

1. Anticloated Faults

. A. Reactivity Disturbances III-l Positive Ramps 15d/sec and Steps 510 '

Startup Flux-Delayed Flux or Startup Nuclear Flux- Pressure

( 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate or l j Flux- Pressure Flux-Total Flow 100% Power Flux- Pressure Flux-Total Flow

[

Full Power High Flux Flux-Total Flow

.L

• Negative' Ramps and Steps Flux-Delayed Flux Modified Nuclear Rate B. Sodium Flow Disturbances Coastdown of a. Single Primary or Primary-intermediate Primary-intermediate Intermediate Pump Speed Mismatch Flow Ratio Loss of 1 HTS Loop Flux-Pressure Primary-intermediate Flow Ratio Loss of 3 HTS Loops HTS Pump Frequency Flux-Total Flow IE aa

=r ,a w

J-

~

,. p ,

\ ._ .

TABLE 7.2-2 (Continued)

Fault Events Primary Reactor Shutdown System Secondary Reactor Snutdown System

-57 C. Steam Side Disturbances Evaporator Module. Isolation Valve IHX Primary Outlet Evaporator Outlet Na Closure Temperature Temperature Superheater Module Isolation Valve Steam-Feedwater Flow Evaporator Outlet Na Closure Mismatch Temperature Water Side Isolation and Dump. IHX Primary Outlet Evaporator Outlet Na of Single Evaporatsr Temperature Temperature Water Side Isolation and Dump Steam-Feedwater Flow Evaporator Outlet Na of Single Superheater Mismatch Temperature

[ Water Side Isolation and Dump of Steam-Feedwater Flow Evaporator Outlet Na 4 Both Evaporators-and Superheater Mismatch Temperature o

Loss of Normal Feedwater Steam-Feedwater Flow Steam Drum Level Mismatch 47l. Turbine Trip with Reactor Trip Steam-Feedwater Flow Steam Drum Level (Loss.of Main Condenser or Mismatch Similar Problem)

Inadvertent Opening of Evaporator Steam-Feedwater Flow Steam Drum Level Outlet Safety Valve Mismatch Inadvertent Opening of Superheater Steam-Feedwater Flow Steam Drum level Outlet Safety Valve Mismatch EE Inadvertent Opening of Evaporator IHX Primary Outlet Evaporator Outlet Na 5$ Inlet Dump Valve Temperature Temperature w ."

4 1

TABLE 7.2-2 (Continued)

Primarv Reactor Seconderv Reactor Fault Events Shutdown Svstem Shutdown System ll. Unlikelv Faults A. Reactivity Disturbances (2)

Positive Ramps 155d/sec and Steps _<60 Startup Flux-Delayed Flux or Startup Nuclear Flux- Pressure 5-40% Power Flux-Delayed Flux or Modifled Nucleat Rate or Flux- Pressure Flux-Total Flow 40-100% Power Flux-Pressure FIex-Total FIow Full Power High Flux Flux-Total Flow w B. Sodium Flow Disturbances Primary Pump Seizure Primary-intermediate Primary-intermediate Flow Speed Mismatch Ratio Intermediate Pump Seizure Primary-Intermediate Primary-intermediate Flow Speed Mismatch Ratio Loss of 2 HTS Loops HTS Pump Frequency Primary-intermediate Flow Ratio C. Steam Side Disturbances (3)

Steam Line Break Steam-Feedwater Flow Evaporator Outlet Na Mismetch Temperature Recirculation Line Break Steam-Feedwater Flow Steam Drum Level Mismatch EE gg Feedwater Line Break (3) Steam-Feedwater Flow Steam Drum Level l

r .o- Mismatch O O O l

, \. ,

O V ,

Table 7.2-2 (Continued)

Fault Events Primary Shutdown System Secondary Shutdown Svsfam~

j Failure of Steam Dump System Steam-Fe.' water Flow ~ Steam Drum Level Mismatch Sodium Water Reaction in Steam Steam-Feedwater. Flow Sodium-Water Reaction Generator Mismatch III.. Extr eelv Unlikelv A. Reactivity Disturbances l

l Positive Ramps 52.0/sec Startup Flux-Delayed Flux Startup Nuclear 5-40% Power Flux-Dolayed Flux or Moditted Nuclear Rate or l Flux- Pressure Flux-Total Flow 40-100% Power Flux- Pressure Flux-Total Flow w

m Full Power High Flux Flux-Total Flow (1) The maximum anticipated reactivity fault results from a single failure of the control system with a maximum insertion rate of approximately 4.1 cents per second.

(2) The maximum unlikely reactivity f aults result f rom multiple control system f ailures leading to withdrawl of six rods at normal j l

speed or one rod at the maximum mechanical speed.

(3) The PPS is required to terminate the results of these extremely unlikely events within the umbrella transient spectried as emergency.for the design of the major components.

E F-

?aa co en

. e-* N

TABLE 7.2-3 ESSENTI AL PERFORMANCE REQUIREMENTS FOR PPS INSTRUMENTATION Accuracy Response Time Plant Parameter (T of span) (msec)

Neutron Flux Primary 11.0 <10 Secondary 11.0 <10 Reactor Inlet Plenum Pressure 12.0 <150 Sodium HTS Pump Speeds i2.0 <20 Sodium HTS Flow 15.0 <500 Reactor Vessel Sodlum Level 5.0 <500 Undervoltage Relay 11 .0 <230 Steam FIow 12.0 <500 Feedwater Flow i2.5 <500 Evaporator OutIet Sodium Temperature 2.0 <5000 Steam Drum Levcl 11 .0 <1000 IHX Primary Outlet Temperature i2.0 <5000 Underfrequency Relay 12.0 <200 i

O 7.2-23 Amend. 67 ,

March 1982 l

O UNDER-FREQUENCY . VARIABLE _

RELtY LOOP 1 TIME DELAY g +

TO NON-PPS c LOOP 1 PUMP PRIM ARY T ACHO-SPEED MISMATCl!

METER LOOP 1 - BUFFER K -

CIRCulTRY INTERME01 ATE i ELECTRONICS H' _

TACHOMETER ABSOLUTE _

LOOP 1 VALUE 9 ELECTRONICS l -

TO NON-PPS m TO NON-PPS -

[ BUFFER H +

- + MATCH PRIM ARY TACHO-METER LOOP 2 1 BUFFER l , K -

BYPASS ,

CIRCulTRY ELECTRONICS l '

+ ABSOLUTE INTERMEDIATE -

TACHOMETER LOOP 2 ELECTRONICS l TO NON-PPS 9 BUFFER l + LOOP 3 PUMP TO NON PPS

• ---]

l BUFFER } ,

K

+ hh ~

BYPASS PRIMARY TACHO- '

CIRCulTRY M ELECTRONICS j + ABSOLUTE ,_

INTE RME DI ATE - , - VALUE FROM TACH 0 METER ELECTRONICS I PERMME LOOP 3 I

TO NON-PPS c

' " ~

Figure 7.2-6. Functional Block Diagrams of the HTS Pump Frequency and Pump Speed Mismatch Protective Subsystems. One Channel of Three is Shown 7.2-34 Amend. 67 Mar. 1982

TO NON-PPS

• BUFFER TEMPERATURE DETECTOR ELECTRONICS

~

+

LOOP =l TO MON-PPS

• BUFFER TEMPERATURE DETECTOR ELECTRON ICS -

LOOP '2 +

g sT TO NON-PPS e BUFFER TEMPERATURE DETECTOR -

ELECTRONICS -

LOOP =3 +

NOTE: TEMPERATURE AND PRESSURE

,pp COMPENSATION SENSORS AND E:.ECTRONICS OMITTED FOR ST EAM CLARITY

.L0wMETER -

LOOP =l - BUFFER

+ BYPASS

+ ~~

FEEDWATER -

ELECTRONICS WD -

K FA -

r CIRCulTRY FL0wMETER LOOP =l ELECTRONICS k C RCU TRY ~1' TO NON-PPS ; - - -

- ~

BUFFER SB TO N ON-PPS -*--

STEAM FLOWMETER - BUFFER h- K SA -

+

LOOP =2 -

-46

+

ELECTRONICS -< M> -

"FA r- CIRCUITRY FEEDWATER - FROM F LOWN

• PERMI SS I V E

_]L BYPASS LOOP 62 ELECTRONICS -16 (FIGURE

+

CIRCUlTRY 7.2-5)

TO NON-PPS

• BUFFER - ~

TO NON-PPS c .

STEAM - -

BUFFER A -

FLOWMETER -

+ BYPASS LOOP 43 *

+'- K r- CIRCUlTRY E W TRONICS FA -

FEEDWATER -

FLOWMETER ELECTRONICS l- CIRCulTRY LOOP = 3 -

- K ^

BUFFER SB l

TO N 6 4-PPS **~~- -

l Figure 7.2 7. Functional Block Diagrams of the IHX Primary Outlet Temperature and Steam to Feedwater Flow Mismatch Protective Subsystem, One Channel of Three is Shown 6678-3 7.2-35 Amend. 57 Nov. 1980

TO NON PPS

• BUFFE2 m g r R E L ECTRONCS *l Flux TOTAL OFF i F-l

- l .,,

CHANNEL ,g STARTUP NUCLEAR EINE AN EVI AII -%i CHANNE L CI RCUlT R Y ll e et g d MOOT F IE D SE TPO4NT 1 NUCLE AR RATE SE TP084 T 2 -

+

n,S

=

, J 3"++ 'N E G A I'VI I F-BYP ASS _ ,

PRIM AR Y F LOW- I CIRCUIT RY d E LE C T RONICS '

PR IM A R Y F ME TE A LOOP NO 1 , PM AUCTIONE E R TO NON-PPS SUFFER h HIGH I -

pg ll - pgon A +

d E LECTRONaCS h>

Y- .

TO NCN-PPS BUFFER PR IM A R Y

, PL FLOW AUCT6ONE E R PRIM AR Y S LOW- '

LOW ME TE R LOOP NU 3 d E LECTRONICS '

TO NON-PPS SUFFEH -

IN T E RME DI A T E FT FLonME TE R d ELECTROhiCS p W

• i c LOOP NO 1 .

TO NON- PPS % eUFFER F .__ K 3

eN T E RME DI A T E q INT E RME DI A TE F L OVWAE T E R M E LECTRONaCS h L(Me NO y FLOW AUC TIONE E R TO NON-PPS SUFFER HIGH ,

- + FLOW 4N T E RME DI ATE -- + LOW h E LECTRONN:S hHi F LOWME T E R -

, sH LOOP NO 1 TO NON PPS % SUFFER H sN TE RME DIA TE FLOW ~

AuC7IONE E R LOW BYP ASS CIRCUf f RV mem-

< PASS

( NtTRY Figure 7.2-8. Functional Illock Diagram of the Flux-Total Flow. Startup Nuclear, 5fodified Nuclear Rate, and Primary to Intermediate Flow Rate Protectise Subsystems, One Channel of Three is Shown

. 4558-1 1

a 7.2-36 Amend. 57 l Nov. 1980 .

1 1

• t

O TO NON PPS + BUFFER a

SL ,,, -

LEVEL DETECTOR ELECTRONICS LOOP NO.1 SL TO NON PPS + BUFFER

~

LEVEL DETECTOR ELECTRONICS LOOP NO. 2 SL TO NON-PPS + BUFFER ',

LEVEL DETECTOR ELECTRONICS LOOP NO. 3 UNDERVOLTAGE VARIABLE ~

RELAY TIME DELAY

+

I oC Figure 7.2-9. Functional Bloc': Diagrams of the Steam Drum level and HTS Pump Voltage Subsystems. One Channel of Three iS Shown.

7.2-37 Amend. 67 O

Mar. 1982

7.5.7 Containment Hydrogen Monitoring The objective of Containment Hydrogen Monitoring is to provide indication in the Control Room of the hydrogen concentration in the upper levels.of containment.

25 7.5.7.1 Design Description ,

The hydrogen instrumentation consists of two fully redundant and independent analyzer channels. The containment atmosphere is sampled through an entry filter located near the top of the RCB. The air samples are pumped to the analyzer which is located in the SGB and operates on the principle of thermal cmductivity. From there signals go to the Control Room where the hydragen concentration readout is provided. This instru-ment is also required to perform functions for events which lie beyond the design basis for the plant. This instrument is further discussed in 57 this capacity in Sections 2.1 and 3.3 of Reference 10b of PSAR Section 1.6.

7.5.8 Containment Vessel Temperature Monitoring The objective of Containment Vessel Temperature Monitoring is to provide indication in the Control Room of the containment vessel 25 temperature.

7.5.8.1 Design Description The temperature instrumentation consists of two fully redundant and independent channels. Each channel consists of eight themocouples mounted at various locations on the inside of the containment wall, with each thermocouple providing a signal to conditioning instrumentation in the SGB.

The instrumentation sends a signal to the Control Room where individual readout is provided. This instrument is also required to perform functions for events which lie beyond the design basis for the Plant. This instru-ment is further discussed in this capacity in Sections 2.1 and 2.2 of 57 Reference 10b of PSAR Section 1.6.

7.5.9 Containment Pressure Monitoring The objective of the Containment Pressure Monitoring System is to provide indication in the Control Room of the pressure inside the containment above the operating floor.

7.5.9.1 Design Description The pressure instrumentation consists of a pressure detector inside the containment vessel. Signals will be provided_to the display 25 and alarm panel in the Control Room so that continuous readout will be provided to the plant. operator. This instrument is also required to perform functions for events which lie beyond the design basis for th'e plant. This instrument is further discussed in this capacity in Sections 57 2.1 and 2.? of Reference 10b of PSAR Section 1.6.

O 7.5-33b Amend. 57 Nov. 1980

7.5.10 Containment Atmosphere Temoerature The objective of the Containment Atmosphere Temperature Monitoring System is to provide indication in the Control Room of the atmosphere temperature inside the containment building.

7.5.10.1 Design Descriotion The temprature instrumentation consists of two f ully redundant and independent channels. Each channel consists of two thermocouples mounted on the RG dome, wtih each thermocouple providing a signal to conditioning instrumentation in the SGB. The Instrumentation sends a signal to the Control Rocm where individual readout is provided. This instrument is also required to perform f unctions for events which lie beyond the design basis for the plant. This instrument is further discussed in this capacity in Section 2.1 and 2.2 of Ref erence 10b of PSAR Section 1.6.

7.5.11 Post Accident Monitoring A discussion of Post Accident Monitoring and the application of R.G. 1.97, Rev. 2 to CRBRP will be provided in a f uture amendment.

O O

7.5-33c Amend. 67 liar. 1982

i l@

.i i

i t

i 4

i i

3 l

l 4

I TABLE 7.5-4 HAS BEEN DELETED.

i t

1 i

i 1

t i

i I@

7.5-42

! Amend. 67 liar. 1902

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

f

[

REACTION PRODUCTS VENT FLOW DETECTION R E ACTION PRODUCTS VENT FLOW r DETECTION j REACTION 2/3 SUPERHEATER PRODUCTS VENT FLOW} r TRIP DETECTION j LOGIC r[( VENT FLOW PRODUCTS PRESENT jA REAGTION REACTION r

PRODUCTS VENT FLOW]

DETECTION REACTION r

PRODUCTS DETECTION VENT FLOW] ---.-

REACTION 2/3 EVAP"A" PRODUCTS VENT FLOW r TRIP r OR DETECTION j LOGIC r[k PRODUCTS PRESENTV _ .-

REACTION PRODUCTS VENT FLOW r DETECTION j REACTION PRODUCTS VENT FLOW r DETECTION j REACTION 2/3 EVAP "B" -

PRODUCTS VENT FLOW TRIP

[ VENT FLOW REACTION ]

DETECTION LOGIC k PRODUCTS PRESENT -

/' REACTION PRODUCTS VENT FLOW DETECTION A IHTS EXP TANK VENT

] &

FLOW DET. j IHTS EXP TANK VENT

}  ;

2/3 TRIP lHTS SODUIM EXP. TANK TO

]

FLOW DET. j LOGIC  :[( SODUIM DUMP TANK VE RE ACTION PRODUCTS PRESENTjg IHTS EXP TANK VENT -g FLOW DET.

MOM N.O.

MAN. "SWRPRS TRIP"\ n,gp . g HS 151

[8A WR RS RESET"  :

HS 151 /g g AND

_ BP Figure 7.5 6 SWRPRS TRIP AND SWRPRS CONTROLLED ISOL S

i.

M I

\.

SWRPRS ISOLATION TRIP BUS g IHTS PONY MOTOR TRIP

~

4h SII.2 fSOLAON VALVE A' AUXILIARY FEEDWATER 1 *@ Si!.2 ISOLATION VALVE "B" RECIRC. PUMP MOTOR CONTROL A wh Si!.7 SWRPRS N 2 PURGO VALVE A MAIN FEEDWATER

• @ SH.3 ISOLATION VALVE STEAM DRUM INLET

--* OS v S11.3 ISOLATION VALVE A

• h SH.3 STEAM DRUM ORAIN VALVES SUPERHEATER

• @ Sil.7 INLET ISOLATION VALVE A

--w h S!! . 5 OUTLET ISOLATION VALVE

--*@ Si! .6 N2PURGE ISOLATION VALVE A STEAM RELIEF VALVE "1" (SRV-Al A

• @ SII.6

• h SII.6 STEAM RELIEF VALVE "2" (SRV-B) A

-

• Sif . 6 STEAM RELIEF VALVE "3" (SRV4) A

--+ S!! . 5 BYPASS VALVE EVAPORATOR "A" (WEST)

• h S!!.4 INLET ISOLATION VALVE A

• Q S!!.4 N2PURGE ISOLATION VALVE A

• g}{ ,4 WATER DUMP VALVE (INNER) A

--> SI{ ,4 WATER DUMP VALVE (OUTER) A R

BP T o

=[s G nS h Sil.4 STEAM VENT RELIEF VALVE (SRV-F) A

-TRIP BUS

- --> --> gyg,4 STEAM VENT RELIEF VALVE (SRV G) A EVAPORATOR "B" (EAST) m SEE

• NOTE F

SII . 4 SEAL IN TRIP v

SIIEET 1 OF 6 ATION VALVES CONTROL LOGIC DIAGR AM 80-433-08 Amend. 67 j 7.5-48 Mar. 1982

.i

Y b

t f

SEL. SWITCH =O "CLOSE4 PEN"\ r HS 135A

[L EMERGENCY = 1 AND TRANSFER SWITCH

\

XS-135

[L SGAHRS _

INITI ATION + +

" RESET" ENERG12E HYDRAULIC OR AND  % OR STEAM + MOTOR VALVE CLOSES DRUM  : [ S.D. LEVEL 8" ABOVE NWL LEVEL k(LSH 135) j O R

[S.D. LEVEL 8" BELOW NWL  :

k(LSL 135) j RESET = 1 SEL. SWtTCH \ OPEN =0 "OPEN-RESET"[B HS 135B P.B SWITCH

, \ CLOSE = 1 (MOM)

HS.132 [g AFW FLOW FK138A AFW FLOW  : [ >150% t.o. -->

k (FSH 132A) 5SEC.

OR O -

STEAM -C '?

DRUM  : [S.D. 200PRESSURE PSIG +- R NOT PR ESSURE k<(PSL 132)] ' '

SH.1 SWRPRS ISO.

3 VALVE TRIP -

P.B. SWITCH

\ RESET = 1 (MOM)

HS. 3I /, D4 AUXILIARY FEEDWATER ISOLATION VALVE "A" Figure 7.5 6 SWRPRS TRIP AND SWRPRS COQ s

1 I

1 0

SE L. SWITCH -

OP O "CLOSE-OPEN"\  :

HS 139A [ EMERGENCY = 1_ ANO

~

TRANSFER SWITCH \

XS-139 [ NORMAL =1 SGAHRS RESET = 1 -

INITIATION

" RESET" ]j=0 TRIP + +

OR AND OR

+ MOTOR VALVE CLOSES S.0. LEVEL

-[12" ABOVE NWL #

STEAM (LSH 139) O DRUM LEVEL n

[S.D. LEVEL 8" BE LOW NWL (LSL 139)

SEL. SWITCH T P =

"OPEN-R ESET"[\

HS 1398 B

P.B. SWITCit "CLOSE" \ CLOSE = 1 (MOM) ;

HS.138

[B F K138A AFW F LOW

[ AFW 150%

FLOW T.D. *

(FSH 138Al 5 SEC.

STEAM

[S D. PRESSURE s ESSUR E ML 138) j ' '

' r SWnPRS ISO. D s 11 . 1 2 VALVE TRIP 4

g P.B. SWITCH k RESET = 1 IMOMI 3 "

',A'i3s / ND AUXILI ARY FEEDWATER ISOLATION VALVE "B" SIIEET 2 OF 6 80 433 07 TROLLED ISOLATION VALVES CONTROL LOGIC DIAGRAM Amend. 67 7.5-49 Mar. 1982

\

4

E MOM N.O.

d "FW. E N ABLE"\

HS-181C [ INPUTS SAME AS ONE SHOWN REFER %

TO TABLE 1 FOR , p n -

r CELL & INSTRU. .--o- OR N OT -*"

2/3 MENT NUMBERS. - -e-DETECTED u _1 AND r SC9 FLOODING r INPUTS SAME AS PROTECTION ONE SHOWN REFER - >

TO TABLE 1 FOR + FLOODING O SYSTEM 3 CELL & INSTRU. -- ,. OR DETECTED NOT+ 1 MENT NUMBERS. -->

A 2/3 MOM N.O.

" BYPASS ENABLE"

\ r,b

^ >500 PSIG = 0 OR  :

DRUM r PRESSURE _I PRESSURE <500 psto j

-G m

9 BELOW NORM = 0 DRUM LEV 2 LEVEL (ABOVE NORMAL] MOM N.C.

"CLOSE..

\ CLOSE=0 HS-152 [8RESET = 0 SWRPRS TRIP ,

SH.1 4

~

MOM N.O.

kOPEN =r1E "OPEN" Q HS 152 [

SEAL IN OPEN :

^ BELOW NORM = 0 '

DRUM  : LEVE 8.."

LEVEL ABOVE NOTIMAL]

(' \ CLOSE = 0 C OS HS 152A [8RESET = 0 SH.1 5 SWRPRS TRIP ]

MOM N.O.

"OPEN" \ " bq HS 152A [B SEAL IN OPEN j:

SGHRS 2 SH.S RESET RESET = 0 SH.1 6 m

SWRPRS TRIP ) _

i MOM N.O.

"OPEN" r O OPEN = 1 _

HS.153A L MOM N.C. f I "CLOSE" \ CLOSE = 0 m AND ' OR LOCAL = 1 i =g HS 153A [

TRANSFER \ LOCAL = 1  : 26 ALARM SWITCH XS.153 [L MOM N.C. 2 -

"CLOSE" \ CLOSE =0

/ gq q y,q, HS 1538 [8 OPEN = 1

( "O m " f  :,

\ HS-150's /8 l (SEAL IN OPENT j:

Figure 1 5 - 6 SWRPRS TRIP AND SWRPRS CONTROLLED 11 1

l l

4 h

TABLE 1 k LOOP NO. 1 2 3 CE LL NO. 241 221 224 207 242 222 225 208 243 223 226 209 f NSTRUMENT NO. 151 152 153 154 251 252 253 254 351 352 353 354 OR  ;

2 a

ENABLE BYPASS EN ABLE AND -

FEEDWATER A

6 OR STM DRUM PRES]S  : CONTROL g NOT <500PSIG SGB FLOODING DETECTION CIRCulT STEAM DRUM LEVE L NOT 8**

, ABOVE NORMAL j u.

ENERG12E R OPEN AND r ENERGlZE SOV MAIN FEEDWATER SOV

w. ALVE OPENS ISOLATION VALVE D E-EN E R- VENT CLOSE

> GlZE AIR __

~-

F.O.h VENT MAIN FEEDWATER ISOLATION VALVE /

1.A. (ACTIVE) 53SGV001A ENERGlZE MIT OPEN ENERGlZE SOV STEAM DRUM INLET SOV VALVE OPENS '

ISOLATION VALVE _

LE-E N E R- VENT G1ZE AIR CLOSE F l F D. VENT STEAM DRUM INLET E_~ _

I /

ISOLATION VALVE AND -

1.A . ,

53SGV004A ENERGlZE OPEN R

ENERGlZE SOV STEAM DRUM

\ SOV VALVE OPENS DRAIN VALVE DE ENER. VENT CLOSE Gl2E AIR p __

FD.f VENT -- --

l .A . - - - > \ /

AND ENERGlZE ADMIT PEN l .C .

AIR 53SGV014 53SGV015 l

- ENERGl2E SOV STEAM DRUM STEAM DRUM

+ g  : VALVE OPENS SOV -

DRAIN VALVES DRAIN VALVE (ACTIVE)

DE-E N E R- VENT CLOSE GlZE AIR F.O.f VENT SHEET 3 OF 6 3LATION VALVES CONTROL LOGIC DIAGRAM 80-433-06 Amend. 67 7.5-50 liar. 1982 j 1

b A

f e SH.1 27 RESET = 0 SWRPRS TRIP J

t EVAP "A" 2H 0] ,

DUMP VALVE ENERGlZf5 OPEN AND -

EVAP "A" H p O] E LECTRO-HYD r 'OR ACTUATOg DUMP VALVE 3 OPEN A E CLOG EVAP "A" SIM  :

VENT RELIEF] :a 1 VALVE OPEN AND -

EV AP "A" SIM VENT RELIEF -

VALVE OPEN j '5 SEL SW N.O. \ CLOSE = 1 SH.1 T "CLOSE/OPEN"[8 HS-160C RESET = 0 13 SWRPRS TRIP EVAP "A" 2H 0]  :

)

DUMP VALVE j OPEN / AND  : m Eh EVAP"A" 2H 0]  ; t O NOT ELECTRi DUMP VALVE p A OPE *!

VA.

EV AP "A" STM )

VENT RELIEF] ;I EVAP '*A" STM VALVE OPEN AND -

VENT RELIEF] -

VALVE OPEN j 'I l SEL SW N.O. \ CLOSE = 1 l SH.1 "CLOSE/OPEN"/0 HS-160A 14 SWRPRS TRIP RESET = 0_ a AND ->

[VAP "A" STEAh VAP "A" STE A f [ OUTLET PRESS r8 OR ENERGi2E SOV 07A R VALVE OPENS,

> 250 PSIG

- MOM N.O.

" RESET"

\ r SEL SW N.O. \

HS-1618 [8 OPEN = 1 "OPEN/CLOSE" \

SH.1 HS-U A 15 SWRPRS TRIP AND + 0 V AP "A" STEAM 078

- [ OUTLET PRESS][

>250 PSIG OR ENERGlZE SOV PSL VAP "A" STE AM VALVE OPENS

R -

07B OUTLET PRESS j 2W NG SEL SW N.O. \ OPEN = 1 EVAP.

OUTLET SH.1 "OPEN/CLOSE"[\

HS S2A B PRESSURE 16  ; [ SWRPRS TRIP) RESET = O_

i AND +

/ EV AP "A" STEAM 0

- OUTLET PRESS  :

107C _

> 250 PSIG ENERGtZE SOV PSL OR 07

[ OUTLET j PRESSEVAP "A" STE Ah R VALVE OPENS 200 N G SEL SW N.O \ OPEN = 1 "OPEN/CLOSE"\

Sg,y HS-1628 II /[SWRPRS TRIP R ESET =0 AND ->

VAP "A" STE AM 0 1 107

[ OUTLET PRESS

> 000 PSIG ENERGl2E SO PSL / EV AP " A" STE A OR

- OUTLET PRESS R VALVE OPENS 7

SEL SW N.O. \

<250 PSIG j s OPCN = 1 "OPENICLOSEa \

i HS-163A [8 4 Figure 7.5 - 6 SWRPRS TRIP AND SWRPRS CONTROL (

I &

I k N

h S

E ti EVAPOR ATOR "A"(WEST)

L RECIRC PUMP F.O. BYPASS VALVE 53SGV018A EVAPOR ATOR "A" (WEST) IN LET ISO. VALVE, N PURGE ISO. VALVE, H O DUMP VALVES INNER 2 2

& OUTER AND STEAM PWR RELIEF VALVES SRV-F &SRV.G TYPICAL.

FOR EVAPORATOR "B"(EAST) REFER TO TABLE FOR APPLICABLE DESIGNATIONS ETC.

ERG 11E b HYDR AULIC TUATOR J l VE OPENS / E '

O

1. EVAPORATOR "A" (WEST)

L INLET ISO VALVE 53SGV000A S ,

~~

EVAPOR ATOR "A" m n 2 N PURGE j 4h

->NOT* S 1.A.W ISOLATION VALVE u __

y

\/

F.O. VENT F.C. N l /.F.C G l.A. ,,

"g ENERGlZE AD '

R OPEN EVA.P "A" H2O ',

,, SOV "A" DUMP VALVE '

l .A. , ., ,,g ADliAIT DE ENER. N ENERGlZE M CLOSE AIR GlZE

\ _ l

SOV "B"

~

/ , .

DE-ENER. VENT CLOSE GlZE AIR ADMIT ENERGlZE OPEN - -

AIR

• g j

F.O. t VENT 1.A. ,7 , 'g

SOV "A.

DUMP VALVE EVAP "A" H2O _

F.C.

b F.C.'

^ D E-E N E R. VENT 53WDV004A S3WDV002A ENERG12E OPEN GlZE AIR CLOSE R (OUTER) (INNER)

% l EVAPORATOR "A"(WEST) WATER

SOV "B" ,' -

DUMP VALVES (INNER & OUTER)

/

DE R ENT g CLOSE EvAP VALVE DESCRIPTION vALvENO.

ZE F.O.T VENT '

-."htilsot__ATION VALVE - 160A 535GVmBA NgPURGE vALvf 107A 161A -

1.A. ,',' ,, c 'w_wy vAyys flNNER) 162A 53wCM2A 4 _"2 ,o>e west "2 uur vALvt touTEm i62s 53wovm4A ENERGlZE OPEN ._sTE3M VENT REUEF v4LvE SRv F 163A S35Gv101 A SIIEET 4 of G AIR STEAM vfMT REUEF valve $Rv4 163C SOV EVAP. "A" (WEST) STM ,, ,,

Rf cmc PUMP BYPASS valve - 160C 535G_v]O3A 535GV018A 82-135-01 -

VENT RELIEF VALVE I INLET [sOLAfioN vaLvs -

_ i6oe 53sovoorA, DE.E N E R. VENT --

PURGE VALVE toBA 1618 -

ClZE AIR CLOSE y"2 ogu ,v g g jin E m ,,, 162]C 53wpfw15 Amend. 67 EAST "20 0we v ALvf louTE m 1820 53wovoo3A F.O. i 53SGV101 A F.C. Mar. 1982 SRVf MW

'T'*"v'"*"'Ut'v"'v"v#

sTE Au vf MT REuf r vALvf SRvo-N -'*3' $3S vi A k

isso 53sGv102A  :

Recmc ruwe syrAss vAtvE 3 y -

isoo s3sovotiA 10 ISOLATION VALVES CONTROL LOGIC DIAGRAM 7.5-51

I SH.1 RESET = 0 18  : SWRPRS TRIP r A O

[V AP "A" STEA 107E ENERGlZE SOV VAP "A"STEA j[k <OUTLET r

250 PSIG PRESS ]  : R ^

SEL SW N.O. OPEN = 1 "OPEN/CLOSE"\

HS-163C [8 EVAP RELil 2 TRIP =0 SGAHRS RESET _

g J l NC

[SUPERHEATER OUTLET PRESS.

\ < H00 NG f l p MOM N.O.

l \ BYPASS = 1 l

" BYPASS" :lAND l -

I ' SEAL IN BYPASS l 9 -

SWRPRS TRIP l MOM N.O. X

\

J _

155 [L M O M N.C.

\ AND l TRANSFER

\

S SSA [L

• SWITCH MAIN CONTROL = 1 XS-155 [L MOM N.C.

l "CLOSE"

\ '

AND MOM N.O. HS-1558 !B "OPEN"

' OR

y l HS.155B B

[ SEALIN OPEN O

26 [SWRPRS TRIP \ RESETNOT =0 -

BP l f MAN CONTROL -

HY 159A _

OPEN = 1 l AND r i MAN CONTROL \

\ OPEN = 1 e

'I BP BP l HY-1598 l

l SH.1 l RESE1 = 07 a 10 SWRPRS TRIP )

f SUPERHEATER AND L SUPERHEATER H p  : OUTLET PRESS- y O OUTLET M D SUPERHEATER PRESSURE i_ PSH r O g

D k >UTLET 250 PSIG PRESS.

M OM N.O.

RESET \ 3 ENERGlZE l HS-1568 !B OR SELECT SW. #

OPEN = 1 l

"OPEN/CLOSE**)\

HS-156 g

s 1

Figure 7.5 - 6 SWRPRS TRIP AND SWRPRS CONTROLLED i

\

l.A. ,,

, q ENER?lZE AR OPEN

• EVAP "A** (WEST) STM ,, ,

SOV '

VENT RELIEF VALVE l._

DE-E N E R. VENT ~~

CLOSE GlZE AIR

~~/

F.O.kVENT j F.C.

)RATOR "A"(WEST) STEAM VENT 535GV103A SRV

) VALVES (SRV F & SRV(a) h 4 7

OSIS CKTS HMALo1 \ LOOP 3 i 7 _.au I ->

g TO LOOPS 2 & 3"AND" GATES l

,] LOOP 2 ,

J \ \

_ tB t_ . ., <=e1

~

LOOP 2 D

n--

INPUTS FROM LOOP -]

2 & 3 "lSOLATORS" LOOP 3 _  !

x =

f-WOT ' . .

ra --

ADMIT

ENERGlZE OPEN AIR

-- ENERGlZE SOV SUPERHEATER AND

~

SV OUTLET ISO VALVE VALVE OPENS ' ',_1_

_ AND  : 1  : DE-EN E R- VENT CLOSE

.s GlZE AIR --

_ /

F.O. f VENT 53SGV012A SUPERHEATER OUTLET ISO. VALVE (ACTIVE) m AND  : I

~

r? ,Y F.O.

u  ?

MANUAL CONTROL \ LY 159 r-ifp g STATION F .C .

53SGV01GA 1.A.) ,P SUPERHEATER BYPASS VALVE

S ' '

SUPERHEATER N2PURGE

(- \j g3 g 31 SOLATION VALVE

-*NOT+ S 1.A. -#-->

(SYS 82) SHEET 5 of 6

" N/

N /

F.O. VENT h- /N /N 82-135-02 l_l Amend. 67 Mar. 1982 ISOLATION VALVES CONTROL LOGIC DIAGRAM 7.5-52 l l

I

)

(

11  : SWRPRS TRIP "hI AND  : 0

[SUPERHEATER OUTLET PR ESS.

'O SUPERHEATER OR

[ OUTLET PRESS. J SUPERHEATER OUTLET l 1028 SELECT SW. < 200 PSIG j OPEN = 1 PRESSURE "OPEN/CLOSE.

l Sg, ,

HS 157A [B 12  : SWRPRS TRIP RESET = O_ a AND  : O j r- 1 [SUPERHEATER}

OUTLET PRESS.  : _

l > 250 PSIG SUPERHEATER OR OMET PRESS. -

10 SE LECT SW.

200 m G j '

ONN=1

! "OPEN/CLOSE"\

SH.1 HS 1578 [8 13  : SWRPRS TRIP

~9E AND  :

o

[SUPERHEATER

'- W - Ed

{  : (> OUTLET 250 PSIG PRESS.j[SUPERHEATER _ R OR y, SELECT SW. \ OUTLET

<200 PSIG PRESS].

"OPEN/CLOSE~ \ OPEN = 1 HS 157C [B SH.1 7 SWRPRS TRIP STEAM j DRUM PRESSURE

[ STEAM DRUM k > PRESSURE 500 PSIG_ j

+

/

CLOSE=1

( , SELECT HS.154 !g SW."OPEN/CLOSE"\

RESET = 0 3

SWRPRS TRIP > _i MOM N.O. \ RESET = 1 -

R OR

" RESET"

/

SELECT SW. k OPEN = 1

• • OPEN/CLOSE" Figure 7.5 6 SWRPRS TRIP AND SWP e

i

I i

I I.A. ,;' " h t

^

ENERGlZE OPEN AIR t SUPERHEATER STEAM RTIZE SOV SUPERHEATER STEAM OV ,;, g, 1- RELIEF VALVES 1.2 & 3 hVEOPENS RELIEF VALVE 1 -

,DE-E N E R. VENT --

I GlZE AIR CLOSE

__j t ' F.C.

• !} ,, g 53SGV106 J k M R4 ftGf 2E OPEN D 9 AIR DRGlZE SOY '1"SOV+ ,,

hVE OPENS DE ENER. VENT --

GlZE AIR CLOSE __f

,. 53SGV107j h Y ENERGlZE OPEN R

ERGlZE SOV SUPERHEATER STEAM L

' VE OPENS SOV "

RELIEF VALVE 3 DE-EN E R. VENT -~

CLOSE GlZE AIR -

f F.O.fVENT F.C.

53SGV108 k ATM h 9 f .A. s ', ' a

ENERGlZE R CLOSE ENERGlZE SOV SUPERHEATER IN LET ,f

• SOV

' (VALVE CLOSES ISOLATION VALVE 1 DE ENER. VENT ~

OPEN

-GlZE AIR --

p F.OfVENT 53SGV011 A

- (ENERGlZE SOV T T

(' 6 EVAPORATOR LINE TO TANK EAST

'(VALVE OPENS '__

d EVAPORATOR LINE TO TANK WEST

,. / ,,

, si S

1.A. ,, -

3, F.O. VENT b.,_ d SUPERHEATER TO TANK EAST SWRPRS NITROGEN PURGE VALVE SHEET 6 of 6 82-135-03 Amend. 67 March 1982 2RS CONTROLLED ISOLATION VALVES CONTROL LOGIC DI AGRAM 7.5-53 l

\.

p 7.8 PLANT DATA HANDLING AND DISPLAY SYSIEM O

7.8.1 Deslan Descriotion l The Plant Data Handling and Display System PDH&DS is a distributive computer system which supports plant operations and performance by monitoring, limit l checking, trending, and displaying plant Information. It supplements other '

monitoring and displaying systems including Plant Annunciation and Plant Control. Other plant systems provide suf ficient instrumentation and control to the operator such that the PDH&DS is not required f or startup, ope'r ation, or shutdown of the plant. However, additional requirements may be placed on the plant operations should the PDH&DS be Inoperative.

The PDH&DS uses redundant Central Processing Units (CPUs) to process plant information acquired f rom numerous Remote Data Acquisition Terminals (RDATs) located in various areas of the plant.- Inf ormation is provided to the operator via Cathode Ray Tube (CRT) and hard copy units located in the control room. A block diagram of the PDH&DS is shown on Figure 7.8-1; Figure 7.8-2 depicts the system arrangement.

Specific functions of the PDH&DS include:

o Monitoring of most plant variables and alerting the operator when any selected variables exceed predetermined limits.

o

[])

(

o Recording of the operating history of the plant.

Performance parameter calculations including:

overall plant heat balance l

l reactor and equipment calorimetries plant protection system channel output monitoring

' shutdown margins control rod worth core assembly exit temperatures reactiv!1y calculations sodium inventory calculations component offIclencies o Providing CRT display units in the control room to present pertinent plant data for survelllance of plant protection and control systems, heat transport systems, SGAHRS, auxiliary systems and balance of plant.

L) 7.8-1 Amend. 67 March 1982

o Displaying measuroments of variables not otherwise displayed in the control room to reduce the number of trips by plant personnel to read local Indicators. This function provides a simple, offIclent method f or the operator to respond to group alarms.

o Providing a mechanism to forewarn the operator of potential harmf ul conditions. Examples of those include high bearing temperaturo, detection of small sodlum leaks and radiation levels. if the condition deteriorates further, the operator will be warned by the annunciator system, o Providing pro and post trip Information for review.

o Providos for acquisition of data for design verification of plant components, o Providing the above information to the operator via an Integrated combination of state of the art human engineered color CRT displays (schematic diagrams, paremeter lists, trend plots and other graphic representations as appropriate) and printed output.

7.8.2 Qesign Analysis Tho PDH&DS is designed for high system availability by utilizing redundancy in processing and display.

The main processing part of this system is located adjacent to the control room in the computer room. Information generated by the DH&DS is presented in the control room. The data acquisition components of the system are located near sensor local panols. The data acquisition components multiplex sensor signals to reduce the number of con 1rrol room panels and associated cabling.

Although the PDH&DS is designed for an extremely high availability, operating proceduros of systems which normally use PDH&DS capabilities are written to allow operation of the plant with manual data recording and calculations should the PDH&DS not be ava!!able.

The oporator Interacts with the PDH&DS to obtain Information, display and calculations via multiple CRTs and hard copy devices. CRTs are positioned on the operatcr's desk and at eye level on the Main Control Panel for convenient uso by the operatcr. Human factors engineering is used in the displays (color, symbology, density, organization, format, etc.) and the displays are Integrated with the Main Control Panol end operating proceduros.

O 7.8-2 Amend. 67 March 1982

CONTRDL ROOM COMPUTER ROOM I BU LD G l I l 8 REMOTE DATA l CRT DISPLAYS a CENTRAL PROCESSING a ACQUISITION  !

l UNITS (2) l TERMINALS I I I I LINE PRINTER /

PLOTTE R l DISCS g l U l REMOTE DATA g _ ACQUISmON l l TERMINALS I I OPER ATOR CONSOLE -

l  ; AG ETIC l l g TAPES l g l REACTOR REFUELING-_-__g g l l COMMUNICATION O g

[ l 3 CENTER (NOT PART OF I g l I SYSTEM (911 l l CARD READER l ~~~~~~~ ~

g , p_ ____ ______ ___

ACQUISITION - 1 I STEAM GENERATOR BUILDING TERMINALS g g REMOTE DATA g -

LINE PRINTER g

- ACQUISITION g g TERMINALS I I I

l r- ----------------

l OlESEL GENERATOR BUILDING TYPEWRITERS -

= CRT DISPLAY g g REMOTE DATA g  ; - A0QUISmON g g TERMINALS I 1

___-___._______J L_ -___-__-_-__----

CONTROL BUILDING l g TUR8INE GENERATOR BU!LDING 1 1 REMOTE DATA l l REMOTE DATA ACQUISITION - 1 I - ACQUISITION TERMINALS l PROGRAMMER I

~ TERMINALS l CONSOLE I I

l

a. . ______________

I CIRCULATING WATER I I PUMP HOUSE I I l FLEXIDLE DISC 1 CRT DISPLAY l REMOTE DATA WITH KEYBOARD

_ l - ACQUISITION l l I TERMINALS I

- ,_ __ _ _ _Ip _ _ _ _ _ _ _ _ _ _ _ _ .I. , <

SWITCH YARD l RIVER PUMP HOUSE l RADIOACTIVE WASTE BUILDING l I REMOTE DATA g REMOTE DATA ACQUISITION - REMOTE DATA l ACQUISITION ACQUISITION TERMINALS TERMINALS l g TERMINALS I I -

O V Figure 7.8-1 Plant Data Handling and Display System Schematic 7.8-3 Amend. 67 Ma r. 1982

illiii- ! !!\!! MAIN j !E Z CONTROL ~ - - -

s J h ., PANEL y' REACTOR -

EU 88 CONTAINMENT 0U' RDAT3 BUILDING LOCAL PANEL -

MPUTER ,- ca AREA SERVICE s M N y CO NTRO L,,, '

YPICAL

- ~

?

=

ROOM d '

CABLE LOCAL ROUTES -

/

's CABLE ,/ PANEL SPREADING 's j AREA

,, ROOM s_ '

REMOTE DATA ACQUISITION TERMINAL (RDAT)

EF

.'=

F f00 Figure 7.8 2 Plant Data Handling and Display System Arrangement.

t 3 42 44910 h0V4 32 7790 h l O O O

.=_ _ _ __ _ _ ._ _. _

l l

o Physical, color, and geometric dif ferentiation of displays and

(~'}

v controls mounted on the board is provided to assure ease of recognition by operating personnel and minimize the chances for Inappropriate actions.

o Arrangement and design of displays and controls is specified to provide arrays which permit determination of proper instrument comparison at a glance, where practical.

o Modular design of switches, controls, and indicators is used to permit

ease of maintenance and minimum Interference with operation. The equipment included on the main control board is summarized below  !

(refer to Figure 7.9-1 and Table 7.9-1).

The arrangement of the instrumentation and control devices on the main control panel is as f ollows:

Section A - Fmargenev. Plant Protection System. Engineered Safety Features and Ptant Control Svstamc:

o Emergency Chilled Water o Emergency Plant Service Water o Reactor Shutdown O o Containment isolation O

o Sodium Water Reaction Pressure Rollef Sub-System Status o Sodium Dump o Control Room Heating, Ventilating, and Air Conditioning o Containment instrumentation o Flux Monitoring '

o Primary and Secondary Manual Scram Switches o Supervisory and Reactor Control o Reactor instrumentation o Rod Control and Rod Position Indication

('a) .

7.9-3 Amend. 67 March 1982

Section B - Primary and Intermediate and Steam Generator System Heat Transoort Systems o Primary Hoot Transport o Intermodlate Hoat Transport o Stoam Generator o Foodwater and Auxillary Flow Section C - Feedwater. Conensate. Auxillarv Steam Generator. Turbine.

Generator arid Switchyard Systems o Condensato l o Auxillary Foodwater Pumps o Protected Air Cooled Condonsor o Steam Gonoratcr Auxillary Heat Removal Vent Controllers o Turbino Control Panois o Turbino Instrumentation o Turbino Steam Bypass o Circulating Water o AC Dus Circuit Breaker Control o Generator Syncroscopo The Following Instramontation and control panels, while not a part of the Main Control Panol, domand rapid operator responso and have been arranged to permit operator scanning f rom the Main Control Panoi:

o Failed Fuol Monitoring o Sodium Look Detection o Sodium Fire Detection o Non-Sodium Firo Detection O

7.9-4 Amend. 67 March 1982

o Control Building Fire Detection v

o Emergency Diesel Generators o Switchyard and Station Electrical Distribution o Direct Heat Removal Servico The layout of Section A of the main control panel is designed to minimize the time required for the operator to evaluate the system performance under accident conditions. Deviations from predetermined conditions are alarmed and/or Indicated so that corrective action may be taken by the operator.

The control room also includes control and Instrumentation equipment that is used Inf requently or for which controlled access is desirable, included in >

this control room back panel area are powur distribution, chilled water, containment Instrumentation, recirculating gas, heat transport, steam generator, heating ventilation and air conditioning, annunciator electronics, turbino, balance of plant, plant control, plant data handling and display system multiplexers, flux monitoring, radiation monitoring, reactor shutdown and containment isolation panels.

7.9.2.4 Main Control Board Design The Main Control Panel is an open U-shaped, stand up vertical panel as shown in Figures 7.9-1 (plen view) and 7.9-2 (side view). There are 3 significcnt f eatures of the control board mechanical design: seismic capability; separation of redundant saf ety related equipment and wiring; and modular construction of switch, Indicator and control equipment.

Q 7.9-5 Amend. 67 March 1982 L

Since the Main Control Panel includes safety related equipment, the sections including this equipment are designed to Selsmic Category I. Structures, wiring, wireways, and connectors are designed and installed to ensure that saf ety related equipmont on the control panel romains operational during and after the SSE. The Main Control Panel is constructed of heavy gauge steel within appropriate supports to provide the requisite stif fness.

Within the bounderlos of the Main Control Panel Sections, modules are arranged according to control functions. Fire retardant wiro is used. Modular train wiring is f ormed into wire bundles and carried to metal wiro ways (gutters).

Gutters are run into metal vertical wireways (risors). The risors are the Interf ace between external wiro trays feeding the panol and Main Control Panel wiring. RIsors are arranged to maintain the separated routing of the exterr.al wiro trays. (Seo Figures 7.9-3 and 7.9-4).

Mutually redundant sa'oty train wiring is routed so as to maintain a minimum of six Inches air separation between wires associated with dif ferent trains.

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

7.9.3 Local Control Stations Local control panels are provided f or systems and components which do not roquire f ull timo operator attendance and are not used on a continuous basis.

In those cases, however, appropriato alarms are activated in the Control Room to alert the operator of e9 equipment mal f unction or approach to an of f-normal condition.

7.9.4 Communications Communications are provided between the Control Room and all operating or manned areas of the plant. In addition to public address and interplant i

communicaons and the privato automatic exchange (used f or ir.-plant and external communications) a sound powered maintenanco communication Jacking system is provided. Rodundant and separate methods of communication between the control room and other TVA generating plants is also provided.

7.9.5 Design Evaluation Following the Three Milo Island accident, a largo task force was formed for the purpose of perf orming a thorough review of the CRBRP Control Room design.

This overall review was divided into throo parts; a planning phase, a review phase, and assessment end implomontation phase. Fol lowing the task f orce ef f ort, NUREG-0700 was issued. NUREG-0700 is similar in intent to the CRBRP Control Room design ovaluation.

7.9.5.1 Planning Phase in tho planning phase the objectivos and scope of the task force were Idonti t led, and critoria were established f or personnel selection. A charter was developed which contained the scopo and objectives, and personnel selection was accomplished.

7.9-6 Amend. 67 March 1982 L

sj The task f orce charter required a review of the Control Room design and the operating procedure outiines to ensure that the systems designs, the integration of the systems, and the man-machine Interf aces properly supported safe operations of the plant during both normal and abnormal conditions. A task analysis was established for observing the operator conducting various duties. Specific 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 for plant operations.

4 Recommendations f rom other Key System Review Task Forces.*

5. Recommendations made by NRC and other parties as a result of the Three MI1e Isiand occurrence.

6. Computer utilization by the operators.

7. Operator training requirements.

8. Remote shutdown capabilities and safety systs status indication in the Control Room.

Cr!1eria were established for personnel selection of those to participate on s The task force. 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 quallfled operators were considered mandatory. Personnel with human factors education and expe.'lence both inside and outside the nuclear Industry were included.

Human f actors considerations were crnphasized in the planning phase. Previous Control Room design of forts had attempted to optimize the man-machine Interface. However, a major objective of the Control Rocrn Task Force was to re-evaluate this interface. Prior to the evaluation of fort a seminar was hold, under the direction of three leading human f actors personnel, to teach the Task Force disciplined methods for considering human f actors. Based on this training and further assistance from human engineers, check lists were prepared to evaluate the man-machine interf ace.

7.9.5.2 Review Phase In the review phase 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 ihan 200 walk-throughs of plant events were conducted.

gs *See Reference 7.9-1 O

7.9-6a Amend. 67 March 1982

lho Control Room design and operating instructions wero thoroughly reviewed in four areas:

1. Proper identification of systems to be operated from the Main Control Room.

2. Proper staf f ing of the Control Room.

3. Proper overall layout of the Control Room to enhance the man-machine interfaces and support the Integrated operation of plant's systems.

4. Proper layout and design of Individual Control Room panels, Instruments, Indicators, and controls to enhance the man-machine Interf ace and support the Integrated operations of the plant's systems.

A full scale mockup of the Control Room was used. The events chosen to be evaluated wero caref ully selected so they would umbrella all of the operations i

that are olther expected to occur or might be postulated to occur over the l life of CRBRP. The of f-normal events include plant responses to single and multiple failures.

The niothodology of perf orming this review consisted of using three groups of peoplo; simulators, operators, and evaluators.

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 I nd i cator 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 f low into the Control Room during the event. This made the walk-through as realistic as possible.

Tho Operators played the part of the Control Room operators and carried out the stops of the proceduro being evaluated. They touched each switch they woro required to operate, and observed each Indicator which was part of the particular event.

Tho Evaluators included a Human Factors Engineer and a Systems Engineer.

Their f unction was to fill out the Operating Sequence Diagram and the evaluntion shcots f or each procedure and event reviewed.

As probicas or concerns were encountered, recommendations were made. These were, in some cases, of a broad nature and reflected the need for reconsideration of decisions mado in the four most important evaluation areas described above. Other problems and concerns related to specific details of the Control Room design or the procedure outlines.

7.9.5.3 Assessment and lmolementatlon Phase Tho 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 overall Control Room and Main Control Panel were made assuming 7.9-6b Amend. 67 March 1982

all recommendations were incorporated into the design. The recommendations

()

s were modified based on the small model to provide a coordinated and consistent set of final recommendations. Senior Project Management reviewed the final set of recommendations and issued them to the Project line 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 procedures, or reject the recommendation nnd provide adequate justification if the recommendation is invalid. Each assessment is reviewed and approved by senior project management.

7.9.5.4 Conclusions The Control Room Task Force Design Review is documented in further detail in Reference 7.9-1. In September 1981, NUREG-0700 entitled " Guidelines for 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 believed met by CRBRP.

O O

J.

s J.

7 . 9-6c Amend. 67 March 1982

Reference:

1. Suminary Report on the Conduct of the Clinch River Breeder Reactor Plant (CRBRP) Key System Reviews, February 1982.

O i

I O

7.9-6d Arnend. 67 March 1982

t (O

\ 59l The use rate of RSB/RG argon by these services is variable and is dependent on operator options. Under start-up conditions, the flow will be maximum, and 59l a minimum automatic supply capability of 94,000 scfd of argon is to be provided.

Argon is to be used for all services involving sodium-wetted components, such as f uel handling, sampling, and maintenance services. This gas also is ultimately exhausted through CAPS to the atmosphere.

Argon is also to be suppIIed for purging and inerting lHTS components for 59 Ii transfer operations and for loop pressure control during normal operations and during all the postulated lHTS design events.

9.5.1.2 Design Descriotion The argon distribution subsystem is composed of liquid argon Dewars with vaporizers, gaseous argon bottles, piping, valves, vapor traps, filters, vessels, rollef systems, freeze vents, and oil traps as necessary to 59l distribute and vent the argon to meet the rec,uirements described in Section 9.5.1.1.

9.5.1.2.1 Reevele Argon Distribution Argon from the primary recycle cover gas storage vessels in the RCB is reduced in pressure to supply cover gas to the reactor vessel, primary sodium overflow vessel, and primary pumps cover gas spaces, which are all Interconnected by a s pressure equalization line. This cover gas system is maintained at a pressure

-59 1 f 6 In. w.g. by a constant purge plus a feed and bleed control system.

There is a continuous transfer of argon cover gas from the reactor and the primary pumps via the equalization line io the primary sodium overflow vessel and then through a 5-scfm vapor trap that removes sodium vapor. This vapor trap consists of a vapor condenser and two parallel aerosol filters (one 59 1 redundant) . - A 1-scfm sample of cover gas is taken from the equalization line and is passed through a 1-scfm sodium vapcr trap to the Failed Fuel Monitoring 59 I S0 1 System. This gas and the reactor cover-gas purge gas and the primary pumps 591 purge gas flow through RAPS for processing.

9.5.1.2.2 Fresh Argon Sucolv at RSB Argon for services in the Reactor Service Building (RSB), the Reactor Containment Building (RCB), and the intermediate Bay (IB) is stored as liquid in two Dewers, located on the RSB pad. These Dewars have a capacity of 1500 gal each and are equipped with fill and vent lines. Normally, only one of the Dewars is in operation. When it is nearly empty, a low-liquid-level instrumentation signal operates automatic controls that shutoff that Dewar and open the other Dewar to the supply header. A hand switch control override allows drawing on both Dewers simultaneously. When the switchover takes place, an alarm signals the operator and he is then required to initiate 4g action to fili the nearly empty dewar.

t O\ t

\ ./

9.5-3 Amend. 59 ,

Dec. 1980 '

+ry

Two ambiont-air vaporizers on each Dewar can evaporate iho liquid argon at a notinal maximum gas flow rate of 2000 scfh each, at 200 to 235 psig. With both Dewers on-line, therefore, approximately 8000 scfh of argon gas can be delivered.

Tho argon from the Dowars passes through a filter and is reduced in pressure to 175 psig. It is then piped to various points in the RSB and the Intermediate boy bef oro it is reduced to the required pressure upstream of each Interface. Bef oro entering the containment, the fresn argon header is reduced in pressure to 50 psig.

9.5.1.2.3 Fresh Argon: RCB Distribution The RC8 fresh argon header enters the building with isolatf or, valves on each sido of the penetration. Valvo status is shown by indicating lights on the Containment isolation System (Cis) panel and the inert Gas Recolving and Processing System local panol.

Upon entering the containment, this argon header manifolds into a number of branches supplying fresh argon through Individual feed and bleed control arrangomonts to the primary sodium storage vessel, head access penetration G-6, IVTM storage pit, makeup pump drain vessel, primary cold trap NaK :orage vessel, primary pump oil supply tanks, to various freeze vents, to the primary sodium plugging temperaturo Indicator (NPTI) and sodium sampiing package, the RAPS cold box purgo lino, and to the RW floor / wall service stations.

9.5.1.2.4 Fresh Argon: IB Distribution A 175 psig pressuro argon supply lino is routed toward the ex-containment primary sodium storago vessels in the intermodlate bay. The argon is reduced in pressure beforo entering the storago vessels, normally to 2 psig.

Alternately, the pressure can be reduced to 40 psig for pressurized sodium iransfor. These vessels are vented through a vapor trap and a pressure control valvo to the Cell Atmosphoro Processing System (CAPS). Additionally, they can be evacuated vla a vacuum station to CAPS.

9.5.1.2.5 Fresh Argen: RSB Distribution The RSB header supplles argon at the required pressures to the fission gas monitor module and the gas sampling trap. A branch line allows argon purge of the RAPS process train in the cold box.

Another 175 psig lino supplies argon through regulators to the Auxiliary Liquid Motal System EVS sodium and NaK components and to the EVS sodium PTl and EVS sodium sampiIng package. The sodium Iines have f reeze vents that are f urnished with argon during startup, maintenance, and sodium drain and fill operations at a nceninal pressure of 2 to 5 psig. An additional gas valve and metal sool flange are provided in the piping associated with the eleva.'ed EVS sodlum loop #3. After fill of the EVS sodium loop #3, the sodlum is frozen in tno frooze vent, orgon gas isolation valves are closed, an argon gas line spool ploco is removed and the argon pipo is sealed with metal seal flange assombilos.

9.5-4 Amend. 67 March 1982

i

! TABLE 9.7-2 NORMAL CHILLED WATER SYSTEM ftAJOR COMPONENTS t

4 APPR0X. CHILLED WATER l

DESCRIPTION QUANTITY FLOW FOR EACH COMPONENT

! Water Chillers 6 3300 GPit

47l Water Pumps 6 3300 GPM 44 i

Expansion Tank 1 N.A.

{

I

. 15 l

!9 1

l l

t

}

i.

Amend. 47 Nov.1978 9.7-19

TABLE 9.7-3 C0tfVNENTS SERVED BY THE EEICENCY DilLLED W ATER SYSTEM HEAT LO/O LOCATION EQUIPENT TITLE B1U/m X 103 BLDG. GLL ELEV ATION Control Room A/C Unit 1375 G 410A 863'-0" Control Roca A/C Unit 1375 G 411A 847'-3" SWGR A/C Unit 852 G 413 86 3'-3" SWGR A/C Unit 828 G 412 847'-3" Aux. Feed Pump Unit Cooler 264 SGB 204A 733'-0" Aux. Feed Pump Unit Cooler 264 SGB 2048 733'-0" Aux. Feed Pump Unit Cooler 311 SGB 202 746'-0" Aux. Feed Pump Unit Cooler 311 SGB 2028 746'-0" Emergency Chillers Unit Cooler 42 SGB 216 733'-0" i

Emergency Chillers Unit Cooler 83 SGB 217 733'-0" l Reactw Heat Transport 1&C Room Unit Coolers 71 SGB 272A 836'-0*

Reactw Heat Transport l&C Room Unit Coolers 82 SGB 2728 836'-0" Rosctr Heat Transport l&C Room Unit Coolers 78 SGB 272C 846'-0" EVS Pump and Cold Trap Cooler 528 RSB 306A 755'-0" e EVS Pump and Pipeways Cooler 728 RSB 325 816'-0" Y FHC 456 RSB 343 779'-0"

$ FHC ABHX Cell Unit Cooler 456 279 RSB RSB 342 327 779'-0" 81 6'-0 "

ABMX Cell Unit Cooler 279 RSB 326 816'-0" Containment Clean-up Filter Cr'l Unit Cooler 283 RSB 359 755'-0" Containment Clean-up Filter C.Il Unit Cooler 283 RSB 3 91 816'-0" Annulus Filter Cell Unit Cooler 75 RSB 398 846 '-0" Annulus Filter Cell Unit Cooler 75 RSB 395 864'-6" Sodlum Make-up Pump and Pipeways 840 RG 105L 752'-8" Sodita Make-up Pump and Pipeways 520 RG 105L 752'-8" El&C Cubicle Unit Cooler 52 RG 165 842'-0" El&C Cubicle Unit Cooler 51 RG 163 824'-3" El&C Cubicle Unit Cooler 47 RG 167 842'-0" c

.' g k

.a CD Ch N N O O O

- .. a --

. -e,s 7MGJ T ann **

llM12r pasm .cw llM fU -

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= , -

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~; 3)};<,

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zi

@e g m@ @e e@ @e m@

w q w e 4 p = e73 x 4

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@m h; != @cla;!a@ @e #::im@

-os  %@ n p W_+pa .Y

. p p k_' P 1 .. -

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llN-561-)

3 i

GENERAL NOTES i

e. se m.s.= orts ur one ve

3. Estas ,C.,s.o r 5 a.aosales. 20 bas 4 OftYtu CLgast leESS CLatssFICaf f 9e CLast C AS a,e S3

• a:'t;",f.M',o"!,';"r'4" a,'""*

."l3:!!.'?,l'i'4.E smar ot eED O Det a. FELO.9s gg ne .F . . . .

• !"!!,'l".o?.?.1,'3"M"A ,

,m, J

'!!! #' ':,;*:':2 M-

,, ' .2.a,?.*,:s*"b.l:.,.'\?J','

re o-ggy,L

_ > ~ . * :P. l:'."".?J *';"oP,'J".l'.

rif.a,

.~ r.'!.;w a n ar.*?,A ma an,as s .sEF i ED

.,as.n.

,a ,

N,. ..;

.I ,y,,

41 , ,

L t 1.. F ipa l_.J 4 __. > m i. -

,e.._..

c c. a sc.. _

= sema.

.. .5 -

i ji

';'.'NI1n.r wuw - MA"O 'n'i.i,:n,c?

u

-Rd El REFEAENQL QAAf INGS

,. .om.a .. .. -

[?{as.q'ig M,W W!(QIesp n .g

- 'p: Figure 9.7-11 EMERGENCY CHILLED WATER

[ ____ SYSTEM SGB, RCB S-?Di

% ,r

-WD i I

-- I 9.7-34 Amend. 67 flar. 1982 1  !

i

t Analysis of core. structures activation and corrosion release is derived from a model containing 153 activation zones to approximate temperature and activation rate variation. The fuel assemblies were divided into seven radial zones, each containing 14 axial regions. The blanket assemblies were described in three (3) radial zones and 12 axial regions. Additional zones were used to describe the control assemblies, removable and fixed radial shield, core barrel and reactor vessel. The axial regions extended from a lower elevation of the inlet modules to the outlet nozzle elevations.

The saturation value of the radioactivity for each of the twelve reactions shown on Table 11.1-2 was calculated using a TRIED 42 group R-Z edit of D0T III W two dimensional neutron flux distributions.

These neutron flux distributions and activation rates were developed from the use of evaluated nuclear data files (ENDF/8-III).

The nominal corrosion rates for each of the zones were calcu-lated using corrosion rate expressions recommended for use in Reference 4.

Two corrosion rate expressions are developed in Reference 4, one for flow rates above 10 feet /second and a second for flow rates below 10 feet /

second. These expressions are:

I 7

R = (6.685 x 10 ) exp

-1f20-k

() i for flow rates >10 feat /second, and

~

6 R = (4) (1.1704 x 107+3.496x10y)exp-f8120"

_ k .

for flow rates 10 feet /second. The quantities in the above equations are defined as follows:

R = mils / year 4 = oxygen content of Na (ppm)

Tk = temperature (OK )

V = velocity (feet /second)

The release of radioactive isotopes to the primary sodium from isothennal regions of the fuel assemblies, blanket assemblies, control assemblies, fixed and removable shield assemblies, core barrel, and reactor 49 vessel is as follows:

Amend. 49 O 11.1-3

A =

A (Pi) (R) (1- ) - (M) (R) (t) (e )

I 3.7 x 10 where: A; = the activity of isotopo i released to the primary sodlum due to corrosion f rom a given component with a usef ul reactor lif e of T, (curies)

S A= surf ace area of zono (cm2 )

P,= equillhritm rate of production and decay of isotopo I (dis /

sec-cm>)

R= corrosion rato (cm/sec)

Ai= decay constant (soc-I) t= timo (soc)

The temperaturo profiles used to evaluate the corrosion release are based on

" plant expected" operating conditions. Parameters associated with this ovaluation are shown on Table 11.1-3. The corrosion release models discussed above were modiflod to account for the following factors:

1. The temporature verlation of the fuel and blanket assemblies as a f unction of operating cycle were included in the model. Fuel assembly outlet temperatures decrease during their residence in the core while blanket assembly temperatures increase during their residcnce in the Coro.

2. The temperature variation of iho fuel and blanket assemblies from inlet to outlot woro considered as described in Table 11.1-3.

3. The radioactivo Isotopo rolcaso f rom high heat flux regions (f uel regions) woro increased by a f actor of two. This factor accounts for increased corrosion rates experimentally observed in high heat flux creas (Referenco 5).

4. The release of Mn 54 and Cr51 are multigled by g additional f actor of two in all regions of the rcector. Mn and Cr are proforentially roloased from steel surfaces.

Tho oxygon concentration 11 sodium was set at two ppm. This oxygon level is considorod to be a f actor of throo higher than the nominal oxygen operating Iovels.

O 11.1-4 Amend. 67 March 1982 a

~

i

[ a

- TABLE 12.2-4 EXPECTED

• ANNUAL EXPOSURE IN NORMALLY ACCESSIBLE CELLS

-Head Access Area Expected Concentration MPC+ Expected Isotope (uCi/ml) (u ci/r,1) Concentration t MPC Xe --131m 9.7 E-13 2.0 E-05 4.8 E-8 133m 3.0 E-ll 1.0 E-05 3.0 E-6 133 5.5 E-10 1.0 E-05 5.5 E-5 135m . 9.4 E-12 1.0 E-06 9.4 E-6 135 1.3 E-9 4.0 E-06 4.5 E-4

~

1 38 1.6 E-ll 1.0 E-06 1.6 E-5 Kr 83m 2.9 E-ll 1.0 E-06 2.9 E-5 85m 1.1 E-14 ~6.0 E-06 1.9 E-5 85 1.7 E-14 1.0 E-05 1.7 E-9 87 6.0 E-11 1.0 E-06 6.0 E-5 88 1.7 E-10 1.0 E-06 1.7 E-4 Ar 39 3.5 E 5.0 E-06 7.0 E-4 49 41 6.4 E-ll 2.0 E-06 3.2 E-5 H3 9.2 E-15 5.0 E-06. 1.8 E-9 l 491 TOTAL 5.84 E-9 0.0015 l 1

Intermediate Sodium Piping Cells 49l H3 4.0 E-09 5.0 E-06 7.9 E-04

+MPC .= MPC for Restricted Areas.

9

'

• Failed Fuel Fraction = 0.1- percent at 1 year operation P

)

!~m~E Amend. 49 idA/ -April 1979.

c

'1 2. 2 -11 t

DETECTOR ASSEMBLY DSPLAY AMO CONTROL UNIT IN CMTRDL ROOM f(DCU)

W DETECTOR AND MNNN SAMPLE CHAMBER AND - TO PL ANT ACCESSORIES FRPTECTIONSWTEM OECK WRCE 4- g (cogTAW6T ISOLATOM i I I I CCW%RATORS, LOGtC AWD SAFETY DISPLAY VISUAL AUDIBLE ORCUITS)

METER ALARMS ALARMS M ER y H1CR0 PROCESSOR 2 SUPPLIES Agp N

= ACCESSORIES COW 1ftlL M - _ _ _ _ _ _ _ _ _ _ _ _ _ . .

CLASS IE y CLA% lE NOPMLASSIE TO PLANT DATA .m LTD SYSTEM CNTROLLER LIN MtALTM PHYSICS AREA 7

HANDLIKa AND onVL AY SWEM qf N SE DesPLKf VISUAL AUDtSLE ETER ALARMS ALARMS CEM1RAL PROCES5a6 Ulai AND MICRO- WTO MAlw COMPUTER CONTROL PMEL ANNUCATORS FLOW m FLOW ELEMENT METER k k l

4 UNE U CRT SOARD PRDITER FLOW ALARV SWITCH PLMP 4 REMOTE PROLE 55 Simow(RP5) SYSTEM (ONTROLLER IN 0041Rik ROOM FGURE 12 2-1 PPS CONTAINMENT ESAUST RAD lAll0N MONITORi% CRANNEl.(CLASS IE) 2d

• 3 W

e O O

) (mv)

Di.1ECTOR As5EMBLY DETECTOR AnD 5 AMPLE 4

• CHAMBER CHELK SOURCE 4-

> TO SYSTEM CONTROLLER IN HEALTH PHYSICS AREA m ~+ MCROPROCE550R SUPPLIES AND ' CEM1RAL PROCESSLE UNIT m y MIE550 RIES AND MICRO-COMPLITER 107Aiw y To PtANT oAtA HANDLlHG AND 0)NTROLFAMEL ANNUNCl418R$

G -

DISPLAY SYSTEM h

C ptsPLAY VISUAL ALOBLE METEg ALARMS ALARMS g g CRT KEYBOWIC tlWE.

FLON .. - FLOW PmWIER ELEMENT METER U

FLOW ALARl4 m PUMP SWITCH CONT E REM 01E PROCESS STATION 951EM CMME lH WM RM yy FIGURE 122-? NON PPS AIR RADIAll0N MON 110 RING CHWEL

?8 P

NO

TABLE 15.1.3-1 PPS SUBSYSTEM TRIP LEVELS OR TRIP EQUATIONS Primary Shutdown System High Flux Positive: g -l 1.01 (irlp at 115% power) )* 1+28S, -0.994(t)+0.1706Np+0.03643 p-50 Fl ux-Delayed Flux Negative: 1.014(t)-[." $(s)I 0.1969Npl

~

~

+0.041610 Fl ux- Pressure 1.3186 - Jr + 0.0425 <0 ,

Primary to Intermediate _ _

Speed Ratio Np(0.147 + 0.0022) + 0.0595 + 0.0007 - AbsVal,

[Np (1+ 0.015) - Ng (1 + 0.015) +0.0075 + 0.0]<0 _

HTS Pump Frequency Trip at 57 Hertz Reactor Vessel Level Trip when level reaches 8.1" above supressor plate Steam-Feedwater Trip at 30% mismatch Flow Ratio lHX Primary Outlet Trip at 8300F Tanperaturo

' Secondary Shutdown Svstem Flux-Total FIow 1.2 Fp - 0.99 / + 0.087 10 Startup Nuclear Trip bef ore 10% powe.-

Primary to Intermediate _ _

Flow Ratio Fp (0.147,+ 0.0022) + 0.050 + 0.0007 - Abs Val ,

[Ep (1 + 0.015) - F7 (1 + 0.015) + 0.0075 + 0.01. <0 Steam Drum Level Trip at i 8" from f ull power steady state level O)

V '

15.1-94 Amend. 67 March 1982

TA81.E 15.1.3-1 (Continued)

High Evaporater Outlet Temperature Trip at 750 0F Sodlum Water Reaction Trip initiated within 3.0 seconds HTS Pump Voltage Trip at 70% of rated voltage DefInttion of Variables

/1 = Reactor Flux P = Reactor inlet Plenum Pressure Np = Average Primary Pump Speed 2F = Total Primary Pump Flow Np = Primary Pump Speed F = Primary Pump Flow N g = Intermediate Pump Speed Fg = Intermediate Pump Flow i

O l

15.1-95 Amend 67 I March 1982

( )

~ _

TABLE 15.1.5-3 SYSTEMS ASSIDED CPERAELE TO MITIGATE TPt CCESEQUEPCES FOLLOwlNG THE OCQFREfCE OF EACH ACCICENT EVENT Required Operable Extnh S sten Pr!=ary Secont rw 15.2.1 Anticipated Events 15.2.1.1 Control Assembly Withdrawal at PPS fo; domed in long Flux-Pressure Flux-Tctal Flow Startup term by decay hea+ Flux-Delayed Flux removal (1) 15.2.1.2 Control Assembly Withdrawal at PPS followed in long High Flux Flux-Total Flow Power term by decay heat Flux-Pressure removal 15.2.I.3 Selsalc React;vity insertions-0BE PPS followed in icng High Flux Flux-Tctal Flow term by decay heat Flux-Pressure removal

- 15.2.1.4 Small Reactivity insertions PPS followed in long High Flux Flux-Total Flow

,# term by decay heat W removal e

m 15.2.1.5 Inadvertent Drop of a Single Control PPS followed in long Flux-Oelayed Flux Modified Nuclear Rod at Full Power term by decay heat Rate removal 15.2.2 Unlikely Events 15.2.2.1 Loss of Hydraulle Holddown PPS followed in long High Flux Flux-Total Flow term by decay beat Flux-Pressure removal e 15.2.2.2 Sudden Core Radial Movement PPS followed in long High Flux Flux-Total Flow term by decay heat Flux-Pressure removal 15.2.2.3 Maloperation of Reactor Plant PPS followed in long High Flux Flux-Total Flow Controllers term by decay heat Flux-Pressure removal kk n

• 3 C.L we W

CO Ch N taJ L

TABLE 15.3-3 (Continued)

Required Operable Events system Pr f r.ary seconefsry 15.2.3 Extreely Unlikely Events 15.2.3.1 Cold Sodlum insertion PPS folic =ed In long Speed Ratlo Flow Patto term by decay heat r eovat 15.2.3.2 Gas Bubble Passage through Fuel, PPS folicwed In long High Flux Flux-Total Flow Padial BlarAet and Control term by decay heat Assemblies removal 15.2.3.3 Seismic Psectivity Insertion-SSE PPS followed In long High-Flux Flux-Total Flow term by decay heat Flux- Pressure r eoval HTS Pump Electrics 15.2.3.4 Control Assembly withdrawal at PPS followed in Long Flux- Pressure Flux-Total Flow Startup-Maximum Mechanical Speed term by decay heat Flux-Delayed Flux

~,

u r eovat 15.2.3.5 Control Assembly Withdrawal at PPS followed in long High Flux Flux-Total Flow 4

a Power term by decay heat removal 15.3.1 Anticipated Events 15.3.1.1 Loss of Off-site Electric Power PPS followed In long HTS Pump Frequency HTS Pump Voltage term by decay heat removal 15.3.1.2 Spurious Primary Pump Trip PPS followed in long Flux to Pressure Flow Ratio term by decay heat Speed Ratio removal 15.3.1.3 Spurious intermediate Pump Trip PPS followed in long Speed Ratto Flow Ratio term by decay heat removal

~:: 3>

Qj@ 15.3.1.4 Inadvertent Closure of One PPS followed In long Steam-Feedwater Evap. Outlet Temp.

- s Evaporator cr Superheater Module term by decay heat

~F to lsolation Valve r eoval

$$ 15.3.1.5 Turbine Trip Long Term by decay heat removal (2)

Steam-Feedwater Loss of Condenser Vacuum e O O

A.56 LSD-2 (Proprietary Burns and Roe)

Line Source Dose - 2 calculates the dose rate behind a multilegged shield due to a number of ganina emitting cylinderical sources. LSD - 2 employs the Rockwell method of approximating the dose rate from a cylindrical source behind a semi-infinite slab shield. The cylindrical source is replaced by an equivalent line source which is located inside of the source to account for source self-attenuation.

Given the radiation sources and the shield material thicknesses along.

the line of sight from the radiation sources to the detector point, the corresponding ganna exponential attenuation is determined for each source energy group to yield the uncollided gamma flux contribution.

The collided flux contribution is calculated by the use of a single material buildup factor. A Lagrangian interpolation of gamma flux to dose rate conversion factors is used to yield the dose rate.

Availability LSD-2 is a Burns and Roe proprietary code.

Veri _ fica tion The LSD-2 code verification has been by hand calculation, as recorded in Burns and Roe, Inc. proprietory documents.

/^'s Application

()

LSD-2 is used to detennine bulk shielding wall thicknesses to meet the CRBRP radiation zone requirements.

(" ; . Amend. 45 C/ A-205 July 1978

A.56. A NUPIPE NUPIPE is a 1incar elastic, static, end dynamic f inite olctnent computer code f or three-dimensional piping systems. The code is prirrarily used to perform thermal expansion, deadweight, and seismic analysis of the Auxiliary Liquid Metal Piping System. The progran performs analysis in accordance with the requiranents of the ASME Section ill Code f or Nuclear Class 1, 2, or 3 components and ANSI B31.1 for power piping at the option of the user.

Availabilltv NUPIPE Version 1.4 is currently available on the IBM 370/3033 computer of RockwelI international, Canoga Park, California.

XerifIcation Since 1976, NUPlPE has been used for piping analyses of over 50 licensed U.S.

nuclear power plants. The NRC example problems have been run and results approved by NRC. Additional hand calculations and cross-checks with other piping analysis prograns have f urther verified the correctness of the program.

Apolication NUPIPE will be utilized for the static and dynamic analyses of piping systems of the Auxillary Liquid Metal System.

Reference Quadrox Corporation, 1700 Del l Avenue, Campbel l, Cal if ornia 95008,

" Verification Manual for NUPlPE Version 1.4," " User's Manual for NUPIPE Version 1.4," and " Programmer's Manual for Version 1.4."

O A-205a Amend. 66 s

March 1982

e j A.57- MAP ~

Q,)

The MAP code is -a radiation transport code employing point kernel techniques with a multigroup angular dependent surface source. The surface source geometry is the cylindrical surface defined by DOTIIIW (see A.22)

~d iscrete-ordinate transport code-problem. - Angular-and energy-dependent surface source data are obtained from the DOTIIInf code on magnetic tape and processed by the f1AP code to provide flux and response data at a surface detector. MAP provides techniques which circumvent the use of discrete ordinate transport codes in calculating radiation transport through voids or near-voids.

Availability The MAP. code is available on the Westinghouse Power Systems CDC-7600

~' computers located at the Monroeville Nuclear Center. The version currently being used was released from WANL in August,1970, and has been updated at ARD i.G satisfy CRBRP shield design analysis needs.

~# '

Verification The MAP code results have been compared to results from hand calculations, results from similar codes, and experimental results. The use of NAP to accurately predict the detector response at various positions external to experimental configurations demonstrate the validity of the MAP technique.

O)

(, Application MAP is 'used to predict neutron spectra and radiation detector responses behind experimental configurations which have.been modeled in D0TIIIW discrete ordinates transport claculations. MAP results are then compared

-to experimental results in order to verify or to define problem areas in CRBRP shielding design analysis methods and/or basic nuclear data.

Reference

.a R. G. Soltesz, R. K. Disney,4 Jedruch, and S. L. Zeigler,

" Nuclear Rocket Shielding iteth'ods, Modification, Updating, and. Input Data Preparation. Volume 5. Two-Dimensional, LDiscrete Ordinates Transport Technique. Final Progress-Report." ,WANL-PR(LL)-34, Vol. 5 (NASA-CR-102968), A'igust 1970.

r

~

33 i'f A2206 Amend. 45 July 1978 x

(

g Ar .4 ,

LIST. 0F TABLES j~

(s

\,-) e mz TABLE 5 CRBRP SODIUM AND COVER GAS SYSTEMS - CLASS 1 Liquid Metal Retaining Welds in Vessels Protected by Guard Vessels G-9 Liquid Metal Rotalning Welds in Vessels Not Protected by Guard Vessels G-10 Cover Gas Retaining Components G-11 Olssimilar Metal Welds G-12 Bolting G-13 Integral Attachments for Vessels G-14 i

Liquid Metal Retaining Wolds in Piping Protected by Guard Pipe or Tank G-15 Liquid Metal Piping Not Protected by Guard Pipe or Tank G-16

) Integral Attachments for Piping and Valves G-17 Liquid Metal Rotalning Valves G-18 Reactor Internal Components G-19 TABLE 5-2 CRBRP SODIUM AND COVER GAS SYSTEMS - CLASS 2 Liquid Metal Retaining Welds in Vessels G-20 Wolds in Guard Tanks and Guard Pipes G-21 Integral Attachments G-22 Bolting G-23 Dissimilar Metal Welds- .G-24 Liquid Metal Piping and Valves. G-25 Steam Generator Tubing G-26

, TABLE 5-3 CRBRP SODIUM AND COVER GAS SYSTEMS - CLASS 3 l ,_s . Class 3 Components- ~G-27

ka l l

G-IV-Amend. 67 March 1982

TN3LE 7-1 CRf3RP COMP 0tJEtJT SUPPORTS FOR SODIUM AtJD COVER GAS SYSTEMS Plate and Shell Type Supports G-30 Linear Type Supports G-31 Component Standard Supports G-32 LIST OF ADDEtJDA ADDINDA EAGE A Justif ications for ASME Code Exceptions G-38 LIST OF APPLICABLE DOCUMEtJTS The below listed documents form a part of Part I of this Appendix to the extent specified herein.

o ASME Boiler and Pressure Vessel Code,Section XI Division 1, 1980 Edition, o ASME Boller and Pressure Vessel Code,Section XI Division 3, Winter 1981 Addenda.

EQIE: Section XI Division 3 is not currently recognized in 10 CFR 50.550.

i e

G-v Amend. 65 Feb. 1982

.. . _=_ - -__ - . .- - ..__. _.- - . .. _. - -__ _ - . _ . . -. -- . .

I e

i i

j AMEN 0 MENT 67

!. LIST OF RESPONSES TO NRC QUESTIONS Response to NRC Questions Received Since the Fall of 1981:

Q471.1-1

, Q471.2-1 Q471.2-2 Q471.3-1 Q471.4-1 Q471.4-2 Q471.4-3 l Q4 71.5-1 Q471.5-2 i Q471.6-1 Q471.8-1 Q760.1-1 I

Q760.2-1 1

O Q760.3-1 0760.4-1 Q760.5-1 t

i L

t O Qi i

% QUESTION CS 471.1 Provido commitments to confccm to the provisions of the following Regulatory Guides, as they apply to CRBRP, cr describe alternative measures to be taken to provide a comparable degree of worker protection.

1.97, " Instrumentation f or Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" 8.12, '" Criticality Accidont Alarm Systems" 8.13, " Instruction concerning Prenatal Radiation Exposure" 8.14, " Personnel Neutron Dosimeters" 8.15, " Acceptable Programs f or Respiratory Protection" 8.26, " Applications of Bloassay for Fission and Activation Products" 8.27, " Radiation Protection Training f or Personnel at Light-Water-Cooled Nuclear Power Plants" 8.29, " Instructions Concerning Risks f rom Occupational Radiation Exposure"

RESPONSE

]v The applicability of the subject NRC Regulatory Guides to CRBRP is summarized as f ollows:

Regulatory Guide 1.97, Rev. 2, is considered to be applicable in principle and intent, however some parameters to be monitored for accidents in CRBRP are dif f erent f rom those f or Light Water Reactors. The appropriate application of R.G.1.97, Rev. 2, to CRBRP wil l be provided in PSAR Section 7.5.11 by May, 1982, which will show a comparable degree of protection consistent with LWR application.

Regulatory Guide 8.12, Rev.1, Regulatory Position 1, is considered to be applicable in principle and Inter.+, The potential for a criticality accident at (RBRP does not exist due to the quantitles and form of the Special Nuclear Material, the geometric spacing and/or permanently installed neutron absorbers. Criticality evaluations for unirradiated f uel assemblies ~ In New

-Fuci Shipping Containers and the New Fuel Unloading Station are discussed in PSAR Section 9.1.1.3. Criticality evaluations in the EVST and the FHC are provided In PSAR Sections 9.1.2.1.3 and 9.1.4.10.3.

Regulatory Guides 8.13, 8.14, 8.15, 8.26, 8.27, and 8.29 are considered to be applicable in principle and ' intent to CRBRP. The explicit appIIcation of these Regulatory Guides will be specified in the FSAR at ths Operating License stage.

O-U -

, Q471.1-1 Amend. 67 March 1982

A CS 471.2 .As specified in Regulatory Guide 8.19, you should use personnel

-V .(12.1.5) exposure data for specific kinds of work and job functions available f rom similar operating plants. Describe Inf ormation you have obtained concerning operating L WBRs, regarding source terms and occupational radiation exposures.

Reply: The number of operating LWBRs is very limited. Most of the existing operating LWBRs which have had suf ficient power operation  ;

to build up residual radiation sources are located in Europe and

' the Sov iet Uni.on. In addition, the results obtained from one operating plant are not necessarily applicable to another plant or plant design. The operating temperatures and plant arrangements are very important f actors in determining] the residual radiation dose rates.

For these reasons, our principal offort has been to develop models which predict the radiation dose rates in all colis requiring-operational or maintenance personnel entry. The source terms used to derive-these radiation dose rates have been compared with the available' LMFBR source term inf ormation. '

Light water nuclear p! ant experience has been used to determine

-maintenance requirements wherever similar components are in use.

The following sources of information have been used to determine source terms and occupational radiation exposures:

Data and Source Term Model Information

1. W. F. Brehm, "Effect of Oxygen in Sodium Upon Radionuclide Release from Austenitic Stainless Steel", HEDL-SA-985, September 1975

2. W. F. Brehm et al, "Radionuclide Release from 316 Stainless Steel into 538 C Sodium", HEDL TME-78-85, March 1979

3. C. Bagnall and D. C. Jacobs, " Relationship for Corrosion of Type 316 Stainless Steel in Liquid Sodium", WARD-NA-3045-23

4. Proceedings of the International Conference on Liquid Metal Technology in Energy Production, Seven Springs, Pa.,1976, "The infIuence .of LMFBR Fuel Pin Temperature Prof IIes on Corrosion Rates",:S. A. Shiels, C. Bagnall, S. L. Shrock, S. J. Orbon.

Sources of Information on Quaratino LMFBRs '

5. V. D. Kfzin, V. L. Polyakov et al, "How BOR-60 Reactor Power Station Equipn.bnt is Services in a Radiation Environment",

' Atomnaya Energlyo, . Vol . 37, No. 6, pp. 471-474, December 1974 a

Q471.2-1 Amend. 67 March 1982

6. V. D. Kizin, V. L. Polyakov et al, " Radioactivity of Long-Lived Nuclides in the Primary Loop of the 80R-60 Reactor During Operation with Defective Fuel Elonents", Lenin Research Institute f or Nuclear Reactors (Nil AR-P-5 299), Dimitrougrad, January 1977.

7. A. W. Thorley et al, " Fission and Corrosion Product Behavior in Primary Circuits of LMFBRs, A Status Review of Work being Done in the U.K.", TGR-Roport-2856(R)

8. "DFR Bohavior of Fission Products, Activation Products and Fissile Material in the DFR Coolant", TGR Memorandum 5131

9. C. A. Erdman et al, "Radionucildo Production, Transport, and Release f rom Normal Operation of Liquid Metal cooled Fast Broeder Reactors", EPA-520/3-75-019 Sources of Information on Radiation Exoosure and Plant Maintenance

10. L. A. Johnson, " Occupational Radiation Exposure at Light Water Cooled Power Reactors", NUREG-0323, March 1978 O

r O

Q471.2-2 Amend. 67 L.

h March 1982

- ~...- - -

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

. 7-~( , QUESTION CS 421,1;(12.2.4.2.1)

, As specified'In Regulatory Guide 1.70, Section 12.3.4, you should describe

, criterla and methods f or_ obtaining representative inplant airborne radioactivity concentrations, including airborne radiolodines and other

-t radioactive, materials. Describe how radiolodines will be determined durir.g .

. use of gaseous and particulate radioactivity monitors. ,

. RESPONSE

• i Section 12.2.4.1 of the PSAR describes the design criteria for airborne .

radioactiv ity . sampl ing. Section 12.2.4.2.1 describes the methods used in processing airborne radioactivity samples for 1) Gaseous activities, 2)

Particulate and Gaseous activities, 3) Particulate, lodine, and Gaseous-activities, and 4) alpha activities. A description of how radiolodines are

, determined when sampled is contained in 3) above, Particulate, lodine, and ,

Gaseous activities.

Samples frum systems (other than general area air samples) are obtained using Isokinetic or other appropriate sampling as the means of receiving a representative. sample. In addition, Instrumentation is located as close as

_ practicable. to the sample point and sample line bends are minimized. Sampl e

. line material is selected to minimize or elim!nate such variables such as plate out, etc.

4 1

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Q471.3-1 Amend. 67 March 1982 +

,-. , ,, ,w.. .. . . - e... _, h .y , _ -_. ,.r -w.. - - . , - , y y .,,49

NUCLEAR REGULATORY CONMISSION QUESTION / RESPONSE CS471.4 Ouestion CS471.4 (12)

As speci f ied in Regul atcry Guido 1.70, Section 12.3.1.2, you shoul d describe any special protectivo f eatures that use shleiding, gocnetric arrangement, or rmote handling to assure that ORES oro ALARA. Describo the spent f uel transf er process, with clear drawings of f uel at each step of that process, showing rolovent shleiding present, access control features, and maximum dose ratos in occuploble spaco nearby. Describo precautions taken to provent inadvertent access to all unshioided potentially very high r adiation areas in the vicinity of the spent f uel transfer pathway.

All accessible portions of the spent f uol transf er pathway must be shielded during f uel transf er. Use of removable shielding for this purposo is acceptablo. This shielding shall be such that the resultant contact radiation lovol s shal l be no greater than 100 rads por hour. All accessible portions of tho spent f uel transfer pathway whero potentially lethal radiation fields are possible during f uol transf er, shall be clearly marked with a sign Indicating 1 hat f act, if removable shielding is used f or the f uel transf er pathway, it must also be explicitly marked as above. If other than permanent shielding is used, Iocal audlblo and visiblo alarming radiatton monitors must bo Installed to alert personnel if temporary f uel transfer pathway shielding is rmoved during f uel transf er operations.

BOSRODSD As specif ied in Regulatory Guido 1.70, Section 12.3.2, special protective features, including shielding, geometric arrangement, and remote handling, assuro that overall radiation exposures are ALARA. All accessible portions of the spent f uel transf er pathway are shioided during f uel transf er. Permanent shielding incorporated in the f uel handling equipment, rather than rmovable shielding, is usod for this purpose. This shielding is such that the resulting contact radiation lovels moot the CRBRP objectivos stated in the PSAR, Section 12.1.1.2.

Rof uoling operations are described in Section 9.1.4.1 of the PSAR. The operations described includo receipt of now core assemblies into the plant (transf er f rom shipping containers to the ex-vessel storage tank), refueling preparations, t of uoling, ref uol Ing termination, and spent f uel shipping. The f I c.: path of now and spent coro assembiles during these operations is shown on the plan view of the RG and RSB in Figuro 9.1-1 of the PSAR. The text and flow path identify several locations in which core assembiles are contained for uso, interim storago, or longer term storage pending use or shipment. These locations are listed below. The equipment and f acilities, and their shielding, are described in the PSAR sections ref erenced and shown in isometric drawings in the PSAR figuros referenced. Shielding of cells in which equipment is contained is described in PSAR Section 12.1.2.

O I

Q471.4-1 Amend. 67 March 1982

O Type of Core Assembiles Eaulomont or Facility _ Section Floure Contained Roactor Vossel 5.2 -

New and Spent Now Fuel Shipping Container 9.1.1 9.1-5 New Only New Fuol Unloading Station 9.1.1 9.1 -3 New Only Ex-Vessel Storage Tank (EVST) 9.1.2.1 9.1 -6 New and Spent Ex-Vessel Transf er Machine (EVTM) 9.1.4.3 9.1-13 New and Spent Fuel Handling ColI (FHC) 9.1.4.10 9.1-7 & Spent Only 9.1-9 Spent fuel Shipping Cask (SFSC) This item is sep- Spent Only arately licensed.

The list includos all locations f or both new and spent core assemblies; however, only spent core assemblies are discussed f urther in this response.

All spont core assemblies (f uel, blanket, control, and rcrnovable radial shleid assemblies) are handled alike. The calculatea radiation dose rates in normally occupied areas f rom spent core assemblies in these locations are listed below.

Equipment or Dose Pete Facilltv Normally Occuoted Area (mrem /hr)

Reactor Vessel Head access area 0.1 to 0.4 (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reactor shutdown)

EVST RSB operating f loor above the EYST 0.05 EVTM EVTM control console and EVTM fIocr 1.8 valve operator locations (on R W or RSB operating floor at least 12 feet f rom the EVTM center!Ine)

FHC FHC operating gallery, RSB operating 0.2 floor, cask corridor (with shipping cask loading port plug installed),

and other colis adjacent to the FHC O

Q471.4-2 Amend. 67 March 1982

/T in cases where transient surf ace dose rates exceed the limits specified in PSAR

(~,) Section 12.1.2 access to the equipment is administratively controlled. In each case, definition of the area and responsibility for ensuring that personnel are excluded w!Il be part of ref ueling operating procedures. The areas are identif led and discussed below.

1. During passage of a spent f uel or blanket assembly through the EVTM closure valve and floor valve, the access to space within 9 f t, of the valve's surf ace (within 12 f t. of the EVTM centerline) is administratively prohibited. The transient dose rate at the valve's surf ace during assembly transf er will be about 1100 mrem /hr. The elevated level will exist for only about 10 sec.

2. During passage of the IVTM port plug through the IVTM Port Nozzle, access to the space within 7 feet of the nozzle centerline is administratively prohibited. The transient dose rate at the equipment surf ace during the IVTM port plug transf er is about 750 mrem /hr. The elevated level will exist for about three minutes.

3. During loading of a spent core assembly into a spent f uel shipping _

cask f rom the FHC, the access door to the cask corridor wil I be locked and posted as a high radiation area. Positive control wilI be made bef ore each entry following the completion of the spent f uel shipping cask loading to ensure that a high radiation level does not exist.

Personnel access to the corridor will be required only for mating the cask to the FHC spent f uel loading port in preparation for core assembly loading and for demating the cask af ter loading has been completed. The transient dose rate at the surf ace of the cask has been estimated to approximately 102-103 Rem /hr based upon SFSC studies. The design surf ace dose rate will be supplied in the FSAR.

The elevated radiation level wilI exist for less than 30 seconds per core assembly transf er.

[ 'l L.)

~ Q471.4-3~

Amend. 67 March 1982

('} QUESTION CS 471 d (12.3)

( As specifled in Regulatory Guide 8.8, you should attain the objectives in Section C to provide assurance that exposures of station personnel to radiation will be ALARA ihroughout the plant, from planning and design through decommissioning.

Describe the f eatures that you have incorporated into your design to maintain occupational radiation exposures as low as is reasonably achievable during the eventual decommissioning of the plant. '

RESPONSE

The ALARA program for design of CRBRP has been described in Section 12A.3.1.

Features incorporated into the design ior maintaining operational, surveillance and maintenance radiation exposures ALARA also assist in maintaining exposures ALARA during decommissioning. The design features include the following:

(1) Permanent Shieldina and Isolation of Comoonents isolation of system components is provided by placing tanks, pumps, valves and process loops in separate shielded cells which will reduce the radiation exposures during decommissioning operations by reducing the exposures from adjacent radiation sources. Furthermore, the thickenss of permanent shielding between cells is based upon reducing h)

L the radiation level in the occupied cell to the objectives specified In PSAR section 12.1 with the adjacent cells operating or filled with radioactive sodium. Since Na-22, Na-24, corrosion and fission products are the predominant radiation source, draining the process systems prior to removal of components during maintenance and decom-issioning will result in lower dose rates f rom adjacent cells than during the operating phase of the plant. Shielding of these cells f or parsonnel protection during the operating phase of the plant is more limiting than exposure considerations for maintenance and decomissioning.

(2) Accessibility for Maintenance and Removal of Eauloment Evaluations of the various plant systems containing radioactive process streams have been conducted to determine the necessary space for portable shielding, the requirements for hoisting equipment and adequate removal paths. Features requis ed f or the instal lation of equipment, such as padeyes, will be retained to f acilitate the removal of components during maintenance activities and decommissioning.

Draining of radioactive sodium from nearby equipment (discussed in (1) above) also f acilitates access for maintenance and decommissioning.

(3) Material Selection Material used in the reactor Internals and primary heat transport ry system have been selected to minimize the presence of activated corrosion / erosion products. As indentified in Section 11.1.5, plate (v)

Q471.5-1 Amend. 67 March 1982

out of corrosion (mass transfer) products in the PHTS system will occur rather than the build up of local crud traps. Hence, the design has inherently avoided the localized high radiation level probi ms associated with crud traps.

(4) Quanuo Fissia Fuel and Corrosion Products bv Cold Traps The cold traps normally operating in the primary and ex-vessel sodium system (see Section 9.2) will collect various activated isotopes (see Tabl e 12.1 -7 ) . As such, these components decrease the overall presence of isotopos in the components to be rmoved during decommissioning. Although cold trap rmoval is not scheduled during the operating life of the plant, an ALARA has resulted in a decision to provide Integral shielding f or these traps to minimize exposures during either unplanned replacement or decommissioning.

(5) Location of eaulomont in lowest oractical radiation zone Where practical, all equipment has been located in low radiation zones to minimize Irradiation of equipment, and to permit maintenance /

removal of the equipment with the lowest resulting personnel doses.

(6) Refueling Sv11em and Comoonents The refueling system is in place during decommissioning. This ensures that alI of the spent core assembiles can be rmoved f rom the reactor and eventually rmoved from the site. The refueling system can also be used to rmove dummy assembiles f rcrn the RV which were used in the core assembly rmoval as welI as remove lower inlet modules from the lower internals of the RV. The core assembly handling machines have been designed for cleaning af ter use since they will be contaminated but not radioactive. The Fuel Handling Celi is lIned wIth stainless steel which aids in cel1 decontamination.

(7) Sodium Removal and Decontamination System and Comoonents Cleaning equipment and decontamination facilities are provided for sodium rutoval and acid solution decontamination and will continue to be available for processing components for disposal during decommissioning. Installed IIquid and gaseous waste processing for the ef fluents is in place and will be available for use during decommissioning. A regulated maintenance shop is available and can be used f or disassembly of low level radioactive components for convenient disposal.

O Q471.5-2 Amend. 67 March 1982

Question CS 47Lft

~

N-As speci f ied in NUREG-0718, " Proposed L Iconsing Requ'/ements f or Pending Applications f or Construction Permits and Manuf acturing License" (ll.B.2), as it relates to CRBRP, you should perf orm radiation and shielding design reviews of spaces around systems that may contain highly radioactive fluids; implement plant design modifications necessary to permit adequate access to vital areas and protect saf ety equipment; and to the extent possible, provide preliminary

-design Inf ormation at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, provide

. a general ' discussion of their approach to meeting the requirements by .

specifying the design concept selected and the supporting design bases and

-criteria; demonstrate that the design concept is technically feasible and within the state of the art, and that there exists reasonable assurance that

, the requirements will be Imptomented properly prior to the issuance of l operating licenses.

I i- Response:

The CRBRP AL ARA program is described in PSAR Section 12. A. CRBRP l implementation of NUREG 0718 is discussed in recently submitted Appendix H to the PSAR ( Amendment 66) . Your attention is directed to Section ll.B.2 of

' Appendix H.

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Q471.6-1 Amend. 67.

March 1982 r . . . . . . .

,m. QUESTION CS 471.8 l \

As specif ied in NUREG-0718, " Proposed Licensing Requirements for Pending App!Ications f or Construction Permits and Manuf acturing License", (lli.D.3.3),

as it relates to CRBRP, you should review your designs to assure that provisions for monitoring inplant radiation and airborne radioactivity are appropriate fcr a broad range of routine and emergency conditions. To the extent possible, provide preliminary design Information at a level consistent with that normally required at the construction permit stage of review. Where new designs are involved, provide a general discussion of your approach to meeting the requirements by specifying the design concept selected and the supporting design base; and criteria. Demonstrate that the design concept is technically feasible ahu within the state of the art, and that there exists.

reasonable assurance that the requirements will be implemented properly prior l- to the issuance of operating licenses.

RESPONSE

Chapter 14 of the PSAR includes detailed inf ormation concerning r.tonitoring of Inplant radiation and airborne radioactivity during routine conditions.

Emergency conditions are also considered. CRBRP Implementation of NUREG-0718 is discussed in Appendix H of the PSAR (Amendment 66). Your attention is directed to Section ill.D.3.3. of Appendix H. Regulatory Guide 1.97, as it l appl ies to CR8RP, is presently being incorporated into the plant design and a revision to the PSAR, Section 7. 5.11, is expected by May,1982.

sm- )

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W (v )

Q471.8-1 Amend. 67 March 1982

O. QUESTfON 760.1-

.[Picase provide] SAS.3D input decks on magnetic tape and microf iche output for E

EOC-4/LOF, BOC-l/LOF, EOC-4 TOP, and 80C-1/ TOP cases corresponding to the best-estimate analyses of mBRPO-GEFR-00523 (draf t dated October .1981).

RESPONSE

i The Information requested has been supplied under separate cover in reference Q760.1-1.

l l

O Referencer 'Q760.1-1 Letter, HQ:S:82:006, J. R. Longenecker to P.S.

Check,' dated February 19, 1982 Q760.1 -1 Amend. 67 March 1982

OUESTl0PL]fL2

[Please provido] a list of all changes to SAS3D Release 1.0 that have been incorporated and used f or each of the cases ref erred to in item 1 above

[Q760.1] If di f ferent f rom Appendix A of CRBRPO-GEFR-00523. In addition, the speci f ic FORTRAN changes f or al l modif ications are needed f or evaluation and updating of the Los Alamos versions of SAS3D.

RESPONSE

The Inf ormation requested has been supplied under separate cover in ref erence Q760.1-1.

l l

/ ^'%.,

I i N. ._.,/

,.-~,.

Q760.2-1 Amend. 67 March 1982 i ,

....--....--.a,,.a. _ . . . - _ . . . - . . _ . _ .

. m . . _ _ . _ . . - . _ . . . - . _ _ __ _ _ _ _. .._

' OUFSTION 760J I (Please 'providel EOC-3 power and neutronics data for constructing an. SAS3D Input deck, i f an f0C-3 deck exists, please provide under item 1 above.

RESPONSE

The neutronics data explicitly requested above has'never been developed by the

~ Project. However, the data could 'be generated by the NRC Consultants using

, the information provided in response to Question 760.5. The NRC and the

] Project agreed on this matter during a teleconference on February 2,1982.

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, ' Amend . 67  :

March 1982 ,

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OUESTION 760.4

[Pleaso provido] SAS3A to SAS3D Input conversion progrm developed at Argone flational Laboratcry. Previous hccogeneous coro analyses were perf ormed with SAS3A. In order to investigato cceparativo accident characteristics, these SAS3A decks need to be converted to the SAS3D f ccmat.

RESPotlSE The inf ormation requested has been supplied under separate cover in ref erence Q760.1-1.

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Q760.4-1  !

Amend. 67  !

March 1982 I

i QUESTf0N 760.5 1

.[Pleaseprovido]isotopicdistributionsforEOC-4andEOC-3. Axial distributions over the f ueled length of the pins f or each subassembly y including blankets and control rods are required f or 01F3DS and SIMMER Input. ,

Documentation on .the generation of these distributions is needed to establish l required data interf aces.

RFRPONSE The neutronics Inf ormation for End of Oycle 4 (EOC-4) based on the three-dimensional ' VENTURE computer code has been supplied under separate cover in Ref erence Q760,1-1.

The two-dimensional isotopic distribution inf ormation from the 2DB computer code f or 800-1, EOC-3, and EOC-4 has been supplied under separate cover in Reference Q760.5-1.

REFERENCE Q760.5-1 Letter HQ:S:82:015, J. R. Longenecker to P. S. Check, dated March 29, .1982.

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x2 Q760.5-1 Amend. 67 3

i March 1982