ML20065S249

Amend 72 to PSAR

ML20065S249
Person / Time
Site:	Clinch River
Issue date:	10/31/1982
From:	ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:
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ML20065S237	List: ML20065S233 ML20065S249
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NUDOCS 8211010154
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:110 October 29, 1982 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Check:

AMENDMENT NO. 72 TO THE PRELIMINARY SAFETY ANALYSIS REPORT (PSAR) FOR CLINCH RIVER BREEDER REACTOR PLANT (CRBRP)

The application for a Construction Permit and Class 104(b) Operating License for the CRBRP, docketed April 10, 1975, in NRC Docket No. 50-537, is hereby amended by the submission of Amendment No. 72 to the PSAR pursuant to 50.34(a) of 10 CFR, Part 50.

g This Amendment No. 72 includes: Responses to U.S. Nuclear Regulatory i Commission requests for additional information contained in letters dated k April 19 and 30, May 14, June 9 and 21, and July 16, 1982; Revisions to Section 11.4, " Process and Effluent Radiological Monitoring System;"

Chapter 12, " Radiation Protection;" and other updates and revisions.

A Certificate of Service, confirming service of Amendment No. 72 to the PSAR upon designated local public officials and representatives of the Environmental Protection Agency, will be filed with your office after service has been made. Three signed originals of this letter and 97 copies of this amendment, each with a copy of the submittal letter, are hereby submitted.

Sincerely, hif.@g rw?M tG Jo R. Longenec(\ er Acting Director, office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy Enclosure SUBSCRIBED AND SWORN to before me this & day of October 1982 cc: Service List Standard Distribution p]

L Licensing Distribution b

7 /> h] r>M]

NOTARYpBLIC 8211010154 021029 , Commission Expires April 28, 1984 gDRADOCK 05000537 AnD

p SERVICE LIST U

Atomic Safety & Licensing Board Dr. Cadet H. Hand, Jr., Director U. S. Nuclear Regulatory Commission Bodega Marine Laboratory Washington, D. C. 20555 University of California P. O. Box 247 Atomic Safety & Licensing Board Panel Bodega Bay, CA 94923 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Lewis E. Wallace, Esq.

Division of Law Mr. Gerald Largen Tennessee Valley Authority Office of the County Executive Knoxville, TN 37902 Roane County Courthouse Kingston, TN 37763 Dr. Thomas Cochran Natural Resources Defense Council, Inc.

1725 I Street, NW Suite 600 Washington, DC 20006 Docketing & Service Station Office of the Secretary U. S. Nuclear Regulatory Commission Washington, DC 20555 Counsel for NRC Staff U. S. Nuclear Regulatory Commission Washington, DC 20555 William B. Hubbard, Esq.

Assistant Attorney General State of Tennessee Office of the Attorney General 422 Supreme Court Building Nashville, TN 37219 Mr. Gustave A. Linenberger Atomic Safety 3 Licensing Board U. S. Nuclear Regulatory Commission Washington, DC 20555 Marshall E. Miller, Esq.

Chairman Atomic Safety & Licensing Board U. S. Nuclear Regulatory Commission Washington, DC 20555 William E. Lantrip, Esq.

n Attorney for the City of Oak Ridge

() 725 Main Street, East Oak Ridge, TN 37830 2/17/82

STANDARD DISTRIBUTION Mr. R. J. Beeley (2) Mr. W. W. Dewald, Project Manager (2)

Prooram Manager, CRBRP CRBRP Reactor Plant Atoinics International Division Westinghouse Electric Corporation Rockwell International Advanced Reactors Division P. O. Box 309 P. O. Box 158 Canoga Park, CA 91304 Madison, PA 15663 Mr. Michael C. Ascher (2)

Project Manager, CRBRP Mr. H. R. Lane (1)

Burns and Roe, Inc. Resident Manager, CRBRP 700 Kinderkamack Road Burns and Roe, Inc.

Oradell, NJ 07649 P. O. Box T Oak Ridge, TN 37830 Mr. Percy Brewington, Jr. (2)

Acting Director Mr. George G. Glenn, Manager (2)

Clinch River Breeder Reactor Plant Clinch River Project P. O. Box U General Electric Company Oak Ridge, TN 37830 P. O. Box 508 Sunnyvale, CA 94086 Mr. Dean Armstrong (2)

Acting Project Manager, CRBRP Stone & Webster Engineering Corp.

P. O. Box 811 Oak Ridge, TN 37830 Mr. William J. Purcell (2)

Project Manager, CRBRP Westinghouse Electric Corporation Advanced Reactors Division P. O. Box W Oak Ridge, TN 37830 Number of copies in parentheses.

1/11/82

LICENSING DISTRIBUTION Mr. Hugh Parris Manager of Power Tennessee Valley Authority 500A CST 2 Chattanooga, TN 37401 Dr. Jeffrey H. Broido, Manager Analysis and Safety Department Gas Cooled Fast Reactor Program General Atomics Company P. O. Box 81608 San Diego, CA 92138 Mr. George Edgar Morgan, Lewis, and Bockius 1800 M Street Suite 700 Washington, DC 20036 O

O 2/17/82

O PAGE REPLACEMENT GUIDE FOR AMENDMENT 72 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT (DOCKETN0.50-537)

Transmitted herein is Amendment 72 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537. Amendment O 72 consists of new and replacement pages for the PSAR text and Responses to NRC Questions.

Vertical margin lines on the right hand side of the page are used to identify changes resulting from NRC Questions and margin lines on the left hand side are used to identify new or changed design information.

The following attached sheets list Amendment 72 pages and instructions for their incorporation into the Preliminary Safety Analysis Report.

O

AMENDMENT 72 PAGE REPLACEMENT GUIDE REMOVE THESE PAGES INSERT THESE PAGES Chapter 1 1.3-21, 22 1.3-21, 22 Chapter 2 2.5-23b, 24 2.5-23b, 24 2.5-61, 62 2.5-61, 62 Chapter 3 3.1-19, 20 3.1-19, 20 3A.8-4, 5 3A.8-4, 4a, 5 3A.8-9a, 9b 3A.8-9a, 9b Chapter 4 4.2-230, 231 4.2-230, 231 4.2-278 thru 281 4.2-2/8, 278a, 279, 280, 281 O,

4.2-304, 305 4.2-304, 305 4.2-410, 411 4.2-410, 411

, 4.2-614 thru 617 4.2-614, 615, 615a, 616, 617 4.2-620, 621 4.2-620, 621 Chapter 5 5.2-4c, 4d 5.2-4c, 4d 5.2-6b, 7, 7a, 8 5.2-6b, 7, 7a, 8 5.2-10b, 10c, 10d, 11 5.2-10b, 10c, 10d, 11 5.2-12, 12a, 13 5.2-12, 12a, 13 5.3-21, 21a 5.3-21, 21a 5.5-12, 12a 5.5-12, 12a 5.5-24, 24a 5.5-24, 24a 5.5-35, 35a, 35b 5.5-35, 35a, 35b 5.5-53a 5.5-53a thru 53e 5.6-5, 6, 6a, 7 5.6-5, 6, 6a, 6b, 7 5.6-42 5.6-42 thru 47 5.7-3, 3a, 4 tnru 8 5.7-3, 3a, 4 thru 8 5.7-14a thru 14k O

v A

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

i REMOVE THESE PAGES INSERT THESE PAGES Chapter 6

'6-1 thru y 6-1, ii, i f a, iii, iv, V 6.2-14 thru 17 6.2-14, 15, 15a, 16, 16a, 17 Chapter 7 7-i thru v, va, vi, via, 7-i thru v, va, vi, via, vii, vii thru xii viii, ix, ixa, x thru xv 7.1-1, 2 7.1-1, 2, 2a 7.1-7 thru 10 7.1-7 thru 10 7.2-1, la 7.2-1, la 7.2-11 thru 14, 14a, 15, 16 7.2-11 thru 15, 15a, 16 7.2-22, 23 7.2-22, 23, 23a 7.3-3, 4, 4a, 5, 6 7.3-3, 4, 5, Sa, 6 7.4-1, 2 7.4-1, 2 7.4-5 thru 8, 8a thru 8e 7.4-5, Sa, 5b, 6, 7, 8, 8a thru 8f 7.4-10c 7.4-10c, 10d 7.5-7, 8, 8a, 9, 10 7.5-7 thru 10 7.5-18a, 18b, 19, 19a, 20 thru 25 7.5-18a, 18b, 19, 20, 21, 21a, 22 thru 25 7.5-33f, 33g 7.5-33f, 33g 7.5-33j 7.5-33j

[]

\.s 7.6-1, 2, 2a, 2b 7.5-54, 55 7.6-1, 2, 2a,.2b 7.6-2e 7.6-2e 7.6-8, 9 .

7.6-8, 9, 9a 7.6-16 thru 19 7.6-16, 17, 18, 18a, 18b, 19 7.6-26, 27 7.6-26, 27 7.7-9, 9a 7.7-9 7.9-5, 6, 6a thru 6d 7.9-5, 6, 6a thru 6d Chapter 9 9.1-60, 61 9.1-60, 61 i 9.6-45, 45a 9.6-45, 45a '

9.6-47, 48 9.6-47, 47a, 48 9.7-3, 4 9.7-3, 4 9.7-7, 8 9.7-7, 8 Chapter 11 11.2-20, 20a 11.2-20, 20a 11.4-1, 2, 3, 3a, 4 thru 13 11.4-1, 2, 3, 3a, 4 thru 15 O 8

REMOVE THESE PAGES INSERT THESE PAGES Chapter 12 12.1-22b 23, 23a, 24 12.1-22b, 23, 23a, 23b, 24 12.1-28, 29 12.1-28, 29 12.1-78, 79, 80, 80s 12.1-78, 79 12.1-81, 82 12.1-81, 82 12.1-85 12.1-85 12.1-87 12.1-87 12.1-89 thru 94 12.1-89 thru 94 12.1-96 12.1-96 12.1-99, 99a, 99b 12.1-99, 99a , 99b 12.1-101 12.1-101 12.2-3, 3a~, 4, 4a, 4b, 4c, 5, 6 12.2-3, 3a, 4, 4a, 4b, 4c, 5, 6 12.2-9 thru 13 12.2-9, 10, 10a, 11, 12, 13 12.3-13, 14 12.3-13, 14, 14a, 14b Chapter 15 15.1-105, 106 15.1-105, 106 15.3-6, 7 15.3-6, 7 15.3-28 15.3-28, 28a, 28b 15.3-34, 35 15.3-34, 35 .

15.3-49 15.3-49 0

O C

AMENDMENT 72 QUESTION / RESPONSE SUPPLEMENT This Question / Response Supplement contains an Amendment 72 tab sheet to be inserted following Qi page Amendment 71, September 1982. Page Qi Amend-ment 72 is to be inserted following the Amendment 72 tab sheet.

This Amendment 72 provides both new and revised Question / Response pages for NRC QUESTIONS RECEIVED SINCE THE FALL 0F 1981.

The following Question / Response pages are to be inserted in numerical order behind the appropriate numbered tabs in PSAR Volume 25 or 26 as appropriate. The parenthesis beside Question / Response shown indicates the number of pages associated with each Question / Response.

• QCS220.25 (84) QCS421.37 (1) QCS760.105 (1)

(1st page is a replacement) QCS421.42 (2) QCS760.110 (2)

• QCS421.9 (1) QCS421.47 (1) QCS760.116 (1)

• QCS421.17 (1) QCS421.48 (5) QCS760.131 (1)

QCS421.22 (3) QCS421.58 (1) QCS760.166 (6)

QCS421.27 (1) QCS721.1 (10) QCS760.172 (4)

QCS421.30 (1) *QCS760.13(1) QCS760.175 (1) 3 QCS421.31 (2) QCS760.28 (28) QCS760.176 (29)

QCS421.34 (1) QCS760.30 (1) QCS760.177 (1)

• QCS421.36 (1) QCS760.36 (7) QCS760.178 (120)

• These are replacement pages.

O D

O O O PSAR CRBRP - 975 Nt FFTF - 400 Nt MONJU*-714 h t Section steam drum to heat rejecticn .

atmosphere. Removes capability is 412%

4 up to s180 N t (18% of rated power.

rated power) (50 N t).

- Long term rejection of decay heat accomplished by condensing of steam in an air cooled 4 condenser. Removes up to 4.5% rated power j (45 Nt). When these systems are unavailable, decay / residual heat is removed by cooling of the reactor overflow y sodium by a Na/NaK heat

! C3 exchanger. NaK heat load i rejected to atmosphere by a NaK/ air heat excha nger, j 41 Removes between 10-11 Nt.

i 3 3 3 No. Loops

9. Plant Protection System 7.2 Reactor Trip Action (1) Release Rods (1) Release Rods (2) Trip Primary (2) Trip Prinary and Inter- and Inter-ay mediate Pumps mediate Pumps 3m E (3) Provides Turbine (3) Programs Dump Trip Signal Heat Exchanger T Guide Vanes D3

UBRP FFTF HONJ U' PSAR 975 Nt 400 Nt 714 N t Sect Ion Reactor Trip Circuits No. Circuits Monitored 24-Prl. System 23-Prl. System -

For Trip Actuation 16-Sec. System 19-Sec. System Basic Signal and Trip Prl.-2/3 Local Prl.-2/3 Local -

Output Signal Logic Coincidence Logic Coincidence Logic 7 Sec.-2/3 General Sec.-1/4 2/3 -

Coincidence Logic Hybrid General Local Colncidence Logic No. External Flux Monitors 3 3 --

Max. RSS Logic Resporse Time 0.200 0.200 -

(Frm time RSS senses w condi tion requiring trip to time when rods are g rel eased. ) (Sec.)

N

10. Containment 3.8 Type / Shape Single steel vessel, Single steel vessel, Single steel vessel, cyiIndrleal shelI wIth cyiIndrleal shelI cyiIndrleal meli wlth flat botta and hemi- with hemi-ellipsoldal hemi-spherical top and ellipsoldal top. Con- top and bottom heads, heel-ellipsoldal bott m.

crete shleldlng InsIde, Concrete shielding Concrete cy1Inder below operating floor, below operating floor. surrounds entire con-Steel containment sur- tainment, rounded by concrete con-f inement bull ding. An annulus space between containment and con-f fnement maintained at negative pressure with respect to outside atmosphere, kN na n

~~ e __ _

e _ _ _ _ _ . _ _ _

Figure 2.5-18, sheet 3 of 4 incl udes alI historical earthquakes within 50 mil es of the site. Maximum Modified Mercalli intensity is shown by open ci rcl es. Unfelt events and events with unreported intensity are shown by the Figure 2.5-18, sheet 4 of 4 al so incl udes all historical earthquakes within 50 mil es of the site. Magnitude is shown by the square symbols. Events with unreported magnitude are shown by the X's.

I l

O O 2.5-23b Amend. 69 July 1982 1

__.._..,y -- . _ , _ _ . . _

y . . , ,, , _ _ , _ , , _ . , _ . , _ . _ _ , . , _,_,m_._.,r_____m_ . , , _ , , _ ,_ , . _ . , , _,, _ _ - , , . - . _ . _ _ , _ . , _ _ - _ , - _ . . _ . _ _ _ . __

Tha lcrgest ecrthquzk:s cvsr recorded in the south:: stern Unit:d Stct:s cre the New Madrid earthquakes of December 16, 1811, January 23,1812, and February 7,1812, and the Charleston, South Carolina earthquake of August 31, 1886. The epicentral intensity at New Madrid, 300 miles west-northwest of the site, is estimated to be Xil MM and the epicentral Intensity of the Charleston j earthquake 315 miles southeast of tb sita * ? estimated to be X MM (Ref.120).

The observed surf ace intensity in the vicinity of the site f rom the New Madrid earthquakes is estimated to be V I-Vll MM (Ref s.121 and 126). Topographic changes reportedly resulted f rom these earthquakes over an area of 30,000 to 50,000 square mil es. At least a two million square-mile area was shaken (Ref.

1 29). Only a very smalI amount of damage was reported, mainly due to Iack of inhabitants. The New Madrid earthquake produced the greatest ground motion at the site of any earthquake in historic time. As previously stated, these earthquakes have been assigned an intensity Xll which is described by " total destruction" at the epicenter.

The observed surf ace intensity in the vicinity of the site f rom the Charleston earthquake is estimated to be VI MM (Ref.111). This earthquake is reported to have been felt over an area of two million square miles. For its reported epi central intensity, the Charleston earthquake was f elt over a large area.

A moderately large earthquake within the southeastern region which was felt at the site was the Giles County, Virginia, earthquake of May 31,1897, with a reported epicentral intensity of V l l-V il l MM (Ref.117). The Giles County earthquake, whose epicenter is about 220 miles northeast, is estimated to have been fel t at the site at about Intensity V MM (Ref.112).

As stated in Section 2.5.2.3, the greatest historic ground motion of the site is estimated to have been intensity VI-Vil MM and was produced by the New Madrid earthquake which occurred about 300 miles f rcrn the site.

2.5.2.6 Correlation of Eoicenters with Geoloolc Structures The tectonic structures which occur in the CRBRP site area and region have been previously described. The tectonic structures or thrust f aults within the Valley and Ridge Province are considered in the Iiterature and by recognized geologic experts as ancient and inactive. Results of the recently can'pleted Law Engineering site investigation substantiate thIs.

Amend. 72 2.5-24 Oct. 1982 1

(119) Meade, B. K. Report of the Sub-Commission on recent Crustal pg 1971 Movements in North America, ~ N.O. A. A., U.S. Dept O

V of Commerce.

(120) 1982 Report by Law Engineering Testing Company, Inc., q' q on CRBRP Earthquake Update, dated May 12, 1982.

( 1 21 ) Nutti I, O. W. Professor at Saint Louis University, Personal Communication to Law Engineering Testing Company. ,,

(122) Seed, H. B. (and Idriss, l. M.; Keifer, F. W.)

1968 Characteristics of Rock Motion During Earthquakes, Earthquake Engineering Research Center, Report No. EE-5C G8-5, College of Engineering University of California, Berkeley, Cal if ornia.

(1 23) Taber, S. Seismic Activity in the Atlantic Coastal Plain 1914 Near Charleston, South Carol ina; Bul letin, Seismological Society of America, Vol. 4, No. 3.

(124) Technical Inf ormation Division, U.S. Atomic Energy Commission 1967 Sw. mary of Current Seismic Design Practice for Nuclear Reactor Facil Itles; John A. Blume and Associates, Engineers, San Francisco, California, Tl D-25021.

(1 25) 1969 Tectonic Map of North America; U.S.G.S. and the American Associatin of Petroleum Geologists.

(126) Tennessee Valley Authority Relationships of Earthquakes and Geology in West i Tennessee and Adjacent Areas.

(1 27) 1972 Preliminary information on Clinch River Site for LMFBR Demonstration Pl ant.

(128) U.S. Coast and Geodetic Survey United States Earthquakes, 1928 - 1970.

(130) U.S. Coast and Geodetic Survey 1956 Earthquake History of the United States.

(131) Gutenberg, B. (and Richter, C. F. ) Earthquake Magnitude, 1956 intensity, Energy, and Acceleration (second paper), Bulletin Seismological Society of America, Vol . 46.

l O

2.5-61 Amend. 72 Oct. 1982

(132) leteu,0 tqoa .2.0 rnom ysoYoyMhy'ihahnglnbrs, Inc. Weston,

,.A.A.0nc f6sY5huldtts,'"SeismicVelocityandElasticModuli Measurements", 1974.

' ant aneqwo enitai va t w (133) sea ,s r p wW" Seismic Refraction Survey", Clinch River Breedereston G wm na ,p R, Reactor Plant, Oak Ridge, Tennessee,1974.

, a.n on t ea r (134) Deere, 0.U. (and Hendron, A. J.; Patton, F. D.; Cording E. J.)

1967 " Design of Surface and Near-Surface Construction in Rock", Failure and Breakage of Rock, Proceedings of the Eighth Symposium on Rock Mechanics, September 15-17, 1966, at the University of Minnesota, The American Institute of Mining, Metallurgical and Petroleum Engineers, Inc.,

New York.

(135)Leonards,G.A. Foundation Engineering, McGraw-Hill, Inc. , New York.

1962 (136)Stagg,K.G. (andZienkiewicz,O.C.)

1968 Rock Mechanics in Engineerino Practice, John Wiley and Sons, New York.

(137) Law Engineering Testing Company in conjunction with Burns & Roe Inc. November, 1975, " Report on Evaluation of Intensity of Giles County Virginia Earthquake of May 31, 1897.

(138) Neumann, Frank 1954 Earthquake Intensity and Related Ground Motion, 12 University of Washington Press, Seattle.

(139) Trifunac, M.D. (andBrady,A.G.)

1954 On the Correlation of Seismic Intensity Scales with the Pcaks of Recorded Strong Ground Motion, Bulletin of the Seismological Society of America, Volume 65, No.

1, pp.139-162.

(140) Seed,H.Bolton (and Silver, Marshall L.)

1972 Settlement of Dry Sands during Earthquakes, Journal l

of the Soil Mechanics and Foundation Division, ASCE, (141) Pyke, Robert (Seed, H. Bolton, and Chan, Clarence K.)

1975 Settlement of Sands under Multidirectional Shaking Journal of the Geotechnical Division, ASCE.

(142) Wong, Robert T. (Seed,H. Bolton, and Chan, Clarence K.)

1975 Cyclic Loading Liquefaction of Gravelly Soils, Journal of the Geotechnical Division, ASCE ,25 2.5-62 Amend. 27 Oct. 1976

i i

Criterion 10 Suppression of Reactor Power'0sc111ations p The reactor and associated coolant, control, and protection systems V shall be designed to assure that power oscillations which can result in condi-tions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

RESPONSE

I 42 The CRBRP is neutronically tightly coupled, preventing any possibility of spatial instability. The main stabilizing feedback is due to Doppler and 42l the CRBRP is inherently stable in response to reactivity perturbations.

42 l The neutronic stability of the CRBRP has been analyzed with point-kinetics techniques (See Section 4.3). The reactor was modelled by a set of coupled linearized first-order differential equations with constant coef.

ficient describing the neutronics and temperature behavior of the system.

The temperature dependent reactivity feedback effects used in this model include Doppler and fuel axial expansion which are fuel temperature dependent and the sodium density effect which is coolant temperature dependent.

These analyses inve shown that CRBRP is a stable, well-behaved system in terms of the response of the reactor to reactivity perturbations about full power. The principal stabilizing feedback mechanism is the Doppler (fuel 42 temperature) effect. The reactor remains a stable system even when the Doppler coefficient is halved and employed in any combination with the n other reactivity feedback coefficients.

o For worst case positive bowing reactivity characteristics, which (d' can occur only in the startup range (0+ to 40% power), a net positive feed-back is possible. With this condition, present control system analyses predict a worst-case (maximum) limit cycle oscillation of 12.2% of full power, comprised of a i 2% dead band plus a 0.2% response turn around on both ends of the dead band. The smallest period associated with the worst-case condition is 500 seconds in that less bowing reactivity would result in a longer oscillation period. Recompensation of the flux control system for the final design may result in a reduction in amplitude of the limit cycle oscillation as well as a reduction in the frequency. Above 40% power and under all per-mitted conditions, where bowing reactivity is always negative, limit cycle oscillation due to this feedback component will not occur. Assurance that the specified acceptable fuel design limits will not be reached is provided 42 throughout the 0 to 100% power operating range by the reactor trip functions.

Amend. 42

, Nov.1977 V 3.1-19

i l

Criterion 11 - Instrumentation and Control l Instrumentation shall be provided to monitor variables and systems over their I anticipated ranges f or normal operation, .for anticipated operational occu. rences, and f or postulated accident conditions as appropriate to assure adequate saf ety, incl uding those variables and systems that can af fect the f ission process, the Integrity of the reactor core, the reactor-coolant boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

RESPONSE

Instrumentation and controls are provided to monitor and control neutron flux, control rod position, temperatures, pressures, fl ows, and l evel s as necessary to assure that adequate plant safety can be maintained. Instrumentation is provided in the Reactor System, Heat Transport System, Steam and Power Conversion System, the Engineered Safety Features Systems, Radwaste Systems and other auxiliaries. Parameters that must be provided for operator use under normal operating and accident conditions are Indicated, in proximity l with the control s f or maintaining the Indicateo paraneter in the proper range.

I The control room is provided as the focal point f rom which the plant can be operated safely during normal operation, anticipated operational occurrences, and f or postul ated accident conditions. The basic criteria for including instrumentation readout and control in the control room is as follows:

o The displ ays or control s necessary to support alI normal plant operating conditions; o The displays and controls necessary to respond to anticipated operational occurrences and accident conditions which Impact on power operations capabil Ity; o The displays or controls necessary to prevent potential radiological hazards to of f site personnel; o The displays necessary to the operator for detection of fire hazards; or o The displ ay and control s necessary to prevent potential damage to the plant.

O 3.1 -20 Amend. 72 Oct. 1982

3 A. 8.3.3 Liner Analvsls 1, O > - ei The iIner system is described in Section 3A.8.2. The Design Requirements, j

Load Categories, Load Combinations, Stress and Strain allowables and Design Analysis procedures are given in paragraphs 3.1 through 3.5 of PSAR Appendix l

3.84. Attachment D to Appendix 3.8-B gives the bases for the strain criteria

• and strain limits adopted for the Postulated Large Liquid Metal Spill (PLLMS)

Loads.

The spacing and size of the Nelson stud anchors in the walI and celling panels and of the floor anchors are designed such that the stresses and strains f all j within the limits specified in Table 3.8-B-1 of Appendix 3.8-8.

1'

! The anchors will resist the shear forces induced when unbalanced forces exist between sections of the Iiner and axial forces caused by the maximum specifled pressure (5 psig) acting on the backside of the lIner under the PLLMS loads.

Since there is a 1/4 Inch gap between the celi Iiner and the insulating concrete, some axial loads in the anchors will be caused by the cell's

! Internal pressure.

The insulating concrete does not act Integrally with the structural concrete and a bond breaker will be provided on the surf ace separating the two material s to reduce shear transfer. The insulating concrete is not considered a main structural element; its main f unction is to provide a thermal shield to prevent degradation of the structural concrete under the elevated temperatures

! of the PLLMS conditions. The adequacy of the insulation thickness has been demonstrated by a preliminary finite element thermal analysis using the computer progran ANSYS. The temperatures calculated at the f ace of the structural concrete did not exceed the 1imits establ ished in Section 3.1.7 of Appendix 3.84. Local hotspots due to heat transfer into the structural concrete through the studs may occur. These of fects w!!I be evalusted by both i analytical and testing methods.

i SpaliIng or degradation of the insulating concrete under the PLLMS Loads wilI not cause a f ail ure of the l iners or l iner anchor system. The anchors w il l be

embedded in the structural concrete to ensure adequate restraint and the j design is such that even if no lateral support to the anchors is provided by the Insulating concrete, the specifled anchor strein Iimits w11I not be

, exceeded.

l Liner f ailure due to behind the Iiner steam pressure is prevented by the l

provision of a venting system on the backside of the l Iner where necessary, to i reduce steam pressure generated f rom heatup of the insulating material and i

structural concrete during a sodium spil l. The 5 psig cell liner vent system

pressure developed behind the cell liner plate is addressed in the analysis of i the I iner system. Two cases of pressure dif ferential across the Iiner are

! considered. The first case considers the 5 psig vent pressure behind the IIner combined wIth the peak Internal celi accident pressure; the second case

a O psig vent pressure combined with the peak internal cell accident i pressure. These two cases provide conservative bounding conditions for the pressure dif ferential across the cell liner under Design Basis Accident conditions.

3A.8-4 Amend. 72 1

Oct. 1982 2

I

)

l Specific vent paths behind the lIner will be provided where analysis and/or testing Indicates they are required. Steam produced would be vented to the non-Inerted areas l l

O l

l l

l t

3.A8-4a Anend. 72 0

Oct. 1982

l within the RCB or RSB consistent with the location of the sodium spill.

Preliminary analysis under Postulated Large Liquid Metal Spill (PLLMS) n conditions indicates that this venting scheme will not require the Q' containment to be purged. A more detailed analysis will be performed to verify these preliminary indications. Since any steam produced during a 5 91 sodium spill would be vented to the non-inerted areas, hydrogen evolution due to sodium / water reactions would occur only following a liner failure.

Failed liner testing is planned and the amount of hydrogen evolved during these tests will be monitored. Even in the unlikely event of the liner failure, purging of containment is not expected to be required.

The liner system will be designed to withstand a backside pressure of 5 psig.

Due to the magnitude of the compressive thermal forces caused by the restraining actions of the concrete structure, buckling of the liner plates is anticipated. Buckling in itself will not produce failure since the thermal deformations are self limiting. However, due to the reduced load carrying capacity of a buckled panel, unbalanced lateral forces can be induced at the anchor. The liner-anchor system will be designed such that under the unbalanced lateral forces due to panel buckling,

" the strains will not exceed the allowable limits. Buckling of panels will improve the stress-strain conditicns at the corner anchors since the unbalanced lateral forces will be reduced.

The dead and live loads, seismic loads, operating pressure and thermal loads, etc., will affect the cell liners through the interaction of the liner-anchor system with the structural concrete. Since the structural concrete is by far more rigid than the liner, the deformations p

V of the concrete under these loads and the restraint it provides to the liner will determine the stress-strain condition of the liner-anchor system for these loads. For these conditions other than sodium spills, the stress levels in the cell lir.ers are expected to be below the yield

{ strength of the material. The maximum normal operating temperature (peak) will not exceed 180 F and no significant stresses and strains will be imposed on the liners under these conditions.

Thecyclictemperaturevariatigninthgcellsduringthe lifetige of the plant (10 cycles from 70 to 140 F,100 cycles from 100 to 140 F and 100 cycles from 140 to 180 F) are within the ASME Code limitations such that the cyclic fatigue should not be a problem. Based on Section NE-3222. 4d of Section III, Division 1 of the ASME B&PV Code, for the specified temperature ranges and number of cycles, no fatigue analysis is required.

.2 Analysis Calculations have been conducted to investigate the adequacy 45!37 of the liner-anchor system under the PLLMS Condition. They consist of elasto-plastic analyses using the computer program ANSYS. The strain values obtained from the finite element analyses under sodium spill conditions are compared against the allowable strains at the exposure temperature. The allowable PLLMS strains are determined using Table 3.8-B-1 and the materials test data presented in 3A.8.4. Table 3A.8-1 summarize the allowable strains under 59 load combination D (PLLMS spill).

v Amend. 59 Dec. 1980 3A.8-5

l l

1 Following selections of the prime sealant material, prototypic electrical I cable penetration assembly performance testing were conducted. The results of this testing program were published in Reference (4).

. Base Material Tests for Liner Steels 59l The objective of this completed testing program was to determine the response of the cell liner plate material (SA-516 Grade 55) gnd its associated weldment material to elevated temperatures up to 1700 F.

The base liner steel will be tested for residual tensile strength (in-cluding stress-strain response), stress-rupture (Creep) and thermal expansion.

59l The weldment material was tested for residual tensile strength (including stress-strain response) and stress-rupture (Creep). Both longitudinal and transverse welds were investigated. The results of the base liner 59 45 steel and weldment material tests have been published in Reference 6.

The material properties information at elevated temperatures which was obtained in this program has been used in the design and analysis 59 of the cell liner system.

O O

3A.8-9a Amend. 59 Dec. -1980

References:

1. McAf ee, W.J. , Sartory, W.K., " Eval uation of the Structural Integrity of LMFBR cel l liners - Results of Prel iminary investigations", ORNL-TM-5145, January, 1976.

2. Chapman, R.H., ORNL-TM-4714, "A State of the Art Review of Equipment Cell Liners f or LMFBR's", February,1975.

3. Sartory, W.K. , McAf ee, W.J. , ORNL-TM-5145, " Eval uation of the Structural Integrity of LMFBR Equipment Celi Liner - Results of Preliminary investigation", February,1976.

4. Humphrey, L.H., Horton, P.H. , Al-D0E-13227 "Sel ection of a Sodi um and Radiation Resistant Sealant for LMFBR Equipment Cell Penetrations",

January 31, 1978.

5. Wireman, R. , Simmons, L. , Muhl estei n, I., HEDL TE 79-35, "Large Scal e l Liner Sodium Spil l Test (LT-1)", December, 1980.

6. Cowgil l, M.G., WARD-0-0252, " Base Material Tests f or Cel l Liner Steel s",

February, 1980.

O 3A.8-9B Amend. 72 O

Oct. 1982

4.?.3.1 Design Basis O

b 4.2.3.1.1 General Safety Design Criteria The General Safety Design Criteria are discussed in detail in Section 3.1.3, subsection lil, Protection and Reactivity Control Systems, and are outlined here f or completeness. Specific criteria which are a part of the design basis for the reactivity control systems mechanical components are:

1. Criterion 20 - Protection System Independence

2. Oriterion 21 - Protection System Faiiure Modes

3. Criterion 23 - Protection System Requirements for Reactivity Control Malfunctions

4. Criterion 24 - Reactivity Controf System Redundancy and CapabilIty

5. Criterion 25 - Combined Reactivity Control Systems Capability These criteria are augmented by the following requirements:

1. The speed of respons.,e of the control rod system, acting as part of the Plant Prctectlon System, shalI be suf fIcient to assure that the Damage Severity Limits of Table 4.2-35 are not exceeded. Specific requirements l for speed of response are presented in Section 4.2.3.1.3.

The allowable damage limits are related to the frequency of the transient condition so that anticipated events do not lead to a reduction in the ef fective f uel lifetime. RDT Standard C16-1T, Dec.1969, is used as the basis for the primary control rod system damage severity limits, without a stuck rod. (See Tabl e 4.2-35.) To provide conservative plant protection, the same primary system damage limits are required to be satisfied under the assumption of a stuck rod. The primary system has the f unction of I imiting f uel damage to design I imits for anticipated events. Failure of the primary system is an extremely uniIkely event. Consequently, the secondary system, which is needed for shutdown only if the primary system f alls, need only limit damage to the major Incident Iimit of an extremely unl ikely f ault. For additional conservatism in ilmiting plant damage for an anticipated event, the lImits of Table 4.2-35 require only minor incident damage for the secondary system. The combined probability of an extremely unlikely f ault event concurrent with f ailure of the primary sysicm is exceedingly low and is not appiled as a design basis.

O 4.2-230 Amend. 72 Oct. 1982

2. For an Operational Basis Earthquake (0BE), an anticipated fault, both control rod systems shall be capable of functioning, including reactor scram, both during and after the earthquake.

Reactor shutdown shall be achieved assuming loss of offsite 51 l power and/or a step reactivity insertion (maximum of 304) coincident with the earthquake (concurrent events defined as an unlikely fault) without exceeding the damage severity limits of a minor incident for primary system shutdown and of a major incident for secondary system shutdown.

For a Safe Shutdown Earthquake (SSE) (extremely unlikely fault),

either control rod system shall be capable of shutting down the reactor during the earthquake but is not required to function after the earthquake other than passively assuring that shutdown is maintained. Reactor shutdown shall be achieved assuming concurrent loss of offsite power and a step reactivity insertion 51 (maximum of 604) coincident with the earthquake without exceeding the damage severity limits of a major incident for both systems.

The requirement for functioning of the secondary system in an SSE provides an additional protective margin beyond that of Table 4.2-35.

3. No electric or other external (to the mechanical control rod system) power shall be required for a scram of any control rod.

4.2.3.1.2 Control Rod System Clearances ,

The specific goal in establishing control rod system clearances is to ensure safe and reliable shutdown and control capability for the reactor. To this end, the basis for establishing clearances fall into the following general categories:

Limit scram retarding forces resulting from misalignment of components.

Limit scram retarding forces resulting from material effects from thermal, radiation, and other environmental characteristics.

Assure normal operation of the control rod systems under misaligned and environmental conditions.

Control Systems clearance requirements and their bases are summarized in Table 4.2-36.

4.2.3.1.3 Mechanical Insertion Requirements This section describes mechanical insertion requirements with regard to scram speed of response, alignment requirements, scram arrest, normal insertion and withdrawal speeds, and coefficient of friction considerations.

4.2-231 Amend. 51 Sept. 1979

4.2.3.3.1.3 Scram Analyst s This section describes the scram analyses performed for the primary control

' rod system to demonstrate the expected rates of reactivity insertion during a reactor scram. Considered in this section are avail able shutdown reactivities, typical rod positions, control rod scram speeds and scram reactivity insertion rates.

Tvolcal Rod Withdrawal Positions Rod positions at the time of the scram may very significantly due to:

withdrawal over the f uel cycle, potential variations in rod bank positions, uncertainties in rod worths and variations in the f uel cycle length between the first and l ater cores. This time to insert the first dollar of shutdown reactivity in the reactor scram is typically of greatest importance as this fIrst dolI ar is suf fIcient to turn around the power peak or f uel temperature increase for most transients. Tabl e 4.2-44 shows typical rod withdrawal positions over the first five operating cycl es. B00-5 has been shown to be l the worst case for the slowest first dollar Insertion and is therefore the basis for the scram insertion analysis.

Control Rod Scram Soeed Control rod insertion speeds are calcul ated by the CRAB-ll computer code which solves the equations of motion considering all the forces acting on the PCRS transl ating assembly, both scram assisting and scram retarding. Section 4.2.3.3.1.1 presents the analysis of the scram retarding forces, and Table 4.2-43 gives the total drag force as a f unction of withdrawal .

Validation of the CRAB-Il code for predicting speed of insertion was done using test data from the PCRS system tests. Figure 4.2-112 shows insertion profiles from various withdrawal positions based on the drag forces given in Tabl e 4.2-43. Figure 4.2-113 demonstrates the abil ity of CRAB-ll to predict actual test data using conditions expected to occur in the core. Also shown in Figure 4.2-113 is the CRAB-ll predicted speed of insertion using the conservative design conditions described in Section 4.2.3.3.1.1. The di f f erence between the two curves represents the conservative margin for speed of insertion used in the scram analysis. Figure 4.2-113a demonstrates the abil Ity of CRAB-li to predict test data over a range of flow rates and insertion times.

f; ram Reactivity insertion Rates Scram reactivity insertion rates have been calculated based on the di spl acement/ time prof il es given in Figure 4.2-112, the cycl e dependent rod positions of Tabl e 4.2-44 and the min? mum and expected benk worths appropriate to each cycl e with the singl e most reactive rod stuck. The resul ts of these cal cul ations are shown in Figure 4.2-114. Although BOC-5 procedures the minimum shutdown, BOC-4 has been inct uded to show the change in scram insertion f rom BOC-5 to BOC-4. All other cases Insert reactivity f aster due to higher worths or f arther initial rod withdrawal .

An evaluation of the inherent shutdown margin can be obtained by comparing the O minimum reactivity insertion with the expected reactivity insertion. The V 4.2-278 Amend. 72 Oct. 1982

minimum reactivity insertion represents a 3d worst case eval uation of maximum excess reactivity and minimun control rod worth, while the reactivity insertion represents nominal core conditions. Additional margin on reactivity insertion for both minimum and expected conditions is included in these curves by using speed of insertion calculated with the conservative design conditions  ;

of drag shown in Figure 4.2-113. l End of cycl e reactivity insertions are signifIcantly greater due to increased ,

shutdown margins and f aster rod speeds due to greater scram assist forces at '

these positions.

AlI curves in Figure 4.2-114 have assumed a del ay of 0.1 seconds f rom the advent of a scram signal tothe start of rod motion. Actual test data from the PCRS system tests has shown this uni atch time to be 0.0486 i 0.0002 seconds.

Thus, on a 36 basis, the scrm insertion curves in Figure 4.2-114 coul d be moved to the I of t by 0.05 seconds.

It is therefore concluded that the primary control rod system satisfies the speed of response requirments given in Section 4.2.3.1.3 for all conditions.

O 4.2-278a Amend. 72 O

Oct. 1982 l

4.2.3.3.1.4 seismic scram Analvsts*

An analysis was performed to determine the ef fect of a saf e shutdown earthquake (SSE) on the CRBRP Primary Control System's scram capability.

Lateral contact forces on the translating assembly were determined for a severe three second segment of the SSE which was then used in eval untion of seram performance under seismic conditlons.

The worst time to Initiate a scram in this 3 sec. time Interval was identified by determining the time required to scram 9 inches. This criterion was used because it represents the required rod travel of the rods to insert approximately one dolIar of reactivity. A 1.2 second Ioad time history whose initial point is the worst scram initiation time was then used repetitively until the rods were fully inserted. A dynamic Impact coefficient of friction of 0.5 was used since this val ue is conservative rel ative to the coef ficient of friction averaged over the i ength of the PCRS (see paragraph 4.2.3.1.3).

The MISYS computer program was used to perform the seismic analysis, using the semi-linear transient dynamic (time history) option of the program. An overall reactor system model was first used to determine the motions of the important components. The gross motions of the system components were then used as input functions in a decoupled primary control rod system model to determine the response of the leadscrew, drivel Ine and control assembly within the PCRDM, shroud tube and control assembiy duct.

The noniinear primary control rod system model and its use in the seismic impact analysis are discussed in Section 3.7.3.15.3. The results of this O analysis used in the scram calcul ations are the contact forces (vs. time) during the seismic event.

• See footnote to Section 3.7.3.15.

O 4.2-279 Amend. 72 Oct. 1982

1 l

The scram analysis was perf ormed using the CRAB computer code (See App. A) l Incorporating the dashpot model and time variant scram retarding force j capability. The results of the SSE scram insertion predictions are compared  ;

with the seismic scram requirements in Figure 4.2-119. BOC-5 was determined to be the time in lIfe which produced the minimum reactivity insertion due to bank position and avail abl e worth. An evaluation of the inherent shutdown i margin can be obtained by comparing the minimum reactivity insertion with the expected reactivih Insertion. The minimum reactivity insertion represents a 36 worst case evaluation of maximum excess reactivity and minimum control rod worth, while the expected reactivity insertion represents nominal core conditions.

It is concluded that the primary control system satisfies the SSE scram insertion requirement of Figure 4.2-93. The reactivity effects of the siIghtly increased scram time are evaluated in Section 15.2.3.3.

The seismic scran analysis is a conservative evaluation of scram capability under SSE environment in that a conservative calculation of loads and scran initiation time was employed.

4.2.3.3.1.5 Control Assembiv Analvses Absorber Pin The primgy control assembly utilizes enriched 8 C3 (approximately 92 atom percent B in Boron). Data on hellum release, thermal conductivity and pelIet swellIng, required for absorber design, are avalIable in References 44 and 44a.

Currently committed 3B C tests providing EBR-Il Irradiation data is support of O CRBRP control assembi? design are given in Table 4.2-46A. The table summarizes each test using the HEDL name for the test. Typical test paraneters for pellet temperature, pellet diameter and B-10 captures completion dates for the EBR-ll Irradiations.

The tests of Table 4.2-46A will extend the irradiation data well above the pellet temperature and pellet sizes anticipated for tbj primary control assembly. The BICM-1 test has provided data to 80X10 B-10 captures /cc, which is comparable of first core burnups for CRBRP. The BV-2 test for vented pins will provide data on pellet swelling for burnups typical of 275 FPD cycle operation. The tests of Table 4.2-46A cover the operating range for the primary control assembly over its required iIfetime.

The planned EBR-Il4B C Irradiation tests do not include in-reactor transient cycl ing of absorber rods. Out-of-pile testing of irradiated pellets has been performed under the HEDL development program to determine gas release under transient thermal conditions. Preliminary results indicate that helium release upon temperature increases occurs over a relatively long time (on the order of 15 minutes) characteristic of a Primary Control Assembly thermal transgent. Since Bg C temperature increases during transients are smalI

(<100 F) the incremental gas increase from a transient is a small ef fect.

incremental gas release 4.2-280 O

Amend. 72 Oct. 1982

O during transients based on the thermal +.ransient tests are included in the pin lifetime analyses. Since the absoroer pins are designed to preclude pellet to clad interactions or B C3 melting under worst case transient conditions, gas release is the oMly B 4C variable required to be assessed in transient analyses.

Further performance data for the PCA will be obtained from the 53l PCA Irradiation Test (see Section 4.2.3.4.1.1) which will provide integrated lifetime performance data for near prototypic environments and operating 51 parameters.

511 Table 4.2-46 sumarizes performance parameters for the absorber pins. The thermal-hydraulic pararraters are discussed in Section 4.4.

For the current design, the plenun. lengths have been established by the maximum available pin length, and the clad stresses at the end of one operating cycle are less than 5,000 psi as shown in Table 4.2-46.

Preliminary strain analyses of the pin have indicated that there is only minimal accumulated strain at the end of the lifetime requirements.

Additional analysis utilizing the cumulative damage function approach 51 has been performed which also verifies the lifetime capability of the pins. Use of the CDF for the absorber cladding requires that the duty cycle be separated into various stress state / time segments superimposed on the steady state operating conditions. This introduces conservatism O in the analysis since conservative estimates of stress and time form the basis for the analysis. Effects such as sodium interaction with the cladding and pin-duct interactions are included in the lifetime evaluation.

B4 C swelling is calculated to assure that no force contact occurs between the pellets and the cladding (see Table 4.2-36) thus reducing the margin for error in the calculations. Figure 4.2-llla shows pellet swelling and associated pellet to clad gap for rod in the Row 7 corner location. Figure 4.2-111b shows axial B-10 burnup profiles for each rod position in the equilibrium cycle.

Based on the results of the preliminary analyses performed, it i is concluded that the pellet / clad gap clearance requirements are satisfied '

for the required 328 FPD lifet.ime with an initial gap of 0.028 inches (Figure 4.2-llla). The initial gap must be increased to allow for 54 53 51 additional pellet swelling over the goal lifetime of 550 FPD.

Structural Evaluation j A preliminary elastic analysis was performed to evaluate the

! structural adequacy of the control assembly outer ducts. Design stress 511 limits were derived using the criteria defined in Table 4.2-37B. Both l

i Amend. 54 flay 1980 4.2-281

4.2.3.4.1 Performance Test Procram V) Extensive testing programs are planned for evaluation of the reliability and design of both reactivity control systems. These tests will incl ude individual component tests and complete prototype systems tests.

4.2.3.4.1.1 Primarv Control Rod System The PCRS testing program consists of the following major testing activities:

A) C =nonent Tests: The following component design test and/or analysis program was established to provide design verification of the PCRS components.

1. Dvnamic Seismic Friction Test This test was performed to evaluate the ef fective coef ficient of friction between a rod and its guide bushings under impact loading conditions. Data obtained are used to provide friction coef fIclents for seismic scram insertion analyses.

2. Control Assembly HvdraulIc (Flow) Test Test results wIlI be used to verify the pressure drops, flow and vibration characteristics of the primary control assembly design under prototypic flow conditions.

3. Control Assembiv Pin Cmnaction Test Test has provided data to determine inter-pin and pin-to-duct loads for the primary controf assembiy analyses.

4. Control Assembly Rotational Joint Test Test has provided performance data on the rotational joint which confirmed the reduction in control assembly wear and rellable operation of the joint.

5. B4 C Data Test The base technology irradiation test program being conducted by HEDL Includes acquisition of data required for design verification of CRBR control assembi les (see Table 4.2-46A).

6. Friction and Wear Tests l The base technology materials test program being conducted at ETEC and ARD provides data for the material couples selected for fabrication of the primary control rod system.

i O Amend. 72 4.2-304 Oct. 1982 I

7. Control Assembiv Analvtical Methods Provides an analytical model calibrated with test results for predicting primary control assembly thermal-hydraul ics perf ormance, lifetime characteristics and scram dynamics behavior.

B) System level Tests: A series of Primary Control Rod System Prototype Tests have been perf ormed to verify that the Primary Control Rod System perf ormance is consistent with its design requirements under design basis operating conditions. The Control Rod Drive Mechanism was eval uated in a CRDM Accelerated Unlatching Life Test. This test program verified the unlatch perf ormance characteristics of a prototype primary control rod drive mechanism over twice the design lifetime travel and scrams. The Accelerated Lif e Test invol ved testing of a f ul l size prototype primary control rod system in sodium, sodium vapor, and argon gas environments that simulate operations in the Clinch River Breeder Reactor Plant. Phase I testing in this series canpleted 1/2 of the PCRS lifetime scrams,1/3 of the leadscrew travel requirement, and about 5 times the PCA travel req uirements. Phase ll of this series will extend total test scrams and travel beyond 01BRP l ifetime requirements.

During System Level Tests of the Primary Control Rod System, each subsystem was al so tested, including the position Indication system and the dash pot. Four prototype systems were tested and the results show PCRS perf ormance incl uding position Indicator accuracy and dash pot

! perf ormance were w ithin th is design requirements.

1 C) PCA Irradiation Test: A PCA Irradiation test is scheduled to be inserted in the FFTF f or 600 FPD. The intent of this test is to provide near-prototypic Irradiation perf ormance data on the PCA absorber assembly to support the PCA l ifetime eval uations. The test assembly will contain 37 pins of enriched Bg C and w fil function as an integral part of the FFTF Secondary Control Assembly Bank. The parameters of the test assembly have been selected to provide data prototypic of the PCA f or burnup, fluence, Bg C and cladding temperatures and cladding strain. Data f rom this test aFe expected to be avail able in 1986.

D) Other Tests: See Appendix C f or Rel iabil ity Test Program.

4.2.3.4.1.2 Secondarv Control Rod System The SCRS testing program consists of the following major testing activities:

A) Latch Tests: Component development tests of the scram latch conf iguration f or the secondary control rod system verified the design of this com ponent.

B) Damoer Tests: Component development tests of the danper conf iguration f or the secondary control rod system verified the design of this component.

I Amead. 72 O

4.2-305 Oct. 1982

TABLE 4.2-44 PCRS CYCLE DEPENDENT W ITHDRM AL POSITIONS (IN INCHES)

R0f 4 R0( 7 CYCLE Minimum Expected BOC & EOC BOC EOC BOC EOC 1 36 16.0 20.9 18.9 24.2 2 36 15.1 22.4 18.1 26.9 3 36 12.9 23.3 16.2 27.7 4 36 12.5 21.1 16.6 26.9

'I 5 36 12.6 15.9

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4.2-41 0 Amend. 72 Oct. 1982 i

O' TABLE 4.2-45 DELETED 51 O

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Sept. 1979 l 4.2-411 .

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0.0 1.0 2.0 3.0 TIME (SECONDS) l Figure 4.1-112 Primary Control Rod System Scram insertion I)istance vs Time 1764-57 4.2-614 Amend. 51 Sept. 1979

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Figure 4.2-113. A Comparison between Scram Predictions by CRAB-II and ARD Sodium Test Data 7089-1 Amend. 72 4.2-615 Oct. 1982 l

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,,3 0.3 0.4 0.5 0.0 0.7 0.8 CRAS II, PREDICTED (SEC.)

Figure 4.2113a. Summary Comparison of the CRAB-Il Predicted Versus Test Observed Scram Time to Reach the Dashpot

7081-2 Amend. 72 O 4.2-615a Oct. 1982

I O O O i $

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(next page is 4.2-620) Jan. 1980

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5.2.1.5 Reactor Vessel Preheat p The Reactor Vessel Preheat System wilI control the dry heat-up and cool down of the Guard Vessel, Reactor Vessel and internals between ambient (700F) and 4000F and if required will provide make-up heat for that lost to the Reactor Cavity during prolonged shutdowns.

The heat will be provided by tubular electrical heaters mounted between the Guard Vessel and insulation. These heaters will be arranged circumferentially around the Guard Vessel and will be grouped and controlled in zones of uniform heat output. Temperature sensing devices will monitor the Guard Vessel temperature in each of these zones and provide the necessary feedback for power level adjustments in the heaters.

The heaters will be mounted to the same framework which supports the Guard Vessel Insulation. Attachment clips will of fset the heaters from the Guard Vessel surface. Convective barriers, reflective sheaths and Guard Vessel insulation will be used to optimize heat input to the Guard Vessel and minimize losses to the Reactor Cavity.

Preliminary preheat, startup, shutdown analyses have been performed on the Reactor Vessel and Guard Vessel to determine the temperature dif ferences which will result in opening and/or closure of the annular gap between the two vessels. By necessity the preheat analysis is very preliminary since no firm preheat procedure has yet been developed. Figures 5.2-4 through 5.2-6 show the temperature dif ferences between the Reactor Vessel and Guard Vessel in the inlet and outlet plenum regions for the three transients in question. As shown the largest positive temperature difference between the Reactor Vessel and the Guard Vessel occurs in the outlet' plenum region during startup (3350F)

,) while the largest negative temperature difference occurs in the outlet plenum region during shutdown (-2140F). The nominal radial gap between the reactor vessel and guard vessel is 8 inches at assembly and at the end of preheat.

This gap decreases to approximately 7.6 inches minimum during start-up and increses to approximately 8.3 inches maximum during shutdown. During preheat the gap also increases but to a lesser value than during shutdown due to the smaller maximum temperature difference.

Variations in the axial gap between the bottom of the reactor vessel and the Inner surf ace of the guard vessel are noted between the states shown in the table. Thus the largest axial gap is 11.0 inches at the dry cold condition and the smallest gap is 6.2 inches at the end of the hea-ling phase of preheat.

5.2.2 Design Parameters Overall schematic views of the reactor vessel, closure head assembly, Inlet and outlet piping, and guard vessel are shown in Figures 5.2-1,1 A and 18.

The top view is given in Figure 5.2-2.

5.2-4c Amend. 72 Oct. 1982

I O

5.2.2.1 Reactor Vessel and Support The reactor vessel and support will be constructed mainly of austenitic stainless and low alloy steels, and consists of six basic sections: the support ring, the vessel flange, the barrel, the core support forging and cone, the inlet plenura, and the vessel thennal liner.

59lThe support ring is an SA 508 Class 2 ste'el forgin welded to the vessel 17 flange. A box ring type of reactor vessel support interfaces with the vessel support ring and the reactor cavity support ledge. Holddown bolts pass through holes in the vessel support ring, the reactor vessel support and the support ledge clamping the three together. The vessel support is a ring structure with a box type cross section. The vertical sides of the box are Inconel 600 to limit the heat flow from the reactor vessel. The top and bottom plates of the box cross-section are 59l58 SA 543 Class 2. The bolts are SA 193 Type B7 with 3.50-8UN threads. The ring supports the reactor vessel and internals and closure head. The vessel flange is a second SA 508 Class 2 steel ring forging welded to an Inconel 600 transition section. The latter is, in turn, welded to the barrel.

Radiation shielding in the form of a boron carbide collar surrounding the vessel near the flange is provided in the annulus between the reactor vessel and the vessel support ledge. The barrel comprises the upper cylindrical portion of the vessel and has an inside diameter of 243 in.

with a minimum wall thickness of 2.38 in. The lower end of the barrel 17 is joined to the core support forging and cone, which provide support for the core support structure. The overall height of the reactor vessel and support is nominally 704 g in. (58 ft. 8 in.). Tge inlet plenum is designed for 200 psig at 775 F and -15 psig at 600 F, the stainless steel 11 portiog of the outlet plenum gis designed for 15 psig plus head of sodium 17 at 900 F and -15 psig at 600 F.

Cgolant enters the reactor vessel through three 24-inch nozzles located 120 apart in the inlet plenum below the core support structure.

Core effluent and bypass flow are mixed in the outlet plenum region above the core, and the Amend. 59 Dec. 1980

The riser has been designed to maintain a maximum temperaturo of 1250F in the region of the elastomer seal s. Thermal analysis has been completed for this design which shows that ihls temperature (1250F) is maintained by the head O- access area cool Ing system.

5.2.2.3 Guard vessel The guard vessel is a bottom-supported, right circular cylindrical vessel l surrounding the reactor vessel. It was f abricated from SA240 Type 304 stainless steel . The purpose of the guard vessel is to assure outlet nozzle submergence in the event of a leak in thu reactor vessel nozzles, piping, or i

O O 5.2-6b Amend. 72 Oct. 1982

I piping connections. To f ul fill this requirem:nt ths guard v:ssel extcnds to appe oximately 6 f t. above the minimum safe sodium level providing fcr sodium shrinkage and pumping head dif ferential. Also, the guard vessel permits p inservice inspection of the reactor yessei by providing a nominal clearance of V 8 in, between the two. The guard vessel is insulated on the outer surf ace to Iimit the heat load into the reactor cavity cell and to reduce heat loads to the Reactor Cavity Heating and Ventilating System. A trace heating system is mounted on the outside surf ace of the vessel for pre-heating and heating durIng prolonged plant shutdown.

Fl ux monitors f or low, intermediate and f ull power operation are provided in the annulus between the guard vessel and reactor cavity cell walls. This annul us wIl I be f Il Ied wIth nitrogen gas Q2% oxygen by vol ume). A discussion of the reactor cavity cell is f ound in Section 3.8 and 3. A.1.

Continuity detectors and aerosol sampiIng iines are mounted inside the guard vessel to detect potential leaks in the reactor vessel or inlet or outlet piping. See Section 7.5.5.1.

5.2.3 Soecial Processes for Fabrication and Insoection 5.2.3.1 Nondestructive Examination Nondestructive examination of material s and wel ds wil I be perf ormed in accordance with the ASE Boller and Pressure Vessel Code and RDT Standards.

The techniques employed, as appropriate for the respective product forms, material s, and wel d conf igurations comprising the reactor vessel, closure head, and guard vessel, are l Iquid penetrant, magnetic particle, ultrasonics, and radiography. Surf ace f inish and cl eani iness w Il I al so meet al l requirements of the ASE Code and the other contract documents. Periodic swab O)

( tests of stainless steel surf aces during f abrication in the shop will be performed to assure that potentially harmful substances such as chlorides do not contact the components in concentrations greater than specifled in appi Icable codes and standards.

5.2.3.2 Controlled Weldino to Maintain Alignments Specifled alignments must be maintained between the core support structure and the upper end of the reactor vessel. Where welds such as girth seams in the vessel and the weld attaching the core support structure to the vessel infl uence these al ignments, special welding procedures and processes util Iz ing proven technology will be used to control the relative alignments of the parts

being joined by welding.

Prior to any welding, the core support structure will be aligned by equalizing the gap begeen the core barrel and thermal lIner support ring at four points located 90 apart. Al so, the weld preps on the core support structure and core support cone wIlI be aligned vertically and radially using the respective wel d l ands as the ref erence surf ace. Four wedges, which have been contour machined to match hal f the weld geometry, are placed in the top of the joint.

Their purpose is to prevent movement during initial welding.

Wel ding w il I be accompi Ished by using f our welders positioned 90 apart.

Movement of the core support structure during welding will be monitored by

) 5.2-7 v Amend. 58 ,

Nov. 1980 l

me:suring the distances bet.:een the core barrel and the thermal Iiner support ring at f our equal ly spaced I ocations. If 1/16 inch or more distortion occurs, welding w11I stop on one side and continue on Tne opposite side until re-alignment occurs. Continuous monitoring w il I be perf ormed untII i/2 inch of weld has been deposited. At that point, the wedges w ilI be removed and periodic monitoring w ill be performed during the rcrnainder of the wel ding.

Af ter wel ding is complete, the weld prep f or the top portion of the vessel at the top of the thermal liner support ring will be machined concentric to the centerl ine of the core support structure. The top portion of the vessel is f abricated so that the bottcrn wel d prep is concentric to the vessel f l ange.

The top portion of the vessel is then assembled to the lower assembly using a ship-lap joint (sometimes known as a spigot f it). With this joint, no special welding techniques are required to maintain al ignment, it is purely manual metal-arc wel ding. By having precise al ignment w ith in the two sub-assembl ies, that is, centerl ine to wel d prep and by using a precision f it-up of these sub-assembi les, the core support structure is located to vessel flange within the required tol erance.

The wel d circumf erence was div ided into f our quadrants, each of which was div ided f urther into 12-inch incrcrnents. The f irst wel d pass was made using f our wel ders working simul taneously, one wel der per quadrant. The position of the core support structure then was measured. If a signif icant movement was found to occur, it was corrected by wel ding 12-inch increments which were sel ected by the wel ding engineer. The subsequent passes were wel ded and corrections made as necessary. This was repeated until movanent of the core support structure ceased.

The sel ective pl acement of wel d passes to control distortion during welding does not resul t in local ization of overl aps or start-stops. The wel d overl aps or start stops are no dif ferent f ran those encountered in normal arc wel ding.

Sensitization is controlled, as it is in other shop f abrication and f ield wel ds, by Iimitin0 the interpass temperature to 350 F maximum per RDT Standard E15-2-fD, which is imposed by appropriate equipment specif Ications.

5.2.3.3 Dimensional Checks All dimensions of the reactor vessel, closure head, and guard vessel will be measured and checked against the dimensions and tolerances specif led on the manufactorIng drawings. Any devlations w il I be documented by Suppl ler 5.2-7a Amend. 41 Oct. 1977

,- 3 Nonconformance Reports. Approval of dimensions not in accordance wIth the drawings will be granted only af ter determining that safety and operabil Ity of (V) the plant will not be af fected adversely. Deviations which do not meet the requirements of the ASE Code wilI not be permitted.

5.2.3.4 ASME Code Pressure Tests Pressure tests will be performed on the completed reactor vessel and on the compieted closure head as required by the ASE Code.

The high-pressure inlet plenum portion of the reactor vessel has been pressure tested to a pressure of 250 psig. This pressure test took place after the installation of the core support structure. The pressure test al so provided structural verif ication of the core support structure, although not required by the ASE Code. Following the pressure test of the inlet plenum, the entire vessel was pneumatically tested; during this test, the upper end of the vessel was sealed by a test head.

The closure head was pneumatically tested to a pressure of 18 psig. A suitable test f ixture was used to retain the head and apply the test pressure to it.

5.2,4 Features for imoroved Reliability 5.2.4.1 Reactor Vessel Thermal Liner and Nozzle Liners in order to protect the pressure boundary of the vessel in the outlet plenum region f rom high temperatures and severe temperature gradients during steady-(o) b state and transient conditions, the reactor vessel is provided with a thermal I iner that extends downward f rom above the sodium pool level to an elevation below the core support horizontal baf fle. Nozzle l iners for th is purpose al so are provided for the three outlet nozzles and for the makeup nozzle.

5.2.4.2 Internal Elbows In Reactor Vessel inlet Plenum in order to promote mixing of the three inlet streams in the reactor vessel inlet plenum and minimize thermal gradients in the pressure boundary of the inlet plenum, each inlet nozzle is provided with an internal pipe elbow that de f l ect s th e f l ow dow nw ard a n d aw ay f rom th e w al l . In this manner, the mixing of the entering sodium occurs in the Interior of the plenum, providing coolant uniform temperature to core components.

O

(.J 5.2-8 Amend. 72 Oct. 1982

1

'J 25 58 The method of obtaining data representative of irradiated permanent component materials consist of 1) selection of coupons of component mat-erials used at locations where it is predicted that detectable change will occur, 2) fabrication of test specimens from the coupons, 3) irradia-tion of the specimens in the reactor in environments which will provide advanced data and 4) withdrawal of the specimens at planned intervals during the plant life and testing of the specimens at component anticipated service temperature. Details of the coupon /speciment selection, irradiation and testing requirements are as follows: 19 Coupon Selection Requirements l

1. The materials of the permanent Reactor System components, which am designed for the full life of the plant, shall be considered for representation by surveillance coupons.

2. Materials surveillance coupons, to monitor radiation effects (9 in the materials of pennanent reactor system components, sh l U be requimd if the predicted fluence is greater than 1 x 10{ c n/cm2, E > 0.0, in the component material .

3. Subject to requirements 1 and 2, base metal and weld metal coupons shall be required.

4. For each location defined by the application of requirements 1 through 3, sufficient material shall be obtained, during fabri-cation, to produce coupons from which 15 test specimens shall be fabricated.

5. The test specimens shall be sub-size tension specimens as indicated in ASTM E-8, having a gage diameter of 1/4 inch and a gage length of 1 inch and an overall length of 2 5/8 inch.

Test Specimen Irradiation Requirements

1. Surveillance test specinens shall be irradiated in the Removable Radial Shields and/or the Fuel Transfer and Storage Assembly as requimd to obtain environmental conditions as noted below.

2. Three test specimens of each component material, defined by the coupon selection requirenents, shall be placed in a capsule set.

(A capsule set shall be one or more individual capsules as required A to obtain environmental conditions as noted below). Four capsule V sets shall be assembled.

Amend. 58

5. 2-10 b Nov. 1980

3. The four capsule sets shall be in place in the reactor at startup. One set shalI be withdrawn at each of 1/4,1/2 and 3/4 of plant iIfe (to the nearest normal ref uel ing interval). The fourth set shall remain in the reactor as a contingency set.

4. Positioning of the capsule sets and the distribution of test specimens within the capsules shall be such that the minimum anticipated fluence on the test specimens shalI be as f olIows:

o Test Speclecn to be withdrawn at 1/4 of pl ant l if e shall have anticipated total fluence at least equal to the anticipated total fl uence on the component material at 1/2 pl ant I if e.

o Test Specimens to be withdrawn at 1/2 of plant life shall have anticipated total fluence at least equal to the anticipated total fl uence on the component material at 3/4 pl ant I if e, o Test Specimens to be withdrawn at 3/4 of plant iIf e shall have anticipated total fluence at least equal to the anticipated total fl uence on the component material at f ul I pl ant i If e.

o irradiation of the contingency test specimens shall essentially dupl icate Irradiation of the test specimens schedul ed f or w ithdrawal at 3/4 of pl ant i If e. ,

5. The test specimens shall be positioned so the anticipated total flux shall not exceed three times the anticipated total fl ux on the component mater ial .

6. The test specimens shalI be positioned to best simulate other component material service conditions af ter fluence criteria are met.

Test Soecimen Testino Reautrements

1. Three test specimens of each component material, def ined by the coupon sel ection requirements, shall be tested in the unirradiated condiflon to provide reference data.

2. Irradiated specimens shall be tested af ter renoval from the reactor according to the schedule def !ned by the irradiation requirements.

-5

3. Specimens shall be tested at a strain rate of 3 x 10 in/In/ sec and at the anticipated service temperature of the component material.

4. Testing procedures shal l incl ude the use of extensometers and other devices to produce a record of load and elongation data.

5.2.4.5.2 In-Service insoection in-service inspection (ISI) equipment is provided to perf orm a visual examination of the outer surf ace of the welds on the reactor vessel and O

5.2-10c Amend. 58 Nov. 1980

nozzles, and the Inner surf ace of the welds of the reactor guard vessel.

These examinations are to be performed during those periods when reactor p coolant temperature is approximately 4000F. The ISI equipment for the reactor

'N vessel / guard vessel annul us consists of a TV camera, transporter, and cabl Ing to provide for cooling and appropriate electrical interfaces.

The overall sensitivity of the TV camera will be such that accumulations of IIquids, liquid streams, liquid drops and smoke are discernible. The TV examination will also be capable of determining the presence of loose parts and debris.

The reactor vessel, guard vessel, guard vessel extension, and support lodge insulation form an assembly designed to provide transporter access to all reactor vessel welds, excepting portions of three reactor vessel longitudinal welds masked by the reactor cavity radiological shield and the reactor guard vessel longitudinal welds covered by the leak detector tubes mounted to the guard vessel. The transporter will be similar to the transporter design empi oyed on FFTF.

5.2.5 Oualltv Assurance Surveillance Qual ity assurance survelilance for the reactor vessel and reactor vessel guard l vessel has been performed by quality assurance personnel who were present at the f abricator's f acil ity during all important phases. Qual ity assurance lpersonnelhavemonitoredall important phases of fabrication for the closure head. The Interf aces between the various QA organizations are given in Chapter 17.0, O

A 5.2-10d Amend. 72 Oct. 1982

5.2.6 Materia!s and insoections The materials used in f abricating the reactor vessel, closure head and guard vessel are summarized in Table 5.2-3. In general these materials (for the reactor vessel and closure head) conform to ASE Boller and Pressure Vessel Code, Section ill, and the supplanental requirements of RDT Standard E15-2tB-T. The materials for the guard vessel conform to ASE Boller and Pressure Vessel Code, Section il1, requirements for Class i Vessel.

Requirements for delta ferrite content are given in the ASE Code, Section 111, ASE Code Case 1592, RDT Standards M1-1T and M1-2T, and the reaciror vessel specification, in all cases, the determination of delta ferrite content will be made f rom chemical analyses of welding materials as appl ied to the Shaef fler Diagram in the ASE Code. There is no requirement that delta f errite determinations be made f rom production wel ds.

The service environment and temperature wilI not result in matertal degradation ef fects in the combination of dissimilar metals and weldments utilized between the Type 304 stainless steel vessel and the SA 508 class 2 ferrite vessel f l ange.

The transition region of the vessel is located in a position which has a total fI uence of Iess than 1015 n/cm2 (E>0.0 MeV). Th I s I ow fI uence I evel Is considered to be below the threshold level for mechanical property degradation of the material s invol ved (Ref.1).

The service temperature for the SA 508 to inconel 600 weld is approximately 4650F and that for the weldment between the inconel 600 and Type 304 stainless steel is about 6500F. At these operating temperatures, the three base metals invoived together with the inconel 82 wel d f IlIer metal are metalIurgIcally stabl e. (Ref. 2 and 3).

Both the internal and external environment are considered benign with regard to degradation of the various material s in the transition region. The internal environment will be argon gas and sodium vapor and essentially no sodium mass transf er or interstitial transf er ef fects occur at temperatures below 7000F, especially when the sodium is present as a vapor or thin condensed l ayer (Ref. 4 and 5). The external environment is reactor cavity gas, nitrogen pl us approximately 2% oxygen, and the interaction of the material s invol ved w ith thIs reactor cavity gas are negl IgIble at the Iow service temperatures.

Selection of the materials for this transition joint was made based on the above considerations coupled with the requirement to minimize thermal expansion dif ferences which could cause high stresses to be built up during thermal cycl ing. In addition, the use of the nickel base alloy filler metal, inconel 82, minimizes the depletion of carbon f rom the f usion zone of the SA 508 during wel ding and subsequent high temperature stress rel ief.

O 5.2-11 Anend. 72 Oct. 1982

i O O l

I b TABLE 5.2-1 SUlemRY OF (X)DE, CDDE CASES AND RDT STANDARDS APftlCELE TO DESIGN AND MANUFACTURE OF REACTOR VESSEL, 0.0SURE HEAD AND GUARD VESSEL 1

! 'e cla--ee : -

Pressure Internals Guerd 1 Component / Criteria Reactor Yessei Boundary (as appropriate) Vossel Section til Addenda thru Winter Addende thru Winter Addenda thru Winter Addenda thru ASE Code, '74 '74 '74 Summer '75 1974 Edition Class 1 Class 1 Class 1 Cl ass 2**

1 AS E Code Cases 1521-1,1592-2,1593- 1682,1690 1521-1 1592-4,1593-1,1594-1 0,1594-1,1595-1, 1592-4,1593-1 If elected lyy sup-1596-1 (1682,1690 pl ter 1521-1 & 1682 Optlonal)

RDT Standards E8-18T, 2/75 E15-2pe-T, 11/74 E15-fe-T, 11/74 EI5-le-T, 11/74

. Mandatory E15-2te-T, 11/74 Amend thru 6/75 Amend thru 6/75 Amend thru 6/76 l Amend thru 1/75

? F2-2, 8/73 F2-2, 8/73 F2-2, 8/73 N F2-2, 8/73 1 l Amend thru 7/75 Amend thru 7/75 Amend thru 7/75 Amend thru 7/75 F3-6T, 12/74 F3-6T, 12/74'*

• F9-4, 9/74 F3-6T, 10/75 i

With Amend.1/75 l) F6-5T, 8/74 F6-5T, 8/74 F6-5T, 8/74 l Amend thru 2/75 Amend thru 2/75 Amend thru 11/75 F7-3T, 11/74 F7-3T, 6/75 F7-3T, 6/75

}

l F9-4T, 9/74 M1-IT, 3/75 F9-4,9/74 M1-2, 3/75

) l Amend thru 7/75

1. For those reactor vessel and closure head components Internal to the pressure boundary special purpose high cycle f atigue curves and creep damage rules have been developed as discussed in Appendix 5.2A.

ON

. r8 P

$M

T/BLE 5.2-1 (Continued)

Clocure Head Pressure Internal s Guard Canponent/Criterla Reactor Vessel Boundary (as appropriate) Vessel ICT Standards M1-IT, 3/15 M1-27, 4/75 M1-4T, 3/75 l Anend. 6/75 M1-4T, 3/75 M1-6T, 4/75 Ane..d 1-7/75 M1-6T, 4/75 M1-10T, 3/75 l Amend. 6/75 M1-10T, 3/75 M1-11T, 3/75 knend 1-7/75 M1-11T, 3/75 M1-17T, 3/75 l Amend. 6/75 m M1-17T, 3/75 M2-2T, 12/74 m M2-27, 12/74 M2-7T, 3/75

O M2-5T, 1/75 M3-10T, 7/75 N hmend 1-2/75 M2-7T, 2/75 M7-4T, 3/75 M2-18T, 4/76 M2-21T, 12/77 M3-6T, 3/75 M3-7T, 4/75 M5-IT, 11/74 M5-2T, 5/73 M5-3T, t 2/74 M5-4T, f/75 M6-3T, 2/75 M6-4T, 2/75 M7-3T, t1/74 l M7-4T, 4/76 Non-Mandatory F9-5T, 9/74 F9-5T, 9/74 F9-5T, 9/74

• • Functionally designated Class 2, and constructed to rules for Class 1, but not hydrostatically tested or code stamped.

es*Except for the three rotating plugs, for which the applicable issues are: F3-6T, 3/69 for LRP & SRP3 F3-6T, 5/74 for IRP.

M2-7T, 2/69 for LRP & SRP; M2-7T, 2/74 for iRP.

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I Both PHTS and IHTS pumps require a shaf t seal to ef fect a zero leak seal from cover gas to ahnosphers. This shaf t seal, shown schematically in Figure O,- 5.3-14A, is an oil lubricated, double rubbing face seal. The seal has a shaf t driven internal oil circuletor and an Integral air to oil heat exchanger, with oil supply to make up f or oil leakage past the rubbing f aces. Oil leakage f rom the seal assembly into the sodium coolant is prevented by two barriers.

The first barrier is an oil dem approximately 1.2 inches above the lower f ace ,

seal. The normal leakage from the lower face seal is diverted by this oil dam '

Into the oil leakage drain passage into the lower seal leakage collection l reservoir. A second barrier is the collar above the drop down seal located just above the purge labyrinth. This collar extends beyond and over the labyrinth, thereby shunting any oil to a drain plenum. For oil to penetrate into the sodium, three things must happen:

o Failure of the oil dam o Failure of the collar to divert oil j o Overflow of the plenum drainage over the drop down seal lip A positive pressure is maintained in the shnft seal oil at all times by means l of an oil supply tank which will be pressurized above the loop operating pressure. The oil feed line to the seal will be oriented to preclude seal drainage in the event of a line break. The seal is capable of many hours of operation on the sel f contained fluid.

The oil system supporting the shaf t seal contains three tanks, each of which will have a level probe, thereby permitting monitoring of total oil inventory, O' its location, and permitting calculation of seal leak rate. The lower seal leakage collection tank is sized to hold the entire system's oil inventory of approximately 41 gallons.

O 5.3-21 Amend. 62 Nov. 1981

Oil vapors which may potentially be drawn from the lower seat leakage collection tank into the tank ullage during draw down (pump speed up) are retarded from such passage by means of a spilt flow purgs gas feed of recycled argon into the purge labyrinth. This gas feed splits and flows up at:d down the shaf t f rom the feedpoint. This gas input is flow controlled at the inlet, and flow controlled at the discharge from the lower seal leakage collection tank. If feed pressure into the tank is detected to be low (by the gas feed system) the discharge of gas from the tank will be closed. in event of gas line rupture at the oil tank discharge, the orificing by the line will retard loss of cover gas pressure.

Radioactive vapors from the tank ullage are prevented from escape to the atmosphere by the two barriers consisting of the gas downflow at the purge labyrinth and the oil lubricated double shaf t seal. Radioactive purging is continuous by means of the bubbiing in the standpipe, which is connected to RAPS.

O O

5.3-21a Amend. 72 Oct. 1982

Functionally, the drum receives a saturated water / steam mixture from the evaporators and subcooled feedwater and produces saturated steam of low moisture content f or the superheater and subcooled water of low steam content f or the recircul ation pump. The water / steam mixture f rom the evaporators enters the drum through the water / steam nozzles and flows into an annular volume along the sides of the Inner drum wall created by a girth baf fle extending along the side of the drum for the length of the cylinder.

Centrif ugal steam separators mounted along the iength of the drum draw from this annular vol ume, separate the mixture into phases, and direct the steam upward and the water downward into the inner vol ume of the drum. The main feedwater enters the drum through a single nozzle which feeds two distribution pipes through a "Y" connection inside the drum. The feedwater is distributed along the length of the drum by rows of orifice holes in the ho pipes which are located along each side of the drum beneath the steam separators. The auxil iary feedwater enters through a separate nozzle and is distributed along the length of the drum by two rows of spray nozzles in a single distribution pipe located above the water level in the drum. To precl ude waterhammer due to injection of highly subcooled water interf acing with saturated steam within a closed volume, the spray line is vented by the nozzles plus 18 7/32" OD hol es. These vents ensure that the feed I ine will remain f ull of water at the temperature of the steam drum Inventory. Feedwater mixes wIth the water from the separators and is drawn downward and out through the water outlet nozzles by the recircul ation pump. The steam passes upward through chevron type dryers in the upper portion of the drum and out through the steam outlet nozzles to the superheater. The dryers remove all but the last f ractional percent of the moisture from the steam and drain this moisture back to mix w ith the resident drum water. Drum drain piping, located along either side of the drum in the region where the water from the separators enters the drum h)

U Inner vol ume, draws water of high impurity concentration f rom the drum.

5.5.2.4 Overoressure Protection Location of Pressure Relief Devices Saf ety/ power rel lef val ves are located in the steam generation system to:

1. Prevent a sustained pressure rise of more than 10 percent above system design pressure at the design temperature within the pressure boundary of ihe system protected by the valve under any pressure transients anticipated; and

2. Provide steam generator modul e bl owdown capabil Ity.

Installation o~f the valves.will comply with the requirements as specified in Section 3.9.2.5. Saf ety/ power rel ief valves are instal led on the outlet l ines f ran each evaporator to provide venting capabil ity and a portion of the required saf ety/rel ief capabil ity. Safety valves are installed on the steam O

5.5-12 Amend. 72 Oct. 1982

drum to provide the remainder of the safety capability for the recirculation loop. Additional safety / power rel lef valves are installed on the steam exit I ine f rom the superheater because the steam I ines to and f rm the superheater have Isol ation val ves. The P&lD for the Steam Generation System, Figure 5.1-4 shows the locations of these saf ety/ power rel ief val ves. Additional detail s of sizes and pressure rating are given in Table 5.5-8.

Pressure-Reilef Devices Water / Steam Side Each safety rel lef valve on the evaporator outlet piping provides a saturated steam (100% qual ity) rel lef capacity of 430,000 lb/hr, or 39% of the rated steam generating capacity of the recirculation loop. Each saf ety rel lef val ve on the steam drum provides a saturated steam rel lef capacity of 410,000 lb/hr, or 37% of the rated steam generating capacity of the recirculation loop. The dif ference in rated capacity of these valves is due to the dif ference in the val ve set pressure. The combined relief capacity of the six valves for the recirculation loop is therefore 230% of the rated steam generation capacity.

This generous margin is provided f or two reasons: (1) the capacity required to rol leve most of the overpressure transients in the recirculation Ioop can be satisfied by opening one or both of the steam drum valves, rel ieving the system with dry steam rather than wet steam; (2) the capacity of the evaporator rel lef valves is based on the capacity required to achieve rapid blowdown of the evaporator modules following a water to sodium leak.

Three safety / power rel ief valves installed on the exit l ine frcm the superheater provide a rellef capacity of 75% of rated superheater steam flow at a pressure of approximately 1800 psig and temperature of 900 F. The remaining 25% of rated flow is rel ieved by the steam drum val ves.

Settings f or the saf ety/ power rel lef valves are in accordance with Code req ui rement s. Setting presently selected are shown in Table 5.5-8.

l l

i l

i 9

l 5.5-12a Amend. 72 l Oct. 1982

Tests and InsoectIons In-service Inspection of the steam generator modules is discussed under in-Service inspection Program, Section 5.5.2.1.3.

Part Load Ooeration Part load operation curves over the range of steam flows f rom 40 to 100 percent are presented in Section 5.7.2.

Design module heat transfer length were used with nminal values of sodium, water or steam, and tube heat transfer correlations for purposes of this analysis. This implied excess area, therefore, results in sodium operating temperatures in the evaporator lower than those used for design. Design heat transfer areas are determined by adding suf ficient margin to the module length to permit operation with fouled tubes at 100% power for nominal sodium conditions. The margin calculated is 10% and is arrived at considering the error-band in heat transfer coefficients and tube wall thickness. Also, included in the 10% margin is a 5% surf ace allowance made for tube plugging.

The steam flow rate is defined by turbine conditions, power level, and feedwater temperature.

The water or steam side taperature and flow rate are essentially the same for both clean operation and fouled operation. However, the presence of fouling will cause an increase in the required sodium operating tmperatures and flow compared to clean operation.

For power levels below about 40 percent a good portion of the inlet sodium end of the superheater and the outlet sodium and of the evaporator wilI operate close to isothermal temperatures with smalI sodium to water temperature differences. This is because most of the heat transfer takes place in other portions of the modules.

5.5.3.6 Evaluation of Steam Generator Leaks A primary design objective for the steam generators is that they be of suf ficiently high quality that leaks in the sodium / water boundary will not occur. Caref ul design and close quality control of materials and manuf acturing processes are expected to yield units which are free of common def ects, and the probability of a leak in a steam generator tube is expected to be quite smalI.

A Steam Generator Leak Detection System, described in Section 7.5.5, has been provided to allow operator action to limit the consequences of a leak. The leak detection system will alert the operator to the existence of a leak rate as low as 2 x 10-5 lb water /sec, which will allow sufficient time for operator action to prevent a significant increase in the leak rate for a broad spectrum 59 of leak rates.

O Amend. 59 5.5-24 Dec. 1980

As a final level of protection against tube leaks in a steam generator, the steam generators and the IHTS are being designed to withstand the of fects of a large sodium water reaction (SWR). The ASME Code categories being appiled in he design of the steam generators and IHTS piping and components for the large SWR event are given in Table 5.5-10.

The design basis leak (DBL) for the CRBRP was selected based upon examination of the physical processes which exist for leak initiation and growth. Two .

types of tests have been reported which provide Information on the leak growth mechanism - small scale tests which model ef fects of a SWR on materials, and large scale tests which model a large water leak in a model of a steam generator. Smaller scale sodium-water reaction tests have been done to develop an understanding of the of fect of a SWR on neighboring tubes in a steam generator. Three mechanisms have been identified for leak growth:

solf-wastage, impingement, and overheating (mechanical damage from pipe whip, although extremely unlikely, could be considered another mechanism, as discussed later in this section). -glf-wastggehasbeenshowntooccurfor S

very small leaks in the range of 10 to 10- lb/sec (Ref. 13). The process is depicted in Figu The result of this process is a leak size of the order of 10 geto15.3.}.3-1.

10- Ib/sec. which can produce wastage on another tube in the vicinity of the leaking tube.

Wastage can occur on the outside of a steam generator tube from a leak in another tube in the vicinity. Tests of thIs mechanism have typically been done by using a water jet directed through sodium to a target material sample.

Water injection rates of approximately 10 4 lb/see to 1 lb/see have been tested. The wastage mechanism results in erosion of the target material at maximum rates of 0.001 to 0.007 inches per second (Ref.14, 29). The wastage rate is found to be a function of the water injection rate, tube spacing, sodium temperature and leak geometry. Wastage occurring on the surface of a CRBRP steam generator tube at these rates could cause a secondary water leak from tube penetration. However, this would require at least 20 seconds to penetrate the 0.109 inch thick tube wall assuming an initiating leak of the proper characteristics to produce maximum wastage.

The size of a secondary water ieak resulting f rom wastage is dif ficut t to quantify since wastage tests are typically done on materials samples rather thanppessurizedjubes. The wastage areas observed in tests have ranged from 0.1 in to 1.5 in . Failure areas corresponding to the highest observed wastage areas would result in water leak rates corresponding to that of a double-ended guillotine tube failure. However, the entire wastage area would not be expected to blow out. The wasted areas are typically pit-shaped with the area of the pit decreasing with depth. It would be expected that the small area at the bottom of the pit would f all, yielding a return water leak l which hal ts the wastage. Therefore, while the size of a secondary failure l caused by wastage is difficult to predict, it is expected to be smaller than l the leak rate corresponding to a double-ended guillotine failure.

5.5-24a Amend. 72 O

Oct. 1982

3 References to Section 5.5

?

2

• 1. W.H. Yunker, " Standard FFTF Values for the Physical and Thermophysical Properties of Sodium" WHAM-D-3, July 6,1970.

• 2. J. A. Bray, "Some Notes on Sodium / Water Reaction Work," CONF-710548, pp.187-205, July 1972.

• 3. Nuclear System Materials Handbook, Hanford Engineering Development Laboratory, TID-26666, Vol ume 1, Section 2-2 1/4 Cr-Mo, pp.1.0-1.2, Rev. O, August 14, 1974.

• 4. R.B. Harty, " Modular Steam Generator Final Project Report," Atomics International, TR-097-330-010, September 1974.

• 5. Nuclear Systems Materials Handbook TID 26666, 1974.

• 6. V.L. Streeter and E.B. Wylle, Hvdraulle Transients, McGraw-Hill, New York,1967, Ch. 2 and 3.

• 7. John Pickford, Analysis of Surge, MacMillan, London, 1969, pp. 32-37.

• 8. D.J. Cagliostro, S.J. Wiersman, A.L. Florence, Stanford Research Institute Final Report P.O.190-01H38GX, " Pressure Pul se Propagation in a Simple Model of the Intermediate Heat Transport System of a Liquid Metal Fast Breeder Reactor," June 1975.

n

( I

• 9. "RELAP4/ MOD 5 a Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reaciors and Related Systems," prepared by Aerojet Nuclear Company for U.S. Nuclear Regulatory Commission and Energy Research and Development Administration under Contract E (10-1)

- 1375, ANCR-NUREG-1335, September 1976.

• 10. J.N. Fox, R. Salvatori, H.J. Thailar G NES), " Experimental Bending Tests on Pressurized Piping Under Static and Simulated Accident Conditions" TRANSACTIONS, ANS Power Division Conference on Power Reactor Systems and Components, September 1-3, 1970.

11. " Draft Design Basis for Protection Against Pipe Whip," ANSI N176, June, 1974.

12. Deleted

• 13. Gudahl, J.A. and Magee, P.M., "Microleak Wastage Test Results,"

GEFR-00352, March 1978.

• References annotated with an asterisk support conclusions in the Section.

Other references are provided as background information.

5.5-35 Amend. 72 Oct. 1982

• 14. Greene, D.A., Gudahl, J.A., Hunsicker, J.C., " Experimental Investigation of the Wastage of Steam Generator Materials by Sodium /

Water Reactions," GEAP-14094, January 1976.

• 15. Dumm, K. et.al., Experimental and Theoretical Investigations on Safety of the SNR Straight-Tube Design Steam Generator with Sodium-Water Reactions, INTAT72.12 (ERDA-TR-27), INTERATOM, Apri l 1972.

16. J.A. Bray, "Some Notes on Sodium / Water Reaction Work," Paper presented at the Specialists Meeting on Sodium Water Reactions, CONF-710548, held May 18-21, 1971, Melekess, USSR.

• 17. B.V. Kuplin, et.al., " Study of Na and H O Interactions in a One 2

Megawatt Modular Steam Generator," Paper presented at the Specialists Meeting on Sodium-Water Reactions, CONF-710548, Hold May 18-21, 1971, Molekess, USSR.

18. Liquid Metal Engineering Center - Failure Data Handbook, LMEC Memo -

69-7, Vol ume 1, Atomics Internat ional, Al-RAR-096-13-00, August 15, 1969.

19. Reactor Primary Coolant System Rupture Study, Quarterly Report No. 23, October-December 1970, GEAP 10207-23, January 1971.

20. J.A. Bray, et al., " Sodium / Water Reaction Experiments on Model P.F.R.

Heat Exchangers- The NOAH Rig Tests," TRG-Reports-1519,1967.

21 . J.A. Bray, "A Review of Some Sodium / Water Reaction Experiments,"

British Nuclear Energy Society Journal, Vol.10, No. 2, April 1971.

• 22. P.B. Stephens, D.N. Rodgers, et.al., "DNB Effects Test Program Final Report," GEFR-00100(L), June 1977.

• 23. J.C. Whipple, et.al., "U.S. Program for Large Sodium / Water Reaction Tests," GEFR-SP-039, November 1977.

24. R.L. Eichelberger, " Sodium-Water Reaction Tests in LLTR Series I, Final Report," ETEC-78-10, July 15,1978.

25. J.0. Sano, et. al., " Evaluation of Sodium-Water Reaction Tests No. 1 Through 6 Data and Comparison with TRANSWRAP Analysis Series l Large Leak Test Program, Volumes I and ll," GEFR-00420, June 1980.

26. J.C. Whipple, et.al. " Evaluation of LLTR Series il Test A-2 Results,"

General Electric Advanced Reactor Systems Depariment, July 1980, Prepared for U.S. Department of Energy under Contract No.

De-ATc,3-76SF70030, Work Package AF 15 10 05, WPT No. SG037.

Y D

e 5.5-35a Amend. 62

{ Nov. 1981 Y

Y

27. J. C. Amos, et.al, " Evaluation of LLTR Series ll Test A-3 Results, Revision 1," General Electric Advanced Reactor Systems Department, (pj May 1982, Prepared for U.S. Department of Energy under Contract No.

DE-ATc3-76SF70030, Work Package AF 15 10 05, WPT No. SG037.

28. J. O. Sterns, " Metallurgical Evaluation of the Modular Steam Generator (MSG) after LLTR Testing," ETEC-78-12, Sept. 1978.

29. D. A. Greene, J. A. Gudahl and P. M. Magee, "Recent Experimental Results on Small Leak Behavior and Interpretation for Leak Detection," CONF-780201, Vol.1, paper No.12 (First Joint U.S./

Japan LMFBR Steam Generator Seminar), February 1978.

30. J. C. Amos, et al, " Evaluation of LLTR Series 11 Test A6 Results,"

prepared for U. S. DOE under Contract DE-AT03-76SF0030, June 1981.

31. D. E. Knittle, et al, " Evaluation of LLTR Series 11 Test A7 Results, prepared for U. S. DOE under Contract DE-AT03-76SF70030, September 1981.

32. J. J. Regimbal, et al, " Evaluation of LLTR Series 11 Test A-8 Results," prepared for U. S. DOE under contract DE-AT03-76SF70030, February 1982.

• References annotated with an asterisk support conclusions in the Section.

Other references are provided as background information.

O l

b a

5.5-35b Amend. 72 Oct. 1982

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l TABLE 5.5.12 MAXIMUM ALLOWABLE STEAM GENERATOR SYSTEM PRESSURE BOUNDARY VALVE LEAK RATES VALVE LEAK RATE, L8/HR Evaporator inlet isolation 4.7 Evaporator inlet Water Dump Isolation .02 Evaporator Outlet Rellef 1.0 Steam Drum Rellef 1.0 Superheater inlet isolation 4.7 Superheater Reilef 1.0 Superheater Outlet Isolation 4.7 Superheater Bypass 1.2 Main Feedwater Isolation 4.7 Steam Drum Drain isolation 4.7 Nitrogen Supply (per Conn.) 1.0 O

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TELE 5.5-13 SUMMMY OF U.S. LARGE S001UM/ WATER REACTION TESTS i

SIGNIFICANT COUNTRY TEST DESIGNATION / TEST VESSEL TEST BU@LE INITI AL PRESS / TEMP. WATER INJECTION OBJECTIV E SODlUM WATER E TH00 DURATION WElGHT RESULTS PSIG of PSIG of SEC LB i

U.S. LLTR Series 11 Test See as A2 Prototypic 125 580 2000 580 See as 30 0 Prototypic rup-Ala, One DEG G SW R-1 ture disk Lower Midspan, essembly used on injected Nitrogen, all Series 11 Prototypic Rupture tests. Served

,! Disk Assembly used to verify RELAP on all Series 11 Tests calibration j

LLTR Series 11, Test S e e as A2 me as A2 125 58 0 2000 580 See as 43 0 Served to verify j

Alb, See as Ala ex- SW R-1 RELAB calibra-

! cept Alb used double tion j

disk and minor difference in leak location, LLTR Series 11, Prototypic Prototypic 125 58 0 1700 580 See as 40 200 No secondary ci Test A2, One DEG 9 Cross-Section SWR-1 failures. Max-T m

Lower Mt dspan, sub- 1/2 Length cooled H2O Imum measured secondary wastage equals 4 N mils. Prototypic double disc assembly served to callt: rate TRANSWRAP rup-ture disc model I

t 1

k kk 4

c=cc G~

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TN3LE 5.5-13

SUMMARY

OF U.S. LARGE S001UN/ WATER REACTION TESTS CDUNTRY TEST DESIGNATION / TEST VESSEL TEST BUNDLE INITIAL PRESS / TEMP. WATER INJECTION SIGNIFICANT OBJECTIVE S001UM WATER ETHOD DURATION WEIGHT RESULTS PSIG OF PSIG of SEC LB U.S. LLTR Series ll, S e e as A2 See as A2 145 580 1700 580 Rapid pull- 145 144 Secondary Test A-3, One sel f- apart of plus f allures (less Wastage Leak prenotched than an EDFG)

Simulation 8 sub- tube to af ter long de-cooled H2O801 expose 0.040" laws (one ibm /secaimedIer dia. hole, alhute and maximum secondary longer).

damage.

LLTR Serles 11, Se e as A2 See as A2 125 580 1700 580 See as 36 200 No secondary Test A6, One DEG 8 SWR-1 fallures.

Lower MIdspan Perl-phery, subcooled H

m 20.

b System modifled as gas-free Actual test con-talned large gas space to S.G.

TRANSWRAP over-predicted measured pressures where comparabie.

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s O O TABLE 5.5-13 SUM 4ARY OF U.S. LARGE SODIUM / WATER REACTION TESTS ODUNTRY TEST DESIGNATION / TEST VESSEL TEST BUNDLE INITIAL PRESS / TEMP. WATEP. INJECTION SIGNIFICANT OBJECTIVE SODIUM WATER ElHOD OURATION WElGHT RESULT 3 PSIG OF PSIG OF SEC LB U. S. LLTR Series 11, Same as A2 S e e as A2 255 580 2000 580 Seeas 2 15 Secondary tubes Test A7, One DEG 8 SWR-1 filled with Lower Midspan, sub- nitrogen 8 400 cooled H20 higher PSIG.

Initial sodiua pressure.

LLTR Series 11, Sane as A2 See as A2 180 900 1550 700 Rapid pull- 40 No secondary Test 2, Intermed- apart of failures deduced late-sized super- prenotched f rom Instrum-heated steam tube to ex- entation and Injection. pose 0.054" post test helium dia. hole. Issk checks.

Final confirne-ation awaits

. post test

• destructive m exaninetton.

<0 LLTR Series ll, See as A2 Sane as A2 50 625 1450 625 Rapid pull 58 TBD Test Report not Te=+ A5, Inter- apart of avai l abl e.

mediate-siz ed tube to ex- Exanination of superheat in- pose 0.25" of test article joction dia. hole in progress.

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44) 5.6.1.2.1.3 Surveillance and In-service Inspection The SGAHRS system will be inspected in accordance with the intent of p 58 Section XI Division 1 of the ASME Code.

5.6.1.2.1.4 Protection Against Accelerated Corrosion and Material Degradation In water and/or steam, carbon steels are susceptible to pitting in s the presence of chloride and oxygen. Furthermore, below 550 F, these materials are susceptible to caustic gouging and, perhaps, caustic stress corrosion cracking. Maintaining the water purity consistent with the requirements for chlorides, caustics and oxygen for short term operation will prevent these forms of localized attack.

Carbon steel is also susceptible to hydrogen embrittlement under SGAHRS operating conditions. However, maintaining the specified water purity will prevent this occurrence. Administrative procedures will be established to assure that water purity will be maintained.

5.6.1.2.1.5 Material Inspection Program i The SGAHRS material inspection program will be based on the require-58[ monts of the ASE Code. Sectfon III. for carbon steel and 2k Cr luo, steel.

5.6.1.2.2 Material Properties The materials used in the SGAHRS are described and discussed in 58l Section 5.6.1.1.4.

5.6.1.2.3 Component Descriptions The major SGAHRS components have been designed with sufficient margin m to assure that they will provide adequate cooling after a plant shutdown from power operation up to 115% of rated power. The decay heat levels shown in Figure 5.6-6 were used for component sizing and system response calculations for SGAHRS. l25 5.6-5 Amend. 58 Nov. 1980

5.6.1.2.3.1 Protected Air Cooled Condensers (PACC)

Comoonent Descriotion l

The PACC is a tube-type steam condenser constructed of carbon steel. Heat is i rejected to the atmosphere by condensing the saturated steam from the steam drums by forced circulation of air over the tube bundles.

Each unit is sized to reject 15 Wt under conditions of forced convection on the air side and natural circulation flow on the steam / water side. Each PACC has two half-size tube bundles, two variable blade pitch fans and two sets of variable position louvers to control airflow and, therefore, heat rejection.

The electrical power supplies and instrument and control circuits for the PACCs are Class 1E. Refer to PSAR Section 7.4 for information on the power sources and l&C.

The arrangement of PACC is il lustrated in Figures 5.6-8 and 5.6-9. Air is delivered from axial fans (one for each tube bundle) into the insulated plenum surrounding each tube bundle. Air flows circumferentially around the tube bundle, then radially inward through the fin tube bundle into a central core.

Air then flows upward through the central core and exhausts through louvers to an exhaust stack.

Each tube bundle consists of 50 finned tubes connected in parallel between vertical pipe headers. Each tube is approximately 100 ft. long and, of the 100 ft. length, 95 ft. Is finned. The individual finned tubes are formed in a conical spiral of approximately four corceniric turns with a slope toward the center. The tubes are connected in parallel between vertical pipe headers. l l The inlet header is on the outside and outlet header is in the center of spiraled coils. The finned tubes are made of 2 inch C.D. tubes'with 0.156 inch minimum walI as shown on Figure 5.6-10. The 0.D.,of the fin is 3.28 inches. The fins are serrated into 0.156 inch segments from continuous strip 0.050 inch thick x 0.75 inch wide. The strip is first formed into the shape of an "L". The strip is then wound around the tube 0.D. to complete the footed fin attachment to the tube. There are two separate tube bundles in each PACC.

Deslan Data Design Conditions:

Pressure 2200 psig Temperature 6500F Thermal Hydraul ic Perf ormance:

Heat Removal 15 W t (7.5 W t per tube bundle)

Steam Pressure 1450 psig Steam Temperature 5920F Moisture 0%

Condensate Temperature 5920F Air Temperature 1000F Air Pressure 14.3 psia 5.6-6 Amend. 72 Oct. 1982

Design Criteria d The power supplies to the PACC fans, instrumentation and controls are Class IE. The Instrumentation and Control System is a saf ety relaied system and as such will meet the requirements of the regulatory guides and standards as Iisted in Tables 7.1-2 and 7.1-3 of the PSAR. The means of compliance are described in Section 7.1.2.

Three PACC units are provided, one for each heat transport loop, each capable of removing the total decay heat approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after shutdown. Each unit is single active f ailure proof in that no single active f ailure will result in the loss of more than 50% of heat removal capability. This is provided by utilizing two tube bundles, two fans, etc., such that at least half capacity is retained following the f ailure. The PACC unit is a Seismic Category I design, hardened against tornado missiles and designed to withstand the pressure loads f rom tornados. The PACC tube bundle design is based upon standard techniques for steam-to-air heat exchangers.

Ooeration and Control The airflow is regulated by the use of variable position inlet louvers and f ans with variable blade pitch. There are separate controls for the air side of each PACC f or each of the two f ans and f or each of the two sets of touvers.

The inlet louvers and fan blade pitch are positioned by controllers which compare steam drum pressure to the setpoint and generate position demand signals to the louvers and f an blade pitch drives as required to maintain p pressure at the setpoint value. In order for the PACC to effcct heat rejection control over the range of operation there are two modes of air side operation:

(1) Forced convection with the louvers open and airflow varle(' by changing the f an blade pitch.

(2) Natural circulation with airflow varied by changing the position of the inlet louvers.

The range of automatic operation is from 15% to 100% heat rejection. From 100% down to approximately 30% (4.5 MWt) the unit is operating in the first mode, and f rom 30% to 15% in the second mode. Control is accompl ished by sensing and maintaining the steam pressure at the desired set point.

U 5.6-6a Amend. 72 Oct. 1982

5.6.1.2.3.2 Auxiliarv Feedwater Pumos (AFWP)

The AFWP w il I be a mul ti-stage, centrif ugal pump selocted f rcm a commercial vendor's equipment I ine. No special requirements should be necessary since these pumps have been proven to be rol table in commercial applications. The turbino driven pump will be sized to deliver a 1432 GPM flow rate at 3927 feet l

developed head, and the two motor driven pumps will be sized to deliver one-hal f of this flow rate each at the same head. The predicted constant speed head / flow curves for, the turbine driven and motor driven AFW pumps are shown on f Igures 5.6-11 and 5.6-12 respectively.

AFWP Motor Drives These motor drives will be synchronous speed squirrel cage induction motors of 980 horsepower. These motors w il l be selected f rom a vendor's standard l ine and no special requirements are anticipated.

AFWP Turbine Drive This component will be obtained f rom an experienced vendor and will be sized to produce 1960 horsepower. The turbine wil l be constructed w ith suf f icient qual ity assurance coverage to assure its rel labil Ity during service.

The auxil iary feedpump turbine is not kept hot for quick start operation. The drive turbine concept selected for the Auxil iary Feed Pump is based on the capabil ity of this turbine to withstand severe service conditions. This is accompl ished by constructing the turbine wheel from a single forging with buckets mil led into the f orging. The start-up procedure is simil ar to that f or the RCIC turbine in a BWR in that it w il l occur w ithout pre-warming.

Pumo Integrity The auxil lary feed pumps wilI be designed to the requirements of ASE B&PV Code, Section 1II, Class 3. In additlon, the pumps and their supports wIlI be designed to Seismic Category I requirements. Allowabl e stress I imits are specif ied in Tabl e 3.9-3 and pressure l imits are specif ied in Table 3.9-4.

5.6.1.2.3.3 Protected Water Storage Tank (PWST)

The FWST hol ds the protected water to be suppl led to the steam drums in the event of loss of normal feedwater or normal heat sink. The size is determined by detailed analysis of the heat removal conditions during the f Irst several hours af ter shutdown and by anticipated component leakage rates. The tank wilI be constructed to the requirements for an ASE Section lIl/ Class 2 vessel and it will operate at low temperature (<200 F) and low pressure (<15 psig).

5. 6-6b Amend. 72 Oct. 1982 i

5.6.1.2.3.4 SGAHRS Piping and Support The SGAHRS piping is described below and is shown in Figure 5.1-5. The SGAHRS piping will be designed in accordance with the ASME Code Section III as specified in Section 5.6.1.1.2. The material 49 specifications are discussed in Section 5.6.1.1.4.

The SGAHRS piping runs can be categorized as follows:

a. PWST Fill Line This 3 inch low pressure, low temperature, Class 3 carbon steel line runs from the 10 inch alternate water supply line through the motor-driven, normally closed PWST fill valve to the PWST inlet.

b. Protected Water Storage Tank (PWST) to Auxiliary Feedwater Pump (AFP) Inlet There are three low pressure, low temperature, uninsulated carbon steel lines from the PWST to the three auxiliary feedwater pump inlets. Two of the lines, each of which leads to a half size, motor-driven pump are 6 inches in dianieter and the third line to the full size turbine-driven pump is 8 inches. All three lines contain a manually 58 l operated, locked open valve and an electrically operated, normally open isolation valve. These lines are Class 2 f] from the PWST to the electrically-operated isolation valve s

v49 l and then Class 3 to the pump inlet.

c. Alternate Supply Line to AFWP Inlet The alternate supply line provides the capability for the AFW pumps to take suction from the condensate storage tank.

A 10 inch carbon steel line runs from the feedwater and cor.densate 58 system junction to the first branch line. An 8 inch branch line 49l passes through an electrically-operated, normally closed isolation valve and tees into the 8 inch turbine pump inlet piping. Two 6-inch branch lines each pass through electrically-49 operated, normally closed isolation valves and then tee into the 6 inch motor-driven pump inlet piping. The total run of piping is Class 3.

d. Auxiliary Feedwater Pump Discharge to Discharge Header (Inclusive)

The 6 inch carbon steel turbine pump discharge line leads to

, a 6 inch discharge header. This header in turn has three dis-charge points, one to each steam drum feedvater supply loop.

a 6 inch carbon steel line from each motor driven pump feeds into a 6 inch header which also has three discharge points, one 43 17 to each drum.

Amend. 58 k N 0 5.6-7

l VENT l

l] TUBE BUNOLE l

3-1/2 FT mI FAN MOTOR 41 FT PACC MA!NTENANCE ISOLATION VALVES _

VENTURI FLOW METER T~1 k ~

~ STEAM DRUM -

1r V y 1r V RECIRC. PUMP HEADER ( )

FIGURE S.6-7 PACC Closed Loop Schematic (Shown During Mormal Plant Operation - PACC Hot Standby)

Amend. 58 5.6-42 80-644-04

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occur during plant operation and are sufficiently severe or frequent to be of possible significance to component cyclic behavior. The transients selected may be regarded as a conservative representation of transients which, used as /^3 a basis for component structural evaluation, provide confidence that the com-(j

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ponent is appropriate for its application over the design life of the plant. Appendix B describes the events which result in transients on heat transport system components. Table 5.7-1 presents a summary of a preliminary selection of those transients. Several events and examples of their affects on the components of the heat transport system are discussed and provided below to illustrate the transient behavior of the Heat Transport System. (More detailed discussions, including plots of temperature, flow, and pressure as a function of time, are included in Chapter 15):

a. Reactor Trip from Full Power A reactor trip from full power results in the release of safety and/or control rods. Sodium pumps coast to pony motor speed. i' The continued transfer of heat results in rapid temperature reductions at the reactor vessel outlet, primary pump, IHX pri-mary inlet, superheater sodium inlet and outlet, and evaporator sodium inlet. The primary hot leg temperatures drop about 300 F in 200 seconds while the superheater inlet sodium temper-ature drops about 200 F in the same time. Superheater outlet and evaporator inlet sodium temperatures fall about 170 F and then increase the same amount in a total of 100 seconds. The latter affect results from controlled dumping of steam to the condenser through the turbine bypass to maintain pressure at I_; the turbine admission valve at 1450 psig to avoid lifting of O safety or power relief valves. The transient is most severe when it occurs with minimum plant decay heat conditions since decay heat tends to slow the rate of temperature reduction.

Figure 5.7-3 depicts the transient at the reactor vessel outlet where the rate of temperature change is the highest. Substantial flow oscillations do not occur following reactor scram as discussed below. The free surfaces in the reactor coolant system are 1) the free s'urface in the reactor vessel, 2) the free surfaces in each of the three primary pump tanks and 3) possibly a free surface at the high point of the primary side of the IHX in the annulus between the outer shell and the tube bundle support cylinder.  ! The only gas which would be under any significant pressure would be that which may accumulate in the IHX. The volume of this gas will deliberately be kept as small as possible by locating the vent line between the IHX and the pump tank as high as possible. The position of the vent line from the IHX is shown in Fig. 5.3-15 and shows the possible trapped gas volume to be extremely small. When the pump is tripped and the pressure in this gas space drops off rapidly from about 165 psia to approximately 15 psia, there will be an expansion of this gas and a lowering of the free surface. The

,     volume of this gas when expanded will be small compared to the gas volume in the reactor vessel and pump tanks and as such, will not significantly (mV) affect sodium levels in either the pumps or reactor vessel.                           2E 5.7-3                          Amend. 25 Aug. 1976

The pump tank cover gas pressure during f uiI fIow conditions wiII be equal to or only siIghtly higher than the reactor vessel cover gas pressure (which is equalized with the overflow tank gas pressure through an equalization line). When the pumps are tripped, the level in the pump tanks will rise and sutmerge the stand pipe bubbler nozzle thereby cutting of f communication of the pump cover gas with the rest of the cover gas in the primary system. The level rise in the tank is limited by the compression of the trapped gas. The increase in pump tank level is at the expense of the level in the reactor vessel but any oscillation in free sur f aces in the pump and reactor vessel is precluded by providing a flow restriction between the pump hydraulics region and the pump tank which wiiI critically damp any potential oscilIation.

b. Ilncontrolled Rod Movement Control systun malfunctions may cause uncontrolled control rod movement resulting in undesired insertion or withdrawal of one or more l

control rods. Uncontrolled insertion of a control rod, which could occur without a compensating reduction in sodium ficws, results in rapid plant temperature reductions similar to those which occur from a reactor tr ip f rom f ul I power. Uncontrolled withdrawal of a control rod nay occur under varicus initial ccnditions. If uncontrolled rod withdrawal occurs f rom 100% power, reactor vessel outlets, lHX primary inicts, and primery sodium pump temperatures wil l increase to values higher than ncrmal and higher than from any other event. When power reaches 115%, a reacter trip occurs. Since temperatures just prior to reactor tr ip are higher than just prior to reactor tr ip f rom f ul l power, a more severe transient will occur. Although the rate of temperature change is about the same as that for a reactor trip f rom f ul l power, the extent of the transient is greater since it starts f rom a tanpera1ure aboui 600F higher than that observed at 100% power. Figure 5.7-4 illustrates the nature of this transient.

5. 7-3 a Amend. 72 O

Oct. 1982

Uncontrolled rod withdrawal during startup also results in an up temperature transient at the reactor vessel outlet although the p] t transient occurs at a lower tcrnperature than when the rod withdrawal starts f rom 100% power. Figure 5.7-5 depicts the transient initiated during startup.

c. Operating Basis Earthauake (OBE)

The operating basis earthquake results in reactive forces acting on the plant components as described in the Seismic Criteria Document. Five OBEs, each with 10 maximum peak response cycles, are assumed to occur over the design life of the plant. Four of these OBE's are assumed to occur during the most adverse Normal Operating Conditions determined on a component and design lImit basis. The other one OBE is assumed te occur during the most adverse upset event determined on a component and design iimit basis, and at the most adverse time in the upset event. Thus, the plant components are simultaneously exposed to the thermal effects of the thermal transients as welI as the stresses of the OBE.

d. Loss of Steam Generator Load Isoletion and dumping of the water / steam sides of both evaporatcrs and the superheater removes the load f rom that loop. This results in up temperature transients on the steam generator modules, the intermediate cold leg, the IHX intermediate inlet, the IHX primery outlet, and the reactor vessel Inlet. The ensuing reactor trip then causes down temperature transients on these components. Tho Intermediate cold leg temperature increases approximately 3500F in 400 seconds; then decreases approximately 2200F in 300 seconds.

This transient is then transported to the IHX primary outlet and reactor vessel inlet. Figures 5.7-6 a-k presents the resulting transient at the Intermediate sodium pump, core & steam generators.

e. Inadvertent Ooening of Suoerheater Outlet Power or Safety Rellef Valve This event results in a large increase in load without an accompanying increase in reactor power or sodium flows, it occurs when a super-heater relief valve inadvertently opens to increase steam flow from 40% to 100%. The event results in a reactor trip but overcooling occurs due to the open relief valve. The steam generators, inter-mediate cold leg, lHX intermediate inlet, primary cold leg and reactor vessel inlet drop in temperature about 150oF in 100 seconds. The reactor vessel outlet, primary hot leg, and lHX primary inlet drop in temperature about 2000F in 75 seconds. Figure 5.7-7 depi::ts the transient at the intermediate pump.

Amend. 72 5.7-4 Oct. 1982

                                         ~.

f. Primarv Pumo Mechanical Failure (O

  \j Primary pump mechanical failure involves the instantaneous stoppage of the impeller of one primary pump due to such reasons as seizure or breakage of the shaft or impeller. Flow in the affected loop repialy goes to zero and a reactor trip occurs almost immediately af ter seizure based on primary to intermediate ficw ratio.       The event is

, characterized by a down transient in the intermediate hot leg and a l' check valve sim in the primary cold leg of the af fected loop. The down-temperature transient in the intermediate hot leg results from the sudden loss of primary sodium flow while intermediate sodium ficw continues. The Intermediate hot leg temperature drops 3000F in about 100 seconds. The check valve sim, which results from the check valve being forced shut by reverse flow from the reactor vessel, results in significant pressure fluctuations at the reactor vessel inlet, the check valve, and the IHX primary outlet. Figure 5.7-8 presents tne temperature transient at the superheater inlet while Figure 5.7-9 depicts the pressure ef fects of the check valve slam at the check valve inlet and outlet.

g. Saturated Steam Line Ruoture A rupture of the saturated steam Iine between the steam drum and the superheater inlet isolation valve results in immediate cessation of superheater steam flow in that loop and initiation of a reactor trip.

The superheater rapidly becomes isothermal at the sodium inlet temperature due to the loss of ccoling. Sodiun leaving the , evaporators of the af fected locp initially drops in tmperature due to h over cooling as the water ficw increases and flashes to atmospheric pressure through the steam drum. Then, as the loop blows dry through - the rupture, ovaporator sodium temperature rapidly increases to the superheater inlet temperature. This transient is the most severe that the evaporator and intermediate pump experience. The transient is propagated through the intermediate cold leg and results in similar severe transients on the intermediate pump, the IHX intermediate inlet, the IHX primary outlet, the primary cold leg and check valve, and the reactor vessel inlet nozzle. Subsequently, these components experience down tmperature transients as a result of the reactor trip. Intermediate cold leg temperature drops 2000F in about 60 seconds and then increases 5000F in about 100 seconds. Figure 5.7-10 illustrates the transient at the intermediate pump. O 5.7-5 Amend. 72 Oct. 1982

h. Loss of One Primary Pumo Pony Motor with Failure of the Check Valve in that Loco to Shut This event occurs subsequent to a shut down or reactor trip and results in reverse flow in the af fected prirrary loop as a result of the head developed by the two operating pumps. The reverse flow of primary sodium at reactor vessel inlet temperatures results in rapid down temperature transients at the IHX primary inlet, the primary pump, and the reactor vessel outlet nozzle of the af fected loop. A core temperature increase occurs as a resul t of the bypassed flow.

Primary hot l eg temperature drops 425 F in about 150 seconds. Figure 5.7-11 depicts a typical transient. 5.7.4 Evaluation of Thermal Hvdraulic Characteristics and Plant Desian Heat Transoort System Design Transient Summary The heat transport system design transients f or the individual heat transport ! system components are described in Appendix B. Table 5.7-1 presents a i preliminary summary listing of design transient events as well as the f requency of each event assigned to the reactor vessel, IHX, primary pump, intermediate pump, primary check valve, evaporator and super-heater. It should be noted that the assigned f requency for a particular event varies among the componcnts in some cases. This is the result of the method used in establ ishing the design transients. The events I isted in Appendix B are the result of grouping less severe events under more severe events and apply Ing the total frequency of alI events in the group to the most severe event in the group. This approach was applled separately to each component so that the most transients discussed in Appendix B (where a particular transient applles to more than one component) do not have the same frequency appi led to each component. This approach was required because each event does not result in the same transient of fect on each component. O 5.7-6 Amend. 72 Oct. 1982

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f i 5.7-7 i c

TABLE 5.7-1 (continued) PRELIMINARY SUMY OF HEAT TRANSPORT SYSTEM DESIGN TRANSIENTS Frequency (Lifetime) Primary DUTY CYCLE y Event Reactor Primary Inter. Check Super-EVENT NUm ER Title Vessel IHX Pump Pump Valve Evap. heater U-11b Water side Isolation & blowdown of - -- - - - 7 7 evaporator module U-11b Adjacent evaporator during water side - - - - - 9 9 isolation and blowdown of evaporator U-21e Adjacent evaporator outlet relief - - - - - 3 3 valves open E-9a Superheater Isolation & blowdown-outlet -- -- -- -- - Note 4 Note 4 valve open E-1g  ? ,dvertent dump of Intermediate sodium -- - -- - -- Note 4 Note 4 OBE Operating basis earthquake 5 5 5 5 5 5 5 E-16 Three loop natural circulation Note 4 l U-21b Inadvertent opening of superheater 42 19 24 14 26 13 13 outlet power or safety relief valve U-23 Inadvertent opening of evaporator - 33 -- 37 - - -- ,m inlet dump valve N U-8 Primary pump pony motor failure #15 5 -- - -- 5 5 s E-1 E-5 Primary pump mechanical failure Loss of one primary pump pony motor with Nate 4 Note 4 Note 4 Note 4 Note 4 -- Note 4 Note 4 Note 4 failure of check valve In that loop to shut E-6 Design basis steam generator sodlum/ - Note 4 -- Note 4 - Note 4 Note 4 water reaction E-7 One loop natural circulation (from Note 4 Note 4 - - -- Note 4 Note 4 Initial two loop operation) E-15 DHRS Activation 24 Hours After Scram 2 2 -- 2 2 2 2 E-16 Three loop natural circulation Note 4 Note 4 Note 4 -- - Note 4 Note 4 Notes: 4. Each component, or part of a component, must accommodate 5 occurrences of the most severe emergency transient for that component or part of a component (one every 6 years) and two consecutive occurrences of the most severe event (or of unlike events if consecutive occurrences of unlike events provide a more severe of f act than two occurrences of the most severe event).

5. See Paragraph 5.7.3(c) l kN nm

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l O Figure 5.7-6a Average Channel Sodium Exit Temperature Top of Active Core vs.  ! Time for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater) l 1200 1100 - 1000 - C D 000 - E 000 - 700 - A _ 500 - 500 0 1000 2000 TIME (SECONDS) O 5.7-14a Amend. 72 Oct. 1982

Figure 5.7-6b Maximum Channel Sodlum Exit Temperature, Top of Active Core for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). 1500 ) 1 1400 - 1300 - 1200 - F~ 1100 u. E a 1000 a:

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000 700 - x 600 - 20 0 1000 2000 TIME (SECOND3) i O 5.7-14b Ame'd. n 72 Oct. 1982

Figure 5.7-60 Blanket Hot Channel Sodium Outlet Tenperature for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both O Evaporators and the Superheater). 1500 1400 - 1300 - l 1200 - - p 1100

           ~

w E Q 1000 - O i# 000 - i j 300 - .i 700 - N _ 600 - I s00 l 0 1000 2000 TIME (SECONDS) O Amend. 72 5.7-14c Oct. 1982

v Figure 5.7-6D Reactor Vessel Exit Temperature for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). 1200 1100 - 1000 - a c 300 _ _ E E E 800 - h 700 x 600 - 500 0 1000 2000 TIME (SECONDS) 1 5.7-14d Amend. 72 Oct. 1982

i Figure 5.7-6E Affected Loop Superheater Sodium inlet Temperature for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). l l l l l i 1200 1 I i 1100 - 1000 - C L 300 - E E O 000 - 700 - 500 - l m 0 1000 2000 TIME (SECONDS) l Amend. 72 5.7-14e Oct. 1982

l Figure 5.7-6F Af fected Loop Evaporator Sodium inlet Temperature f or Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). l l 1200 1100 - 1000 - I

        ;    900     -

f g r I 800 3 l i 700 600 - 1 500 2000 1000 0 TIME (SECONDS) l Amend. 72 5.7-14F Oct. 1982

Figure 5.7-6G Affected Loop Evaporator Sodium Exit Temperature for Loss of O Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporatcrs and the Superheater).  ; 1000 000 - - 800 - C t E o O ie  ; e 600 - 500 - 400 O 1000 2000 TIME (SECONDS) O Amend. 72 5.7-14G Oct. 1982

Figure 5.7-6H Intermediate Pump Sodium Temperature Vs. Time for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). 1000 900 - 800 - - C E E 700 - e m E 3 600 - 500 - 400  ! 0 1000 2000 TIME (SECON DS) O Amend. 72 5.7-14H Oct. 1982 l

l l Figure 5.7-61 Af f acted Loop Drum Steam Temperature 'T Loss of Steam I O Generator Loed (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). l 4 1 i m3  ; I I i 700 - 600 - C O E m 4 m 500 - 1 t' m - W - 400 -

• j 300 -

200  ! l 0 1000 2000 TIME (SECONDS) O Amend. 72 5.7-141 Oct. 1982

1 Figure 5.7-6J Af fected Loop Evaporator inlet Water Temperature for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the Superheater). 800 l 700 - l 600 - i C

        ?

E

        =

400 - 300 - i 200 O 1000 2000 TIME (SECONDS) O Amend. 72 5.7-14J Oct. 1982

I Figure 5.7-6K Af fected Loop Drum Pressure for Loss of Steam Generator Load (Dumping of Water / Steam Sides of Both Evaporators and the l Superheater). 3000 2000 -

                                                                                                       ~

l 5 E e a N i E 1000 i l l o e 1000 2000 j TIME (SECONDS) Amend. 72 5.7-14K Oct. 1982

OiAPTER 6.0 - ENGINEERED SAFETY FEATURES O b TABLE OF CONTENTS PAGE NO. 6.0 ENGINEERED SAFETY FEATURES 6.1-1 6.1 GENERAL 6.1-1 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 Confinement / Containment Functional Destgn 6.2-1 6.2.1.1 Design Bases 6.2-1 6.2.1.2 System Design 6.2-2 l 6.2.1.3 Design Evaluation 6.2-3a l 6.2.1.4 Testing and Inspection 6.2-7 l 6.2.1.5 Instrumentation Requirements 6.2-9 6.2.1.6 Materials 6.2-9 6.2.2 Containment Heat Pemoval 6.2-9 b V 6.2.3 Containment Air Purification end Cleanup 6.2-9 l6.2.4 Containment isolation Systems 6.2-10 6.2.4.1 Design Bases 6.2-10 6.2.4.2 Systems Design 6.2-12 6.2.4.3 Design Evaluation 6.2-13 , 6.2.4.4 Tests and Inspections 6.2-14 6.2.5 Annulus Filtration System 6.2-14 6.2.5.1 Design Bases 6.2-14 l 6.2.5.2 System Design 6.2-14 l6.2.5.3 Design Evaluation 6.2-15 6.2.5.4 Tests and Inspections 6.2-15 N,) 6-1 Amend. 72 Oct. 1982

TABLE OF CONTENTS (CONT. ) PAGE NO. O l6.2.6 Reactor Service Building (RSB) Filtration Systm 6.2-16 l 6.2.6.1 Design Basis 6.2-16 l 6.2.6.2 Systm Design 6.2-16 l 6.2.6.3 Design Evaluation 6.2-16 Test and inspection 6.2-17 l 6.2.6.4 6.2.7 Steam Generator Building Aerosol Release 6.2-17 Mitigation System Functional Design 6.2.7.1 Design Bases 6.2-17 6.2.7.2 Syst m Design 6.2-17 6.2.7.3 Design Evaluation 6.2-18 6.2.7.4 Testing 6.2-19 6.2.7.5 Instrumentation Requironents 6.2-19 6.3 HABITABil.lTY SYSTEM 6.3-i 6.3.1 Habitability Systs. Functional Design 6.3-1 6.3.1.1 Design Bases 6.3-1 l6.3.1.2 System Design 6.3-2a 6.3.1.3 Design Evaluation 6.3-4 Testing and inspection 6.3-5 l l6.3.1.4 6.3.1.5 Instrumentation Requirment 6.3-7 6.3.1.6 Ef fects of Sodium Cmbustion Products or Other Toxic Gases on the Habitability System 6.3-7 6.3.1.6.1 Sodium Ccvnbustion Products 6.3-7 6.3.1.6.2 Toxic Gases 6.3-7a O 6-!I Amend. 72 Oct. 1982

TABLE OF CONTENTS (CONT.) PAGE NO. l 6.4 CELL L I NER SYSTEM 6.4-1 1 6.4.1 Design Base 6.4-1 6.4.2 System Design 6.4-1 l 6.4.3 Design Evaluation 6.4-1 6.4.4 Tests and inspections 6.4-1 6.4.5 Instrumentation Requirements 6.4-1  ! 6.5 CATCH PAN 6.5-1 6.5.1 Design Base 6.5-1 6.5.2 System Design Description and Evaluation 6.5-1 - 6.5.3 Tests and inspections 6.5-1 i 6.5.4 Instrumentation Requirements 6.5-1 > t 4 6-il-a Amend. 72 Oct. 1982

 ,,,-c .

LIST OF TABLES b PAGE NO. TABLE NO. 6.1-1 List of Engineered Safety Features in CRBRP 6.1-2 l 6.2-1 Reactor Containment Design Basis Sodium 6.2-21 Pool Accident l6.2-2 Reactor Containment Design Basis Accident 6.2-22 l 6.2-3 Constituents of Containment Aerosol FoiIowing 6.2-23 Failure of in-Containment Primary Sodium Storage Tank During Maintenance l 6.2-4 Summary Description of Heat Sihks Used for 6.2-24 Containment Pressure / Temperature Analysis l 6.2-5 Lines Penetrating Containment 6.2-25 l 6.2-5A Summary of Containment isolation Valving 6.2-34 Categorles and Applicable GDC l 6.2-6 Table of Bypass Leak Paths 6.2-35 6.3-1 Conformance of the Control Room Filtration 6.3-8 System With Respect to Each Position of U.S. NRC Regulatory Guide 1.52 6.3-2 Free Air Space Volume Serviced by The Control 6.3-17 Room Emergency Ventilation System , 6. 3-3 On Site Toxic Material Storage 6.3-18 O 6-1ii Amend. 72 Oct. 1982

LIST OF FIGURES FIGURE NO. Page b l 6.2-1 Sodium Burning Rate-Primary Sodium In-Containment 6.2-28 Storage Tank Faiiure During Malntenance l 6.2-2 Containment Gas Pressure-Primary Sodium In- 6.2-39 Containment Storage Tank Failure During Maintenance l 6.2-3 Containment Gas Tanperature-Priraary Sodium In- 6.2-40 Containment Storage Tank Failure During Maintenance l 6.2-4 Containment Wal l Tanperature-Primary Sodium In- 6.2-41 Containment Storage Tank Failure During Maintenance l 6.2-5 Containment Aerosol Concentration-Primary Sodium 6.2-42 In-Containment Storage Tank Failure During Maintenance l 6.2-6 Gas Pressure in Storage Tank Cell Following 6.2-43 Primary Sodium Storage Tank Failure in Containment During Maintenance l 6.2-7 Gas Temperatu-e in Storage Tank Cel l Fol low ing 6.2-44 Primary Sodium Storage Tank Failure in Containment During Maintenance l 6.2-8 Cell Liner (hall) Tanperatura in Storage Tank Cell 6.2-45 foiiowing Primary Sedium Storage Tank Faiiure In Cortaintrent During Maintenance l6.2-9 Coli Liner (FIocr Surface) Temperature FcllowIng 6.2-46 Primary Sodium Storage Tank Failure in Contaltunent During Maintenance l6.2-10 Containment isolation Valve Conf Igurations 6.2-47 O 6-iv Amend. 72 Oct. 1982

LIST OF REFERENCES PAGE N,(L References to Section 6.2 6.2-20 1 1 1 1 l l

                                                        )

i I l l l 1 l l l 1 l l

                                                        )

i Amend. 72 ) 6-v Oct. 1982 l i I

The argon and nitrogen supply line valves provide a double barrier which is automatically activated on loss of the ex-containment boundary. The valves

 /'    and associated actuators are located in protected areas and are testable.

k )/ ss Remote and local manual Initiations are provided. The nitrogen exhaust line to CAPS has two automatically initiated valves. The valves provide two barriers following closure. The valves and associated actuators are located in protected areas and are testable. For the renainder of the penetrations, two valves are provided as barriers to release. Manual Initiation will be adequate to prevent releases exceeding the guideline values. For lines of closed systems penetrating containment, one isolation valve located outside of containment as close as practical to containment is provided. A single valve meets Criteria 48 and provides the necessary capability to limit the release of activity. For lines which do not contain radioactive fluids, the closed system provides the first boundary while the Isolation valve provides the second boundary to release of activity. Therefore, in all cases, there are two boundarles which ef fectively limit the release of activity from a postulated event. The valves and associated actuators are located in protected areas and are testable. Manual initiation of isolation is provided. i 6.2.4.4 Tests and Insoections The periodic test capability is described in Section 7.3. 6.2.5 Annulus Filtration System n\ 6.2.5.1 Design Bases The Annulus Filtration Systen is designed to ensure than an accepteble upper limit of leakage of radioactive material is not exceeded under the site suitability source term conditions. , The functional design and evaluation of the Annulus Filtration System is based upon the site suitability source term, as identified in Section 15.A. The design capability of the annulus filtration system as described in the following section will provide a large margin of safety over the containment design basis accident identified in Table 6.2-1. A 6.2.5.2 System Design The RCB annulus filter system design shall satisfy the following criteria: (1) The containment / confinement annulus space shall be maintained under 1/4 inch W. G. negative pressure during normal plant operation and accident conditions. < (2) Capability shall be provided to filter the containment / confinement annulus exhaust during normal operation. Amend. 64 Jan. 1982 < 6.2-14 1

(3) Capabii Ity shalI be provided to filter the RG ventiIatIon exhaust alr through the annul us f II ter system during ref uel ing operations, when the RG/RS8 ref uel ing hatch is open. s ,) Capabil ity shall be provided to f ilter and recircul ate the annul us air during accident conditions. For every 1000 CFM f il tered exhaust air (required for the maintenance of 1/4 in. W. G. negative pressure) not less than 3500 CFM air shall be recircul ated through the f liters. (5) The recircul ating duct system shall be designed to accompl ish proper mixing in the annul us in accordance w ith USNRC Standard Review Pl an Section 6.5.3. (6) The Annul us Fil tratton System shalI f ul ly compiy with USNRC Regul atory Guide 1.52. (7) The filter system shall be designed to achieve a minimum of 99% particul ate and 95% adsorbent ef fIclency. Radiation monitoring equipment associated w Ith the annul us f Il tratIon system is described in Section 12.2. By maintaining the annul us at a minimum of 1/4" water gauge negative pressure w ith respect to the outside atmosphere, the bypass leakage (that f raction of annul us radioactivity which leaks f rom the conf inement buil ding w ithout being f iltered) can be maintaired at less than 1%. 6.2.5.3 Design Evaluation The Annulus Filtration System features of the design provide the necessary assurance ihat the radioactivity released as a result of the site suitabil ity source term wilI not exceed the guidelinos of 10CFR100. The annul us pressure n.aintenance f ans have been sized at ?000 CFM, which has conservatively been deTcrmined to be greater than the tctal leakage into the annul us f ran al l sources, incl uding the dampers (vents and cap) provided at the top of the Conf inement Buil ding and the dampers provided at ihe 816' - 0" I elevation f or the Annul us Cool ing System (leakage based on a negative 1/4" I w.g. pressure). Analysis wilI be conducted to substantlate that the annulus space wIlI rmain under a 1/4" W.G. negative pressure considerirg the of fects of heat transf er, barometric pressure change, Inl eakage and w ind l oads. The resul ts of th is analysis will be provided in the FSAR. Two 100% redundant f il ter-f an units consisting of a demister, heating coll, pref 11ter bank, HEPA filter bank, adsorber bank, pressure maintenance and exhaust f an, annul us recircul ation f an, wIth associated ductwork and accessories, are provided f or the annul us exhaust, recircul ation and filtering. This insures that no single active f all ure wilI prevent 100% operation of the annul us f il tration system. The Annul us Fil tration System is described i n Sect ion 9.6.2.2.4. O 6.2-15 Amend. 72 Oct. 1982

6.2.5.4 Tests and insoections The annulus filtration system shall be tested per the requirements of Regul atory Guide 1.52. Containment penetrations shall be tested per Appendix J to 10CFR50 in order to verify bypass leakage assumptions used for radiological accident analyses. O O 6.2-15a Pmend. 72 Oct. 1982

6.2.6 Reactor Service Building (RSB) Filtration System V 6.2.6.1 Design Basis The RSB filtration system is designed as an Engineered Safety Feature (ESF) to filter the RSB exhaust air in order to mitigate the consequences of the Site Sultabil Ity Source Term (SSST) event. The system is designed to f unction continuously. 6.2.6.2 System Design l The RSB is maintained at a minimum 1/4" negative water gauge pressure as descri bed i n Section 9.6.3.1.1. 1 The RSB Filtration System is used and designed to maintain the RSB at a l minimum of 1/4" negative water gauge pressure and filter the RSB exhaust under l < all conditions except when the railroad door is open. A network of ducting is utilized in supplying and exhausting air to various floor elevations and/or cel l s in the RSB. This mode of operation exhausts 18,000 CFM of air through l the missile protected exhaust on the Reactor Service Building (RSB). During accident conditions the RSB Filtration System will automatically shif t to an alr recirculation node of operation exhausting that amount of alr (41700 CFM) required to maintain a minimum of 1/4" negative water gauge pressure. The f Il ter sy stem w ll l be designed as a Safety Class 3 system and w11l meet

                 tne requirecents of Regulatory Guide 1.52. The f Il ter sy stem w il I be designed to achievo a minimum of 99% particulate and 95% adsorbent ef ficiencies.

6.2.6.3 Design Evaluation The RS8 filter system is designed to filter 18,000 CFM of air of which 1700 CFM oi air is exhausted while 16,300 CFM of air is recirculated during accident condi tions. The exnausted air is designed to of fset building in leakage air while naintaining 1/4" negative water gauge pressure. Analysis will be conducted to substantiate that the RSB will remain under a 1/4" W.G. negative pressure considering the ef fects of heat transfer, barcmetric pressure change, Inleakage and wind loads. The results of this analysis will be provided ir. the FSAR. Two (2) 100% redundant filter f an units consisting of a demister, heating coII, pre-fil ter bank, adsorber bank, HEPA f il ter bank, cleanup f il ter f an, with associated ductwork and accessories, are provided for the RSB exhaust, recircul ation, and f il tering. This insures that no single active f ailure will prevent 100% operation of the RSB filtration system. The RSB filtration 4 system is described in Section 9.6.3.1.1. The system ducting is designed to exhaust air f rom all potentially radioactive areas. Capabil ity ext sts to Isolate the supply and exhaust alr fIow to the areas where an accident has occurred and to maintain these areas at a greater i O(/ negative pressure than other areas. l 6.2-16 Amend. 72 Oct. 1982

This capability is designed to prevent the spread of airborne radioactivity f rom contaminated to cl ean areas w ithin the buil di ng. O O 6.2-16a Amend. 72 Oct. 1982

n 3 6.2.6.4 Test and'Insoection The RSB filtration system w11I be tested per the requirements of Regulatory Guide 1.52. Visual inspection will be conducted on installation. 6.2.7 STEAM GENERATOR BUILDING AEROSOL RELEASE MITIGATION SYSTEM FUNCTIONAL DESIGN 6.2.7.1 Design Bases The Steam Generator Building Acrosol Release Mitigation System is designed to assure that release of a maximum of 630 lbs of sodium aerosols f rom the Steam Generator Building is not exceeded in the event of a design basis leak in one of the three loops in lHTS piping. This limit is obtained by releasing through a controlled vent area a maximum of 440 lbs of aerosols during the first five minutes of ihe accident. Between 5 minutes and 5000 seconds, 90 lbs of aerosols may be released through building cracks. Beyond 5000 seconds, an additional 100 lbs of aerosols could be released through building cracks. A release of aerosols through the controlled vent area stack is required to maintain large (360 bullp)ing Ft overpressures steam vent louvers.below the 0.7 psig sotpoint for opening of the The functional design and evaluation of the SGB Aerosol Release Mitigation Festures are based upon the design basis accident described in Section 15.6.1.5 of 1ho PSAR. 6.2.7.2 . System Design Controlled relesse of aerosols frou the Steam Generator Buildir.g (SGB) is accomplished by closure cf SGB HVAC cutlets and venting through a controlled area vent stack, t:oth actions being Initiated f rom either of a r(dundant set of saf ety-related aerosol smoke detectors located in the SGB HVAC exhaust stack. Aerosols are released from the controlled area vent stack for five

!      minutes to assure until peak pressures in the SGB are within acceptable limits, at which time the vent path is closed to the external atmosphere.

The SGB Aerosol Release Mitigation Features consist of redundant sets of l saf ety-related aerosol detectors (see Section 9.13.2) located in each SGB loop j HVAC exhaust duct, redundant rollef dampers to each loop controlled area vent

;       stack, and redundant closure dampers in each controlled area vent stack.

Each aerosol detector set consists of three detectors provided power by three 1E uninterruptible power sources. These detectors gip when the sodiun v aerosol concentration in the SGB HVAC exhaust is 10 gm/cc. Whenjwoofthe three detectors in either set sense an aerosol concentration of 10 gm/cc, a signal is provided to activate the I&C logic for the SGB aerosol release mitigation features. Within 10 seconds of receipt of an aerosol detection signal, the SGB buil ding HVAC system will be closed to the outside atmosphere, the relief dampers to the controlled vent area will open, the controlled vent area closure devices will ranain in their normally open position, and the g remaining nuclear Island building (RCB & RSB) HV AC systems w il l be cl osed to the outside atmosphere. The controlled vent area closure devices close five tLj minutes af ter receipt of the trip signal from the aerosol detectors, with a 6.2-17 Amend. 64 Jan. 1932

 /                               CHAPTER 7.0     INSTRUMENTATION AND CONTROLS

(, TABLE OF CONTENTS PAGE NO.

7.1 INTRODUCTION

7.1 -1 7.1.1 Identification of Safety Related Instrumentation 7.1-1 and Control Systems 7.1.2 identif ication of Saf ety Criteria 7.1 -1 7 .1. 2.1 Design Basis 7.1 -2 7.1.2.2 Independence of Redundant Saf ety Rel ated Systems 7.1 -3 7.1.2.3 Phy sical identif ication of Saf ety Rel ated Eq ui pment 7.1-4 7.1.2.4 Conf ormance to Regulatory Guides 1.11 7.1 -4

                          " Instrument Lines Penetrating Primary Reactor Containment" and 1.63, " Electric Penetration Assembi les in Containment Structures f or Water-Ccol ed Nuct car Power Pl ants" s_-    7.1. 2. 5        Conf ormance to IEEE No. 323 "lEEE Standard f or        7.1-4 Qual ifying Cl ass IE Equipment f or Nuclear Power Generating Stations" 7.1. 2. 6        Ccnformance to IEEE No. 336 " Installation, i ns pec- 7.1 -4 tion and Testing Requirements for instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations" l 7.1.2.7           Conf ormance to IEEE No. 338-1971 " Periodic Testing of              7.1-5 Nuclear Power Generating Station Protection Sy stem" 7.1.2.8          Conf ormance to Regul atory Guide 1.22 " Periodic          7.1 -5 Testing of Protection System Actuation Functions" 7.1. 2.9         Conf ormance ,o Regul atory Guide 1.47 " Bypassed         7.1 -6 and inoperable Status Indication f or Nuclear Power Pl ant Saf ety Systems" 7.1.2.10         Conf ormance to Regul atory Guide 1.53 " Appl Ication       7.1 -6 of the Single Failure Criterion to Nuclear Power Pl ant Frotection Systems" O

7-1 Amend. 72 Oct. 1982

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

TABLE OF CONTENTS (CONT.) PAGE NO. 7.1.2.11 Conf ormance to Regul atory Guide 1.62 " Manual 7.1 -6 initiation of Protective Functions" 7.1.2.12 Regul atory Guide 1.89 " Qual if ication of Cl ass IE 7.1 -6 a Equipment f or Nucl ear Pcwer Pl ants" 7.1.2.13 1 & E inf ormation Notice 79-22 " Qual if Ication of 7.6-6a Control Sy stems 7.2 REACTOR SHUTDOWN SYSTEM 7 . 2-1 7.2.1 Descri pt ion 7.2-1

7. 2.1.1 Reactor Shutdown System Description 7.2-1 7 . 2.1. 2 Design Basis Inf ormation 7.2-6 7 . 2.1. 2.1 Primary Reactcr Shutdown System Subsystems 7.2-7 7.2.1.2.2 Secondary Reactor Shutdown System Subsystems 7.2-9

7. 2.1. 2. 3 Essential Perf ormance Requirements 7.2-11 7.2.2 Anal y si s 7.2-13 7.3 ENGINEERED. SAFETY FEATURE INSTRUMEPHATION AND C0ffTROL 7.3-1 7.3.1 Containment isolation System 7.3-1 7.3.1.1 System Description 7.3-1 7.3.1.2 Design Basis Information 7.3-2 7.3.1.2.1 Containment Isolation System Subsystems 7.3-2 7.3.1.2.2 Essential Perf ormance Requirments 7.3-3 7.3.2 Anal y si s 7. 3-3 7.3.2.1 Functional Perf ormance 7.3-3 7.3.2.2 Design Features 7. 3-3 l 7.4 INSTRUMENTATION AND CONTROL SYSTEMS 7.4-1 REOUIRED FOR SAFE SHUTDOWN 7.4.1 Steam Generator Auxil iary Heat Removal 7.4-1 instrumentation and Control Sy stem 7-il Amend. 72 Oct. 1982

TABLE OF CONTENTS (Cont.) - PAGE NO. 7.4.1.1 Design Description 7.4-1 7.4.1.1.1 Function 7.4-1 7 . 4.1.1. 2 Equipment Design 7.4-1 s 7 . 4 .1.1. 3 Initiating Circuits 7.4-3 7.4.1.1.4 Bypasses and Interlocks 7.4-3 7.4.1.1.5 Redundancy / Diversity 7.4-4 7.4.1.1.6 Actuated Devices 7.4-4 7.4.1.1.7 Testabil Ity 7.4-4 7.4.1.1.8 Separation 7.4-4 7.4.1.1.9 Operator Inf ormation 7.4-5 7 . 4.1. 2 Design Analysis 7.4-6 7.4.2 Outlet Steam isolation Instrumentation 7.4-6 and Control System 7.4.2.1 Design Description 7.4-6 7.4.2.1.1 Function 7.4-6

7. 4 . 2.1. 2 Equipment Design 7.4-7 7.4.2.1.3 Initiating Circuits 7.4-7 7.4.2.1.4 Bypasses and Interlocks 7.4-7 7.4.2.1.5 Redundancy and Diversity 7.4-7 7.4.2.1.6 Actuated Device 7.4-8 7.4.2.1.7 Separation 7.4-8 7.4.2.1.8 Operator Inf ormation s 7.4-8 7.4.2.2 Design Analysis 7.4-8 7.4.3 Pony Motors and Control s

• 7.4-8 7.4.3.1 Design Description 7.4-8a O Amend. 72 7-III Oct. 1982 a ,, _ _ - _ _ - --.

TABLE OF CONTENTS (Cont. ) PAGE NO. 7.4.3.2 initiating Circuits 7.4-8a 7.4.3.3 Bypasses and Interlocks 7.4-8a 7.4.3.4 Araly ses 7.4-8a l 7.4.4 Remote Shutdown System 7.4-8b l 7.4.4.1 Design Description 7.4-8b l 7.4.4.1.1 Function 7.4-8b l 7.4.4.1.2 Design Bdsis 7.4-8b l 7.4.4.1.3 Remote Shutdown Operations 7.4-8c l7.4.4.1.4 Equipment Design 7.4-8d l 7.4.4.2 Design Analysis 7.4-8f 7.5 INSTRUMENTATION AND MONITORING SYSTEM 7.5-1 7.5.1 Fl ux Monitoring System 7.5-1 7.5.1.1 Design Description 7.5-1 7.5.1.1.1 Source Range 7.5-2 7.5.1.1.2 Wide Range 7.5-3b 7.5.1.1.3 Power Range 7.5-3b 7.5.1.2 Design Analysis 7.5-4 7.5.2 Heat Transport Instrumentation System 7.5-5 7 . 5 . 2.1 Description 7. 5 -5 7 . 5 . 2.1.1 Primary and Intermediate Sodium Loops 7. 5 -5 7 . 5 . 2.1. 2 Sodium Pumps 7.5-8 7 . 5 . 2.1. 3 Steam Generator 7.5-9 7.5.2.2 Analy si s 7.5-12 7.5.3 Reactor and Vessel Instrumentation 7.5-13 7.5.3.1 Description 7.5-13 7-Iv Amend. 72 Oct. 1982

TABLE OF CONTENTS (Cont.) 7.5.3.1.1 Sodium Level 7.5-13 7.5.3.1.2 Temperature 7.5-13 7.5.3.1.3 Non-Rept aceabl e instruments 7.5-13 7.5.3.2 Analy sis 7.5-14 1 7.5.4 Fuel Fail ure Monitoring System 7.5-14 7.5.4.1 Design Description 7.5-15 7.5.4.1.1 Cover Gas Monitoring Subsystem 7.5-15 l 7.5.4.1.2 Reactor Del ayed Neutron Monitoring Subsystem 7.5-16 l l 7.5.4.1.3 Failed Fuel Location Subsystem 7.5-17 7.5.4.1.4 Tests and Inspection 7.5-17 7.5.4.2 Design Analysis 7.5-18 7.5.5 Leak Detection Systems 7.5-18 7.5.5.1 Sodium to Gas Leak Detection System 7.5-18 7.5.5.1.1 Design Bases and Design Criteria for the 7.5-18a Liquid Metal - to - Gas Leak Detection Sy stems 7.5.5.1.1.1 Design Description 7.5-19 7.5.5.1.2 Design Analysis 7.5-22 7.5.5.2 Intermediate to Primary Heat Transport 7.5-24 System Leak Detection 7.5.5.2.1 Design Description 7.5-24 l 7.5.5.2.2 Design Analysis 7.5-25 7.5.5.3 Steam Generator Leak Detection System 7.5 -25 7.5.5.3.1 Design Description 7.5 -26 7.5.5.3.2 Design Analysis 7.5-27 a 7.5.6 Sodium-Water Reaction Pressure Rel lef 7.5-30 System (SWRPRS) Instrumentation and Control O d l 7.5.6.1 Design Description 7.5-30 7-v Amend. 72 Oct. 1982

TABLE OF CONTENTS (Cont.) PAGE NO. 7.5.6.1.1 Function 7.5-30a 7.5.6.1.2 SWRPRS Trip Logic 7.5-30a 7.5.6.1.3 Bypasses and Interlocks 7.5-32 7.5.6.1.4 Sodium Dump 7.5-32 7.5.6.1.5 Monitoring instrumentation 7.5-32 7.5.6.1.6 Sodium Dump Tank Instrumentation 7.5-33 7.5.6.1.7 Water Dump Tank Instrumentation 7.5-33 7.5.6.2 Design Analysis 7.5-33a 7.5.7 Containment Hydrogen Monitoring 7.5-33b 7.5.7.1 Design Description 7.5-33b 7.5.8 Containment Vessel Temperature Monitoring 7.5-33b 7.5.8.1 Design Description 7.5-33b 7.5.9 Containment Pressure Monitoring 7.5-33b 7.5.9.1 Design Description 7.5-33b 7.5.10 Containment Atmosphere Tstnperature 7.5-33c 7.5.10.1 Design Description 7.5-33c 7.5.11 Post Accident Monitoring 7.5-33c 7.5.11.1 Description 7.5-33c 7.5.11.2 instrumentation Design and Qual if ication 7.5-33 d 7.5.11.2.1 Category 1 7.5-33d 7.5.11.2.2 Category 2 7.5-33f 7.5. 1.2.3 Category 3 7.5-33g l 7.5.11.2.4 General Requirements to Category 1,2, and 3 7.5-33g 7.5.11.3 Instrument identif ication 7.5-33h 7.5.12 Inoperable Status Monitoring System 7.5-33i 7-va Amend. 72 Oct. 1982

rm PAGE NO. U 7.5.12.1 Design Description 7.5-33i 7.5.12.2 Design Analysis 7.5-33i 7.6 OTHER INSTRUMENTATION AND CONTROL SYSTEMS REOUIRED FOR SAFETY 7.6-1 7.6.1 Emergency Plant Service Water instrumentation 7.6-1 and Control Systems 7.6.1.1 Emergency Plant Service Water system (EPSW) 7.6-1 7.6.1.2 Design Criteria 7.6-1 7.6.1.3 Design 7.6-2 7.6.1.3.1 Control System 7.6-2 7.6.1.3.2 Monitoring Instrumentation 7.6-2 7.6.1.3.3 Inputs to PDH & DS 7.6-2a 7.6.1.3.4 Design Analysis 7.6-2a 7.6.2 Emergency Chilled Water (ECW) System 7. 6-2b 7.6.2.1 Design Criteria 7.6-2b 7.6.2.2 Design 7.6-2c 7.6.2.2.1 Control System 7.6-2c 7.6.2.2.2 Monitoring instrumentation 7.6-2c 7.6.2.2.3 Inputs to PDH & DS 7.6-2e 7.6.2.2.4 Design Analysis 7.6-2e 7.6.3 Direct Heat Removal Service instrumentation 7.6-3 and Control 7.6.3.1 Design Description 7.6-3 7.6.3.1.1 Function 7.6-3 7.6.3.1.2 Design Criteria 7.6-3 7.6.3.1.3 Equipment Design 7.6-3a 7.6.3.1.4 initiating Circuits . 7.6-3c 7.6.3.1.5 Bypass and Interlocks 7. 6-3c 7-vi Amend. 72 Oct. 1982

I PAGE NO. 7.6.3.2 Design Analysis 7.6-3d 7.6.4 Heating, Ventilating, and Air Conditioning Instrumentation and Control System 7.6-4 7.6.4.1 Design Criteria 7.6-4 7.6.4.2 Design Description 7.6-5 7 . 6. 4 . 2.1 Control System 7.6-5 7.6.4.2.2 Monitoring Instrumentation 7.6-7 7.6.4.3 Design Analysis 7.6-8 7.6.5 SGB Flooding Protection System 7. 6 -9 7.6.5.1 Design Basis 7.6-9 7.6.5.2 Design Requirements 7.6-9 7.6.5.3 Design Requirements 7.6-9 7.6.5.3.1 Instr umentation 7.6-9 7.6.5.3.2 Control s 7.6-9 7.6.6 Recircul atIng Gas Cool ing (RGC) 7.6-10 instrumentation and Coltron System 7.6.6.1 Design Criteria 7.6-10 7.6.6.2 Design 7.6-11 7 . 6. 6. 2.1 Control System 7.6-11 7.6.6.2.1.1 Saf ety-Rel ated Subsystem Operation 7.6-11 7.6.6.2.1.1.1 Fan Operation 7.6-11 7 . 6 . 6. 2.1.1. 2 Automatic isol ation Val ve Operation 7.6-13 7 . 6 . 6 . 2.1.1. 3 Drain Val ve Operation 7.6-13 7.6.6.2.1.1.4 Chil Ied Water Val ve Operation 7.6-13 7.6.6.2.1.2 Saf ety-Rel ated Subsystem EB 7.6-14 7.6.6.2.2 Monitoring Instrumentation 7.6-14 7.6.6.2.3 Inputs to PDH & DS 7.6-15 O 7-vla Amend. 72 Oct. 1982

(g PAGE NO. t

   ") l7.6.6.2.4     Design Analysis                                   7.6-16 7.7         INSTRUMEN'TATION AND CONTROL SYSTEMS NOT          7.7-1 REOUIRED FOR SAFETY 7.7.1       Pi ant Controf System Description                 7.7-1 7.7.1.1     Supervisory Control System                        7.7-2       j 7.7.1.2     Reactor Control System                            7.7-3       l 7.7.1.3     Primary and Secondary CRDM (Control Od Drive      7.7-4 Mechanism) Controller and Rod Position Indication 7.7.1.3.1   Primary CRDM Control                              7.7-4 7.7.1.3.2   Rod Position Indication System                    7.7-6 7.7.1.4     Sodium Flow Control System                        7.7-7 7.7.1.5     Steam Generator Steam Drum Level Control System   7.7-8 l7.7.1.5.1     Feedwater Flow Control Valve Control              7.7-8 7.7.1.5.2   Main Feedwater Isolation                          7.7-9 l7.7.1.5.3    Operational Considerations                        7.7-9 7.7.1.6     Recircul ation Flow Control System                7.7-10 7.7.1.7     Sodium Dump Tank Pressure Control System          7.7-10 7.7.1.8     Steam Dump and Bypass Control System              7.7-11

> 7.7.1.9 Fuel Handl ing and Storage Control System 7.7-12 7.7.1.10 Nuclear isi and Auxil lary Instrumentation 7.7-15 and Control Systems 7.7.1.11 Bal ance of Pl ant Instrumentation and Control 7.7-15a 7.7.1.11.1 Treated Water instrumentation and Control System 7.7-15a 7.7.1.11.2 Waste Water Treatment Instrumentation and 7.7-16 Control System 7.7.1.11.3 Remaining Systems 7.7-16 7.7.2 Design Analysis 7.7-16 g 7.7.2.1 Supervisory Control System 7.7-17 7-vil Amend. 72 Oct. 1982

PAGE NO. 7.7.2.2 Reactor Control Sy stem 7.7-18 7.7.2.3 Sodium Flow Control System 7.7-18 7.7.2.4 Steam Generator Feedwater Flow Control System 7.7-19 7.7.2.5 Bal ance of Pl ant Instrumentation and Control 7.7-19 7.8 PLANT DATA HANDLING AND DISPLAY SYSTEM 7.8-1 l 7.8.1 Design Description 7.8-1 i 7.8.2 Design Analysis 7.8-2 7.9 OPERATING CONTROL STATIONS 7.9-1 7.9.1 Design Basis 7.9-1 7.9.2 Control Room 7.9-1 7 .9 . 2.1 General Description 7.9-1 7.9.2.2 Control Rocrn Arrangement 7.9-2 7.9.2.3 Main Control Board Arrangement 7.9-2 7.9.2.4 Main Control Board Design 7.9-5 7.9.3 Local Control Stations 7.9-6 7.9.4 Communications 7.9-6 7.9.5 Design Eval uation 7.9-6 7.9.5.1 Planning Phase 7.9-6 7.9.5.2 Rev iew Phase 7.9-6a 7.9.5.3 Assessment and Implementation Phase 7.9-6b l 7.9.5.4 Conclusions 7.9-6c Amend. 72 O 7-vill Oct. 1982

I q LIST OF TABLES TABLE NO. PAGE NO. 7.1 -1 Safety Related instrumentation end Control Systems 7.1-7 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 Rei ated instrumentation and Control Systems l 7.1-4 Deleted 7.1-5 Del eted 7.1-6 Del eted 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 Perf ormance Requirements f or 7.2-23 PPS Instrumentation m) 7.2-4 List of IEEE Standards Applicable to the 7.2-23a Reactor Shutdown System Logic 7.3-1 Containment Isolation System Design Basis 7.3-5 7.3-2 List of IEEE Standards Appl Icable to the 7.3-5a Containment isolation System Logic 7.4-1 Sequence of Decay Heat Removal Events 7.4-9 7.4-2 SGAHRS Naninal Set Points 7.4-10a 7.4-3 List of IEEE Standards Applicable to SGAHRS 7.4-10d and OSIS instrumentation and Control Systems 7.5-1 Instrumentation System Functions and 7.5-34 Summary 7.5-2 Reactor and Vessel Instrumentation 7.5-39 7.5-3 Summation of Sodium / Gas Leak Detection 7.5-40 Methods 7.5-4 Safety Functions and Primary Systems Monitored 7.5-42 by ISMS 7-Ix Amend. 72 Oct. 1982

LIST 0" TABLES (Cont.) TABLE NO. PAGE NO. Symbol s 7.6-17 l 7.6-1 7.6-2 List of IEEE Standards Applicable to Emergency 7.6-18a I Pl ant Service Water, Emergency Chil led Water, l HV AC, and Recircul ating Gas Instrumentation and Control Systems

7. 6-3 List of IEEE Standards AppiIcable to SGB 7.9-18b FIooding Protection Subsystem Use of Ref uel ing Interlocks 7.7-19a l 7.7-1 7.9-1 Control Roctn Arrangements 7.9-8 O

O 7-Ixa Amend. 72 Oct. 1982 [

LIST OF FIGURES [Vh FIGURE NO. PAGE NO. 7.2-1 Reactor Shutdown System 7.2-24 l 7.2-2 HTS Pump Breaker Logic Diagram 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 7.2-2B Primary PPS Logic Diagram 7.2-28 7.2-2C Typical Secondary PPS Instrument Channel 7.2-29 Logic Diagram 7.2-2D Secondary PPS Logic Diagram 7.2-30 7.2-3 Typlcal Primary Subsystem 7.2-31 7.2-4 Typical Secondary Subsystem 7.2-32 7.2-5 Functional Block Diagrams of the Fl ux-Del ayed 7.2-33 FI ux, High FI ux, FI ux-Pressure, and Reactor Vessel Level Protective Subsystems 7.2-6 Functional Block Diagrams of the HTS Pump 7.2-34 Frequency and Pump Speed Mismatch Protective Systems 7.2-7 Functional Block Diagrams of the IHX Primary 7.2-35 Outlet Temperature and Steam to Feedwater FIow Mismatch Protective Subsystems 7.2-8 Functional Block Diagrams of the Flux-Total 7.2-36 Flow, Startup Nuclear, Modifled Nuclear Rate, and Primary to Intermediate Flow Rate Protective Subsystems 7.2-9 Functional Block Diagrams of the Steam 7.2-37 l Drum Level and HTS Pump Voltage 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 Isol ation System Biock Diagrun 7.3-6 O 7-x Amend. 72 Oct. 1982 l

LIST OF FIGURES (Cont.) FIGURE NO. PAGE'NO. 7.3-2 Conteinment Selection System Logic Diagram 7.3-7 7.4-1 SGAHRS Initiation Logic 7.4-11 7.5-1 CRBRP Fl ux Monitoring System Block Diagram 7.5-43 7.5-2 00RP Fl ux Monitorir.g System Instrument Range 7.5-44 Coverage l 7.5-3 Fuel Fail ure Monitcring System 7.5-45 7.5-4 Main Sodium Streem First Pass Hydrogen 7.5-46 Concentration Change vs. Leak Rate 7.5-4a Main Sodium Stream First Pass Oxygen 7.5-46a Concentration vs. Leak Rate 7.5-5 Hydrogen Concentration vs. Time for Various 7.5-47 Water Leak Rates 7.5-6 SWRPRS Trip & SWRPRS Controlied Isol ation 7.5-48 Val ves Control Logic Diagram 7.6-1 Emergency cool ing Tcwer Fan 7.6-19 7.6-2 Emergency Pl ant Service Water Makeup Pump 7.6 -20 7.6-3 Emergency Pl ant Service Water Pump Start 7 .6 -21 7.6-4 Emergency Pl ant Service Water Pump Stop 7.6-22 7.6-5 Emergency Chil led Water Pumps Logic 7.6 - 23 7.6-6 Emergency Chilled Water Chiller Start 7.6 -24

7. 6-7 Emergency Ch il led Water Chil lers Stop 7.6-25 7.6-8 Emergency Ch il led Water isolation Val ves to 7.6 -26 Secondary Cool ant Loop 7.6-9 Emergency Ch il Ied Water System NW to EOl 7.6 -27 Isolation Val vos 7.6-10 Emergency Chilled Water System Loop A & Loop B 7.6-28 A0V's Normal & Emergency Operation 7.6-11 Saf ety Cl ass Equipmeny V ital Bus Hookup 7.6 -29 7.6-12 Functional Controf Diagram Typlcal HV AC 7.6-30 Exhaust Fan 7-xi Amend. 72 Oct. 1982

LlSI OF FIGURES (Cont.) FIGURE NO. PAGE NO. 7.6-13 Functinal Control Diagram Typical HV AC 7.6-31 Unit Return Fan 7.6-14 Functional Control Diagram Typical Filter 7.6-32 Unit Supply Fan 7.6-15 Functional Contrel Diagram Annul us Cool ing Fan 7.6-33 7.6-16 Functional Control Diagram Containment 7.6-34 Cleanup Scrubber Exhaust Fan 7.6-17 runctional Control Diagram Control Room HV AC 7.6-35 Unit Supply Fan 7.6-18 Functional Control Diagram Diesel Room 7.6-36 Emergency Supply Fan 7.6-19 Functional control Diegram 1 of 2 Redundant 7.6-37 Supply Fans for SGB-IB Air Handl ing Unit 7.6-20 Functional Control Diagram Typical Unit Cooler 7.6-38 _ Serving Cell Containing Saf ety Rel ated Equipment 7 .6 -21 Functional Control Diagram Typical Unit Cooler 7.6-39 Fan Serving Cel l Containing Saf ety-Rel ated Equipment Where Redundant Coolers are Required 7.6-22 Functional Control Diagram Typical Unit Cooler 7.6-40 Fan Serving Cell Containing Containment Cleanup Eq ui pment 7.6 -23 Functional Control Diagram Containment 7.6-41 Purge and Vent Val ves 7.6-24 Functional control Diagram Containment Cleanup 7.6-42 Scrubber Fan Discharge & Bypass Valves 7.6-25 Functional Controf Diagram Controf Room Outside 7.6-43 Air Exhaust & HV AC Unit Outside Air intake Valves 7.6-26 Functional Control Diagram Control Room FiIter 7.6-44 Unit Air intake Valves 7.6-27 Functional Control Diagram Control Room Main 7.6-45 Air Intake Isol ation Val ves 7.6-28 Functinal Controf Diagram Control Rocrn Remote 7.6-46 Air intake isolation Val ves O 7-xil Amend. 72 Oct. 1982

I LIST OF FIGURES (Cont.) FIGURE NO. PAGE NO. 7.6-29 Functional Control Diagram PG Supply & Exhaust 7.6-47 Containment isolation Val ves 7.6-30 Functional Control Diagram Annul us Cool ing 7.6-48 Exhaust & Fan Discharge Dampers 7.6-31 Functional Control Diagram Annul us 7.6-49 FiitratIon Recirculation Dampers 7.6-32 Functional Control Diagram Annul us Fii tration 7.6-50 Exhaust Dampers 7.6-33 Functional Control Diagram RSB Cleanup Discharge, 7.6-51 Exhaust, Decircul ation & Cel l isolation Dampers 7.6-34 Functional Control Diagram Typical Process 7.6-52 Parameter Control of Damper 7.6-35 Functional Control Diagram Diesel Generator 7.6-53 Emergency Supply Fan Tcmperature f bdul ated Dampers 7.6-36 Functional Control Diagram Annul us Pressure 7.6-54 Maintenance Fan Pressure ibdul ated Damper 7.6-37 Functional Control Diagram Typical Flow 7.6-55 Modul ated Vortex Damper 7.6-38 Loop Diagram Recircul ating Gas Cool ing Sys. 7.6-56 Subsy stem bM 7.6-39 Logic Diagram Subsystem MA Supply & Return 7.6-57 I sol at ion V al ves 7.6-40 Logic Diagram Subsystem MA Supply & Return 7.6-58 i sol at ion V al ves 7.6-41 Logic Diagram Subsystem MA Fan 7.6-59 7.6-47 Logic Diagram Subsystem MA Supply & Return 7.6-60 I sol at ion V al ves 7.6-43 Logic Diagram Subsystem MA Cooler Drain Valves 7.6-61 7.6-44 Logic Diagram Subsystem MA Emergency Chilled 7.6-62 Water (to Cooler) I sol ation Val ve 7.6-45 Logic Diagram Subsy stem MA Mal f unction Al arm 7.6-63 7.6-46 Logic Diagram FGCS Saf ety-Pel ated Subsystem 7.6-C? Mal f unct ion Common Al arm 7-xiii Amend. 72 Oct. 1982

LIST OF FIGURES (Cont.) 7.7-1 Plant Control System 7.7-20 7.7-2 Supervisory Control System 7 .7 - 21 7.7-3 Reactor Control 7.7-22 7.7-4 CRDM Controller and Power Train for Primary Rods 7.7-23 7.,7-5 Block Diagram of Primary Rod Group Control 7.7-24 7.7-6 General BIock Diagram for the Rod Misal ignemnt 7.7-25 Rod B lock System 7.7-7 Sodium FIow Control System FIow/ Speed Contrel 7.7-26 7.7-8 Fuel Handl ing and Storage Control System 7.7-27 7.8-1 Plant Data Handl ing and Display System Schematic 7.8-3 7.8-2 Pl ant data Handl ing and Displ ay System Arrangement 7.8-4 ! 7.9-1 Control Roan Layout 7.9-11 l g 7.9-2 Typical Control Panel (Side V lew) 7.9-12 7.9 -3 Main Control Panel Pl an V lew 7.9-13 7.9-4 Typical Control Panel Wiring Layout 7.9-14

                                                                       ?

O 7-xiv Amend. 72 Oct. 1982

LIST OF REFERENCES i Section 7.5 7.5-33J Section 7.9 7.9-6d l O 7-xv Anend. 72 Oct. 1982

7.0 INSTRUMENTATION AND CONTROLS

7.1 INTRODUCTION

This chapter includes a description of the Instrumentation and Control Systems prov ided f or the CRBRP. Particular emphasis is placed on the description of saf ety-related systems, which inct ude the Piant Protection System and the safety-related display instrumentation required to maintain the plant in a saf e shutdown condition. The Plant Protection System includes all equipment to initiate and carry to completion reactor heat transport and balance of plant shutdown, decay heat removal and containment isolation. Saf ety-rel ated display instrumentation assures that the operator has suf ficient information to perf orm required manual saf ety f unctions and monitcr the saf ety status of the pl ant. Major control systems not required for safety are described and analysis is incl uded to demonstrate that even gross f ail ure of those systems does not prevent Plant Protection System action. Analysis is al so incl uded to demonstrate that the requirements of the NRC General Design Criteria, IEEE Standard 279-1971, applicable NRC Regulatory Guides and other appropriate criteria and standards are satisfied. 7.1.1 Identification of Safetv-Related Instrumentation and Control Systems Tabl e 7.1-I lists the Safety-Related instrumentation and Control Systems and i incl udes the def inition of Saf ety-Rel ated Equipment f rom Section 3.2.1. The entire Pl ant Protection System, incl uding the Reactor Shutdown System, the Containment Isolation System and the Shutdown Heat Removal System is saf ety-rel ated. The Reactor Shutdown System input variables are described in Section p 7.2. The Containmen+ 1 solation instrumentetion and Contrcl System is described in Section 7.4 and Section 7.6. The Instrumentation which provides Q signal input to the Plant Protection System is also safety-related and is described in Section 7.5. Safety-Rel ated Displ ay Instrumentation, which assures that the operator has suf f Icient Information to monitor the saf ety status of the plant end maintain it in a safe shutdown condition, is discussed in Sections 7.5 and 7.9. Other safety-related instrumentation and control systems incl uding Emergency Chilled Water System, Emergency Plant Service Water System, and Fuel Handling and Storage Interlocks are described in Section 7.6. 7.1.2 Identification of Safety Criteria in addition to meeting the requirements of the CRBRP General Design Criteria (refer to Section 3.1), the saf ety-rel ated l&C systems wil l be designed to meet the appl icable requirements of the Regulatory Guides and IEEE Standards l isted in Tabl es 7.1-2 and 7.1-3. The means of compliance with the guides and standards applicable to alI saf ety-related instrumentation and control equipment are described in paragraphs 7.1.2.2 through 7.1.2.11. Compl iance with guides or standards appl icable to specif ic l&C systems or equipment are described in the paragraphs relaied to those systems. The instrument error and other perf ormance consideration are addressed in the description of Individual subsy stems. O 7.1 -1 Amend. 72 Oct. 1982

with guides or standards applicable to specific l&C systems or equipment are described in the paragraphs related to those systems. The Instrument error and other perf ormance consideration are addressed in the description of individual subsystems. 7 .1. 2.1 Deslan Basis The Plant Protection System (PPS) incl udes the Reactor Shutdown System (RSS), the Containment isolation System and the Shutdown Heat Removal Systems. The Reactor Shutdown System consists of a Primary and a Secondary System either of which is designed to initiate and carry to completion trip of the control rods and sodium coolant pumps to prevent the results of postulated f ault conditions f rom exceeding the allowable l Imits. Table 4.2-35 shows the basis f or Primary and Secondary RSS perf ormance f or the def ined f ault categories. The perf ormance I imits f or the f uel and cl adding are identif ied in Section 4. The Reactor Shutdown Systems are descrited in Section 7.2. Two diverse Reactor Shutdown Systems have been provided f or CRBRP to ensure that the reactor is protected f rom the consequences of all anticipated and uni Ikely events even if one of the Reactor Shutdown Systems f all s. The two Reactor Shutdown Systems have been made diverse in order to reduce the probabil Ity that a common mode f ail ure w ill prevent a reactor shutdown f rom taking pl ace. This diversity extends f rom the sensors used as input to the two systems, through the logic util ized, to the actuation devices required to trip the two dif ferent control rod designs. Tabl e 7.1-4 l ists the principal diverse design f eatures present in the two h sy stems. These dif ferent design f eatures are discussed in more detail in Section 7.2.1.1, When combined w Ith the separation, qual if ication and other design requirements arising f rom the Regulatory Guides I isted in Tables 7.1-2 and 7.1-3, these designs provide protection against degradation of perf ormance arising f rom common mode initiators. The Containment Isol ation System (CIS) is designed to react automatically to prevent or I imit the release of radioactive material to the outside env ironment. The system acts to isolate the interior of the contair. ment by closing the containment isolation valves in the event that radioactive material is rel eased w ith in the contai nment. Radiation monitors w ith in the containment boundary are used to activate the CIS. A description of ibis system is given in Section 7.3. The Shutdown Heat Ranovel Instrumentation and Control System is designed to provide assurance against exceeding acceptable f uel and reactor coolant system damage I imits f ol l ow ing normal and emergency shutdowns. The description of this instrumentation and control is given in Section 7.4 for the ranoval through the auxil iary steam / water system (Steam Generator Auxil iary Heat Removal Sy stem ( SG AHRS) and Outlet Steam isolation System (OSIS) and Section 7.6 for removal through the NaK to air system (Direct Heat Removal System (DHRS)). Sufficient instrumentation and associated displ ay equipment w il l be provided to permit ef fective determination of the status of the reactcr at any time. Section 7.5 provides a description of the instrumentation provided. The 7.1 -2 Amend. 72 Oct. 1982

      .-.             .--___                  _ . . - ~ _ - . .    - _ _ _ -  _            . _ _ - - _ ~      _ _ - _ - . . . _ _ = _ - .                   . . _ - - -            .-

above design bases have been applled to the PPS Instrumentation Iisted in ,

Table 7.5-1 and described in Section 7.5. In Sect!on 7.9, a description of O the control room, control room layout, operator-control panel Interface, Instrument and display groupings and habitabil Ity are given. In the areas where the rupture of the steam or feedwater Iines can occur, the fleid Instrumentation and control shall be quellfled to survive the resulting higher temperature and pressure transient. l O l l l i O 7.1-2a Amend. 72 Oct. 1982

A T/BLE 7.1 -1 V SAFETY REL ATED INSTRUENTATION AND CONTROL SYSTEMS

• Reactor Shutdown Systems incl udes al l RSS sensors, signal conditioning calcul ation units, comparators, buf fers, 2/3 logic, scram actuators, scram breakers, control rods, back contacts on scran breakers, HTS shutdown logic, coolant pump breakers, and mechanical mounting hardware (equipment racks).

Containment Isolation Svstem incl udes radiation monitoring sensors, signal conditioning, comparators, 2/3 logic, containment isol ation val ve actuators and val ves. Decav Heat Removal Svstem Instrumentation and Control System incl udes initiating sensors, signal conditioning, calculation units, comparators, logic, auxil iary feedwater pump actuators and controls incl uding f eedwater turbine pump, PACC DHX actuators and control s, steam rel ief val ve actuators and valves; sensors, signal conditioning, logic and actuators rel ated to decay heat removal functions of DHRS including control of sodium and NaK pumps and air bl est heat exchangers; and sensors, signal conditioning, logic and actuators related to removal of heat from the EYST. Other Safety Related Instrumentation and Control incl udes Instrumentation and Control s f or portions of the f ollowing f unctions to assure the plant is maintained in a saf e shutdown condition: o Emergency Chil led Water System o Emergency Plant Service Water System o Instrumentation necessary to assure plant is maintained in saf e shutdown status (See Table 7.5-4) l o Fuel Handl ing and Storage Saf ety Interlocks o Heating, Ventilating, and Air Conditioning System o Recircul ating Gas Cool Ing System

     *The Cl inch River Breeder Reactor Pl ant (CRBRP) saf ety-related structures, systems, and components are designed to remain f unctional in the event of a Saf e Shutdown Earthquake (SSE). These incl ude, but are not I frr.ited to, those structures, systems and components which are necessary:

To assure the integrity of the Reactor Coolant Boundary; To shut down the reactor and maintain it in a safe shutdown condition; To prevent or mitigate the consequences of accidents which could resul t in potential of f-site exposures comparabl e to the guidel ine exposures of 10CFR100. NOTE: Class IE equipment loads are identif ied in Chapter 8. 7.1 -7 Amend. 72 Oct. 1982

l TABLE 7.1-2 LIST OF REGULATORY GUIDES APPLICABLE TO SAFETY RELATED INSTRUMENTATION AND CONTROL SYSTEMS 1.6 Independence Between Redundant Power Sources and their Dis-tribution Systems (as discussed in Sections 8.3.1.2 and 8.3.2.2) 22 1.12 Instrumentation for Earthquakes 1.17 Protection of Nuclear Power Plants Against Industrial Sabotage 1.22 Periodic Testing of Protection System Actuation Functions 1.28 Quality Assurance Program Requirements (Design and Construction) 1.29 Seismic Design Classification 1.30 Quality Assurance Program Requirements for the Installation, 5 Inspection, and Testing of Instrunentation and Electric Equipment 1.32 Use of IEEE Std 308-1971 " Criteria for Class lE Electric Systems for Nuclear Power Generating Stations" 1.40 Qualification Tests of Continuous Duty Motors Installed Inside the Containment of Water Cooled Nuclear Power Plants 1.47 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems 1.53 Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems 1.62 Manual Initiation of Protective Actions 1.63 Electric Penetration Assemblies in Containment Structures for Water-Cooled Nuclear Power Plants 1.64 Quality Assurance Program Requirements for the Design of Nuclear Power Plants 1.73 Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants 1.75 Physical Independence of Electric System T.7S Control Room Habitability During Chemical Release (as discussed in Section 6.3). 1.89 Qualification of Class IE Equipment for Nuclear Power Plants (as discussed in Section 7.1.2.5). 22 Amend. 22 7.1-8 June 1976 l

TM3 LE 7.1 -3 LIST OF IEEE STANDARDS APPLICABLE TO SAFETY RELATED INSTRUfENTATION AND CONTROL SYSTEfG IEEE-279-1971 IEEE Standard: Criteria for Protection Systems for Nuciear Power Generating Stations IEEE-308-1974 Criteria for Cl ass IE Power Systems f or Nucl ear Power Generating Stations I EEE-317-1976 Electric Penetration Assembi les in Contairenent Structures f or Nuclear Power Generating Stations IEEE-323-1974 Qual ifying Ciass IE Electric Equipment for Nuciear Power Generating Stations I EEE-323-A-1975 Supplement to the Foreword of IEEE 323-1974 IEEE-336-1971 lEEE Standard: Installation, inspection, and Testing Requirements for instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations } IEEE-338-1977 Criteria for the Periodic Testing of Nuclear Power Generating Station Saf ety Systems IEEE-344-1975 IEEE Std. 344-1975, IEEE Recommended Practices for Seismic h d Qual if icaticn of Cl ass 1 Equipment f or Nuclear Power Generating Stations IEEE-352-1975 General Principles for Rel Icbil ity Analysis of Nuclear Power Generating S1ation Protection Systems I EEE-379-1972 IEEE Trial-Use Guide f or the Appl ication of the Singl e-Fail ure Criterion to Nuclear Power Generating Station Protection Systems I EEE-383-1974 Standard for Type Test of Class 1E Electric Cables, Field Spl ices, and Con'nections f or Nuclear Power Generating Station. IEEE-384-1974 IEEE Trial Use Standard Criteria for Separation of Class IE Equipment and Circuits I EEE-420-1973 Trial-Use Guide for Class IE Control Switchboards for Nuclear Power Generating Stations IEEE-494-1974 IEEE Standard Method f or identif Ication of Documents Rel ated to Class IE Equipment and Systems f or Nuclear Power Generating Station O 7.1 -9 Amend. 72 Oct. 1982 i

T/BLE 7.1-4 RSS DIVERSITY Primary Secondarv Logic: Local Coincidence General Coincidence Sensors: Iniet Plenum Pressure Primary Loop Flow Primary Pump Speed Primary Loop FIow Intermediate Pump Speed Intermediate Loop Flow HTS Bus Frequency HTS Bus Voltage Steam Flow Steam Drum Level Feedwater Fl ow Reaction Products Flow IHX Primary Outiet Evaporator Outiet Temperature Sodium Temperature Logic I sol at lon: Light Coupling Direct Coupied Equipment: o Circuitry Integrated Circuits Discrete Components o Power Supplles Separate vendors utilized o Potentimeters Separate vendors util ized o Buf fers Light Coup!Ing Magnetic Coupling o Control Rod Circuit Broekers in Solenoid Operated Rel ease 2/3 Logic Arrant;ement Pneumatic Val ve in a 2/3 Logic Arrangment 1 l Amend. 72 O 7.1-10 Oct. 1982

I m 7.2 REACTOR SHUTDOWN SYSTEM { l

 \

7.2.1 Descriotion I 7.2.1.1 Reactor Shutdown System Descriotion The Reactor Shutdown Systern (RSS) consists of two independent and diverse systems, the Primary and Secondary Reactor Shutdown Systems, either of which is capable of Reactor and Heat Transport System Shutdown. All anticipated and unl ikely events can be terminated without exceeding the specif ied l imits by either system even if the most reactive control rod in the system cannot be inserted. In addition, the Primary RSS acting alone can terminate all extremely uni Ikely events without exceeding specified limits even if the most reactive control rod in the system cannot be inserted. To assure adequate independence of the shutdown systems, mechanical and electrical isolation of redundant components is provided. Functional or equipment diversity is included in the design of instrumentation and electronic equipment. The Primary RSS uses a local coincidence logic configuration while the Secondary RSS uses a general cc s ncidence. Suf ficient redundancy is included in each system to prevent single random f ail ure degradation of either the Primary or Secondary RSS. As shown in the block diagram of the Reactor Shutdown System, Figure 7.2-1, the Primary RSS is composed of 24 subsystems and the Secondary RSS is composed of 16 subsystems. Figure 7.2-2A is a typical Primary RSS Instrument channel logic diagram. Each protective subsystem has 3 redundant sensors to monitor a

    ~ physical parameter. The output signal from each sensor is ampiifled and converted for transmission to the trip comparator in the control room. Three physically separate redundant instrument channel s are used. When necessary, calcul ational units derive additional variables f rom the sensed parameters, with the calculational units inserted in front of the comparators as needed.

The comparator in each instrument channel determines if that instrument channel signal exceeds a specif ied l imit and outputs 3 redundant signals corresponding to either the reset or trip state. The 3 outputs of each comparator are isolated and recombined with the isolated outputs of the redundant instrument channels as inputs to three redundant logic trains. The recombination of outputs is in a 2 out of 3 local coincidence logic arrangement. Operating bypasses are necessary to allcw RSS functions to be bypassed during main sodium coolant pump startup, ascent to power, and two loop operation. Operating bypasses are accompi ished in the instrument channel s. For bypasses associated with normal three locp operation, the bypass cannot be instated unless certain permissive conditions exist which assure that adequate protection will be maintained while these protective f unctions are bypassed. Permissive comparators are used to determine when bypass conditions are satisfied. When permissive conditions are within the allcwable range, the operator may manual ly instate the bypass. if the permissive condition goes O V 7.2-1 Amend. 57 Nov. 1980

i out of the allowable range, the protective f unction is automatically reinstated. The trip f unction will rmain reinstated until the permissive conditions are again satisf ied and the operator again manually initiates the - by pa ss. Operator manual bypass control is not of fective unless the bypass comparator Indicates that permissive conditions are satisfied. A functional diagram of the Primary and Secondary bypass permissive logic is shown in Figure 7.2-2AA. Two loop bypasses are establ ished under administrative control by changing the hardware conf iguration w ith in the l ocked comparator cabinets. These bypasses are al so under permissive control such that the pl ant must be shutdown to estabi Ish two loop operation and if the shutdown loop if activated the bypass is automatical ly removed. Bypass f eatures incl uded w ithin the Primary and Secondary RSS hardware for two Ioop operation wilI be deactivated durIng alI three Ioop operating modes so that the three Ioop operating conf iguration can not be af fected by these bypass features either by operator action or by two loop hardware f ail ure. Bypass permissives are part cf the Pl ant Protection System (PPS), and are designed according to the PPS requirements detailed el sewhere in this section of th e PS AR. Continuous Iocal and remote indication of bypassed instrument channels wilI be prov ided in conf ormance w ith Regul atory Guide 1.47, ' Bypassed and i noperabl e Status Indication f or Nucl ear Power Pl ant Saf ety Systems". O O 7.2-la Amend. 72 Oct. 1982 i

Evaporator Outlet Sodium Temperature O)

          \s-                                           The Evaporator Outlet Sodium Temperature subsystems (Figure 7.2-10) compare the sodium temperature at the outlet of the evaporator in each HTS loop to a fixed set point.                                                                If this temperature exceeds the set point, a reactor trip is initiated.                                                                There are three of these subsystems, one per
  ,                                  loop. These subsystems detect a large class of events which impair the heat removal capability of the steam generators. These subsystems are never bypassed.

Sodium Water Reaction The Sodium Water Reaction subsystems (Figure 7.2-10) detect the occurrence of a sodium water reaction within a superheater or evaporator module. There are three of these subsystems, one per loop. Each subsystem 571 receives nine signals from the sensors in the reaction products vent lines of a steam generator. These subsystems are never bypassed. 7.2.1.2.3 Essential Performance Requirements In order to implement the required protective functions within the appropriate limits, PPS equipment must meet several essential performance requirements. These essential performance requirements and the PPS equip-ment to which they apply are sumnarized below. The PPS instrumentation will meet the essential performance require-57lmentsofTable7.2-3. This table defines the minimum accuracy and time O constants which will result in acceptable performance of the PPS. Analysis of worst case PPS functional performance is based on the values given in Table 7.2-3. . The maximum delay between the time a protective subsystem indicates the need for a trip and the time the rods are released is 0.200 second. l l This time includes the delays due to the calculational units, comparators, logic, 41 scram breakers, and control' rod release. The maximum delay between the time a protective subsystem indicates the need for a trip and the time the HTS sodium pumps are tripped is 0.500 i second. This time also includes the delays due to the logic and HTS scram breakers. The PPS is designed to meet these essential performance require-ments over a wide range of environmental conditions and credible single events to assure that environmental effects do not degrade the performance 1 i  % ' Amend. 57 7.2-11 Nov. 1980

    ,_,-__ .   . , . , - . . - .       . . . , ,          ,  - - . . . - - . - - . + - - , _ , , . - - , - - . - . - . - - , . ,                 - - .          .-,.,,..-.,r,   ,- r   ,    . . , , - - - -

o Environmental Chances All electrical equipment is subject to perf ormance degradation due to major changes in the operating environment. Where practical, PPS eq uipment is designed to minimize the of fccts of environmental changes; if not, the performance at the environmental extremes is used in the analysis. Measures have been taken to assure that the RSS electronics are l capable of performing according to their essential performance requirements under variations of temperature. The range of temperature environment specif led f or all the electronic equipment considered here is greater than is expected to occur during normal or abnormal conditions. Electronics do not f all catastrophically when these l Imits are exceeded even though this is the assumed f ail ure mode. The detailed design of the circuit boards, board mounting and racks includes f ree ventilation to minimize hot spots. Ventil ation is a resul t of natural ceavecticn air fIow. The RSS is designed to operate under or be protected f rom a wider range of relative humidity than that produced by normal or postulated l accident conditions. Vibration and shock are potential causes of f ailure in electronic components. Design measures, including the prudent location of equipment, minimize the vibration and shock experienced by RSS el ectrcn i cs. The equipment is qual if led to shock and vibration specif ications which exceed al l normal and of f-normal occurrences. The RSS comparators and protective logic are designed to cperate over a power source voltage range of 108 to 132 VAC and a power source f requency range of 57 to 63 HZ. The maximum variation of the source vol tage is expccted to be 110%. More extrcrne variations in the power source may result in the af fected channel comparatcr or logic train outputting a trip signal. In addition, testing and monitoring of RSS equipment is used, where appropriate, to warn of irrpending equipment l degradation. Theref or e, it is not expected that changes in the environment w il l cause total f ail ure of an instrument channel or logic train, much less the simultaneous f ail ure of al l Instrument channel s cr logic trains. l l The majority of the RSS electronics is located in the control l buil ding, and is not subjected to a radioactive environment. Any PPS equipment located in the radioactive areas (such as the head access i area) will be designed to withstand the level of activity to which it w 11 I be subjected, if its f unction is rcquired. O 7.2-12 Amend. 72 Oct. 1982

o Tornado The RSS is protected f rcrn the ef fects of the design basis tornado by locating the equipment within tornado hardened structures. o Local Fires All RSS equipment, including senscrs, actuators, signal ccnditioning l equipment, wiring, scram br aa). ors, and cabinets housing this equipment is redundant and separated. These characteristics make any credible f ire of no consequence to the saf ety of the plant. The separation of the redundant components increases the timo required f or f Ire to cause extensive damage and also allows time for the fire to be brought to the attention of the operator such that corrective action may be initiated. Fire protection systems are also provided as discussed in Section 9.13. o Local Exoloslons and Missiles All RSS equipment essential for reactor trip is redundant. Physical l separation (distance or mechanical barriers) and electrical isol ation exists between redundant components. This physical separation of redundant components minimized the possibil ity of a local explosion or missile damaging more than one redundant component. The remaining redundant components are still capable of performing the required protective functions, o Earthauakes , All RSS equipment, incl uding sensors, actuators, signal condif f oning l equipment, wiring, screm breakers and structures (e.g., cabinets) housing such equipment, is classed as Seismic Category 1. As such, all RSS equipment is designed to remain f unctional under CBE and SSE l conditions. The characteristics of the CBE and SSE used f or the evaluation of the RSS are found in Section 3.7. l 7.2.2 Analvsis The Reactor Shutdown System meets the saf ety related channel perf ormance and l rel labil Ity requirements of the NRC General Design Criteria, IEEE Standard 279-1971, appl Icable NRC Regulatory Guides and other appropriste criteria and standards. The RSS Logic is designed to conform to the IEEE Standards lIstea in Table 7.2-4. General Functional Reautrement The Plant Protection System is designed to automatically initiate apprepriate protective acilon to prevent unacceptable plant or component damage or the release or spread of radioactive material s. O 7.2-13 Amend. 72 Oct. 1982

Single Failure No single failure within the Plant Protection System nor removal from service of any component or channel will prevent protective action when required. 57l Two independent, diverse reactor shutdown systems are provided, either o7 which is capable of terminating all excursions without allowing plant param-eters to exceed specified limits. Each system uses three redunda.nt instru-ment channels and logic trains. The Primary RSS is configured 57 using local coincidence logic while the Secondary RSS uses general coincidence logic. To provide further assurance against potential degradation of protection due to credible single events, functional and/or equipment diversity are included in the hardware design. Bypasses Bypas;es for normal operation require manual instating. Bypasses will be automatically removed whenever the subsystem is needed to provide protection. The equipment used to provide this action is part of the PPS. Administrative procedures are used to assure correct use of bypasses for infrequent operations such as two loop operation. If the protective action of some part of the system has been bypassed or deliberately rendered inoperative, this fact will be continuously indicated in the control room. Multiple Setpoints Where it is necessary to change to a more restrictive setpoint to provide adequate protection for a particular normal mode of operation or set of operating conditions, the PPS design will provide automatic means of assuring that the more restrictive setpoint is used. Administrative proce-dures assure proper setpoints for infrequent operations. Fcr CPERP, power operation cn two-loops will be an infrequent occurrence, and will only be initiated fron a shutdcwn condition. t!hile the reactor is shutdown, the FPS equipment will be aligned for tuo-loop operation which will include set down of the appropriate trip points. Sufficient trip point set down is being designed into the FPS equiprent to adequately ccver the possible range (conceptually fror 2% to 10GX) of trip point adjust. Tent required. In addition, administrative procedures (specifically the pre-critical chechoff) will te in/cked durino startup to ensare that the proper PPS trip pcints have l'een set. The anclysis of plant performance during two-loop operation tas not been conpleteJ to date. Therefore, the exact trip point settir.gs for two-loop operation cannot be specified at this time. !iowever, the range of trip point settings indicated above is adequate to ensure that trip points apprcpriate for the anticipated lowest two-lcop operating power can be achieved. In surrary, the design of the PPS equiprent trip point adjustnents and other features for two-loop operation coupled with the anticipated two-loop operating power level anc administrative procedures assure full  ! conpiiance with Branch Technical Position EICSD 12 and satisfy Section 4.15 of IEEE s td 279-1971. h; 7.2-14 Amend. 57 Nov. 1989

Access m Administrative control of access to all setpoint adjustments, module calibration adjustments, test points and the means for establishing a bypass permissive condition is provided by locking cabinets and other access design features of the control room and the equipment racks. Information Read-Out Indicators and alarms are provided as an operating aid and to keep the plant cperator informed of the status of the RSS. Except for the IHX primary outlet temperature analog Indicators which are part of the accident monitoring . system, all indicators and alarms are not safety-related. The following items are located on the Main Control Panel for operator information. Analoa Indication A. Secondary Wide Range Log MSV Power Level B. Secondary Wice Range Linear Power Level C. Primary Power Range Power Level D. Reactor Vessel Level E. HTS Pump Speeds F. HTS Loop Flows G. Reactor inlet Pressure H. lHX Prirrary Outlet Temperature

1. Evaporator Outlet Temperature J. Steam Flows K. Feedwater Flows b L. Steam Drum Level .

Indicating Lights A. Instrument Channel Bypass Permissive Status B. Instrument Channel Bypass Status C. Logic Train Trip / Reset Status D. HTS Loop Trip / Reset Status E. HTS Loop Test Status Annunciators d A. Instrument Channel Trip / Reset information is provided for each function listed in Table 7.2-1 B. Logic Train Power Supply Failure C. Two Loop Bypasses instated Most Information is also available to the operator via the Plant Data Handling and Display System. Annunciator for PPS Alarm Trios A visual and audible indication of all alarm conditions within the PPS will be provided in the control room. These alarm ccnditions include any tripped PPS comparators in the Primary RSS, Secondary RRS, Containment Isolation System and Shutdown Heat Removal System. The Plant Data Handling and Display system 7.2-15 Amend. 72 Oct. 1982 N

I I alerts the operator to significant deviations between redundant RSS analog instrumentation used to monitor a reactor or plant parameter for the RSS. Control and Protection System Interaction The Plant Protection System and the Plant Control System have been designed to assure stable reactor plant operation and to protect the reactor plant in the event of worst case postulated Plant Control System failures. The Plant Protection System is designed to protect the plant regardless of control system action or Iack of actlon. Isolation devices wiII be used between protection and control functions. Where this is done, all equipment common to both the protection and control function is classified as part of the Plant Protection System. Equipment sharing between protection and control is minimized. Where practical, separate equipment (sensors, signal conditioning, cabling penetrations, raceways, cabinets, monitoring etc.) is provided. The sharing of components does not lead to a situation where a single event both initiates an incident through Plant Control System malfunction and prevents the appropriate Plant Protection System. Periodic Testing The Plant Protection System is designed to permit periodic testing of its functioning including actuation devices during reactor operation. In the Primary RSS, a single instrument channel is tested by inserting a test signal at the sensor transmitter and verif ying it at the comparator output. A logic train is tested by inserting a very short test signal in 2 comparator inputs and verifying that the voltage on the scram breaker trip coils decrease. Because of the time response of the undervoltage relay coils of the scram breakers and very short duration of the rest signal, the reactor does not trip. In the Secondary RSS, an instrument i l l O 7.2-15a Amend. 72 Oct. 1982

channel can be tested from sensor to scram actuator by inserting a single test signal because of the general coincidence configuration of the 3 (]/ w redundant channels. The primary and secondary rod actuators cannot be tested during reactor operation since dropping a single control rod will initiate a reactor scram. Scram actuators and control rod drop will be tested and maintained when the plant is shutdown (See Section 7.1-2). When-ever the ability of a protective channel to respond to an accident signal is bypassed such as for testing or maintenance, the channel being tested is placed in the tripped state and its tripped condition is automatically indi-cated in the control room. Failure Modes and Effects Analysis A Failure Modes and Effects Analysis (FMEA) has been conducted to identify, analyze and document the possible failure modes within the Reactor Shutdown System and the effects of such failures on system performance (see Appendix C, Supplement 1). Components of the RSS 40 analyzed are: 41 57 e Reactor Vessel Sodium Level Input e PPS Sodium Flow Input e Pump Electric Power Sensor 40l e Compensated Ion Chamber Nuclear Input ( e Fission Chamber Nuclear Input e Primary Loop Inlet Plenum Pressure Input e Sodium Pump Speed (Primary and Intermediate) e Steam Mass Flow Rate Input e Feedwater Mass Flow Rate Input e Steam Drum Level Input e Primary Comparator e Secondary Comparator e Primary Logic Train e Secondary Logic Train e Primary Calculational Unit e Secondary Calculational Unit O' Amend. 57 Nov. 1980 7.2-16

4 C [ ( \ I i Table 7.2-2 (Continued)~ Fault Events Primary _3hutdown System Secondary Shutdown Svsiem i Failure of Steam Dump System Steam-Feedwater Flow Steam Drum Level Mismatch Sodium Water Reaction in Steam Steam-Feedwater Flow Sodium-Water Reaction Generator Mismatch

lli. Extemelv Unlikely A. Reactivity Disturbances Positive Ramps $2.0/sec i

Startup Flux-Delayed Flux Startup Nuclear n

j 5-40% Power Flux-Dolayed Flux or Poditled Nuclear Rate or ] Flux- Pressure Flux-Total Ficw i 40-100% Power Flux- Pressure Flux-Total Flow i N ! tu 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 rare of approximately 4.1 cents per second. (2) The maximum unlikely reactivity faults result from multiple control system follures leading to withdrawl of six rods at nor mal j speed or one rod at the maximum mechanical speed. (3) The PPS is required to terminate the results of these extremely unlikely eveqts within the umbrella transient specir sed as emergency for the design of the major components. i l h

5ao.

I to hb i

1 T/BLE 7.2-3 ESSENTI AL PERFORMAN REQUIREENTS FOR RSS INSTRUtENTATION CHANNELS O Accuracy Response Time Plant Parameter (f of scan) (msec) Neutron Fl ux Primary 11 .0 <10 Secondary 11 .0 <10 Reactor Inl et Pl enum Pressure 12.0 <150 Sodium HTS Pump Speeds 12.0 <20 Sodium HTS Flow 15.0 <500 Reactor Vessel Sodium level 15.0 <500 Undervoltage Roley 11 .0 < 23 0 Steam FIow 12.0 <500 Feedwater Fl ow 12.5 . <500 Evaporator Outl et Sodium Temperature 12.0 <5000 Steam Drum level 11 .0 <1000 IHX Primary Outiet Tanperature 12.0 <5000 underf req uency Rei ay 12.0 <200

• Note that these accuracy and response times relate to the perf crnance of the instrumentation channel s f rom the sensors up to 1he signal conditioning output.

In addition, as noted in Section 7.2.1.2.3, the reactor shutdown system logic, actuators and rod unl atch features require a f urther response time del ay of 200 msecs. O l 7.2-23 Amend. 72 Oct. 1982

TE LE 7.2-4 LIST OF lEEE STANDARDS APPLICABLE TO THE REACTOR SHUTDOWN SYSTEM LOGIC (1) IEEE 279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations IEEE 308-1974 Criteria for Class 1E Power Systems for Nuclear Power Generating Stations IEEE 317-1976 Electric Penetration Assembl les in Containment Structures f or Nuclear Power Generating Stations lEEE 323-1974 IEEE Trial-Use Standard: General Guide for Qualifying Class 1E Electric Equipment for Nuclear Power Generating Stations IEEE 323-A-1975 Supplement to the Foreward of IEEE 323-1974 IEEE 336-1971 IEEE Standard: Installation, inspection and Testing Requirements for Instrumentation and Electric Equipment 4 During Construction of Nuclear Power Generating Stations 1EEE 338-1977 IEEE Trial Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices for Seismic Qual ification of Class 1 Equipment for Nuclear Power Generating Stations IEEE 352-1975 IEEE Guide f or General Principles for Rel labil Ity Analysis of Nuclear Power Generating Station Protection Systems IEEE 3791972 lEEE Triel-Use Guide for the ApplIcaton of the SIngie Fail ure Criterion to Nuclear Power Generating Station Protection Systems IEEE 384-1974 IEEE Trial Use Standard Criteria for Separation of Class IE Equipment and Circuits IEEE 494-1974 IEEE Standard Method for identif ication of Documents Related to Class 1E Equipment and Systems for Nuclear Power Generating Station (1) lEEE Standards applicable to the instrumentation and monitoring systems are l Istea in Section 7.5. ( v 7.2-23 a Amend. 72 Oct. 1982

Head Access Area Radiation [} \s_- The Head Access Area Radiation Subsystem initiates closure of the containment Isolation valves in the event of large radiation releases in the head access area. Three radiation sensors are located in the head access area to provide early initiation and closure of the isolation valves to assure that releases f rom design basis events do not exceed the guidel Ine val ues of 10CFR100.  !

7. 3.1. 2. 2 Essential Performance Reaufrements To implement the required isolation f unction within the specified limits, the Cls must meet the f unctional requirements specif ied below:

The closure time requirement f or the inlet and exhaust isolation valves is 4 seconds with a three second or less detection time in the heating and ventil ating system. A 10 second transport time f rom sensing point to the valve exists (see Section 15.1.1). The 3 seconds incl udes sensor time response, comparator and logic time del ays. The CIS is designed to meet these requirements f or the environmental conditions described in Section 7.2.1. 7.3.2 Analvsis The design of the CIS provides the necessary design features to meet the f unctional and performance requirements as described below. The CIS logic is designed to conform to the IEEE Standards l isted in Table 7.3-2. ("} 7.3.2.1 Functional Performance The analyses in Sections 15.5 and 15.6 shows the results of the postulated f aul t conditions. These analyses assumed a closed containment where the , events occurred w Ith the containment hatch closed. For the I imiting event, primary drain tank fire during maintenance, scoping analyses have been perf ormed to determine the required closure time of the containment isolation val ves. For the primary drain tank fire, closure within 20 minutes is adequate. Further, analyses to determine the required closure time under postulated accident conditions have been perf ormed and are discussed in Sect ion 15.1.1. These analyses are used to determine the available design margin. The results of this assumed condition do not exceed the guideline values of 10CFR100 if the main exhaust and inlet valves are closed within 4 seconds assuming the normal air transport time f rom the detectcr to the val ve is 10 seconds or more, a 14,000 Cfm normal ventil ation rate. Since the automatic Containment isolation System is designed to isolate within the above time response requirements, all of the design basis conditions are terminated within the necessary limits f or the present design concept. 7.3.2.2 Deslan Features The CIS instrumentation, controls and actuators are designed to meet the requirements of IEEE-279-1971. The analyses of compl iance with 1hese are summarized bel ow. O 7.3-3 Amend. 72 Oct. 1982

l l Single Failure No single failure within the CIS nor removal from service of any component or channel will prevent protective action when required. There are three independent Instrument channels for each necessary measurement, two independent 2/3 logics, and two independent actuators provided (as shown In Figure 7.3-1). Hypasses No bypasses are provided. Multiple Setpoints Multiple setpoints are not required. Comoletion of Protective Action The automatic CIS is designed sc that, once initiated, protective action at the system level must go to completion. Return to normal operation requires manual reset of the CIS breakers by the operator. Manual Initiation The CIS Includes means for manual Initiation of containment isolation at the system level. No single failure will prevent manual Initiation of the containment isolation action. Control and Protection interaction There are no shared components between the control system and the CIS. The provisions for access, Information read-out, annunciation of trips, and periodic testing are as specified for the Reactor Shutdown System in Section 7.2.2. Physical Separation The following criteria assure physical separation for the CIS. There will be at least one containment penetration for each of the three Primary PPS instrument channel conduits and each of the three Secondary PPS instrument channel conduits which exit containment. AlI requirements for separation of PPS wirleg through rnduits will also apply to separation of PPS wiring through containment penetrations. O 7.3-4 Amend. 62 Nov. 1981

O O O TABLE 7.3-1 CONTAINMENT ISOLATION SYSTEM DESIGN BASIS Applicable Event Federal Regulation Limit Anticipated Fault 10CFR20 5 105 <2 millirem in any one hour No examples of anticipated faults which lead <100 millirem in any one week to release of activity have been identified. Unlikely Fault 10CFR20 5 403b <5 rem in any two hours No examples are presently identified for the automatic containment isolation system design basis.

. Extremely Unlikely Faults & Design Margin

• 10CFR100 <25 rem in any two hours
• Examples include major sodium fires <300 rem iodine doses in the thyroid in any two hours

                                                                                  <75 rem to the lung 44                                                                             <150 rem to the bone
      *The design basis for the CIS includes limiting the results of postulated accidents within the guideline values of 10CFR100. See Section 15.1.1.

Eb' le ~."

T/BLE 7.3-2 LIST OF IEEE STANDARDS APPLICABLE TO THE CONTAINfENT ISOL ATION SYSTEM LOGIC IEEE 279-1971 IEEE Standard: Criteria for Protection Systems for Nuciear Power Generating Stations IEEE 308-1974 Criteria for Class 1E Power Systems f or Nuciear Power Generating Stations IEEE 317-1976 Electric Penetration Assembiles in Containment Structures for Nuclear Power Generating Stations IEEE 323-1974 IEEE Trial-Use Standard: General Guide f or Qual ify ing Cl ass 1E Electric Equipment f or Nuclear Power Generating Stations IEEE 323-A-1975 Supplanent to the Foreward of IEEE 323-1974 IEEE 336-1971 IEEE Standard: Installation, inspection and Testing Requirements f or instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations IEEE 338-1977 IEEE Trial Use Cri1eria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE 344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Seismic Qualif Ication of Class 1 Equipment for Nuclear Power Generating Stations IEEE 352-1975 IEEE Guide for General Principles f or Rel labil ity Analysis of Nuclear Power Generating Station Protection Systems IEEE 379-1972 IEEE Trial-Use Guide f or the Appi icaton of the Single Fail ure Criterion to Nuclear Power Generating Station Protection Systems IEEE 384-1974 IEEE Trial Use Standard Criteria for Separation of Class 1E Equipment and Circuits lEEE 494-1974 IEEE Standard Method f or identification of Documents Related to Cl ass IE Equipment and Systems f or Nuclear Power Generating Station l l l l l O l l 7.3-5a Amend. 72 Oct. 1982

c O 4.

v. LOGIC TRAIN 1

                    -                       CONT AINME NT VENTil ATION EXHAUST DETECTORS                                         llSOL

[ SENSOR ~ s COMPARATOR 2/3 Hi i CON T0 ING llSOL l i TEST llSOL B TYPICAL < v - BREAKER OR VER l l

                                                                                                        ' ISO L                                      INSIDE TYPICAL                                                            L                        CONT AINME NT Cl                                                                                '                                  VALVES l                                            2/3 -
                                                                                                        ' .I_SO L          LI e

BUFFERED OUTPUTS , ISO L TO Cl3 METERS HE AD ACCESS ARE A LOGIC TRAIN 2 q

             ;                                       DE T E CT OR$                                      l1 SOL SENSOR       -

CON Tl ING

                                                                              '^"^ "   '                                          '

TEST SOURCE l 4 SOL TYPICAL --<> V B < g - BREAKER , I j O UTSIDE i lSOL CONT AINME NT C T YPICAL L VALVES M 2/3 -

           *D                                                                                            ISO L l
          $$s BUFFERE0 00TPUTS TO CIS METERS i          -."

e co m

liipure 7. 4-1. Containment Imlation Splem Hlock Diagram i

I f

7.4 INSTRUMENTATION AND_ CONTROL S,YSTEMS REOUIRED FOR SAFE SHUTDOWN (~'} The Instrumentation and Control Systems necessary for safe shutdown are those \/ associated w Ith monitoring of core critical ity, decay heat romoval (SGAHRS portion), outlet steam isolation, and control room habitabil Ity. Monitoring of core criticality is ef fected by the Flux Monitoring System (Section 7.5.1). The control room habitabil Ity is covered in Chapter 6. Thus, this section treats the control and instrumentation needs f or decay heat removal by the Steam Generator Auxil lary Heat Removal System (SGAHRS) and outlet steam isolation by the Outlet Steam Isolation System (OSIS); control and Instrumentation f or Direct Heat Removal Service (DHRS) is discussed in Section 7.6. 7.4.1 Steam Generator Auxiliarv Heat Removal Instrumentation and Control System 7.4.1.1 Design Descriotion 7.4.1.1.1 Function The SGAHRS (fluid system and mechanical components as described in Section 5.6.1, and electrical components as described below) provides the heat removal path and heat sink f or the nuclear steam supply system following upset, emergency, or f aulted events which render the normal heat sink unavailable. The SGAHRS Instrumentation and Control System in conjunction with the PPS detects the need f or, initiates, and control s the alternate heat rar. oval path when the normal heat sink is unavailable. The SG AHRS nominal control [s-}' m setpoints shown in Table 7.4-2 are discussed in the following subsections. The SGAHRS Instrumentation and Control System is designed to the IEEE Standards l isted in Tabl e 7.4-3. 7.4.1.1.2 Eautoment Design The mechanical system f or which the SGAHRS I&C is provided is briefly described bel ow. When actuated, the SGAHRS draws water from a Protected Water Storage Tank and pumps it to each steam drum. Two supply lines are provided f or each steam drum. One I ine is suppl ied by two hai f-sized, motor-driven teedwater pumps while the other is suppl led by a f ull-sized, turbine-driven pump. Each supply line provides a flow control valve and an isolation valve at 1he inlet to each steam drum. The isol ation val ves are provided to isolate the auxil iary feedwater system f rom the steam generator system during power operation and to provide leak (solation during SGAHRS operation. in addition, a Protected Air Cooled Condenser (PACC) supplied with each steam drum is placed into operation. This system rejects heat to the atmosphere via convection. Saturated steam is suppl led to the condenser f rom the steam drum O Amend. 72 7.4-1 Oct. 1982

and saturated water is returned. This steam and water loop is driven by natural circulation. Each PACC unit consists of two tube bundles, two sets of louvers and two f ans. Regulation of heat rejection is accomplished by controlling the air flow across the condensing tubes through adjustment of inlet louver and fan blade pitch positions. The air side flow is driven by either forced or natural convection. The arrangement of SGAHRS equipment is shown in Figure 5.1-5 (SGAHRS P&lD). Instrumentation and controls are provided for the components described below: o Auxiliarv Feedwater Pumo Control - Upon receipt of the SGAHRS initiation signal, (see Section 7.4.1.1.3), the two motor driven pumps are started, resulting in both pumps coming on lIne and operating at constant speed. In addition, the Isolation valves in the steam supply lines f rom the steam drums to the turbine driven pump are opened. At the turbine inlet a pressure regulating valve reduces the steam supply pressure to the 1000 psig required by the turbine drive. The turbine drive mechanism is equipped with a governor to provide speed regulation. Each auxil iary f eedwater pump can also be actuated manually at the operator's discretion. Each pump control includes a " Normal Long Term Cooldown (LTC)" mode selector, in " normal" mode, the pumps start on SGAHRS Initiation, in the "LTC" mode, the operator may shutdown any or al l AFW pumps provided the steam drum water level is above the trip point setting. When in the "LTC" mode, the pumps come on lIne autcmatically when the steam drum water level drops to a low level trip point. o Auxiliary Feedwater Flow Control - The Auxiliary Feedwater Isolation Valves are opened upon receipt of the SGAHRS initiation signal. During SGAHRS operation, these valves close automatically upon Indication of a sodium / water reaction, a high steam drum level, a steam drum pressure less than 200 psig, or AFW flow greater than 150% of full flow for 5 sec. This autanatic closure occurs only in the affected loop. If the valves are closed by a high drum level signal they will reopen autcmatically when the drum level falls to the low drum level trip point. The flow to the steam drum is controlled with a control valve that is positioned by a single controller. Manual control of the Auxiliary Feedwater Flow Control valves is provided at the main control panel and at the local SGAHRS panel. O 7.4-2 Amend. 71 Sept. 1982

l l 7.4.1.1.9 Ooerator Information l b(j indicators and alarms are provided to keep the plant operator informed of the status of the SGAHRS. The following items are located on the Main Control Panel for operator inf ormation. Analog Indication o Protected Water Storage Tank Level o Protected Water Storage Tank Temperature o Auxil iary Feedwater Flow (each loop) o Auxiliary Feedwater Pump Discharge Pressure o Drive Turbine Steam inlet Pressure o Drive Turbine Speed o PACC Outlet Air Temperature o PACC Outlet Water Flow and Temperature o PACC Inlet Louver Position o PACC Fan Blado Pitch Position o Steam Drum Pressure and Water Level Indicating Lights o PACC Outlet Louver Position o Position of all isolation and Control Valves o Operating Status of all Motors o SGAHRS Initiation Logic Reset ( C Annunciators o Low Protected Water Storage Tank Level o Low Low Protected Water Storage Tank Level j o High PWST Temperature j o Simultaneous Opening or Closure of the AFW Pump Inlet Valve and the AFW Pump Alternate inlet Valve o Flow Limiting of AFW o High AFW Supply Temperature o High/ Low Drive Turbine Speed

o High Drive Turbine Steam inlet Pressure

o Drive Turbine Group Alarm (Bearing and Lube Oil System) o AFW Pump Group Alarm (Bearing Temperature and Seal Cavity Pressure) o High Motor Bearing Temperatures

o Transfer Switches on Local SGAHRS Initiation Logic Trip

                                                                                       ^

1 o o Na Aerosol Concentration High o Na Aerosol Control Bypassed o PACC Startup on Reactor Trip "on Test"

l o PACC Start-up Delay Additional indicators and alarms are provided at the local instrumentation and control panels. Most information is also available to the operator via the Plant Data Handling and Display System (PDH&DS).

7.4-5 4 Amend. 64 Jan. 1982

7.4.1.1.10 Instrumentation Protected Water Storage Tank (PWST) Level The PWST level is measured to monitor the water Inventory available to be supplied to the steam drums in the event of loss of normal feedwater or the normal heat sink. The level is redundantly measured by two differential pressure sensors mounted across tap Iines near the top and bottom of the tank. A PWST level measurement signal is provided to the Plant Control System (PSC), PDH&DS, Plant Annunciator System (PAS) and to a PAM recorder. PWST Temperature The PWST water temperature is measured to monitor the capacity of the water inventory to provide an efficient heat sink. The temperature is measured by a single chromel-alumel thermocouple. The temperature signal from the transmitter is provided to the PCS, PDH&DS and PAS. Auxiliarv Feedwater (AFW) Flow The AFW flow is monitored to provide input to (a) restrict maximum flow through each control valve to less than 105 5% rated AFW flow, and, (b) initiate automatic closure of AFW isolation valves in locps with AFW flow greater than 150% for 5 seconds. The flow in each of the AFW iines is redundantly measured by two differential pressure sensors across one venturi. This provides capacity for four flow measurements per loop. A flow measurement signal is provided to the PCS, PDH&DS, PAS and to a PAM recorder. AFW Pomo Discharge Pressure The pressure of the water in the discharge line of the AFW pump is measured to provide the control of the valve in the recirculation line for AFW pump reduced flow operation. One pressure transmitter monitors the line pressure on the discharge side of each AFW pump. The pressur e measurements are provided to the PCS and PDH&DS. Drive Turbine Steam Inlet Pressure The AFW Drive Tubine steam inlet pressure is measured to provide a control signal to modulate the pressure control valve. A single pressure transmitter is located between the turbine inlet and the control valve. The signal is provided to the PCS, PDH&DS and PAS. Onlye Turbine Speed The AFW Drive Turbine speed is measured to provide a signal tc the turbi..e speed governor and for initiating an overspeed trip. A single magnetic pickup provides signals to the PCS, PDH&DS and PAS. O 7.4-5a Amend. 72 Oct. 1982

i Protected Air Cooled Condensor (PACCl o PACC Outlet Water Flow - sensed by one differential pressure sensor per loop across a venturl. Signals are provided to the PCS and the PDH&DS. o PACC Outlet Water Tamnerature - sensed by one chromel-alumel thermocouplo per loop. Signals are provided to the PCS and the PDH&DS. o PACC Outlet Air Tamnerature - sensed by three chromel-alumel thermocouples per loop. Signals are provided to the PCS from only the "A" outlet, o PACC Inlet Louver Position - sensed by two louver position sensors per loop. Signals are provided to the PCS and the PDH&DS. o PACC Outlet Louver Position - sensed by two switches per louver. Signals are provided to the PCS and the PDH&DS. o PACC Fan Blade Pitch Position - sensed by one pitch position sensor per fan (i.e. - two per loop). Signals are provided to the PCS and the PDH&DS. Isolation and Control Valve Positions The position of each valve is sensed by two limit switches; one indicates the valve is open, one indicates the valve is closed. The "Open/ Closed" position signal is provided to the PCS and tne PDH&DS. The monitored valves are:

   '-        - PWST Fill Valve
             - Alternate AFW Supply Valve AFW Pump Inlet Valve
             - AFW Pump Alternate inlet Valve AFW Pump Recirculation Valve
             - AFW Control Valve
             - AFW isolation Valve AFW Pump Test Loop Isolation Valve
             - Drive Turbine Steam Supply isolation Valve
             - Drive Turbine Pressure Control Valve
             - Superheater Vent Control Valve
             - Steam Drum Vent Control Valve
             - Turbine Drive Governor Valve
             - PACC Noncondensible Vent Valve i

O t 7.4-5b Amend. 72 Oct. 1982 l

7.4.1.2 Desion Analysis To provide a high degree of assurance that the SGAHRS will operate when necessary, and in time to provide adequate decay heat removal, the power for the system is taken f rom energy sources of high rel labil ity which are readily avai l abl e. As a safety related system, the Instrumentation and control s critical to SGAHRS operation are subject to the safety criteria identifled in Section 7.1.2. Redundant monitoring and control equipment will be provided to ensure that a single f ail ure will not impair the capabil ity of the SGAHRS Instrumentation and Control System to perform its intended safety function. The sy stem w il l be designed for f all safe operation and control equipment where practical and w il l, in the event of a f ailure, assume a f ailed position consistent with its intended saf ety f unction. Because there are three redundant decay heat removal loops, the instrumentation and control s associated w ith each Individual loop (e.g., auxil iary feedwater flow and air cooled condenser control systems) do not independently meet single f ailure criteria. However, when taken col lectively as a system, they provide the single f ail ure capabil Ity required. 7.4.2 Outlet Steam isolation Instrumentation and Control System 7 . 4 . 2.1 DesIan DescrlotIon 7 . 4 . 2.1.1 Function The Outiet Steam Isolation Subsystem (OS1S) provides Isolation of steam system pipe breaks. Steam system isolation is a necessary f unction f or saf e shutdown in those pipe break conditions af fecting the three steam supply systems and is provided if needed on a per loop basis. By def inition, this zone of protection w il l incl ude the high pressure steam supply system downstream frcm the Individual loop check val ves. The OSIS Control s are designed to the IEEE Standards l isted in Table 7.4-3. l l l t l l l 7.4-6 Amend. 72 Oct. 1982

7.4.2.1.2 Eauf orcent Design /N A high steam flow-to-feedwater flow ratio is indicative of a main steam supply (V ) leak down stream from the flow meter or Insufficient feedwater flow. The superheater steam outlet valves and superheater bypass valves shall be closed with the appropriate signal supplied by the heat transport instrumentation system (Section 7.5). This action will assure the Isolation of any steam system leak common to all three loops and also provide protection against a major steam condenser leak during a steam bypass heat removal operation. 7.4.2.1.3 Initiating Circuits The OSIS is initiated by the SGAHRS Initiation signal coincident with either a low superheater steam pressure signal or a high feedwater header pressure signal. The SGAHRS Initiation signal is described in 7.4.1.1.3. This initiation signal closes the superheater outlet isolation valves in all 3 loops when a high steam-to-feedwater flow ratio or a low steam drum level occurs in any loop. In each Steam Generator System loop, the three trip signals for high steam-to-feedwater flow ratio and the low steam drum level are input to a two of three logic network. If two of three trip signals occur in any of the 3 loops, the OSIS is Initiated, and all 3 loops are Isolated from the main superheated steam system by closure of the superheater outlet isolation valves and superheater bypass valves. 7.4.2.1.4 Byoasses and Interlocks Control interlocks and operator overrides associated with the operation of the superheater outlet isolation valves have not been completely defined. V Bypass of OSIS may be required to allow use of the main steam bypass and condenser for reactor heat removal. In case the OSIS is initiated by a leak in the feedwater supply system, the operator may decide to override the closure of certain superheater outlet isolation valves. 7.4.2.1.5 Redundancv and Diversity Redundancy is provided within the initiating circuits of OSIS. The primary trip function takes place when a high steam-to-feedwater flow ratio is sensed by two of three redundant subsystems on any one SGS loop. The low steam drum level sensed by two of three O U Amend. 72 7.4-7 Oct. 1982

redundant channels in any one loop provides a backup trip f unction, j Additional redundance is provided by three independnt SGS steam supply locos serving one common turbine header. Any major break in the high pressure steem i system external from the individual loop check valves will be sensed as a steam feedwater flow ratio trip signal in al l three l oops. 7 . 4 . 2.1. 6 Actuated Device The superheater outlet isolation and superheater bypass valves utilize a high rol labil ity electro-hydraul ic actuator. These valves are designed to f all closed upon loss of electrical supply to the control solenoid. 7.4.2.1.7 Seoaration The OSIS instrumentation and Control System, as part of the Decay Heat Renoval System is designed to maintain required isolation and separation between redundant channel s (see Section 7.1.2). 7.4.2.1.8 Ooorator Information Indication of the superheater outlet isolation valve position is supplled to the control room. Indicator lamps are used f or open-close position indication to the pl ant operator. 7.4.2.2 Design Analysis To provide a high degree of assurance that the OSIS will operate when necessary, and in time to provide adequate isolation, the power for the system is taken f rom energy sources of high rol labil ity which are readily available. lh As a saf ety rel ated system, the ir.strumentation and control s critical to OSIS operation are subject to the safety criteria identif ied in Section 7.1.2. Redundant monitoring and control equipment w ill be provided to ensure that a single f ail ure w ill not impair the capabil ity of the OSIS instrumentation and Control System to perf ccm its intended saf ety f unction. The system w il l be designed f or f all saf e operation and control equipment, where practical, wil l assure a f ailed position consistent with its intended saf ety f unction. 7.4.3 Pony Motors and Controls There are six pony motors, one in each primary and intermediate heat transport l oop to prov ide sodium fl ow for decay heat removal . These motors through the use of a gear box are capable of providing f ive to ten percent sodium flow in f ive discrete steps by gear changes. Section 5.6 describes the interaction of the primary and Intermediate heat transport loops with the SGAHRS to provide decay heat removal . 7.4.3.1 Design Descriotion The pony motors are 75 horsepower, 480 VAC, 3 phase, 60 Hz, total ly encl osed f an cooled Cl ass 1E motors. These motors are mounted on top of the sodium pump vertical drive motor. They are 1600 rpm motors which del iver power to the sodium pump via a reducing gear, an overrunning cl utch, and the vertical motor shaft. 7.4-8 Amend. 72 Oct. 1982

The overrunning clutch allows the pony motor to run continuously during all ('_ ' modes of plant operation and automatically drives the pump when the vertical \m motor speed decreases below the output speed of the reducing gear. Thus, af ter a reactor trip and pump (vertical drive motor) trip sodium flow does not decrease below pony motor fIow. During normal operation at pony motor speeds the external oil cooling system is in operation. However, the vertical dirve motor bearings are designed to start and operate continuously at pony motor speed without the external oil cool ing system or high pressure l if t pump. The pony motor is control led using both Non-class 1E and Class 1E circuit. The Non-class IE circuit is isolated f rom the Class 1E circuit and is overricden by the Class 1E circuit. Normal pony motor start is through a Non-1E permissive sequence circuit which f irst starts the vertical drive motor external circuit which f irst starts the vertical drive motor external lubricating oil cooling system and high pressure l ube oil pump. When the oil system achieves flow and pressure the pony motor starts. Once started the Class 1E circuit takes over and the loss of the external lubricating oil system will not result in a pony motor trip. This method of starting is not classified as safety-related and is used f or starting the pony motor during reactor shutdown periods af ter maintenance which requires the pony motor to be of f. The Class IE controls start the pony motors without the use of the external lubricating oil cooling system or high pressure lube oil pump. This f unction

     ) is carried out by a start-stop switch on the main control panel in the control

(/ x- room. Once started by either the Class IE or Non-class 1E control the pony motor w ill automatical ly restart f ol lowing the loss of of f-site power on the Class 1E diesel s. 7.4.3.2 initiatina Circuits The pony motor runs continuously during all modes of plant operations except during reactor shutdown f or maintenance. During maintenance only one loop is permitted to be out of service. Theref ore, there is no need f or automatic or manual Initiation circuits. However, the Class 1E start-stop switch is l ocated on the main control panel. 7.4.3.3 Bvoasses and Interlocks There are no bypasses in the Class 1E control circuit. The only condition which results in an interlock / automatic pony motor trip is a sodium-to-water leak in the steam generator modules. This results in an automatic trip of the af fected Intermediate heat transport loop pony motor only. The sodlun-to-water leak trip is describe in 7.5.6. 7.4.3.4 Analvsis The pony motor and the Class 1E control circuit is designed to the IEEE Standards Iisted in Table 7.4-3 and is qualified in accordance wIth Section (,) (~'s 1.6 Reference 9. 7.4-8a Amend. 72 Oct. 1982

i Remote Shutdown System l7.4.4 l 7.4.4.1 Design Descriotion Function l7.4.4.1.1 The Remote Shutdown System provides the means by which (1) saf e shutdown conditions of the reactor plant can be establ ished and maintained f rom locations outside of the Control Room in the event that the Control Room must be vacated; (2) hot shutdown conditions can be achieved and maintained; and, (3) If desired, the plant can be cooled to and maintained at the ref uel ing temperature. l 7.4.4.1.2 Design Basis The Remote Shutdown system is designed to use equipment located outside of the Control Room to place the reactor and plant into a saf e shutdown condition under the f ol Iow ing conditions: (a) The evacuation of the Control Room is not coincident w ith any other abnormal plant condition with the one exception that loss of of f site power may cccur. (b) No severe natural phenomena such as earthquake, tornadoes, hurricanes, floods, tsunami and seiches (free 10CFR50, Appendix A, Criterion 2) occur coincidently with the excavation of the Control Room. (c) The plant remains in an orderly shutdown status f rom the initiation of the evacuation of the Control Room to the time that command of the shutdown is re-estabi ished outside of the Control Room. (d) The remote shutdown operations wil l be commanded f rce one location and w il l use pl ant systems operated in their local mode to ef fect the shutdown and decay heat removal . (e) Pl ant instrumentation and control systems required f or rcoote shutdown operations w Il I have transf er sw itches I ocated at the I ocal panel s to permit the plant operating personnel to select to operate f rom the local panel s while isolating the remote controls or, conversely, to operate f rm the control room whil e isol ating the local control s. The transfer of control of a plant system f rom the rcrrote to the local mode is annunciated in the control rocm. (f) Communications between the Rmote Shutdown Monitoring Panel (RSMP), the command location f or remote shutdown operations, and toe SGAHPS panel s and other Iocal panel s durirg remote shutdown operations wil I be by the Maintenance Ccrnmun ication Jacking (MCJ ) system util iz ing a sound-powered telephone. O 7.4-Ob Amend. 72 Oct. 1982

l 7.4.4.1.3 Remote Shutdown Ooerations v/ The RSMP will be located in Cell 271 of the 836'-0" level of the SGB. The l RSMP will have indications (see Section 7.4.4.1.4) from which an operator can assess the progress of the shutdown, and it will be the location f rom which that operator will command the operation of the plant systems being operated in their local mode to ef fect shutdown. The Division 1,11 and III SGMRS (Section 7.4.1) local panel s wil l be located in Cells 272A, B and C respectively, in close proximity to the RSMP, on the 836 '-0" l evel of the SGB-18. The SGAHRS, operated in its local mode, will be used to control the removal of heat from the reactor plant to achieve and stabil Ize the plant at the desired plant temperature (hot shutdown or refueling temperature). The Iocal SGMRS panels wIlI have alI of the controls and indications necessary to completely control the system. All signals f rom the Control Room to thc SGMRS panels are buf fered to prevent f aults occurir.g in the Control Room from propogating back to the SGAHRS panel s. All SGMRS component controls can be transferred to local at the local SGMRS panels. Placing the transfer switches in " local" overrices all control functions in the Control Room. The Division 1,11 and III OSIS local panels are located in SGB Cells 272A, B and C wlth the SGMRS panels, and wIlI be operated in the Iocal mode when required to control heat removal from the plant in conjunction w ith the operation of SGAHRS. Isolation of OSIS panel controls f rom the Control Room is incorporated in the design. Steam drum drain and superheater outlet isol ation val ve control s can be transf erred to local at the local OSIS panel s.

  \

v' Whenever any SGAHRS component control transf er sw itch Is pl aced in the "Iocal" position an alarm is infilated in the Control Room to alert the Control Room operator. The same statement is true for the steam drum drain controls and superheat outlet isolation valve controls on the OSIS panels, if of f site power is lost coincident wIth having to achieve a safe shutdown condition in the reactor plant f rom outside of the Control Room, the diesel generators will start and f unction in accordance with the design provided by the Bull ding Electrical Power System. Any operator actions required in conjunction with operating and loading the diesel generators will be done in the local operating mode at the DG local panels. In the event that the Control Room must be vacated, reactor scram and SGAHRS operation w ll l be initiated manually. The operating personnel wilI move to the 836'-0" level of the SGB where the SGMRS in the local mode will ef fect teat removal and stabil Ization of the pl ant temperatures. Operation of the SGAHRS in the local mode will ef fect heat removal and stabil ization of the plant temperatures. The plant shutdown wIlI be directed by the operator at the RSMP who w il I al so assign operating personnel not continuously occupied in operating SGAHRS to oversee or operate other systems as required. Movanent of personnel within the plant and access to building cells and local panels will be controlled by the f acilities and procedures of the industrial Security Sy stem. O G 7.4-8c Amend. 72 Oct. 1982

f 7.4.4.1.4 Eautoment Desian The RSMP is the only piece of equipment provided by the Renote Shutdown Sy stem. It will be a vertical sided, non-Class 1E cabinet assembly containing meters and a phone Jack panel. The meters wil l receive buf fered signal s f rm, the initiating systems and, thus, do not require transf er switches to isolate them f rm the Control Room. The phone Jack panel will permit the operator at the RSMP to communicate with ihe f ive NSSS or Nuclear Island buildings by means of any of the three MCJ circuits provided in each of the buildings. In addition, communications among the bulldings can be estabiished through the phone Jack panel on the RSMP. The indications provided on the RSMP are as f ollows: o For each prirrary heat transport system locp, 1 - Pump outlet sodium temperature indicaton (3 total) 1 - Reactor iniet sodium temperature indication (3 total) 1 - Sodium pump shaf t speed Indication (3 total) o For each intermediate heat transport system locp, 1 - IHX outlet sodium temperature indication (3 total) 1 - lHX inlet sodium temperature indication (3 total) 1 - Sodium pump shaf t speed Indication (3 total) o For each superheated steam Iocp, 1 - Tmperature indication (3 total) 1 - Steam f l ow indication (3 total) o One reactor vessel sodium level meter (long probe) o For each Diesel Generator (3 total) 1 - Wattmeter 1 - Frequency meter 1 - Varmeter 1 - Vcl tmeter w Ith phase sel ector sw itch 1 - Ammeter w ith phase sel ector sw itch in addition to the foregoing indications, other indications used during ranote shutdown operations that are not on the RSMP will be available as f ollows: o SGAHRS Controls and indicators used for the operation of each SGAHRS division are located on the three seperate SGAHRS panelt. In cel l s 272A, B, and C. Each SGAHRS division is separate and redundant f rm the other divisions. See the response to Question CS421.04 for additional inf ormation about SGAHRS division assignments. O 7.4-8d Amend. 72 Oct. 1982

The f ol low ing control s, Indicators and alarms are on each SGAHRS panel." () Controllers Auxil iary Feedwater Fl ow AFW Steam Turbine Steam inlet Pressure PACC Inlet Louver Position PACC Fan Blade Position Steam Drum Level Steam Drum Vent Superheater Vent Analoo Indicators Protected Water Storage Tank Level Protected Water Storage Tank Temperature Auxil lary Feedwater Flow Auxil lery Feedwater Pump Discharge Pressure Steam Driven Turbine Steam inlet Pressure Steam Driven Tubrine Speed PACC Outlet Air Temperature PACC Outlet Water Flow and Temperature PACC Inlet Louver Position , PACC Fan Blade Pitch Position Steam Drum Pressure and Water Level Annunciators

     )  Protectec Water Storage Tenk Level V     FWST Temperature AFW Supply Temperature Steam Driven Turbire Speed Driven Turbine Steam Inlet Pressure Steam Driven Turbire Bearing and Lube Oil Temperature High Motor Bearing Temperatures SGAHRS Initiation o         Diesel speed and f uel oil indications will be available at the diesel generator local control panel s in the Diesel Generator Buil ding Cells 511 and 512.
        *Each indicator, al arm and control ler is repeated on each of the SG AHRS panel s except f or those associated w Ith the AFW pumps. Panels A and B have the control s, alarms and indicators f or motor dr iven AFW pumps A and B; Panel B has those associated w Ith the steam driven AFW pump.

O 7.4-8e Amend. 72 Oct. 1982

7.4.4.2 Desion Analvsis The Remote Shutdown System provides the RSMP from which an operator can assess the progress of the plant shutdown and command the local operation of the pl ant systems (prirrarily SGMRS) to ef fect the shutdown, it shoul d be noted that the PACC subsystem of SGMRS is autanatically initiated by cli reactor trips, and it rernains in operation for the duration of the pl ant shutdown or as long as the reactor generates significant decay heat. The Remote Shutdown System imposes no special requirements on the pl ant systems, but takes advantage of the folIowing system design features: o The abil ity to operate in both local and remote modes with isol ation f ran and annunciation in the Control Room when operating in the local mode. o The redundancy diversity, separation, isolation and rol labil ity of the saf ety grade systems. o The design and location of saf ety grade systems equipment that minimize the probabil ity and of fect of fires and explosions en the abil Ity of the systems to perf orm their saf ety f unction. o The redundant saf ety grade SGMRS provides the capabil ity to achieve and maintain hot shutdown and, if desired, to cool the pl ant to and maintain the pl ant at ref uel Ing conditions. o When transf erring SGMRS to the local mode, the operator manually starts SGMRS. Once started, SGMRS automatical ly control s those parateters used to remove decay heat. The RSMP is a non-Cl ass IE Seismic Cl ass ill assembly and theref ore, is not subject to the separation requirements of IEEE 364-1974, or to the sei sm ic qual if Ication requirements of IEEE 344-1974, or to any of the other IEEE Standards I isted in Tabl e 7.1-3. 7 a-8f knend. 72 Oct. 1982

TABLE 7.4-2 NOMINAL SET POINTS (Cont'd) l NOTES:

1. The capability ' for the operator to assume manual control of the Indicated f unctions from either the control room or the local panel is provided.  ;

2. Valves will reopen should steam drum level fall to the low level trip (-8 in, from normal water level). Valves in the motor-driven AFW pump loops close at +8 In. from normal water level while the valves in the turbine-driven AFW pump loops close at +12 in from normal water level.

3. In the long term cooldown mode, the second motor driven pump automatically restarts af ter a 1-minute delay if steam drum level remains at -7 in, or lower.

4 . Steam drum pressure must be above 1000 psig to initiate turbine operation.'

5. PACC vent control valves are controlled by the temperature dif ferential between the noncondensible gas collection pipe and the steam saturation temperature measured in the PACC outlet header.

6. Normal steam drum water level is 1 inch above drum centerline.

O O 7.4-10c Amend. 71 Sept. 1982

T/BLE 7.4-3 LIST OF IEEE STANDARDS APPLICABLE TO SGAHRS AND OSIS INSTRUENTATION AND CONTROL SYSTEMS lEEE-279-1971 IEEE Standard: Cri1eria for Protection Systems f or Nuciear Power Generating Stations 1EEE-308-1974 Criteria f or Cl ass 1E Power Systems f or Nuciear Power Generating Stations I EEE-323-1974 IEEE Trial-Use Standard: General Guide f or Qual ify Ing Class 1E Electric Equipment for Nuclear Power Generating Stations I EEE-323-A-1975 Supplanent to the Foreword of IEEE 323-1974 IEEE-336-1971 lEEE Standard: I nstal l ation, i nspect i on, and Testing Requirements f or Instrumentat ion and Electric Equipment Durir.g Construction of Nuclear Power Generating Stations I EEE-338-1977 Crlierta for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Selsnic Qual if ication of Cl ass 1 Equipment f or Nuclear Power Generating Stations IEEE-352-1975 General Principles for Rellabil Ity Analysis of Nuclear Power Generating Station Protection Systems I EEE-379-1972 IEEE Trial-Use Guide f or the Appl Ication of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems IEE-382-1980 IEEE Standard f or Qual if ication of Saf ety-Rel ated Val ve Actuator s IEEE-384-1974 IEEE Trial Use Standard Crlieria for Separation of Class 1E Equipment and Circuits I EEE-494-1974 IEEE Standard Method for Identif ication of Documents Related to Class IE Equipment and Systems f cr Nuclear Power Generating Station O 7.4-10d Amend. 72 Oct. 1982

3 provide the required time response. The thermowell is also swaged at the tip. [d The thermocouples are spring loaded against the bottom of the well. Although f ailures of the wells are not expected, as confirmed by tests and analysis, the head of the thermowell, including the cable penetration, is sealed to provide a secondary boundary for the sodium. Tests have shown that this system will provide a time response less than 5 seconds. Flexible mica, polylmide and fiberglass insulated thermocouple extension wires in conduit are used to bring the signal s out of the Heat Transport System Cel L. The signal s are then routed to the containment mezzanine into reference junctions and signal conditioning equipment. The conditioned signals are transmitted to the control room for the Reactor Shutdown System logic. The Reactor Shutdown System provides buf fered signals to the PCS and PDH & DS. Primarv and Intermediate Hot and Cold Leg Temoerature The primary and Intermediate hot and cold leg temperature's are measured to determine and record operating conditions and to calorimetrically calibrate the permanent magnet flowmeters. The measurement is made by two duplex element resistance temperature detectors (RTDs) per loop, installed in thermowel I s. Although f alIures of the welIs are not expected, as confirmed by tests and analysis, the head of the thermowell, including the cable penetration, is sealed to provide a secondary boundary for the sodium. The signal s f rom the RTDs are routed to signal conditioning equipment which converts the resistance variation to a standard signal level for transmission to the PDH & DS. Primarv and Intermediate Pumo Discharge Pressure The primary ar.J intermediate pump discharge pressure measurements monitor pump performance. In addltion the primary pump outIet in conjunction wIth the intermediate IHX outlet pressure provide the primary loop / intermediate loop differential pressure. The measurements are made by pressure elements Installed in the elevated section of the drain Iine from the discharge piping of the sodium pump. NaK filled capillaries f rom the pressure elements are connected to pressure transducers which develop electrical signals proportional to the pressure. These pressure transducers provide a secondary boundary if the bellows in the pressure elements should f all. The conditioned signal is supplled to the PDH & DS. Since this pressure element is located in an inerted celi and replacement would require entry into the celi and draining of the loop, two pressure elements per loop are provided. Intermediate IHX Outlet Pressure The Intermediate IHX outlet pressure measurement is used to monitor the loop and lHX operational performance history. The measurements are made by pressure elements installed in the intermediate loop piping between the IHX and the superheater. NaK filled capillaries from the pressure elements are connected to pressure tranducers which develop electrical signals proportional to the pressure. The pressure transducers provide a secondary boundary if the bellows in the pressure elements should f all. The conditioned signal is suppl led to the PDH and DS. O 7.5-7 Amend. 72 Oct. 1982

                                                                           -r.- -

IHX Differential Pressure The primary sodium pump discharge pressure and the IHX Interneciate Loop outlet pressure detectors are used to provide a dif ferential measurement of the IHX Primary / Intermediate pressure dif ference, which is maintained above 10 psi during normal operating conditions. The dif ferential pressure treasurement is alarmed if the internediate loop pressure drops to 10 psi above the primary loop pressure io alert the operator for corrective action to assure intermediate to primary dif forential pressure is maintained above the minimum req u i red. Intermediate Pumo inlet Pressure The ir,termediate pump f olet pressure measurements provide a signal to monitcr pump perf ormance. lised w Ith the pump outlet pressure, the dif ferential pressure across the pump is obtained. in the primary loop, the reactor pressure is used f or th is surveil lence. The measurements are made by pressure elcments installed on the piping between the evaporators and the pump f riet. NaK fIlIed capilIarles f rom the pressure elements are connected 1o pressure transducers which develop electrical signal s proportional to the pressure. The pressure fransducers provide a secondary boundary if the bellcws in the pressure el anents shoul d f all . The conditioned signal is suppl led to the PDH & DS. Intermediate Exoansion Tank Level Two separate level measurement channel s are provided; both channels are used f or indication in the control room and DH & DS and f cr al arm. Al arn channel s proviae a broad range measurement that covers possible high and Icw l evel s during plant operation as well as the IHTS t ill level. The PDH & DS uses measurements f or irternediate loop sodium inventory (see al so Section 7.5.5). The l evel probes are designed to be repl aceable. Evaoorator Sodium Outlet Temoerature , Three thermocouple (a , described above in the paragraph on lHX outlet temperature) channels are provided to measure the sodium tempera 1urc at the outiet of the evaporators in each Iocp. The thermocoupl es are pl aced just af ter the pipes f rom each evaporater join to f orm two singl e I ines. These three signal s are conditioned separately and provided to the Reactor Shutdown Sy stem l ogic. The Reactor Shutdown System in torn provides buf fered signal s to the PDH & DS. 7 . 5 . 2.1. 2 Sodium Pumos Sodium Level Sodium level is measured in each pump tank. The signal provides indication and al arm. The alarm is used to notify the operater of abnorrnal operation and allcw initiation of action to prevent pump damage. The signal is also provided to the PDH & DS where it can be used in calculation of sodium inventory . 7.5-8 Amend. 72 Oct. 1982

Pony Motor The pony motors are Class 1E motors and are supplied power from the Class IE {JT 480 VAC busses. Non-class 1E signals are provided to the main control room to Indicate the i pony motors are running. These signals are pony motor speed indicators which are locat'ed in the main control panel and pony motor current which is available through the PDH & DS. Al so start and stop l ights are on the main control room panel which are from the pony motor starters. During pony motor operation Indication is available on the main control roon panel f rom sodi um fl ow. 7 . 5 . 2.1. 3 Steam Generator Sodlum Flow Venturi flowmeters are provided, one loop only, to accurately measure the sodium flow rate through each of the superheater outlet ports. The accurate flow data is used f or determination of the performance characteristics typical of the superheaters and evaporators. Sodium Temoerature The evaporator and superheater outlet temperature is monitored, on all three loops, by Resistance Temperature Detectors (RTD). The superheater inlet is monitored, on one loop only, al so by an RTD f or purposes of steam generator perf ormance eval uation. These temperature sensors provide signals f or the O' PDH & DS. The evaporator bulk outlet temperature is measured with three thermocouples and are part of the Reactor Shutdown System. Sodlum Pressure For the purpose of steam generator perf ormance evaluation, pressure is measured, in one loop only, at the superheater inlet, superheater outlet (both legs) and evaporator outlet. The type of pressure sensor used is the same as the one f or Intermediate pump Inlet pressure. These pressure measurements provide pressure signals to the PDH & DS. Steam and Water Flow o Feedwater Mass Flow - sensed by three dif ferential pressure et enents across one venturl in the inlet i Ine to each steam drum, a 7.5-9 Amend. 72 Oct. 1982

• The temperature corrected feedwater flow signals are supplied to the Reactor Shutdown System logic. The Reactor Shutdown 59l System provides buffered signals to PCS and PDH & DS.

e Steam Mass Flow - sensed by three differential pressure elements across one venturi in the outlet of each superheater. The temper-ature and pressure corrected mass signals are supplied to the Reactor Shutdown System logic. The Reactor Shutdown System pro-59l29 vides buffered signals to PCS and PDH & DS. e Steam Drum Blowdown Flow - sensed by flow orifice (differential pressure) in the blowdown line for each steam drum. The signal is 59l provided to the PDH & DS. e Evaporator Inlet Flow - sensed by a differential pressure element across a venturi in the inlet line to one of the evaporators in one loop only. This is to aid in the performance evaluation of a typical evaporator module. Steam and Water Temperature e Feedwater Temperature - sensed by three resistance temperature detectors in the steam drum inlet line. The signal provides temperature compensation for the feedwater flow signal. Buffered 59l signals are supplied to the PDH & DS. o Recirculating Water Temperature - sensed by a thermocouple detector in the recirculation pump discharge header. The signal is pro-59l vided to the PDH & DS. e Saturated Steam Temperature - sensed by,a thermocouple detector in the outlet header from the steam drum. The signal is provided 59 l to the PDH & DS. e Superheat Steam Temperature - sensed by three resistance tempera-ture detectors in the superheater outlet line. The signal pro-vides temperature compensation for the steam flow. Buffered sig-59l nals are supplied for PCS and PDH & DS. e Evaporator and Superheater Inlet and Outlet Temperature - sensed by RTDs located at the inlet and outlet nozzles for one evap-l orator and superheater in one loop only. Used for performance evaluation for a typical generator module. e Steam Drum Blowdown Temperature - sensed by a thermocouple located on the blowdown line. The signal provides temperature compensation for the steam drum blowdcwn flow and is also supplied 59l to the PDil a DS. Amend. 59 7.5-10 Dec. 1930 l

7.5.5.1.1 Design Bases and Design Criteria For the Liquid Metal-To-Gas h3J Leak Detection System v The design bases of the Liquid Metal-to-Gas Leak Detection System arises f rom the need to protect plant equipment, considerations of maintenance and plant availability, and the corrosion ef fects of sodium compounds on stainless steel s at high temperatures. Considering the significance of corrosion with respect to piping integrity, it is appropriate that the design criteria assure that the Liquid Metal-to-Gas Leak Detection System provide rellable detection for the Primary and Intermediate Heat Transport in-Containment Systems in a smalI fraction of the nominal time to penetrate the pipe by local corrosion. The offects of corrosion on the CRBRP FHTS piping have been thoroughly assessed in WARD-D-185

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

Design Criteria have been established to guarantee reliable plant operation w ith pipe temperatures greater than 700 F. These incl ude:

1. The PHTS and in-containment lHTS shall be monitored for leaks by diverse methods each capable of providing the required time response.

2. Capability shall be provided to procure a filter sample for laboratory analysis to provide a highly rel lable confirmation method. Fil ter samples should be analyzed a minimum of once every 1000 hrs.

3. The Liquid Metal-to-Gas Leak Detection System must operate af ter an ope'r ating basis earthquake (OBE).

4. The leak detection system shall be equipped with provisions to readily permit testing f or operabil ity and cal ibration during pl ant operation.

5. A rel iable sel f-monitoring provision shall be provided to detect component f alI ure.

7.5-18a Amend. 72 Oct. 1982

6. Upon loss of abil Ity to f ul fil l the specif ied time response, the pl ant will be placed in a hot shutdown condition.

7. The system shall be qual ified to operate in its environment.

Additional Design Criteria of the Liquid Metal-To-Gas Leak Detection System required to protect plant and capital investment, Iimit maintenance and protect pl ant avalI abII Ity are outl Ined below:

1. The Liquid Metal-To-Gas Leak Detection System shall detect and locate l iquid metal-to-gas leaks throughout the pl ant between the temperatures of 375 to 1000 F as required to ful fili continuous monitoring requirements of Appendix G, "CRBRP Plan For Inservice and Preservice Inspections."

2. The Leak Detection System shall be able to identify the general location of the leak.

This system is not needed for Initiation of plant shutdown, for removal of decay heat or for reduction of of f-site radiation exposure to acceptable l evel s; theref ore, it is cl assif ied as a non-saf ety system. The saf ety rel ated Instrumentation provided to accommodate l iquid metal leaks is described in Section 7.5.3.1.1. The passive engineered saf ety features provided to mitigate the ef fects of I ! quid metal leaks are described in Section 3.8. O 1 l l l 7.5-1gy Amend. 72 Oct. 1982 l

The electrical sensing types of detectors (cables and contact) respond wIth an ' b V alarm when i Iquid metal causes an electrical short between the electrode and Its protective sheath. The Sodium lonization Detectors (SIDs) pgide an alarm when the aerosol concentration reaches a level of about 10 gm/cc. The PFADs, which are Integrating devices, respond with an alarm when the dif ferential pressure across a filter has increased by 2 inches of water. The tima for this response is relaty to aerosol concentration as shown on Figure 7.5-7. For exampl e, at 1 x 10- gm/ce, the time response is approximately 250 hours. Both SIDs and PFADs have filters which are chemically examined for sodium on a monthlgbasis so that leaks which result in aerosol concentrations l ower than 1 x 10- gm/cc w il l al so b e . A le msulmg N a concentration of approximately 2 x 10-i3 gm/cc is detectable by chemical examination of the filter pads. The sodium aerosol concentration resulting f rom a 100-gm/h leak in inerted CRBRP cel ls ranging in vol ume f rcrn 15,000 to 115,000 ft3 is shown on Figure 7.5-8. In the operating temperature range of 700-1000 F, the leak detection criteria are easily met with either SIDs or PFADs. In addition, during reactor operation, the radiation particulate monitoring systemgill detect leaks resulting in aerosol concentration of approximately 10- gm/cc in those cells containing primary sodium. The aerosol detectors are connected tc the PDH&DS so that the rate at which the signal is changing can be checked af ter a leak alarm is obtained. A rapid increase in PFAD dif ferential pressure (less than I hour f rcm normal reading to alarm) accompanied by leak alarms f rom other detectors in the same area woul d indicate a l arge l eak (greater than 1 gpm). Conversely, a leak signal that took 10 to 100 hours or more to reach the alarm level would Indicate a m smal I (100-1000 gm/h) leak. The SIDs are calibrated so that aerosol concentration can be related to the signal level. Instruments are set to al arm at specif ic aerosol concentrations. The l iquid metal-to-gas l eak detection system is designed to f unction af ter an OBE. The radiation particulate monitoring system is designed to f unction af ter an SSE. All leak detection equipment w il l be tested periodically to demonstrate operabil ity. The increase in cell atmosphere temperature and pressure in the event leaks larger than 20 kg/ min as detected by temperature and pressure sensors can provide an additional source of leak detection. The abil Ity to detect smalI leaks ( 100 gm/hr) by several methods in hours pl us the abil Ity to detect I arge Ieaks (>kg/ min) in minutes wilI provide a highly rel table leak detection system that provides the operator information to enable shutdown to repair defects without extensive time for cleanup operations. Af ter a sodium or NaK leak has occurred, the Liquid Metal-to-Gas Leak Detection System equipment impacted by the leak will be either replaced or cleaned (pneumatic system rinsed with alcohol) to remove sodium leak residue products. The system will then be acceptance tested and cal ibrated in accordance w ith the preoperational test specif ication criteria util ized prior to inital pl ant startup. O V 7.5-20 Amend. 72 Oct. 1982

Table 7.5-3 gives the primary and back-up methods of leak detection f or the principal sodium systems and components in the plant. The methods shown in the table are related to the three sizes of leaks defined in Section 7.5.5.1.2. The principal methods of leak detection are described below. Aerosol Monitorina Aerosol monitoring will be perf ormed by measuring the pressure drop across a membrane f il ter w ith a constant flow of gas sampl ed f rom the annul ar space between major piping and its insulation, from the space within guard vessels, and f rom cel ls containing i Iquid metal systemt. Another cell aerosol monitoring method uses a sodium Ionization detector. Liquid Metal aerosols or vapor are Ionized by a hot f il anent and the ion current is measured. increases in the ion current indicate a leak. Based upon the experimental results, these methods provide f or detection of leaks of 100 gm/hr and less, with a response time depending on temperature and the vol ume being monitored. The major f unction of this instrumentation will be to provide indication of the presence of small leaks which do not present a significant contamination hazard, but which might result in undesirable long-term corrosion. Contact Detectors (Soark-Plua) Contact detectors consist of a stainless-steel-sheathed, mineral oxide-Insulated, two-wire probe with the sensing end open and the wire ends exposed. Contact detectors are instal led, for example, on bel lows sealed val ves w Ith the sensing end between the bellows and the mechanical backup seal. A leak is detected by the reduction in circuit electrical resistance caused by sodium contacting the wire ends. Cable Detectors Cabl e detectors consist of stainl ess-steel-sheathed, mineral-oxide-Insul ated, cable wIth holes penetrating the sheath to permit leaked liquid metal to come in contact w ith the conductors. Cabl e detectors w il l be pl aced, f or exampl e, in the bottom of guard vessels and below large tanks. Other Detection Methods Pressure and temperature measurements available in the inerted cells (Section 9.5.1.5) will provide immediate Indication of the presence of large leaks over the 20 kg/ min size. In the case of systems containing radioactive sodium, the detection of airborne radioactivity arising f rom Na-24 or Na-22 in the aerosol s wil l be perf ormed by particul ate radiation monitoring equipment (Section 11.4.2) which provipgs a sensitive detection method f or aerosol concentrations as l ow as 10 gm/cc. Chemical analysis provides positive concentrations of approximately 10 ,jetection capebil ity for aerosolgm/cc, depending period. l O 7 .5 -21 Amend. 72 l Oct. 1982 I

7.5.5.1.1.1 Deslan Descriotion General Detection equipment is provided to monitor the primary and intermediate sodium coolant boundaries to identify comparatively small leaks when they occur. The leak detection methods selected for the following installations are:

1. Particulate monitors (radiation detectors), Sodium lonization Detectors (aerosol detectors), and chemical analysis for atmosphere monitor.ing in selected cells.

2. Plugging filter aerosol detectors (PFADs) for Main Heat Transfer System piping and guard vessels, major components, and f or inerted cel l atmosphere monitoring.

3. Contact detectors in the space between the bellows and the stem packing of the bellows sealed sodium valves.

4. Cable detectors in guard vessels and under major liquid metal components.

Of the types of leak detection devices that comprise the Leak Detection System, only sodium aerosol leak detection devices show a dif ference in their response when operated in an air atmosphere as opposed to an Inert atmosphere. The time for a detector to respond to a leak in air is generally shorter than in an inert atmosphere. Tho electrical sensing types such as cable and (. contact detectors shew no dif ference in response due to operating atmospheres. t However, the potential for higher moisture content in air can result in greater inhibition to sodium flow when the leak is very small. Considerations which materially af fect detection times include: sodium leak rate, sodium temperature, and cel I size. Test data (See Reference 5) confirm that sodium leaks of 100 gm per hour in an air or inert atmosphere can be detected by aerosol detection over the operating temperature ranges, within the detection time periods identifled in Figure 5.1.1 of WARD-0-0185,

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

D 7.5-19 Amend. 72 Oct. 1982

4 Other Rackun Detection Method f Liquid Sodlum Level Sensors In the reactor, the EVST, the IHTS expansion tank, and sodium storage tanks will provide Indications of large leaks. Smoke detectors (Fire Protection System) will detect combustion products originating from sodium leaks in air (See Section 9.13.2). i I i i l .: l i O t 7.5-21a Amend. 72 Oct. 1982 j I

Indication in Control Room An audible group alarm is sounded in the control room upon Indication of a leak or certain f ailures of contact, cable, or aerosol channels. The channel number producing the alarm and the location of the region covered by this channel are displayed on an annunciator on a local panel. This information will identify the leak as occurring in a specific major comonent or series of pipe sectl'o ns, or specific belIow-sealed valve, or the celi containing the leaking system. The leak detection system uses the Plant Data Handl ing System for channel f ailure monitoring, data and trend logging; the sampiIng time interval wIlI nominally be approximately 30 seconds. No automatic isolation f unctions or reactor scram are initiated by the Liquid Metal-To-Gas Leak Detetion System. Isol ation or shutdown of a system show ing a leak will be performed manually, following verification of the leak and review of the operating conditions. 7.5.5.1.2 Design Analvsis The Liquid Metal-to-Gas Leak Detection System will meet the appropriate requirements of CRBR Design Criterion 30, " Inspection and Surveillance of Reactor Coolant Boundary and Criterion 33, " Inspection and Surveillance of Reactor Cool ant Boundary. Criterion 30 requires that means be provided for detecting and identifying the Iocation of the source of reactor coolant leakage f ran the reactor coolant boundary to the extent necessary to assure that timely discovery and correction of leaks which could lead to accidents b V whose consequences could exceed the Iimits prescribed for protection of the heal th and saf ety of the publ ic. Criterion 33 requires that means be provided f or detecting intermediate coolant leakage from the intermediate coolant boundary. In order to demonstrate how the intent of the criteria will be satisfled, the Instrumentation requirements met by this system for three different ranges of leaks are discussed. These ranges have been selected to analyze situations which cover the complete range of leak detection I n str ument s. Section 15.6 discusses the consequences of leaks for the health and saf ety of the publ Ic. Large Leaks This category covers f ailures up to those resulting in a leak of 30 gpm or 100 kg/ min. A significant physical characteristic of leaks of this size is that they would result in pressure and temperature changes in the primary celIs if the leak occurs in PHTS pipe sections. This feature sets the lower boundary of the leak at about 20 kg/ min; this being an estimte of the amount of sodium which would result in measurable changes in cell pressure and temperature. If the leak occurs in a guard vessel, continuity detectors will provide detection m 7.5-22 l Amend. 72 l Oct. 1982 i ._. _

of these large leaks. Leaks of this magnitude would be detected in five minutes or less for the primary and intermediate heat transport system. The operator would then be able to initiate and complete plant shutdown within ten minutes after the start of the leak. The pressure and temperature measuranents available in the inerted cells will, in conjunction with the aerosol detectors, continuity detectors and radiation monitors, provide the response required for proper operator action in case of leaks of this magnitude. Intermediate Leaks Intermediate leaks were defined as those leaks which would not result in signifIcont changes in celi pressures and temperatures but where the extent of the resulting contamination and plant maintenance makes plant shutdown desirable. The range of leak rates covered extends from the lower limit of the large leaks previously considered down to a leak of 100 gm/hr. The detection times for the wide range of leaks in this group would vary from a few minutes to several hours depending on the rate of leakage. Based upon experimental results, it is concluded that several systems would detect a leak of this magnitude in several hours at least and possibly in minutes, instrumentation capable of detecting leaks of this magnitude include radiation monitors, continuity detectors, and the dif ferent types of aerosol detectors. l SmalI Leaks Small leaks at or below 100 gm/hr were defined as those events resulting in O releases of sodium which do not pose a contamination or maintenance problem but might result in undesirable long-term corsosion (see Section 5.3.3). The methods for detecting leaks of this range are aerosol detectors and radiation monitors in the case of the primary systan. In the course of test programs, aerosol concentrations produced by leaks of down to 5 gm/hr were found to be within the detection capability of both a Sodium lonization Detector and a Plugging Filter Aerosol Detector in test chambers. The test results show that leaks of this size can be detected in the range of one hour to 24 hours by annuti monitors depending upon the sodium temperature and gas environment. It is deduced from the test results that very small leak (<1 gm/hr) will be detected by annull monitors in several days. O 7.5-23 Amend. 72 Oct. 1982

Tests during 1975 and 1976 showed that under environmental conditions typical of LMFBR operation, small leaks from typical piping configurations can be [-) detected by both Sodium lonization and Plugging Filter Aerosol Detectors. O Continuity (cable or contact) detectors did not reliably detect small pipe leaks under these conditions. Testing in 1978 verified the performance of aerosol detectors using prototypic CRBRP cell atmosphere recirculation as well as pipe / insulation design. It is deduced from the test results that the sodium vapor / aerosol systems will, in conjunction with existing radiation monitoring technology, provide adequate Indication of the smal test sizes of leaks of interest. Sodium Leaks into an Air Atmosohere Test results indicate that the methods applicable to sodium leaks in inerted cells will also operate when applied in an air atmosphere. The additional use of smoke detectors and the accessibility of piping located in an air atmosphere to visual inspection assist in the selection of an ef fective sodium-to-air leak detection system. 7.5.5.2 Intermediate to Primary Heat Transoort Svstem Leak Detection 7.5.5.2.1 Design Descriotion The IHTS pressure is maintained at least 10 psi higher than the Primary Heat Transport System at the IHX to prevent radioactive primary sodium from entering the IHTS in the event of a tube leak. Maintaining a positive n pressure dif ferential across the IHX is a limiting condition for operation of ( i the plant (Chapter 16 - Technical Specifications). This prov ides assurance that a zero or negative dif ferential will not exist during any extended interval. A loss of this pressure or a reversal of it is not expected to occur except during accident conditions. Such an occurance would necessitate an orderly plant shutdown to correct the problem. Since a reverse dif ferential cannot occur for a significant interval. the potential leakage of primary , sodium into the intermediate system, through an lHX tube leak, is smal1. Leakage of primary sodium into the IHTS, should it occur, wil I be detected by radiation monitors provided on the IHTS piping within the SGB. The radiation monitor systan will provide an Indication of the radiation level and will provide alarms for conditions of excessive radiation indicative of ingress of primary sodium. Since the only activity expected in the IHTS is a low level of tritium, the radiation monitors will be very sensitive to the presence of significant amounts of radioactive primary sodium in the intermediate system. For accidents which involve a loss of IHTS boundary integrity the radiological ef f ects have been evaluated. The results of these evaluations are presented in Sections 15.3.2.3, 15.3.3.3 and 16.6.1.5. 7.5-24 Amend. 71 Sept. 1982

Maintaining a positive pressure dif ferential across the IHX assures that the leakage across the lHX tube barrier will result in an inflow of sodium into the primary system causing a loss of sodium Inventory in the IHTS. The sodium inventory in the IHTS is monitored by tracking the sodium levels and correcting for loop temperature effects. Alarms are provided in the control room to alert the operator upon detection of a Iarge 1oss of IHTS sodium inventory. 7.5.5.2.2 Desicn Analysis Intermediate to Primary Heat Transport System leak detection is provided to comply with CRBRP General Design Criterion 36 " Inspection and Surveillance of Intermediate Coolant Boundary". In order to demonstrate now the Intent of this criterion will be satisfied, an analysis of the minimum detectable leaks in the IHX is provided below. The minimum detectable level change of sodium in the IHTS pump and expansion tank is approximately 3 inches which corresponds to about 150 gallons. In the event of a f ull-circumferential break of an lHX tube, the leak rate of intermediate sodium to the primary side of the IHX would be approximately 150 gpm. At this leak rate, the detection time would be about one minute assuming steady state temperature conditions. Based upon a 3-inch level change, leakage of as low as 6.25 gph would fall within the detection threshold. Over long time periods, the sensitivity of the detection system will be reduced by an insignificant amount due to other potential leakages from the system. If leakage occurs due to piping or

 . component leaks, the external leak detection system will detect the leakage.

A second potential source is leakage through the four sets of dump valves which has a, maximum expected rate of one to two gallons per day. Since this leakage rate is essentially two orders of magnitude smaller than the leakage threshold, It wIII not have a consequential effect.on the detection sensitivity. 7.5.5.3 Steam Generator Leak Detection System (as low as ASjeamGeneratorLeakDetectionSystemisprovidedtodetectsmall 10- Ib/sec) water-io-sodium and steam-to-sodium leaks in the steam generator modules, to identify the module in which the leak has occured, and to alert the control room operator enabiIng him to take manual corrective action to prevent the leak rate from increasing. Leak detection instrumentation is provided for:

1. Sodium exiting from the superheater and the evaporators.

l

2. Sodium filled vent lines from the evaporator vents and the superheater vent.

7.5-25 O Amend. 64 Jan. 1982

f-s, features, or their power sources, concurrent with tne f ailures that ['- are a condition of, or a result of a specific accident, will prevent the operator from being presented the required information, o The Principal instruments from sensor to indicator, and the Redundant Backup instruments from sensor through the indication device will be qualif ied in accordance with PSAR Section 1.6 Ref erence 13,

                " Requirements for Environmental Qualification of Class IE Equipment."

They are quallfled to provide the Information needed by the operator to assess plant and environs conditions during and following design basis events. o instrumentation will continue to read within the required accuracy following, but not necessarily during, a Safe Shutdown Earthquake (SSE). o The Principal Instrument (from sensor to indicator) and Redundant Backup instrument (from sensor through the isolation devices) will be energized from Class IE power and be supplied with battery backing where momentary interruption of the indication is not tolerable. l7.5.11.2.2 Categorv 2 o Each Category 2 Instrument signal, will be, as a minimum, processed for display on demand. js o The Category 2 Instrument indicators will be located to ef fectively () support normal and emergency plant operations. , o The Category 2 instruments from sensor to indicator will as a minimum be qualified in accordance with Reference 13, PSAR Section 1.6,

      -          " Requirements for Environmental Qualification of Class IE Equipment" except for seismic. They will be quallfled to provide the information needed by the operator to assess plant and environs conditions during and following design basis events.

o The instrumentation will be energized from a highly reliable power source (not necessarily a Class IE power supply). Where interruption of the power supply is acceptable station AC power may be used. Where momentary interruption is not tolerable, the non-1E UPS is used. i O\

  \_/

7.5-33f Amend. 72 Oct. 1982

7.5.11.2.3 Cateaorv 3 o Each Category 3 instrument signal, wili be, as a minimum, processed f or dI spl ay on demand. o The location of the Category 3 Instrument indication will be chosen to support normal and of f-normal operati ons. o The Category 3 Instrumentation will be a high quality commercial grade. 7.5.11.2.4 General Reauirements to Cateaorv 1. 2. and 3 o Servicing, testing, and calibration programs wil l be specified to maintain the capability of the monitori ng Instrumentation. For those instruments where the required interval between testing shalI be less than the normal time interval between generating station shutdowns, a capability for ter, ting during power operation shal I be provided. o Whenever means f or removing channels f rcm service are included in the plant design, the plant design wil l f acilitate administrative control of the access to such removal means. o The piant design wiii facilitate administrative control of the access to alI setpoint adjustments, module calibration adjustments, and test points. o The montToring instrumentation design wiIi minimize the development of conditions that would cause meters, annunci ators, recorders, al arms, etc., to give anomalous indications potentially conf using to the operator. o The instrumentation will be designed to f acilitate the recogni tion, location, replacement, repai r, or adjustment of mal functioning components or modules. o To the extent practicable, monitoring instrumenta-tion inputs wilI be f rom sensors that directly measure the desired variables. An Indirect measuranent wilI be made only when it can be shown by analysis to provide unambiguous information. 7.5-33g Amend. 71 O Sept. 1982

I References to Section 7.5

1. Ford, J. A., "A Recent Eval uation of Fcreign and Domestic Wastage Data f rom Sodium Water Reaction investigation", APDA CTS-73-05, January,1973.

2. Morejon, J. A., " Sodium-to-Gas Leak Detection Mockup Tests",

W O7-TR-520-004, September 17, 1975. (Atomics International)

3. Greene, D. A., J. A. Gudahl and J. C. Hunsicker, " Experimental investigation of Steam Generator Materials by Sodium-Water Reactions, Vol ume 1, GEAP-14094, January 1976.

4. Gudahl, J. A. and P. M. Magee, "Microleak Wastage Test Resul ts",

GEFR-00352, March 1978.

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

i O l l

O 7.5-33J Amend. 72 Oct. 1982

Figure 7.5-7 Liquid Metal /Ges Leak Detection System Response Time vs. Sodium Aerosol Concentration (inerted CelIs) 1 l l l I l l 1 PLUselefG FILTER AEROSOLDETECTORB 4  : s  : i

                       .h'%
                       ...g                                                                                       PL Uuf = OOOO CC/Mlld h                                                                                       AP          = 21s.H2O
                        .. , . . t.                                                                                                                        ,

4. y teO =  : e ,,, , n; g g . , s .* .* l' N . .,*.,a... w sN .,3.>......'..

                                               , . . . M ; ..
 ~

g .

                                            .. . .

• li r '. . , e s

                                                    . ,l.  .. ..

.. .- . 3. . .

                                                          .a.     .
                                                                     .,...r.

I

                                                                            . .5 g                                                                -

t le -

                                                                                            .,   n ,-
                                                                                               . .. . E y ls..\ . .
                                                                                                        . a t.e.e
                                                                                                                . *,;s g e ;s,!.8.;

p.- ..

                                                                                                          .'- , ; 1,h g

I f f f tell te4 te4 tr8 to4 1o4 800fUM AER000L CONCENTRATION lyAnd l l 7.5-54 Amend. 72 Oct. 1982 1

Figure 7.5-8 Liquid Metal / Gas Leak Detection System Predicted Aerosol Concentration for 100 g/HR , Leak in N2(1% 02) Atm sphere j l 10*7 , , PREDICTED AERCOOL CONCENTRATION FOR 100 g/hr LEAK IN N2(1% O 2) ATMOSPHERE TEST CELL - RANGE OF

      ~

r CRBRP CELI.- l 8IZES l s - 3 ir _ 2 n 8

                                                     /(

h l O a 4 - g 10'8 - 2 o S I ' 10 10 3 108 %d 10 3 CELL OROUP VOLUME (ft ) 9 7.5-55 Arnend. 72 Oct. 1982

 ,/ ^    7.6 OTHER INSTRUMENTATION AND CONTROL SYSTEMS REOUIRED FOR SAFETY h       The additional instrumentation and control systems required for safety which have not been discussed earller in Chapter 7 are identified as the Emergency Pi ant Serv ice Water, Emergency Chil Ied Water, Recircul ating Gas Cool ing, Heating, Ventil ating, Air Conditioning, instrumentation and Control Systems, and the Direct Heat Removal Service Instrumentation and Control. The l Radiation Monitoring System also contains saf ety related components which are l discussed in Chapter 11. The Emergency Plant Service Water, Emergency Chilled Water, Recirculating Gas Cooling, Heating Ventilating, Air Conditioning Systems, Fuel Handl ing, and DHRS Instrumentation and Control are discussed in this Section. Review of the f unctional control diagrams will require reference to the symbols, notes and abbreviations as shown in Table 7.6-1.

7.6.1 Emergency Plant Service Water and Emergency Chilled Water Instrumentation and Centrol Svstem 7.6.1.1. Emergencv Plant Service Water Svstem (EPSW) The EPSW System consists of two redundant divisions which supply cooling water to the diesel generators, the Emergency Chilled Water System and seismically qual if led Non-Sodium Fire Protection System. The Instrumentation and C6ntrol System is provided for automatic control of the Emergency Plant Service Water System, to monitor and indicate system process parameters during normal and of f-normal conditions, and to provide signal inputs to Pl ant Data Handl ing and Displ ay System. Functional Control Diagrams f or Emergency Pl ant Service Water System are identified in Figures 7.6-1, 7.6-2, 7.6-3 and 7.6-4. 7.6.1.2 Design Criteria Design criteria that are applicable to Emergency Plant Service Water Instrumentation and Control System are as f ollows: A. EPSW System is provided with Class IE power supply, and is backed up by Diesel Generators to provide power during of f-normal conditions. B. No singl e f ail ure of an instrument, interconnecting cable or panel will prevent a key process variable f rom being controlled or monitored in both redundant divisions. C. Physics! and electrical separation of redundant portions of Emergency Fi ant Serv ice Water is provided. D. System level manual initiation capabilItles are provided in both divisions to perf orm all the actions perf ormed by automatic initiation. E. Instrumentation used in the control of Emergency Pl ant Service water w il l function during and af ter an SSE. 7.6-1 Amend. 72 Oct. 1982

F. Instrumentation used in the control of Emergency Plant Service Water will function during normal environmental conditions and during environmental conditions created by any design basis accident. G. Capabilities for periodic testing and calibration of all instruments are provided. H. Capabilities are provided for rmote shutdown, should the control room become uninhabitable.

l. Capabilities are provided to monitor the inoperable status of components in accordance with Reg. Guide 1.47.

J. Capabilities are provided to monitor the process variables to assess plant and environs conditions during and following an accident. 7.6.1.3 Design Instrumentation and controls are provided for the following equipment in the EPSW System: EPSW Pumps; EPSW Makeup Water Pumps; Emergency Cooling Tower Fans and Temperature Control Valves. For a complete description of the EPSW System refer to Chapter 9.9.2. 7.6.1.3.1 Control Svstem A. Remote, auto and manual controls are provided in Control Room for EPSW Pumps, EPSW Makeup Water Pumps; Emergency Cooling Tower Fans. B. Local, auto and manual controls are provided in Local Panels for EPSW Pumps; EPSW Makeup Water Pumps; Emergency Cooling Tower Fans; and Temperature Control Valves. C. EPSW wiIl start autcmatically under the following conditions:

1) On an Emergency Chilled Water System start demand;

11) 20 seconds after the Diesel Generator Load Sequencer is actuated; ill) when the system level manual control is initiated from Control Room.

7.6.1.3.2 Monitoring Instrumentation The following process variables are monitored through indication and alarms: A. EPSW Pump Discharge Tmperature B. EPSW Pump Discharge Pressure C. EPSW Pump Pit Level D. EPSW Makeup Pump Flow G 7.6-2 Amend. 71 Sept. 1982

~ E. Operating Basin Overflow p F. EPSW Makeup Pump Discharge Pressure G. Emergency Cool Ing Tower Basin Level H. EPSW Flow to Emergency Chillers

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

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

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

A. EPSW Discharge Tunperature B. Emergency Cooling Tower Basin Level C. EPSW Temperature at the Discharge of Emergency Chiller D. EPSW Temperature at the Discharge of Diesel Generator Heat Exchanger E. EPSW Flow to Emergency Chiller inoperable status of EPSW Pumps; Makeup Pumps; and Cooling Tower Fans is al so monitored through inoperable Status Monitoring System. 7.6.1.1.3.4 Deslan Analvsis 1 EPSW System is designed to operate autanatically. The system is operated only during emergency conditions. EPSW System components are cascaded to operate in sequence. Starting of EPSW Pumps wIlI operate EPSW Makeup Pumps and Cool ing Tower Fans. System will not operate when the EPSW Pump pit level is low or when electrical f ault exists. The design of the EPSW System is in conformance with the following IEEE standards l isted in Tabl e 7.6-2. O 7.6-2a Amend. 72 Oct. 1982

7.6.2 Emergencv Chilled Water (ECW) System The ECW System consists of two redundant divisions which supply chilled water. Controls are provided for the following equipment in the ECW System: Emergency Chillers, Circulating Pumps, Expansion Tank Valve, Normal-to-Emergency isolation Valves, Temperature Control Valves, and "Rocirculating Gas Cooling System Heat Exchanger and Secondary Coolant Heat Exchanger" Leak isol ation Valves. Detailed description of these controls is given in the following paragraphs. For a complete description of the ECW System, refer to Chapter 9.7.2. The ECW System cannot operate without support from the Plant Electrical Power System and the Emergency Plant Service Water (EPSW) System. The ECW System power supply is Class 1E and requires a diesel generator back-up. A detailed description of the diesel generators and the Plant Electrical Power System is given in Chapter 8. The EPSW System supplles service water to the ECW Chiller. A detailed description of the EPSW System is given in Chapter 9.9.2. Functional Control Diagrans for Emergency Chilled Water System are identified in Figures 7.6-5, 7.6-6, 7.6-7, 7.6-8, 7.6-9 and 7.6-10. 7.6.2.1 Design Criteria Design criteria that are applicable to Emergency Chilled Water instrumentation and Control System are as foilows: A. ECW System is provided with Class 1E power supply, and is backed up by diesel generators to provide power during off-normal conditions. B. No single failure of an instrument, interconnecting cable or panel will prevent a key process variable from being controlled or monitored in both redundant divisions. C. Physical and electrical separation of redundant portions of Emergency Chilled Water is provided. D. System level manual initiation capabilities are provided in both divisions to perform all the actions perf ormed by automatic initiation. E. Instrumentation used in the control of Emergency Chilled Water will function during and after an SSE. F. Instrumentation used in the control of Emergency Chilled Water will f unction during normal environmental conditions and during environmental conditions created by any design basis accident. G. Capabilities for periodic testing and calibration of all Instruments are provided. H. Capabilities are provided for remote shutdown, should the control roan become uninhabitable. O 7.6-2b Amend. 71 Sept. 1982

I 7.6.2.2.3 Inouts to PDH&DS The following process variables are provided as inputs to Plant Data Handling O- & Display System (Non-Safety System): A. Em Temperature at the inlet of Emergency Chiller B. ECW Temperature at the Discharge of Emergency Chiller C. EW Flow from Emergency Chiller D. ECW Chiller Trip Status i E. ECW Containment isolation valves Status F. Secondary Cool ant Expansion Tank DT-J Leakage 7.6.2.2.4 Design Analvsis ECW System is designed to operate automatically. The system is operated only during emergency condition. ECW System components are cascaded to operate in seq uence. Low flow of NOW to ECW loop signal will align ECW isolation valves and operate ECW Pumps, Emergency Pl ant Service Water Loops, and ECW Chil lers. System will not operate when the EPSW flow through chiller is not established or when electrical fault exists. The design of the ECW System is in conformance with the IEEE Standards listed in Tabl e 7.6-2. O l l I l 7.6-2e Amend. 72 Oct. 1982

l

3. Unit cooler or WAC unit supply air temperature hIgh or alr O temperature entering cooling co!l Iow.

Smoke, anmonia, chlorine, fluorine or radiation present in Control 4. Room main or remote air intake.

5. Control switch in the local mode (Control Room only alarm only).

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

1. Control Room and computer room humidity.

2. Containment dif ferentIal pressure.

3. Annul us dif ferential pressure.

4. RSB confinement dif ferential pressure (four dif ferent cells).

5. Control Room dif ferential pressure.

6. Air temperature entering and IeavIng each filter unit.

7. air temperature entering and ieavIng each WAC unit.

8. Cell temperature of each area being serviced by a unit cooler or HV AC un it.

9. Inoperable or bypass status of components.

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

1. Annul us to atmosphere dif ferential pressure.

2. RSB conf inement to atmosphere dif ferential pressure.

3. WAC units discharge air temperature.

4. Filter units adsorbent filter leaving air temperature.

5. WAC and filter units air flow low.

6. Damper and valve position Indication.  !

7. Fan operation status Indication.

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

Non-Cl ass IE circuits. The design of the HVAC Instrumentation and Control sy stem is in conf ormance w ith the IEEE Standards and l isted in Table 7.6-2. Ref er to PSAR Section 7.1.2 for conf ccmance to appl Icable IEEE Standards. 7.6.5 Steam Generator Building (SGB) Floodina Protection Subsystem 7.6.5.1 Design Basis The SGB Flooding Protection Subsystem is provided to prevent flooding of SGAHRS equipment resulting f rom postul ated SGS water / steam line ruptures, thereby assuring the avail abil Ity of SGAHRS for reactor decay heat removal follow ing water / steam l ine rupture events. The SGB Flooding Protection Subsystem is designed to the IEEE Standards iIsted in Tabl o 7.6-3. 7.6.5.2 Design Reautrements The SGB Flooding Protection Subsystem is designed to perform the following functions: a) Detect the presence of large steam / water piping ruptures (see Section 15.3.3.1) by temperature and moisture sensors In each cel1. b) Detect water level flooding conditions in each celi by water level sensing el ements. c) Provide the signals to initiate the alarms and activate the equipment which provides the SGB flooding protection. 7.6.5.3 Deslan Descriotion 7.6.5.3.1 Instrumentation Instrumentation provided fcr this subsystem consists of Class IE temperature, and moisture transducers. In addition, non-Class IE level transducers are provided. The transducers and associated control logic are located in the SGB cells containing main f eedwater or recirculation piping. Three independent moisture and temperature measurements in each cel I are util ized for identifying a majcr water / steam line rupture. Water level measurements in each cell confirm a flooding condition and are annunciatad in the main control room. 7.6.5.3.2 ControIs Each heat removal loop isolates the main f eedwater supply upon detection of a major pipe rupture. The start-up and main feedwater control va!ves close upon activation by a two-out-of-three logic using measurements of moisture rnd temperature in each cell. The main f eedwater isol ation val ve is Independently closed upon activation by a two-out-of-three logic using the same three moisture and temperature measurements f rom each col 1. Separation and isolation is maintained between the control valve and isolation valve activation logic. 7.6-9 Amend. 72 Oct. 1982

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

;                       l evel . Manual corrective control of flooding is initiated by the operator upon annunciation in the main control room.

4 i 1 i O O 7.6-9a Amend. 72 Oct. 1982 <

Inoperable status of subsystem f ans (MA, M3, EA, EB) and Isolation val ves (two per subsystem) is al so monitored through inoperable Status Monitoring System. (} 7.6.6.2.4 Design Analys ts Ref er to PS AR Section 7.1.2 for conf ormance to appl icable IEEE Standards. RGC system is designed to operate automatically. The system and its safety-related subsystems are operated during normal as well as emergency conditions. The RGC system components are cascaded to operate in sequence. Starting a fan will open associated supply and return gas isolation valves. A subsystem will not cperate when high water vapor or cooler high water level or electrical f aul t exists. As discussed in Section 9.16 each subsystem of RGCs supplies cooling to redundant components, so no additional redundancy is provided in its components and Instrumentation. The systems are designed f or f all safe operation and control equipment wIll assume a f ailed position consistent with its Intended safety function. The coolant supply ^o saf ety-related subsystems MA, M3, EA, EB is provided by Emergency Chilled Water System. The f an motors f or these subsystems are provided with AC power from Class 1E power sources to continue operating during loss of of f site power, except f or the booster f an of the subsystem EB which is not required to operate during loss of power condition. Subsy stems MA and EA and the EM pumps cooled by these two subsystems are served by Class

   . 1E power supply Division 1. Al so, subsystems MA and EA are served by Emergency Chilled Water Loop "A". Subsystems M3 and EB are served by O'    Emergency Chilled Water Loop '13", and Class 1E power supply Division 2. The EM pumps cooled by subsystems NB and EB are also connected to Class IE power supply Div ision 2. Automatic isolation valves are designed as f all open valves so as to be in their safety position upon loss of power.

Fan and Isolation Valve control switches are located in the local panels as wel l as in the back panel s f or subsystems MA, MB, EA and EB, except f or booster f an. Thus, in case of control room evacuation the f ans and valves can be control led f rom outside the control rooms, using l ocal panel s. The design of the Recirculating Gas System is in conformance with the IEEE Standards I isted in Tabl e 7.6-2. O l v} 7.6-16 Amend. 72 Oct. 1982

TABLE 7.6-1 SYMBOLS b ALARM INOPERABLE STATUS MONITORING

                                   %     RED IND LITE                          .
                                   %     GREEN IND LITE M     WHITE IND LITE h     COMPUTER INPUT NOTES:

1) Control switches are spring return to auto from start with a maintained stop unless otherwise stated.

ABBREVlATIONS SSPL S - Solid State Programmable Logic System CR - Control room (remote) L - Local (not control room) T. D. - Time delay N. C. - Normally closed l F. C. - Fall closed r i S. O. V. - Solenoid operated valve l A.O.V. - Alt operated valve MOD - Mover operated damper ZS ?c,sition switch CIS - Containment isolation signal l PPS - Plant Protection System E/H - Electro-hydraul c O 7.6-17 Amend. 71 Sept. 1982

TABLE 7.6-1 (Continued) TE - Temp eiement TT - Temp transmitter tlc - Temp Ind controlIer OAl - Outside air intake TMD - Temp modulated damper RA - Return air PDI - Pressure dif ferential Indicator PDC - Pressure dif ferential controller PMD - Pressure modulated damper PDISH - Pressure dif ferential indicating switch high FR - Flow recorder FIC - Flow Indicating controller FSL - Flow switch low FT - Flow transmitter FMD - Flow modulated damper FE - Flow element M - Moisture PB - Pushbutton MUX - Multiplexing l AHU -- Air handling unit O 7.6-18 Amend. 71 Sept. 1982

TM LE 7.6-2 LIST OF lEEE STANDARDS APPLICABLE TO EERGENCY PLANT SERV ICE WATER, EMERGENCY QilLLED WATER, HV AC, AND RECIRCULATING GAS INSTRUENTATION AND CONTROL SYSTEM c) lEEE Standard 279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations b) IEEE Standard 308-1974 Criteria f or Cl ass 1E Power Systems f or Nucl ear Power Generating Stations c) IEEE Standard 323-1974 Qual ifying Class IE Electrical Equipment f or Nuclear Power Generating 5tations d) IEEE Standard 338-1977 Criteria for Periodic Testing of Nuclear Power Generating Station Saf ety Sy stems o) IEEE Standard 379-1972 lEEE Trial-Use Guide for the Appl Icabil Ity of the Single-Fail ure Criterion to Nuclear Power Generating Station Protection Systems f) lEEE Standard 383-1974 Standard f or Type Test of Cl ass IE Electric Cables, Fiel d Spl ices and Connections for Nuclear Power Generating Stations g) lEEE Standard 384-1974 IEEE Trial-Use Standard Criteria for Separation of Class 1E Equipment and Circuits. O 7.6-18a Amend. 72 Oct. 1982

TABLE 7.6-3 LIST OF lEEE STANDARDS APPLICABLE TO SGB FLOODING PROTECTION SUBSYSTEM IEEE-279-1971 IEEE Standard: Criteria for Protection Systems for Nuclear Power Generating Stations lEEE-323-1974 IEEE Trial-Use Standerd: General Guide for Qualifying Cless IE Electric Equipment for Nuclear Power Generating Stations I EEE-323-A-1975 Supplement to the Foreword of IEEE-323-1974 IEEE-336-1971 lEEE Standard: Installation, inspection, and Testing Requirements for Instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations IEEE-338-1971 lEEE Trial-Use Criteria for the Periodic Testing of Nuclear Power Generating Station Protection Systems IEEE-344-1975 IEEE Standard 344-1975, IEEE Recommended Practices f or Seismic Qualification of Class 1 Equipment for Nuclear Power Generating Stations l EEE-352-1972 lEEE Trial-Use Guide: General Principies for Rel labil ity b Analysis of Nuclear Power Generating Station Protection Sy stems IEEE-379-1972 lEEE Trial-Use Guide for the Appi ication of the Single-Failure Criterio to Nuclear Power Generating Station Protection Systems IEEE-384-1974 IEEE Trial-Use Standard Criteria for Separation of Class IE Equipment and Circuits IEEE-494-1974 lEEE Standard Method for identification of Documents Related to Class IE Equipment and Systems for Nuclear Power Generating Station O 7.6-18b Amend. 72 Oct. 1982

CONO! TION CONTROL ACTION RESULTANY MON]YOR EPSM PUMP a STARTED j kPN0 >-- Cp t F . CONTROL SW \ ^ FAN "/ [ START PNO > AUTO

                                 \                 *

((AFTERSTOP3 [

                                  \                                m

( STOP f d OR - y AND -> FAN START

                                                                                                -4 >-     R L

LOCAL SIGNAL

          '" "                                                                    DE-ENERCIZE SSeLS    J                                                           +       o" S" ace l                                      "o' SEA 1ER D-- AND     >
                                                   >                                                         CR REMOTE SIGNAL                                 ,                    >                                    CR N         SSPLS    j                         ll          ANO     b-      m

?

                                                   ?                    -

AND CR l U-k AND > (CONTROLSW S T

                                  /

AUTO

                                  \

(AFTER STOP) [ > m FAN L AND >d STOP ( STOP [ O ENERCIZE

                                       -- >  NOT
                                                                              +   MOTOR SPACE HEATER MOTOR m

THERMAL - OVERLOAD j TF + g -

                                                                                                      ' iS

?] -+ v FIGURE 7.6-1 EMERGENCY COOLING TOWER FAN O O O

r. %./ 'v' '%) CONDITION CONTROL ACTION RESULTANT MONITOR CR LOCALSIGNAL) j( FROM 3 SsPLs t m 7 h AND + [ l g R CR ( m OR (AFTER PEN [ AND + 4 ENERG!ZE AIR OPEN H CLOSE AOv (FO)

                                                                 +  ENERGIZE  AR     CLOSE H
                       +   NOT
                                                                                             ~B' l

m EXP TK LEVEL LOW OR b m OTJLEAKAGEj D v ENE GIZE AIR OPEN H T H H O AND + SOY AOV (FO) SW k l 7 ENERGIZE R CLOSE f (CONTR I =

                                                         +                                          e k                h  AND

( '"

                        "AUTOo'e'EN
                                 /
                                 \

REMOTE s!GNAL ( CLOSE D Ssets )

 %'t As 8%                                                 FIGURE 7.6-8 N

EMERGENCY CHILLED WATER ISOLATION VALVES TO SECONDARY COOLANT LOOP

CONDITION CONT ROL ACT!ON RESULTANT MONt'OR CR

                                                                                                                                                               >/A 1
                                                                                                                                 =

LOCA SCNAL\ g ( 0  ; - j -- SSPLS t ME RGf NCY

                                                     /       CONTROL SN   k(                     AND ->      ;                                  OPE RATON

[ -> ENERGIZE AIR OPEN -O-AUTO AFTER NORMAL /EMER [ OR 4@ ENERCIZE ADMIT MIR CLOSE -* O CR EMER ALIGN NOHMAL REMOTE SIGNAL [ OPER A TION

                                                                                                                                                             ~

0 L 1 , SSPLS j lfl AND A '

• cp C
• m a w CR N[W TO ECW LOOP A , - 7.6-10 g f

m ( FLON LON OR

                                                                                              >  AND ->                                   E ME RGENCY 4

" (CR INITIATION I

                                                                                              >                                           OPE R ATION (sg;astig\'"    [                   >
                                                                                                          @         > ENa%E SOY Y1R CLOSE
                                                                                                                                                          +@'
                                                                          \            $4 AND A N0RmA6 ems [
                                                                          \

• ENERGrzE AT DEEN NORMAL OPERAT Q

                                                                                                                                                          *d"
                                                                                              >                                                                      L (EMERALIGN      [                      RND +

llI1 hA~+

                                                                                                             -       ~,

l> AND R

                                                                                  ->        30kEC Rir Fa P

e gg FIGURE 7.6-9 EMERGENCY CHILLED WATER SYSTEM NCW TO ECW ISOLATION VALVES S O e

[^g h ]j ,f [N t I t

                                                                                                                     \ ,,/
                                                                 \j                                                    s

\d CONDITION CONTROL ACTION RESULTANT MONITOR CR LOCA GNAL SSPLS j t , 7 OPEN k > AND + (CONTROLSH / = R CR

                                        \                 I m           'OR (AFT /    PEN                        AND  +       4 ENE GIZE  AIR    OPEN  H Oy CLOSE                                                          AOV   (FO)
                                                                              +   ENERGIZE  AR     CLOSE H L

o I

                            +     NOT

? O l 7 f EXP TK > AND + R $ D A E OR DTJ VAPOR DE- VENT L y ENERGIZE AIR OPEN -4 R CON r E NT HIGH m AND + SOY A0V (FO)

                          !  "       SN \         l                             E ENERGIZE  AR     CLOSE H
                          \             /

AUTO k AND + l ( OPEN [

                                        \

REMOTE SIGNAL FROM ( CLOSE

                                        /

f f

                                                            =

2 AND + SSPLS ] Et Ra P 8% FIGURE 7.6-8 ho EMERGENCY CHILLED WATER ISOLATION VALVES TO SECONDARY COOLANT LOOP

COPO! TION CONTROL ACTION RESULTANT M_0N 1 ' OR CR Du, A T - LOCAL SIGNAL FROM v 1-3 kr i l-SSPLS L ~ &, t ME RM NC v COM ROL SN #NO 04 "AI D k '

                                                             /                                        ->     ENERb2E   AIR         OPEN AUTO AFTER n->         ENERGIZE             CLOSE     HF EMER ALIGN                                                                          N REMOTE S!GNAL
                                                                        ,                               FIG. ,,

SSPLS j lfl ANO -h ' I'0 CR G oSM N b TO ECH LOOP A $- - l 7.6-10 g g f { FLON LON OR '

                                                                               >   ANO ->                                   E VE RGE.NC Y N   (CRINIT!ATION                                                             >                                            OPE R ATION TR hk ENERb2E        AIR       CLOSC     H
                        \                                    [                 >                                SOY
                                                             \          $--> ANO -->

( NORMAL [ EMER 4 ENERGlZE R OPEN H& (EMERALIGN / + " ll> l AN0 OR 0

                                                                               >                 -        ~,

ANO + + R l, __

                                                                     ~>      30kEC

RF r2 P

e gg FIGURE 7.6-9 EMERGENCY CHILLED WATER SYSTEM NCW TO ECW ISOLATION VALVES 9 9 e

I l p) ( o Steam Flow - Steam flow is sensed at a flow element in the outlet line from the superheater by a differential pressure transmitter. The dif ferential pressure signal is compensated for temperature and pressure variations and linearized to provide a mass flow signal. o Feedwater Flow - Feedwater flow is. sensed at a flow element in the inlet iIne to the steam drum by a dif ferential pressure transmitter. The differential pressure signal is corrected for temperature variations and linearized to provide a mass flow signal. 7.7.1.5.2 Main Feedwater Isolation isolation of the main feedwater supply is provided to mitigate the consequence of the loss of feedwater to a steam drum, a steam line break, or to prevent superheater flooding. Isolation of tho feedwater supply to the affected loop in the event of a steam generator system feedwater leak will ensure integrity of the feedwater supply to the two unaf fected loops and mitigates the consequence of flooding damage to other equipment. This protection is provided by automatic closure of the steam drum isolation valve and both feedwater control valves upon sensir.g a low steam drum pressure (500 psig) signal and automatic closure of both feedwater control valves and feedwater valve isolation upon sensing a steam generator building flooding (temperature and humidity) signal. In the event of a steam iIne break, steam drum dryout may occur and would (] result in damage to the steam generator loop upon re-introduction of C/ feedwater. Protection against the re-introduction of feedwater is provided by the closure of the feedwater Isolation, the steam drum isolation, and control valves on low steam drum pressure (500 psig) signal. In the event of a f ailure in the drum water level control components, an overfilling condition might resukt in flooding of the steam drum and superheeter. Protection against this is provided by three redundant water level sensors and by trip functions which close the feedwater valves at two steam drum levels. The first trip level, 8 inches above normal water level, closes the feedwater steam drum isolation valve, and the feedwater control valves. The second trip level,12 Inches above normal water level, closes the feedwater isolation valve. Protection against flooding of the superheater during steam generator auzillary heat removal is discussed in Section 5.6.1. 7.7.1.5.3 Ooerational Considerations Normal Ooerations The steam drum level controller utilized for feedwater control valve operation is located in the control room back panels. The operator control station for the controller is located on the main control panel in the control room.

  )

v 7.7-9 Amend. 71 Sept. 1982

o Control Building Fire Detection b V o Emergency Diesel Generators o Switchyard and Station Electrical Distribution o Direct Heat Removal Service 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 f rom predetermined conditions are alarmed and/or indicated so that corrective action may be taken by the operatcr. The control room also includes control and instrumentation equipment that is used Infrequently or for which controlled access is desirable. Included in this control room back panel area are power distribution, chit led water, containment instrumentation, recirculating gas, heat transport, steam generator, heating ventilation and air conditioning, annunciater electronics, turblne, balance of plant, plant control, plant data handling and display system multiplexers, flux monitoring, radiation monitoring, reactcr 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 (plan view) and 7.9-2 (side view). There are 3 significant p features 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. O 7.9-5 Amend. 67 March 1982

Since the Main Control Panel incl udes safety related equipment, the sections incl uding this equipment are designed to Seismic Category I and qualified in accordance w ith IEEE Std. 323 and IEEE Std. 344 Structures, wiring, wireways, and connectors are designed and installed to ensure that safety related equipment on the control panel remains operational during and af ter the SSE. The Main Control Panel is constructed of heavy gauge steel within appropriate supports to provide the requisite stif fness. Within the boundaries of the Main Control Panel Sections, modules are arranged according to control functions. Fire retardant wire is used. Modul ar train airing is f ormed into wire bundles and carried to metal wire ways (gutters). Gutters are run into metal vertical wireways (risers). The risers are the interf ace between external wire trays feeding the panel and Main Control Panel airing. Risers are arranged to maintain the separated routing of the external a ire trays. (See Figures 7.9-3 and 7.9-4). Mutually redundant safety train wiring is routed so as to maintain separation in accordance w ith the criteria of IEEE Std. 384. A minimum of six inches air separation is maintained between wires associated with dif ferent trains. Where such air separation is not available, mechanical barriers are provided in Ileu of air space. The Main Control Panel protection system circuits are designed and selected to ensure that system performance requirements are met and channel Integrity and independence are maintained as required by IEEE Std. 279. Power division separation and isolation are maintained in accordance with the requirements of IEEE Std. 308. 7.9.3 Local Control Stations Local control panel s are provided for systems and components which do not require f ulI time operator attendance and are not used on a continuous basis, in these cases, however, appropriate alarms are activated in the Control Room to alert the operator of an equipment mal function or approach to an of f-normal conditlon. 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 interpiant communications and the private autm..stic exchange (used for in-plant and external communications) a sound powered maintenance communication Jacking sy stem is provided. Redundant and separate methods of communication between the control room and other TV A generating pl ants is al so provided. 7.9.5 Design Evaluation FoiIowing the Three MIIe Isiand accident, a i arge task f orce was f ormed f or the purpose of performing a thorough review of the CRBRP Control Room design. This overall review was divided into three parts; a planning phase, a review phase, and assessment and implementation phase. Follow ing the task f orce effort, NUREG-0700 was issued. NUREG-0700 is similar in intent to the CRBRP Control Room design eval uation. 7.9-6 Amend. 72 Oct. 1982

l 7.9.5.1 PiannI'no Phase O Q in the planning phase the objectives and scope of the task force were identif led, and criteria were establ ished f or personnel selection. A charter was developed which contained the scope and objectives, and personnel selection was accompl Ished. The task force charter required a review of the Control Room design and the operating procedure outl ines 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. Specif ic items incl uded in the review are:

1. OveralI Control Room and Individual panel designs and features, and their interf ace wIth the operator.

2. System and overall plant operating procedure outl ines.

3. Administrative approaches for plant operations.

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

       's . Recommendations made by NRC and other parties as a result of the Three Mile Isl and occurrence.

6. Computer util Ization by the operators.

V 7. Operator training requirements.

8. Remote shutdown capabil ities and safety system status Indication in the Control Room.

Criteria were established for personnel selection of those to participate on 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 qual if led operators were considered mandatory. Personnel with human f actors education and experience both inside and outside the nuclear industry were included. Human f actors considerations were emphasized 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 Room Task Force was to re-eval uate this interf ace. Prior to the evaluation ef fort a seminar was held, under the direction of three leading human f actors personnel, to teach the Task Force discipl ined methods f or considering human f actors. Based on this training and f urther assistance f rom human engineers, check Iists were prepared to eval uate the man-machine interf ace.

  *See Ref erence 7.9-1 O

7.9-6a Amend. 72 Oct. 1982

1 7.9.5.2 Review Phase in the review phase extensive analysis of plant events were conducted. h Functional analyses were made of the operator in his response to automatic equipment actions, manual actions which had to be perf ormed in the Control Room, and manual actions required by operators external to the Control Room. More than 200 walk-throughs of plant events were conducted. The Control Room design and operating instructions were thoroughly reviewed in four areas:

1. Proper identif ication of systems to be operated f rom the Main Control Room.

2. Proper staf f ing of the Control Room.

3. Proper overal l l ayout of the Control Room to enhance the man-machine interf aces and support the integrated operation of plant's systems.

4. Proper layout and design of Indiv idual Control Room panel s, i n str uments, indicators, and control s to enhance the man-machine interf ace and support the integrated operations of the plant's sy stens.

A f ull scale mockup of the Control Room was used. The events chosen to be eval uated were caref ully selected so they woul d umbrel la al l of the operations that are either expected to occur or might be postulated to occur over the l if e of CRBRP. The of f-normal events incl ude pl ant responses to single and mul tiple f ail ures. The methodology of perf orming this review consisted of using three groups of people; simul ators, operators, and eval uators. The Simulators analyzed the events which were to be evaluated prior to the wal k-throughs and then, during the walk-through eval uations, simul ated the control panel indicators. 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 woul d appear to the operator. The control panels were mocked up by the Simulators to represent the changing plant conditions and the inf ormat ion fl ow into the Control Room during the event. This made the wal k-through as real istic as possibl e. The Operators played the part of the Control Room operators and carried out the steps of the procedure being evaluated. They touched each switch they were required to operate, and observed each indicator which was part of the particul ar event. The Eval uators included a Human Factors Engineer and a Systems Engineer. Their f unction was to fill out the Operating Sequence Diagram and the eval uation sheets f or each procedure and event reviewed. O 7.9-6b Amend. 72 Oct. 1982

p As problans or concerns were encountered, recommendations were made. These V were, in some cases, of a broad nature and reflected the need for reconsideration of decisions made in the four most important evaluation areas described above. Other problems and concerns related to specific details of the Control Room design or the procedure outl ines. 7.9.5.3 Assessment and imolementation Phase The eval uation and implementation of the recommendations started with a check of the consistency of alI of the recommendations by the task force. SmalI model s of the overall Control Room and Main Control Panel were made assuming all recommendations were incorporated into the design. The recommendations 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 iIne organization for assessment and impl ementation. 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 and provide adequate justification if the recommendation is i nval i d. Each assessment is reviewed and approved by senior project management. 7.9.5.4 Conc lusions The Control Room Task Force Design Review is documented in f urther detail in Reference 7.9-1. In September 1981, NUREG-0700 entitled "Guidel ines for p, Control Room Design Review" was issued. A comparison between ihese two (l documents leads to the concl usion 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 Revlew, and the intent of NUREG-0700 is bel leved met by CRBRP. O 7.9-6c Amend. 72 Oct. 1982

Reference:

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

O l l 7.9-6d Amend. 67 March 1982

N11 in. diameter opening. The increased thickness of the EVTM floor valve in radial and axial direction provide the additional shielding required for the much higher radiation source which passes through an EVTM floor valve (spent O 20 fuel assembly) compared to an AHM floor valve (IVTM port plug). The stepped upper and lower steel plates of the floor valves, con- ' 44 centric to the valve port, (see Figure 9.1-18) prevent diffusion and radia-tion streaming through the minimal mating surface gaps. These design fea-tures limit the transient dose rate at the surface to less than 200 mrem /hr 44l during transfer of radioactive components, and 5 mrem /hr when closed over the reactor ports. The floor valve is sealed to the fuel transfer port adaptor by double seals, and bolted to the adaptor flange. The movable circular disk which closes off the port opening in the valves is also sealed by double 20 seals. 9.1.4.6.3 Safety Evaluation The radial shielding limits the dose rate on the floor valve surface to less than the criteria in Sections 12.1.1 and 12.1.2 during transfer of the l highest powered spent fuel assembly (for the EVTM floor valve).The floor valve 44l 49 is considered a piece of equipment whose main function is to permit transfer of radioactive components, both fueled and non-fueled, between a machine and a facility. The radiation source is transient and short tenn (less than 1 min per transfer) in nature. Hence, it results in a low integrated dose. Another function of the floor valves is to provide axial shielding O- to replace that normally provided by the port plugs. The axial shielding limits the dose rate to personnel to 5 mrem /hr when placed over a reactor 44 port and to 0.2 mrem /hr when placed over EVST or FHC ports. Personnel cannot receive a direct axial dose because of the large diameter of the floor valve. In addition, the valve is covered by a mating machine much of the time. In all cases, sufficient axial and radial shielding for the EVTM and AHM floor  ! valves is provided to limit the integrated' dose to less than 125 mrem / quarter, 44l and dose rates to the zone criteria of dection 12.1. The floor valve has adequate seals to prevent excessive radioactive release to the RCB and RSB operating floors. Accidental cover gas release through inadvertent opening of a floor valve in the absence of a mating fuel handling machine (EVTM, AHM) on top of the floor valve is prevented by inter- . locks. One interlock prevents energizing the valve operating motor unless a mating machine is on top of the floor valve. (Electrical power to the floor

59) valve motor is supplied by connection to the mating machine.) Other interlocks prevent (1) depressurizing the buffer gas zone, and (2) raising i the closure valve extender, unless both the closure valve and the floor valve i are in their closed positions.

l l Amend. 59 O 9.1-60 Dec. 1980

                                                                                                   ,_,m._,,,

As discussed in Section 15.5.2.4, an unl ikely accident releasing radioactive cover gas f rom the reactor leads to a site boundary dose well below the guidel ine yal ue of 10CFR20. 9.1.4.7 Safety Asoects of the Reactor Fuel Transfer Port Adactor and Fuel Iransoort Port Cooling Inserts The reactor f uel transf er port adaptor (see Figure 9.1-19) is positioned on top of the reactor f uel transf er port and extends f rom the reactor head to the bottom of the floor valve which is located at the elevation of the RCB operating fl oor, it serves as an extension of the reactor cover gas containment and provides. shielding when irradiated core assemblies are removed f rom the reactor. The adaptor al so gu i des cool i ng a i r f rom an ai r bl ower to a cooling insert inside and below the adaptor. The f unction of the cool ing inserts, located around the EVST and FHC f uel transf er ports as well as the reactor port (see Figure 9.1-19), is to remove decay heat should an irradiated core assembly in a sodium-filled CCP become immobilized in a f uel transfer port during transfer between the reactor vessel, EVST or FHC and the EVTM. 9.1.4.7.1 Deslan Basis The design bases f or shiciding and radioactive release of the f uel transf er port adaptor are the same as f or the EVTM (see 9.1.4.3.1). The reactor, EVST, and FHC f uel transfer port cool ing inserts have the capacity to remove decay heat from 20 KW irradiated core assembl ies in sodi um-fil led CCPs to prevent exceeding the 1500 F spent f uel cl adding temperature l imit specif ied f or unl ikely or extremely unl ikely events (Table 9.1-2). t 9.1.4.7.2 Design Descriotion The reactor f uel transf er port adaptor extends f rom the upper surf ace of the fuel transf er port in the reactor head to the operating floor, see Figure 9.1-19. The upper surf ace of the reactor f uel transfer port adaptor consists of a flange which is bolted to the floor valve and provides the sealing surf ace f or the doubl e seal s on the l ower surf ace of the fl oor val ve. Shielding is provided by a thick, annular lead cylinder surrounding the adaptor cover gas containment tube over its entire length to limit the dose l rate at the shiel d surf ace to less than the l imits given in Sections 12.1.1 and 12.1.2. The lower part of the adaptor is bolted to the reactor head during ref uel ing only. The reactor f uel transf er port cooling insert extends f rom the top flange of the adaptor to the f uel transf er port nozzle. The cool ing insert uses a col d wLi l cool ing concept, simil ar to the EVTM. The CCP containing a spent f uel assembly is cooled by thermal radiation and conduction across the argon gas gap to the cold wall which forms the confinement barrier for the reactor cover gas, knblent air is blown down the outside annulus of the cooling insert, and discharges into the reactor head access area. Ai r f l ow from the blower is adequate to l imit the cl adding temperature of a 20 KW f uel assembly to less than 1500 F. 9.1-61 O Amend. 72 Oct. 1982

9.6.5.4 Testing and insoection Recuirnmants AlI components are tested and inspected as separate components and es integrated systems. Velometer readings are taken to ensure that alI systems are balanced to deliver and exhaust the required air quantities. All water coils are hydraulically tested for leakage prior to being placed in. service. Capacity and performance of the fans are tested according to the Air Moving and Conditioning Association requirements prior to operation of the plant. .I 9.6.6 Steam Generator Building HVAC Svstem 9.6.6.1 Design Basis 9.6.6.1.1 Steam Generator and Auxillarv Bav HVAC System The Steam Generator and Auxiliary Bay HVAC System is a safety-related system designed to provide filtered and conditioned air io the Steam Generator Loop Celis, the Auxillary Bay CelIs and the Intermediate Bay IHTS CelIs to permit continuous routine personnel access during normal operation and to ensure operability of the equipment under all conditions. The HVAC System serving these areas is designed to: a) Maintain 100 F max. within all areas during normal operation. b) Maintain 120 F max. within all areas under single failure of an HVAC System component, c) Maintain 120 F max. within alI areas under loss of cooling from the Normal Chilled Water System.

       , /      d) Maintain the ventilation rate within all areas under all operating conditions.

e) Comply with the single failure criterion of Regulatory Guide 1.53. f) Operate from the Class IE AC power supply during loss of off-site l power. g) Maintain 120 F max. within the Auxiliary Feedwater Pump Cells during off-normal conditions. h) Provide ducted cool air directly to the lobe oil cooling panels.

1) Provide exhaust ductwork for the Intermediate Sodium Pump Drive to exhaust hot discharge air directly outside to atmosphere.

J) Provide ducted exhaust from the intermediate Sodium Cold Trap. k) Provide design features to mitigate the consequences of a sodium fire. 9.6.6.1.2 Steam Generator Building Intermediate Bav HVAC Svstems The Steam Generator Building intermediate Bay HVAC Systems are safety-related l systems designed to provide filtered and conditioned air throughout the O 9.6-45 Amend. 72 Oct. 1982

intermediate Bay (except lHTS cells) to permit continuous routine personnel access and to ensure operabil ity of the equipment during normal operation. The HVAC System serving these areas is designed to: a) Maintain 95 F max. within all areas during normal operatior, b) Maintain 120 f max. within all creas under single f ailure of an HVAC system c.omponent. O 1 l l l e Amend. 59 9.6-45a De c.1980

The Air Handling Unit with its two (2) 100% capacity redundant supply fans are located in Steam Generator Loop 1 Cell on a platform at E1. 852'-6". p identical arrangement of Air Handling Units and their supply fans exist in ( Loops 2 and 3 respectively. Each Air Handling Unit is connected to an independent missile protected air intake structure located on the north side of the SGB-lB roof by ductwork with redundant fire dampers. Each air handling l unit consists of a mixing plenum with an outside and return air intake damper, pre and after filter, cooling coil and access sections. Downstream of the cooling coil sectlon, a sufficiently long end access sectIon is provided for the connection of the 100% supply f ans. The length of the end access section is selected to permit equalization of the air flow through the cooling coils required by the off-center location of the fans. The length of the other access sections is determined by the maintenance requirements of the individual components. The fan sections are connected to the end access section and are followed by manual dampers (normally locked open), flexible connections, fans, flexible connections and automatic isolation dampers. The "Y" duct section connects with the supply ductwork which serves the respective coll. In cells 244, 245, and 246, a branch duct is connected to the sodium pump drive tube oil cooling panels from the main supply duct. The air is then relleved to the colI. Two (2) Unit Hester s are located in each SG Loop Cell at E1. 816'-0". One Unit Heater is located in each Steam Drum Cell at E1. 846'-0". The return air is transferred from cell to cell to one of two (2) 100% l redundant exhaust fans, located in each of the three (3) SGB Loop Cells at E1. 851'-6" and 861'-0". The discharge side of each fan is connected to a flexible conr. action and followed by ductwork and an automatic isolation N damper. The damper section is connected to the discharge ductwork which

  /

either returns the air to the cell for recirculation in the system or exhausts it to atmosphere Through a missile protected exhaust structure. The exhaust ductwork is provided with redundant fire dampers to create a controlled vent path. The exhaust stack is provided with redundant motorized dampers to allow closure after Initial venting. A tritium sampler, monitors the air discharge for release of tritium. Exhaust from the two (2) cold traps located in each Steam Generator Cell at El. 806'-0" connects to the Steam Generator Celi exheust ductwork. The IHTS pump motor draws the air required for cooling from the celi. Exhaust ducts are provided with redundant fire dampers and are connected to the air discharge flanges of the IHTS pump motor and discharge the ho1 air to the atmosphere through the steam vent structure. The Steam Generator Building Aerosol Release Mitigation System, an engineered safety feature whose operation is described in Section 6.2.7, is located in each SGB loop. The ESF consists of redundant safety-related closure dampers, rollof vent dampers, and aerosol detectors. The closure dampers are fire da.npers preceding each loop's air handling unit, following each loop's IHTS pump motor and exhaust fans, and clutch-type motorized exhaust dampers. Controlled release of aerosols from the SGB is accomplished by closure of the fire dampers and the controlled venting of the aerosols through the vent stack until terminated by the clutch-type motorized exhaust dampers. Additionally, the HVAC systems of the remaining Nuclear Island Buildings will be isolated D 9.6-47 Amend. 72 Oct. 1982 l t . ._

i i from the outside atmosphere by either closing dampers or shutting off supply and exhaust fans. These actions are initiated by redundant sets of safety-rolated aerosol smoke detectors in the SGB. The rellef vent dampers are also ftre actuated dampers. The P&lD for SGB Loop 1, 2 and 3 is Figure 9.6-12, 9.6-13, and 9.6-14, respectively.

 .Two (2) Unit Coolers provide conditioned air for the electric driven Auxiliary Feedwater Pump cells (one cooler for each cell).

The supply air is distributed to the cells by an independent ductwork system to satisfy the cooling requirements. The unit cooler filters maintain the cleanliness of the air supply (for initial start-up only). The cooling coils provided in the unit coolers, along with their instrumentation and controls, maintain the Indoor Design Conditions. 9 9 9.6-47a Amend. 72 Oc t. 1982

The Unit Coolers are located in the Auxiliary Feedwater Pump ( / Cells 204A and B at El. 733'-0". The unit coolers consist of disposable filters, cooling coils and a V-belt driven centrifugal fan. The fan discharge is connected to an independent ductwork that serves the SGAHRS water storage tank at El. 746'-0", Auxiliary Feed Pump Cells 204A and B and Cell 204 at El. 733'-0". The return air to the unit coolers is not ducted. An outside air duct connected to a missile protected outside air intake located at SGB auxiliary bay wall at El. 880'-0" with a supply fan provides the necessary ventilation for the auxiliary feedwater pump (electric driven) cells. An electric duct heater is installed in the outside air duct downstream of the supply fan to preheat the outside air during winter operation. An outside air filter is installed in the outside air duct to maintain the cleanliness of the outside air supply. Two (2) 100% capacity unit coolers provide conditioned air to the Turbine Driven Auxiliary Feedwater Pump cell. The supply air is distributed to the cell by an independent ductwork system to satisfy the cooling requirements. The unit cooler filters maintain the clean-liness of the supply air (for initial start-up only). The cooling coils provided in the unit coolers along with their instrumentation and con-trols, maintain the supply air temperature to satisfy the Indoor Design Conditions. The 100% unit coolers are located on a platform at El. 746'-0" O in Cells 202 and 2028. One of the unit coolers is enclosed with V a 3 hour fire rated wall. Each unit coaler consists of a disposable filter, cooling coil, and a V-belt driven centrifugal fan. Eacn unit cooler discharges air to the cell independently. Return air to the unit cooler is not ducted. An outside air duct connected to a missile protected outside air intake located at the SGB auxiliary bay west wall at El. 880'-0" with a supply fan provides the necessary ventilation for the auxiliary feedwater pump (turbine driven) cells. An electric duct heater is installed in the outside air duct downstream of the supply fan to preheat the outside air during winter operation. An outside air filter is in-stalled in the outside air duct to maintain the cleanliness of the out-side air supply.

2. Intermediate Bay (SGB-IB) System One of the two recirdulating type Air Handling Units with two (2) 100% capacity supply fans provides conditioned supply air to the SGB-IB areas above El. 836'-0" and the normal chiller rooms. The supply air is distributed by supply ductwork to the various areas to satisfy the Ventilation Requirements. The Air Handling Unit Amend. 49 April 1979

All Normal Chilled Water piping and piping components located outside of the ( ) RCB are built to the requirements of ANSI B31.1, " Power Piping", whereas heat \m / exchangers and pressure vessels outside the R0B are built to the requirements of ASME Boller and Pressure Vessel Code, Section V lli. The Normal Chilled Water System is terminated by two sets of ASNE, Section lil, Class 3 isolation valves, where cross-connections are made to the Emergency Chilled Water System. Upon loss of Normal Chilled Water Supply to the Emergency Chilled Water System headers, the isolation valves are closed automatically, and the Emergency Chilled Water System starts. Where the Normal Chilled Water System penetrates the RCB, one remote manually actuated ASME Section lil, Class 2 isolation valve is provided on each line. The piping on the RCB side of this valve up to the next manual isolation valve is ASME Sect i on l i l, Cl ass 2. The components served by the Normal Chilled Water System are listed in Table 9.7-1. The major system component design data are l isted in Table 9.7-2. 9.7.1.3 Safetv Evaluation One 20 percent capacity standby chiller unit is provided to ensure continuous cool ing capabil ity in case of a mal function of a chiller unit. One 20 percent capacity standby chilled water circulation pump is al so provided f or the same purpose. The diversity of the cooling loads provides additional ref rigeration margin f or the system.

   -g   in addition to these considerations, Section 9.7.3 lists system design g';   f eatures intended to prevent water /sodi um interactions. Pipe break analysis f or th i s moderate energy f l ui d sy stem w il l be prov i ded i n the FS AR.

9.7.1.4 Tests and insoections The system is tested and inspected as separate canponents at the manuf acturer's f acilities and as an integrated system prior to plant operation. All water flow rates are balanced and set to the design flow conditions. Periodic inspection of the equipment is scheduled to ensure the proper operation of the system. All chilled water lines penetrating the containment shall be provided with vents and drains to permit drainage. Normal chilled water supply and return headers immediately upstream and downstream of the containment isolation val ves shal l be drainable. Vents and drains will be opened to permit drainage and to permit transmission of containment test pressure to the closed Isolation valves. v 9.7 -3 Amend. 72 Oct. 1982

9.7.1.5 Instrumentation Application Chilled water system control panels are located in the area of the water chillers. These panels include control switches, monitors, and system alarms. Local alarms are provided for the following condi-tions:

a. Expansion tank high water level

b. Expansion tank low water level

c. Leak detection and isolation (described in Section 9.7.3)

d. High chilled water discharge temperature i

e. Water chiller trip alarm (includes following chiller mal-functions ):

1. low chilled water temperature

2. high condensing pressure

3. low refrigerant temperature or pressure

4. low chilled water flow

5. low condenser water flow

6. low oil pressure

7. high shaft vibration

8. high bearing temperature

9. high motor temperature 15e A common system annunciator for "a" through "e" above is provi-59 ded in the control room to indicate trouble in the Normal Chilled Water System. In addition, an annunciator alarm is provided for condition "e" in the control room, with first out indication locally for condi-tions "e.1" through "e.9" above. 44 E9l 9.7.2 Emergency Chilled Water System 9.7.2.1 Design Basis The function of the Emergency Chilled Water System is to provide 59l  ! chilled water for systems listed in Table 9.7-3. The Emergency Chilled Water System has a chilled water operating temperature of less l

iSl!than 60 0 F and an operating pressure of less than 150 psig. The system 44 l is designed to meet the following design criteria: Amend. 59 g g,4 Dec. 1920

( If during normal operation normal chilled water supply is interrupted, fl ow U switches in the emergency chilled water supply header wIlI close the ASE III Class 3 Isolation valves between the two systems and automatically start the Emergency Plant Service Water System and then the Emergency Chilled Water Sy stem. In addition to these considerations, Section 9.7.3 lists system design f eatures which are provided to prevent a water / sodium reaction. Pipe break analysis for this moderate energy fluid system will be provided in the FSAR. 9.7.2.4 Tests and insoections Af ter testing each individual component of the system, the entire system is tested prior to pl ant operation. Instruments and control s are provided f or periodic' ally testing the perf ormance of the system during normal plant operation or scheduled shutdown. All water flow rates are balanced and set to the design flow conditions. Periodic inspections of equipment are scheduled to ensure the proper operation of the system, in-service inspections w.il l be conducted according to ASE XI, as described in Section 9.7.2.1.g. AlI chiiIed water iInes penetrating containment shalI be provided wIth vents and drains to permit drainage. Emergency chilled water supply and return l ines immediately upstream and downstream of the containment isol ation val ves shal l be drainable. Vents and drains will be opened to permit drainage and to permit communicetiot of containment test pressure to the closed Isolation valves.

9. 7. 2. 5 Instrumentation Aoolication ChIIied water system contral panels are Iocated in the area of the water chillers. The panel s incl ude the control switches, monitors, and system al arms. Local al arms are provided f or the f ol lowing conditions:

a. High chIIIed water temperature

b. Low ch il l ed waier f l ow

c. Normal chil led to emergency chil led changeover val ve mal f unction

d. Expansion tank high water level

e. Expansion tank low water level

f. Leak detection and isolation (described in Section 9.7.3)

g. Water chil ler trip al arm (incl udes f ol lowing chil ler mal f unctions):

O

9. 7-7 Amend. 72 Oct. 1982

1. low chilled water temperature  !

2. high condensing pressure l

3. low refrigerant temperature or pressure

4. low chilled water flow

5. low condenser water flow

6. low oil pressure

7. high shaft vibration

8. high bearing temperature

9. high motor temperature Individual annunciator alarms are provided for conditions "a" through "c" above in the control room main control board and locally for glbothloopsAandB. A common system annunciator alarm for conditions "d" through "g" is provided on the control room main control board and locally for both loops A and B. In addition, an individual annunciator alarm is provided for condition "g" or back panel with first out in- 44 59l dication locally for conditions "g.1" through "g.9" listed above.

O I i l l Amend. 59 l Dec. 1930 9.7-8 l

__ . _ . . . - - _ _ _ - __ _ --- _ _ . - - - . - =m. - - - . - - - - - - - - - ~ . - -- - . - - - - . - - s , l I l l TABLE 11.2-5  ! EQUlFRNT DES 0tlPTION OF LIQUID RADWASTE SYSTEM , Quality Seismic Design Equipment Capacity Number of Throughput Class Category Te per- Design Description (gal) Components Rate (gpm) (RG 1-26)s (RG l-29)* Codes Meterial ature of Pressure

Piping and Valves - -

1-125 D lli ANS B31.1 SS 200 150 PSI lALL/LALL Filters - 2/2 125/50 D lli ASE V i ll SS 200 150 PSI l ALL Collection Tank 24700 2 - D 111 API 650 SS 200 Atmos. , LALL Collection Tank 2400 2 - D lli API 650 SS 200 Atmos.

Evaporator Pref il ters - 4 10 D lli ASE V i ll $$ 200 150 PSI f

Evapor ator s - 2 10 D 111 ASE Vill SS 200 150 PSI i Distillate - 4 10 D lli ASE V ill SS 200 150 PSI i Domineral iz er 1 Resin Traps - 4 10 D 111 ASE V ill SS 200 150 PSI I I ALL Distillate 24700 2 - D 111 API 650 SS 200 Atmos. 7 N Storage Tank

!                         O     L AL L Moni tor ing Tank            2400            2                 -           D           lli    API 650                 SS      200     Atmos.

Pumps - - I-125 D lil Manufactu- SS 200 Atmos. t ers Std.  ; I Caustic Neutralizing / 2500 1 - D 1II API 650 SS 200 Atmos. Storage Tank j Caustic Feed Tank 150 Antifoam Tank, l- Resin Feed Tank , Acid Feed Tank 700 1 - D lll API 650 HNA 200 Atmos.

• RG - Regu l ator y Gui de 4 I F8a -

w. ,

CD N j NN i i  ! 1 1 e- w

TABLE 11.2-5A INDOOR RADI0 ACTIVE WASTE TANKS - PROVISIONS TO PREVENT AND CONTROL OVERFLOW CONDITIONS Ianks Provisions 1._lntermodlate Activitv Level (a) Liquid Level I ndicator in Radwaste Llauld (IALL) Collection Control Room. Tank A & B (b) High and Low Liquid L vvci Annunciatcr Alarms in Radwaste Con 1 ol Room. (c) lALL Collection Tanks sized so that the two of them can hold the entire system inventcry overf ivw for both tanks is connected to the radwaste sump. (d) Cell walls are capable of containing any leakage. The contained liquid is returned to the radwaste sump vie floor drains. (c) A common Main Control Room alarm from the Radwaste Control Room to annunciate abnormal system conditions in the Radwaste Area.

2. Low Activity Level Llauld (a) Liquid Level Indicator in Radwaste (LALL) Collection Tank A & B Control Room.

(b) High and Low Liquid Level Annunciator / Alarms in Redweste Control Room. (c) Overflow for both tanks is connected to the radwaste sump. (d) Floor drains are provided to collect and return any leakage to the radwaste sump. O 11.2-20a Amend. 65 Feb. 1982

i l 11.4 FROCESS AND EFFLUENT RADIOLOGICAL MONITORING SYSTEM 11.4.1 Design Obtectives Process radiation monitors are provided to allow the evaluation of plant equipment perf ormance and to measure, Indicate and record the radioactive concentration in plant process and ef fluent streams during normal operation and anticipated operational occurrences. The monitors are provided in l accordance with CRBRP (Section 3.1) Design Criterion 56. Radiation monitoring of process systems provides early warning of equipment mal f unct ions, Indicative of potential radiological hazards, and prevents release of activity to the environment in excess of 10CFR 20 limits. Each monitor will be equipped with a loss-of-signal Instrument f ailure alarm and a high l evel al arm, (a high-high level alarm is also provided when required). These alarms alert operating personnel to channel mal function and excessive radi oactiv i ty. Corrective action will then be manually or automatically performed. Monitoring of liquid and gaseous ef fluents under normal operating conditions will be in accordance with NRC Regu!atory Guide 1.21 and any activity released will be within limits established n 10CFR20. The number, sensitivities, ranges, and locations of the radiation detectors will be determined by requirements of the specific monitored process during normal and postulated abnormal (accident) conditions. All monitors will be O designed so that saturation of detectors during a severe accident condition will not cause erroneously low readings. bbnitoring during severe post accident conditions will be accomplished by the high-range gunma area monitors discussed in Section 12.1.4, in conjunction with the sampling lines described in Section 11.4.2.2.1. Radioacilvity in the low level waste releases will be integrated and recorded. Control signals will be provided by the radiation monitor (s) to tenninate l iquid or gaseous ef fl uent if an out-of-limit signal is recorded. The monitcring and control exerted by the process radiation monitoring equipment and the operator during any release will al so be verified by periodic manual sampling and laboratory analysis in accordance with Technical Specifications. For tritlated process liquids, tritium surveillance will be by sampling and lab analysis. All detectors will be shielded against ambient background radiation levels so that roquired activity measurements can be maintained. Monitors associated with accident conditions are al so discussed in 3. A.3.1. Area monitors and airborne radioactivity monitors are discussed in 12.1.4 and 12.2.4, respectively. The radiological ef fluent sampling program is discussed in Section 11.4.3 and meets the reporting requirements of Regulatory Guide 1.21. O V 11.4-1 Amend. 72 Oct. 1982

11.4.2 continuous Monitoring /Samoling 11.4.2.1 General Descriotion The descriptive tabulation of the varIous continuous monitcrs/sampiers f or process and of fluent radioactivity monitoring, which ircludes those gas and iIquid monitcring devices in or associated with lIquid or gas process streams consi dered i n thi s di scussion, is found in Table 11.4-1. The basis fcr selecting the locations as well as the control f unctions associated with the monitor, are described below. Each cont inuous moni tor w il I be equipped w Ith power suppl les, micro-processor and accessories, Indication and local alarm indicator lights. Each monitor w il l transmit radioactivity level and alarm status Information f or display and logging by Radiation Monitoring equipment located in the Control Room with redundant display and logging equipment located in the Health Physics Area of the Pl ant Service Buil ding. The alarms are provided to Indicate instrument l mal f unctions or a radioactivity level in excess of the monitcr's alarn setpoint. Each continuous monitor has a local indicatcr at the detector location to f acil itate the testing and/or calibration of the equipment. The lowest scale division of each continuous monitor's range is the maximum detector sensitivity deemed appropriate for the intended service. The range of the monitor will be a minimum of five decades above the maximum sensitivity level ; and w il l al low for a minimum of one decade span above the monitor high-high setpoint (when high-high setpoints are employed). The of fl uent al arm setpoint corresponds to the al arm annunciation level dictated by the CRBRP Technical Specifications (Chapter 16.) For each monitcr, a sample ch mber and/or detector is selected and will be installed in such a way as to minimize sampi ing losses and electromagnetic and background interf erences. The output of all ef fluent monitors will be continuously sampled and recorded by the CRBRP Pl ant Data Handl ing and Di spl ay System. The Reactor Contairment Isol ation Monitcrs (PPS), Control Room Air intake monitcrs and other saf ety-rel ated monitors w il I be powered by Cl ass IE, redundant 120 VAC power. 11.4.2.2 Gaseous System Descriotion 11.4.2.2.1 Post-Accident Containment Atmosohere Monitors l The capabil Ity to monitor the coniainment atmosphere radioactivity level follcwing containment isolation during an accicent condition shall be provided. Three pair of penetrations, located 120 apart around the contai nment structure w il l al lcw ai r sampl es to be taken by mobil e or pcrtabl e moni tors and sampl ing equi pment. The penetrations design and locations w il l consider the fofIowing criteria:

1. The penetration opening on the inside of contair. ment w il l te positioned to obtain a representative sample.

! 2. The penetration opening on the outside of contairment will be l positioned in an acces.,lble area to enable connection of the , monitoring and/or sampi Ing equipment. l O l 11.4-2 Pnend. 72 Oct. 1982

3. Each penetration will have two isolation valves; a remote manual controlled valve inside containment and a manual, locked valve outside containmen1 with a blind flange.

(} 4 The penetration design will comply with CRBRP Design Criteria Numbers 45 and 47 (Section 3.1) Each pair of penetrations can be connected to a mobile moniter which will be utilized for continuous monitoring of the containment atmosphere. The sample is withdrawn f rom containment, passes through the monitcr for radiation detection and returned to containment. Grab samples will also be obtained f or f urther laboratory analysis. 11.4.2.2.2 Reactor Containment Isolation Monitors The radiation level in the head access area will be monitored by three detectors f or direct gamma activity. The output of these detectors is routed to the plant protection system to initiate closure of containment isolation valves if a preset 1 Imit is reached by two out of three of the detectors, in addition, the radiation level in containment exhaust, upstream of the isolation valves will be Isokinetically monitored for gaseous activity by three gas monitors. Their output will also be provided to the PPS for initiation of containment isolation when a preset radiation level is reached by two of the three detectors. The monitoring system will be designed to comply with IEEE 279-1971. The overall containment isolation system design and protection logic is discussed O in Section 7.3. Figure 12.2-1 shows a typical block diagram of these channels and Figure 7.3-1 shows the trip logic configuration. 1 1.4. 2. '!. 3 Buildino Ventilation Exhaust Monitors The number and location of building exhaust plenums f rom which potentially radioactive plant gaseous release may amenate are: One located in the intermediate Bay (SGB-IB), nine located near the top of the RCB dome, two located in the Reactor Service Building (RS3), one located in the Radweste Area (Bay), one located in the Plant Service Building (PSB), fourteen in the Turbine Generator Building (TGB), and three located in the Steam Generator i Buil ding (SGB). Continuous monitoring will be perfcrmed at those exhausts which could conceivably undergo a significant increaso in detectable levels in ! radioactiv ity. The remaining exhausts will be sampled periodically. l The exhaust plenum located in the IB receives ventilation exhaust air f rom the Intermedl ete Bay area. A continuous air monitor (CAN) will be provided to detect particulate, radiolodine and gaseous activity in the ef fluent stream. The air sample will be obtained Isokinetically from the exhaust, on a continuous basis. The operation of the three-channel CAM unit is described in Section 12.2.4.2.1. O 11.4-3 Amend. 72 Oct. 1982

The exhaust plenum located on the Radweste Building receives ventilation exhaust air f rom the radwaste area. A Continuous Air Monitor (CAM) will be provided to dotect particul ate, lodine and gaseous activity in the effluent st ream. The air sample wil I be obtained isokinetically from the exhaust, on a continuous basis. The operation of the three channel CAM unii is described in Section 12.2.4.2.1. The two RS3 exhausts will be continuously monitored f or radioactivity roleases. The f irst exhaust plenum located on the RSB roof which receives ventilation exhaust f rom the RCB will be continuously monitorec f or particul ates, radio gases, and radiolodino activity in the ef fluent stream. The second exhaust plenum located on the RSB which recolves ventilation exhaust f rom the RSB via the RSB cl ean-up f il tration units wil l al so be continuously monitored f or particul ate, gaseous and radiciodine activity. The exhaust plenum located near the top of the RCB dome, which receives exhaust f rom the Containment Clean-up and Annul us Pressure Maintenance cnd Fil tration System w Il l be continuously monitored f or particul ate, radiolodine, radiogas, and pl utonium activity in the ef fluent stream. The 8 exhausts located at the top of the RCB dome for the Annul us Cooling Air becomo potential rcdioactivity release poir.ts only in tFe event cf very low probabil ity accidents beyond the design basis (e.g., Thermal Margin Beyond the Design Base). On l ine monitoring f or particul ates, radiolodires and radiogases hcvc been provided for these exhausts in the evcnt of such en accident. TGB areas wil l be periodical ly grab sampl ed and sempl es w il l te analyzed f or tr i t i um act i v i ty. lf The exhaust in the PSB receives ventilation f rom the combined laboratory. Samples w il l be col lected i sokinetically by a particul ate (and lodine, if required) fil ter and analyzed f or isctopic content in the Counting Room. Certain ef fluent radiation monitors are identified as Accicent Monitoring instrumentation in Table 11.4-1. As such, these monitors w il l meet the requirements of Section 7.5.11 of the PS AR. The reporting of of fl uent radioactivity released f rom ihe CR3RP w il l be consistent w ith the guidel ines establ ished in Regul atory Guido 1.21. This reporting will be based upon the results of Counting Room analysis of ef flueni samples obtained at each location I isted above. 11.4.2.2.4 Condenser Vacuum Pumo Exhaust. Deaerator Continuous Vents and Turbine Steam Packing Exhauster Tritium Samolers A gas sample will be continuously withdrawn f rom each one cf the condenscr vacuum pump air, deacratcr exhaust, and turbine steam packing exhouster air into tri ti um sampl ers compr i sed of sil ica get dessicant col umr. to enabl e determination of tritium activity to indicate unacceptable tritium dif f usion in the steam generators. The sample will be analyzed using liquid scintillation techniques in the counting recm. O 11.4-3a Amend. 72 0' . 1982

11.4.2.2.5 Control Room inlet Air Monitors The main and remote Control Rcom air intekes will each be continuously monitored by two redundant monitors. These three channel (particulate /radiolodine/radiogas) CAMS will detect radioactivity in the air intakes and will determine which intake should be used during the Control Rom isolation condition. Detalls concerning the sequerce of operation during Control Room isolation are given in Section 9.6.1.3.4.13. A f if th three channel CAM wIlI be Installed downstream of the parallel INAC make-up air filters to monitor tne performance of the HEPA filter trains. A detailed description of the operation of each of these CAM units is given in Section 12.2.4.2.1. 11.4.2.2.6 Inerted Celi Atmosohere Monitors The capability for monitoring the atmosphere of each individual inerted cell for high radioactivity will be accomplished by three methods. One method is the sequential sampling of groups of cells with on-line gas monitors as described in 3.A.1.4.2. Each monitor shalI have a trip signal determined by the process system to initiate activation of cell purging equipment. In l addition, mobile particulate, Iodine and gas monitors are provided to sample any indivicual inerted cells atmosphere, as described in 12.2.4 Finally to provide a sensitive method of sodium leak detection, particulate l monitors are provided for continuous monitoring of inerted cells within the RG containing components contacting radioactive sodium. These monitors will 3' alarm for activated sodium present in the cells atmosphere. The Individual Inerted cells that are continuously monitored for sodium leak detection are l isted i n Tabl e 3. A.1-3. 11.4.2.2.7 RAPS and CAPS Monitorina Gas monitors will be provided for the Radioactive Argon Processing System (RAPS) and Cell Atmosphere Processing System (CAPS). A monitor will be located at the CAPS Inlet for controlling the rate of radioactivity input. Monitors will also be located at the output of these systems to ascertain that the radionuclide activity of the processed gas is within limits for reuse in RAPS or within 10 CFR 50, App. I and ALARA limits for those gases exhausted to the H&V system by CAPS. 11.4.2.2.8 Safetv-Related Monitors Certain monitors which provide control signal s to saf ety related process systems or are used to monitor safety related systems are classified as safety rel ated monitors. These monitors will be supplied with Class IE power f rom redundant vital AC buses and will meet the requirements described in Section 7.1. Saf ety rel ated monitors are identi f ied in Tabl e 11.4-1. These monitors will each have a dedicated Display and Control Unit (DCU) in the Control Rocan. The DCU will also meet the requirments described in Section 7.1 and wil l be suppl ied with Cl ass IE power. The DCU's wil l be located in the back panel area of the Control Room adjacent to the Radiation d Monitor Consolo (computer). 11.4-4 Amend. 72 Oct. 1982

11.4.2.3 Liquid Systems Description CJ 's 11.4.2.3.1 Radwaste Disposal System Liquid Effluent Monitor Effluents from the Liquid Radwaste Disposal System are discharged into the cooling tower blowdown. A liquid radioactivity detector will continuously monitor, record, and control the activity released to the cooling tower blow-down stream. The blowdown flow rate available for liquid waste dilution and compliance with 10CFR20 will be considered in establishing a high radiation set-point for this monitor. A high radiation signal will automatically close the isolation valve in the discharge line and alarm in the control room. Frequent composite samples of the blowdown downstream of the radio-active liquid input will be taken for radionuclide determination including tritium. 11.4.2.4 Maintenance and Calibration On completion of the monitoring system installation, each process monitor will be checked for proper operation and calibrated against a radia-tion check source (s) traceabie back to the National Bureau of Standards or from an equally acceptable sourcc. This initial calibration, and sub-sequent calibration at six month intervals will verify the electronic s 541 operation of both local and Control Room indications and also all annunciation points (loss-of-signal), loss-of-sample flow, high radiation, etc. In addition, each monitor is supplied with a built-in check 6 source to provide rapid functional tests at periodic intervals. O 11.4.3 Sampling This section provides information on the CRBRP process and' effluent sampling program. Process sampling provides the means for determining and monitoring various plant systems containing radioactive and potentially radio-active fluids. Effluent sampling provides the means for the reporting of radioactive releases to the environment. The effluent sampling will meet the reporting requirements of Regulatory Guide 1.21 and will provide data necessary for the semiannual report required by 10CFR50.

                                                               .~

Amend. 54 May 1930 11.4-5 l l

11.4.3.1 Process SamolIna Periodic sampiIng is conducted to alert the operator of any abnormal condition that may be devel opi ng. Both local and remote liquid samples are taken. Gaseous samples are taken directly at the sample station adjacent to the gas analyzer. The locations f or gaseous sample instrumentation are given in Section 11.3.3.3. Operating procedures and performance tests of gaseous samples are discussed in Section 11.3.4. S aplIng of primary sodium, secondary sodium, ex-vessel sodium and cover gas is discussed in detail in Section 9.8, entitled " Impurity MoniforIng System". This sectIon al so discusses the location of samples, expected composition and concentration, sampi ing f requency and procedures. The basis f or seiecting the 1ocations f or sampie stations Is to prcyIde an Indication of the of fectiveness of key process operations. Analyses of these samples are related to the process sequence f rcm which they were obtained to eval uate spect f ic equipment perf ormance. Gaseous samples are monitored f or gross activity and periodically analyzed for Isotopic content. Tables 11.3-1 through 11.3-15 list inventories of the expected concentration and composition of the of fluent gas samples. Sections 11.4.3.1.1 through 11.4.3.1.5 describe in detail each of the lIquid sampiIng points in the Radioactive Waste Systems. Sampi ing procedure, analytical procedure, and sensitivity for each sample point are the same and are discussed in detail in the f ol lowing paragraphs. Sampl ing Procedure: Samples are collected in a sampling station located on the operating floor of the radwaste building. Sample circulating lines run l thrcugh this sampling station. The upstream side of the sample lines are connec1ed to the discharge of the pumps serving the tanks. After passing through the sampling station, the circulating sample fluid is returned to the tank f rcm which it was drawn. Analytical Procedure and Sensitivity: The quantity of sample to be counted f or gross beta-gamma is pipetted onto a pl anchet. The planchet is placed on a turntable and evaporated to dryness under an inf rared bulb. The rotation Insures a uni f ormiy di stributed dried sampie f or reproducible counting. The height of the inf rared bulb is adjustable to obtain a moderate rate of evaporation. Counting is done by means of an internal proportional counter. The isotopic analysis is perf ormed by a compietely autcmated Pulse Hetght Analysis System. A shielded Ge (LI) detector is used with a computer-based i pul se height analysis system. The system satisf ies the reporting requiren'ents of Regul atory Guide 1.21. Provisions wil l al so be made f or al pha and tritium assay. O 11.4-6 Amc nd. 72 Oct. 1982

11.4.3.1.1 Intermodlate Level Activity Llauld Waste Collection Tanks 1 hose tanks receive decontamination waste f rom the Large Cceponent Cleaning Cel I . The analysis of this waste provides a check on the decontamination procedure. The composition is expected to be sodium hydroxide solution, nitric acid sol ution and water rinses. Af ter neutral ization a sol ution of sodi um sul f ate or sodium nitrate results. Activity will be SI uCl/cc. The quantity to be measured is the gross 6-y activity. Additional rinses would be required if the activity of the component is higher than expected. Additional passes through the purification equipment would be required if the activity of the product f rom the evaporator is too high. Corrective action woul d be taken if the DF is lower than the expected val ue. The expected recirculation flow through the sample line is 10 gpm. 11.4.3.1.2 Process Distillate Storace Tanks These tanks receive the distillate f rom the Process Waste Evaporator. The sample provides the check on the DF of the evapcratcr and purity of the product to be recycled for plant uses or released to the environment afier dil ution w ith cool ing tower blowdown. The composition is expeged to be very dii ute sodium sul f ate or sodium nitrate wIth an activity 4 10 pCl/cc. O The quantity to be measured is the gross 6~Y activity, if no excess inventory exists. If excess inventory exists and a portion of the content is to be

   . released to the low activity liquid system, an isotopic analysis will bo l perf ormed consistent w ith reporting rcquirm.ents of Regul atory Guide 1.21.                                            If the activity of the sample is unacceptably high, the contents of the tenk are reprocessed through another evaporator-lon exchange cycle. Corrective measures would be taken if the DF is much lower than the expected value.

The expected recirculation flow through the sampl ing line is 10 gpm. 11.4.3.1.3 Low Level Activity Liauld Waste CoIIection Tanks These tanks receive l aboratory drai ns, fl oor drains, lavatcry drains, and shower drains f rom areas that reay contain radioactivity. An activity check at these points determines the possibility of the need for f urther processing. It al so permits a check on the DF of the purif ication equipment by comparir.g it with the activity of the purif ied waste. These tanks receive waste f rca several sources, hence the composition is not wel l def ined. The conductivjty will be measured to determine impurliy level. The expected activity is 10- u CI/cc. The quantity to be measured is the gross 6-Y acti v ity. O 11.4-7 Amend. 72 Oct. 1982

l The sampling f requency will be in accordance with reporting requirments of Regul atory Guide 1.21. ' Higher sample activity indicates abnormal operations elsewhere in the plant. Corrective measures at those locations would be taken. Al so, higher activity indicates that a second pass through the equipment would be required. The expected recircul ation flow of the sampi lng l ine is 10 gpm. 11.4.3.1.4 Low Level Activity Distillate Monitorino Tanks Since these tanks are holding tanks f or the purified product f rca the low level waste evaporator, pendi ng rel ease to the di scharge canal, sampl e analysi s i s mandatory. The composition is expected to be equivalent to grade C water or comgy with f ederal and state regulations and have an average activity of 10 p Cl/cc. A gross 6-y-a count is made bef ore releasing to the environment. Tritium content wiII also be sampied. An i sotopic analysis is perf orned f or record purposes as required by Regul atory Guide 1.21. Sanpl ing f requency wil l be determined by reporting requirments of Regulatory Guide 1.21. High sampic activity indicates the need for reprocessing the batch. Corrective measures woul d be taken i f CF is Iower than the expected I evel . No particul ar process flow is associated with this sample point. 11.4.3.1.5 Concentrated Waste Co!!ection Tank The material in this tank is intended to be solidified and shipped to the disposal site. To determine the type of packaging and degree of shielding required to meet the shipping regulation CFR Title 49, the analysis of sample is necessary. The composition is expected to be a solution of sodium sul f ate or sodium nitrate and an activity of N 50 pCi/cc. The quantity to be measured is the gross B~Y activ ity. The sampl ing f requency wil l be determined in the FSAR. No process flow is associated w ith this sampi ing procedure. 11.4.3.2 EffIuent Samoting The radioactive ef fluents are continuously monitcred or sampled as Indicated in Section 11.4.2.2.3 by activity and by flow. The sampi Ing system is designed to obtain a representative ef fluent sample to establish l concentrations of radioactivity and to f acilitate radioisotopic analysis to assure compliance with recognized codes and standards f or radiation protection. The samples are taken before the ef fluent release to the environment. The gaseous of fluents are discussed in detail in Section 11.3 and lIquid of fluents are discussed in Section 11.2. 11.4-8 Amend. 72 Oct. 1982

The Cooling Tower blowdown, wastes and drains and other normally non-fi radioactive l iquid ef fluent streans wil l be sampled f or suspended /dissol ved V activity including tritium. The probim associated with continuous monitoring of low level $ activity in tritium is recognized and therefore, pericdic batch samples f ran each IIquid ef fluent strean will be taken and analyzed in the Iaboratory. Building Storm drains and Plant Service Building liquid ef fluents are normally non-radioactive and w ll l not be monitored, but wilI be periodically sampied f or radioisotopic analysis as necessary. . To satisfy Regulatory Guide 1.21 requirements for gamma spectroscopy and sensitivity, a high resolution automated radioisotopic analysis system will be proviced at the plant site to f acilitate precise identification and analysis of compl ex radionuct Ide concentrations. 11.4.4 Pecorting An automated Report Processor wilI be provided which wlll generate the Ef fluent Radioactivity Release Reports in accordance with Appendix B of f;RC Regulatcry Guide 1.21. This computer based processor wilI be interf aced wIth the Radiation Monitoring System Controllers and the CRBRP Environmental Computer. The Report Processor wilI also accept rr.anual ertry of analyses performed by the Health Physicist. O l i O 11.4-9 Amend. 72 Oct. 1982 1

TMILE 11.4-1 PRO SS & EFFLUENT M)NITORING NfD SAMR.lW Seple or Range Expected Quant. Description Oldg. Elev. Cont. pCI/cc)U0S Concent. Meas. Rev= =r ks Reactor Contalrrnent isolation Monitors (PPS):

       -Contaltnent Ventilation (3)    RG         842      Continuous   10'I-10 -2   Cs I37         See Section Gross     Saf ety-rel ated Exhaust (Gaseous) CAM                                                                      11.3.2.6     Concent. C1 ass IE PPS Related
       -Head Access Area (3)           RG         802      Continuous   10'I-10 ~4 mR/hr                         Direct   See Section Direct Ganma                                                                                            Gama     7.3.1.2 Redweste Monitors NA                  Continuous   4x10~7 4x10  -2 Cs I37                   Gross
       -I ALL Evaporator, Heating                 775 Elment : Heating Water                                                                                  Concent.

Out (Liquid) Continuous 4x10-7 4x10 -2 Cs I37 Gross

       -LALL Evaporator, Heating       RfA        775 Elment; Heating Water                                                                                   Concent.

Out (Liquid)

~
       -lALL Evaporator, DistIlI.      RfA        775       Continuous  4x10~74x10   -2 Cs I37                  Gross
-        Cooler; Cooling Water                                                                                   Concent.

3 Out (Liquid)

       -LALL Evaporator, Distill       RfA        775       Continuous  4x10-7 4x10  -2 Cs I37                  Gross Cooler; Cooling Water                                                                                   Concent.

Out (Liquid) RfA Continuous 4x10-7 4x10 -2 Cs I37 Gross

       -LALL Ef fluent                            795 Concent.

RAPS & CAPS Process Monitors:

       -Gas Entering RAPS              RG         733       Continuous   2.7-2.7x10 5 g,85                       Gross Cold Box (Gaseous)                                                                                      Concent.

4

       -Coolant Leaving RAPS.          RSB        779       Continuous   2.7x10    -2.7x 10~I Kr 85              Gross    in-Line Cold Box (Gaseous) (2)                                                                                  Concent. Monitoring 85              Gross
        -Cas leaving RAPS              RG         733       Continuous   2.7x10-3-2.7x 10+ 2 Kr Cold Box (Gaseous)                                                                                      Concent,

.s c.

                                                                                ~4 E'      -Gas leaving CAPS               RSB        779       Continuous   2.7x10    -2.7g10" k{85                 Gross Qy        Surge Vessel (Gaseous lodine)                                              10~ -10* I 3I                Concent.
                                                                                            ~I   83 Gross
        -CAPS Header Serving           RO3                  (bntinuous   2.7x10    -2.7x 10 Kr RG Cells (Gaseous)                                                                                      Concent.

TM3LE 11.4-1 PRORSS & EFFLUENT FONITORING AND SAFf' LING Septe or Range Expected Quant. Description Bldg. Elev. Cont. (ACl/ce) 00S Concent. Meas. Rm arlts

        -Gas Frcm Nitrogen Cell       R2          755       Continuous   2.7x 10-6 -2.7x10'I kr 85 Gross Atmosphere Sept Ing Unit                                                                             Concent.

(Gaseous ) 4

        -Gas Frca Nitrogen Cell       RW          752       Continuous   2.7x10    -2.7x10'I Kr0 '             Gross Atmosphere SamptIng Unit                                                                             Concent.

(Gaseous 1 CAPS Process Gas Ef fluent to FN AC (Gaseous) ( 2) RSD 779 Continuous o 85 Gross 2.7x 10-5,g,7,9 I (lodine) 10-3 -10' I Concent. Ef fluent Gas Frm (2) Inerted Cells to IN AC RSB 800 Continuous 2.7x10-6-2.7x10~I Kr 85 Gross (Gaseous ) Concent. tNAC Duct Monitoring (CAM g of RAPS / CAPS Cells:

        -RAPS Cold Box & Valve        RG          733       Continuous   2.7x104 -2.7x10 -I    Kr0 '           Gross    in-line a        Gallery Cells (Gaseous)                                                                              Concent. Nnitoring &

Celi 1 solation

                                                                                -6                 0
        -RAPS Noble Gas Storage       RG          733       Continuous   2.7x10 -2.7x10"I Kr                   Gross    in-line Vessel Cell (Gaseous)                                                                                Concent. Monitoring &

Cell isolation

        -RAPS Cmpressor and                                                                        85 RG          733       Continuous   2.7x10'0-2.7x10'I Kr                  Gross    In-Line Aftercooler Cells (2)                                                                                Concent. Monitoring &

(Gaseous) Celi Isolation

        -RAPS Vessel s (Gaseous)                                                -6       ~I      85 RG          733       Continuous   2.7x10 -27x10        Kr               Gross    in-Line Concent. Monitoring &

Cell i sol ation Continuous -6 ~I 85

        -RAPS / CAPS Pipeway          RG          780                    2.7x10    -2.7x 10 Kr                 Gross    In-Line (Gaseous)                                                                                            Concent. Monitoring &

Og Cell isolation

   $    -CAPS Cold Box Cell (Gaseous) RSB         792       Continuous   2.7x10-6-2.7x10'I Kr '    0 Gross    In-Line

,_ , P Concent. MonttorIng & O Celi isolatton 03 N

        -CAPS V essel Cel l s         RSB         755       Continuous          -6 2.7x10 -27x10
                                                                                         ~I   Kr 0

Gross In-line

          & Gallery (Gaseous)                                                                                  Concent, mnitoring &

Cell i sol ation O O e

TMLE 11.4-1 PROCESS & EFFLUENT K)NITORING AND SAM 1.ING Sample or Range Expected Quant. Description Bldg. Elev. (bnt. (nCl/cc) 00S Concent. Meas. Ramnarks

       -CAPS Compressor & (2)                               Continuous    2.7x104-27x10'I      Kr '             Gross    In-Line Af ter Cooler Cells (Gaseous)                                                                          Concent. Monitoring &

Cell isolation

                                                                                            -I     85
       -RAD Water Holding                                   Continuous    2.7x10'0-27x10       Kr               Gross    In-tim Vessel & Ptsnp Cell (Gaseous)                                                                           Concent, penitoring &

Cell lselation 0

       -Access Areas (4)                                    Continuous    2.7x10 -27x10
                                                                                           -I Kr '             Gross    In-Line (Gaseous )                                                                                             Concent. Monitoring &

Cell isolation B5

       -Cover Gas Monitoring                                Continuous    2.7x10 -27x10'I Kr                    Gross    In-Line CelIs (Gaseous)                                                                                        Concent. MonttorIng &

Coll isolation 85

       -Pipe Osase & Vapor             RW          772      Continuous    2.7x104-27x10'I      Kr               Gross    In-Line Trap Cell (Gaseous)                                                                                    Concent. Monitoring &

Cell Isolation

 -     -W AC Common Header             RW          766      Continuous    2.7x104-27x10 -3     Kr 85 Gross    In-Line F       For various Cells                                                                                     ,Concent. Monitoring &

A (Gaseous) Cell isolation 8 N Main HV AC Duct Continuous 2.7x10 Gross Frm All RAPS / CAPS Cells (Gaseous) CAM RSB 779 10 7x10

                                                                                        -10     Kh 1              Concent.

(lodine) Soditsn Leak Detection For FollowIng Recirc. Gas Cooling Subsystems: (All Particulate) Reactor Cavity RG 733 Continuous 2.94x10'I3-2.94x10-5 Na24 Gross Alarm Only Concent. FHTS Loop I RG 766 Continuous 2.94x10'I3-2.94 x 10 -5 Na 24 Gross Alarm Only Concent. litTS Loop 2 RG 766 Continuous 2.94x10'I3-2.9 4x 10 -5 Na 24 Gross Alarm Only Concent. FNTS Loop 3 Rm 766 Continuous 2.94x10 -I3

                                                                                       -2.94x 10-5 g,24         Gross    Alarm Only Ok      Na Makeup Pump & Vessels        RG          752       Continuous   2.94x10'I 3-2.94 x 10 -5 g,24 Concent.

Gross Alarm Only rtQo. Concent.

                                                                                                -5     24 g        Na Makeup Pump & Pipeway       RG          752       Continuous   2.94x10'I3-2.9 4 x10 Na               Gross    Alarm Only cn y                                                                                                             Concent.

NN

T/ULE 11.4-1 PRORSS & EFFLUENT MX41TORING ANO SAWLING S epte or Range Expected Quant. Description Bldg. Elev. Cont. (xCl/cc) U05 Concent. Meas. Rear k s Col d Trap, Nak Cel ls RG 794 Continuous 2.94 x 10~ 3 3-2.9 4 x 10 -5 g,24 Gross Alarm Only Concent. Control koom Main (2) G 863 Continuous See Section Ginss inttlate C/R Air intehe (Gaseous) CAM 3x10 - x10 2 kr 3, 12.2 Concent. Isolation, see ( l odi ne ) 4x10 p-4x10'7 i Sec. 7.6.4.5.6 (Particul a te) 2x10

                                                                                           -10
                                                                                                -2x10 -5 Cs I                              Saf ety-Related (IE)

Control Rom Rmote (2) SGB 851 Continuous See Section Gross initiate C/R Air intake (Gaseous) CAM 3x10~7-3x10 -2 Kr 85 12.2 Concent. Isolation, see I I (lodine) 4x10'I2-4x10'5 I Sec. 7.6.4.5.7 (Par t i cul ate ) 2x10-10-2x10 - Cs I3 Safety-Related (IE) Control Room Cm mon G 847 Continuous See Section Gross Monitcr Only Duck Downstream of 12.2 Concent. Filter Units (Gaseous) CAM 3x10~7-3x10-2 kr 85 (lodine) 4x10-12 4,9o-7 g 137 [ (Particulate) 5x10-10-5x10-5 Cs I37 p (HTS Loop 1 SGB 765 Continuous 10'2-103 mR/hr Gross (Olrect G ema) Act iv i ty w 3 IHTS Loop 2 SG8 765 Continuous 10~2-10 mR/br Gross (Olrect G mmal Activ i ty IHTS Loop 3 SG8 765 Continuous 10~2-10 mR/hr 3 Gross (01 rect G ama) Act Iv i ty Large Camponent RG 756 Continuous 10'I-104mR/hr Gross Clcaning CetI (LCCC) Acti vi ty LCCC Cooling Water RG 733 Continuous -7 -2 3 4x10 -4x10 Cs Gross (Liquid) Concent. LCCC Process Gas RG Continuous 10 10'I Kr 0 Gross Efiluent (Gaseous) Concent.

           @@     Fuel Handling Cell (FHC)        RSB         779       Continuous      -       ~

r+g Argon Gas (Gaseous) E 10,6g10 ,13 kr,53, Concent, a (lodine) 10 137 (Particulate) 10 -1010-5C 's "o" - NN EVST Argon Cover RSB 842 Continuous 10 -10 kr 4 85 Gross Gas (Gaseous) Concent.

                 #                                                                 9                                                                      e

f (' TM3LE 11.4-1 PRO &SS & EFFLUENT FONITORING AND SAWLING Seple or Range Expected Quant. Description Oldg. Elev. (bnt . ( Cl/cc) 005 Concent. Meas. Rm arks THC Util i ty Moni tor RSB 779 Continuous 10'I-10 mR/hr Gross (Direct G ama) Act Iw Ity Radweste Oullding M 867 Continuous Gross initiate I

                                                                                -1 3 85                            Concent. Filtering of Exhaust (Gaseous)                                                3xtf0           gI CAM       (lodine)                                               10- -10       1 10 Cs I3I Effluent (Part icul ate)                                        10                                                      from M RSB Oper at i ng Fl oor ( 2)  RSD        816       Continuous                                                 Gross      initiate RT IN AC Exhaust (Gaseous)                                                                                                                       3x10~I px10
                                                                                       -2     85
                                                                                        ~7Kr I33                    concent. Conf Inement see CAM       (todine)

I Section (Particul ate) 4xjl-10'f 10 x10 Cs 137 7.6.4.3.3 ( 4) Safety related (IE) Fuel Handling Cell (2) R2 779 Continuous Gross - saae - IN AC Exhaust (Gaseous) 3x10 Kr 05 Concent. (lodine) 4xg' Igx 10-2-fx10'I3III I (Particulate) 10 -10~ Cs Annulus Filter (2) RSB 840 Continuous Gross Select Filter i Discharge (Gaseous) 851 4.4x104 -7

                                                                                  -4x4x10           0 Kg3I               Concent. train Section
~                 (todine)                                             1.1x10 -1                                               7.6.4.2.2 (1)

I37 1.2x10-10 1x10" 5lCs

 #                                                                                  -1.2x10                                    Saf ety Related (Particulate)

(IE) Annulus Filter Inlet /(2) RSD 840 Continuous Gross 1) Start Filter Annulus Cooling Exhaust 861 Concent, see 7.6.4.2.2

                                                                            -7 3x10 -

d 05 (6) 2) Monitor CAM (Gaseous) lx10-10_1x,,10 fr 131 Exhaust see ( lodi ne ) jg gI 15.4.2.2.3 (Particul ar) 1x10 -1x10 Cs (Accident (Monitor) RSB Clean Up Fil ter RSD 816 Continuous Gross Select Fil ter Discharge (Gaseous) 794 3x10~7-3x10-2 Kr 85 Concent. Train See 1x10 1x10 -5 g 131 $,c,gn, (lodine) I3

                                                                             -6        -I 7.6.4.3.3(1) og                   (Particulate)                                      1x10 -1x10         Cs Safety Qm

.3 Related (IE) ct b' Radweste Ventilation $M Exhaust Effluent (Gaseous) RS8 867 Continuous 1x103 r See Section Gross Effluent, Acci-(lodine) 1x10~ 1x10 1x10 1I3 11.3.6 Concent. dent Monitor (Par ticul ate) 1x10 -Ix10 Cs

TFOLE 11.4-1 PRO SS & EFFLUENT 10NITORING #4D SA41.1NG Se ple or Range Expected Quant. Oldg. Elev. Cont. (etC1/cc) 005 Concent. Meas. Remarks Description RT Ver,tilation RSD 861 Continuous See Section Cross Exhaust Ef fl uent (Gaseous) 1x10~0-1x10"I Kr 85 11.3.2.6 Concent. 1x10 1x10 -5 g 131 (todine) -5 Cs I37 (Particulate) 1xt0 ~I O-1x10 840 Continuous Accident Rm Annulus /TFBW (2) RSD 3 85 Monitor Effluent (Gaseous) 861 1x10 - r 101x10 I3I Saf ety Related I (lodi ne ) 1x10 10 1x10 Cs I I37 (1E) (Par 1iculate) 1x10 I2 1x10~7 239 (PIutonIum/AIpha) 1x10" 1x10 Pu R2 816 Continuous Accident RSD Exhaust Ef fIuent (Gaseous) 1x104 -1x10 85 I3I MonttorIng l (lodine) 1x i G"I1x10 1x10~IO -

                                                                                               ? -1 x10 l37                                                                         f (Particulate) 83 6     Continuous                               See Section           Accident 5G0-18 Exhaust                        SG8 A       Ef fluent (Gaseous )                                                     1x10-6-1x103 rr85I         11.3.2.6              Moni ring                                        j

[-. ( lodi ne ) 1x10-10-1x102 3 131

 *                   (Particulate)                                                1x10-10-1x102 Cs I37 Se pte n                                              Gross Hot Laboratory, Counting              PSD Concent, i

Roon, and Decont mInatIon ' Area v entilation Exhaust Particulate Smpler Sept e 8' See Section Concent. Pl ant Olscharge YARD - Canal Liquid S e pler 11.2.5

          ** Particulate collection on f I! ter, analyst s by proportional counters and spectroscopy system.

Liquid $mples collected in container. Analysis by proportional and liquid scintillation counters and spectroscopy system.

   ?m5 n

• 3 e

Co N NN

• O e---

LALL Distillate Demineralizers The Activity inventory of the LALL demineralizers is provided in Table 12.1 -39a. This inventory is assumed to contain the activity inventory of the LALL Collection Tank (Table 12.1-35). This is based on accumulation in the demineralizers of the activity in one 2400 gal, batch of filtered LALL process fluid (assuming the evaporator are bypasses).

       .lALL and LALL Resin Traps l  The purpose of the resin traps downstream of the distillats domineralizers is to catch resins which may be contamirated which have broken away from the demireralizer beds. The source term for each resin trap is assumed to contain 6% of 1he activity Inventory of the demineralizers (Tables 12.1-39 and 12.1-39A). This reflects the activity that would be present in the resin traps, should a rupture of the demir.eralizer resin retention devices occur.

The activity inventory of the l ALL resin trap is provided in Table 12.1-40, and the Inventcry cf the l ALL resin trap is provided in Table 12.1-40A. foncentrated Waste Tank The Concentrated Waste Tank in the SRWS receives the concentrated radioactive wastes from the lALL and LALL evaporators. The activity inventory is given in Tabl e 12.1-43. Decantino Tank The Decanting Tank collects the powdered resin waste f rom the spent l ALL and LALL Distillate Domineralizer resins. The activity inventcry for the Decanting Tank is based on isotope inventory cf the spent Distillate l Demineralizer resins in Tabie 12.1-39 and 12.1-39A. The activity inventcry is given in Table 12.1-44. Decantate Filters The Decantate Filters rcmove undissolved solids from the liquic decanted off the Decanting Tank. These filters are assumed to contain 1% of the activity of the Decanting Tank. The activity inventcry is given in Table 12.1-45. 101Id Radwaste Drums Concentrated liquid radwaste and spent resins will be drumtred, solidified and stored in the SRS in the Radwaste Buildirg. Up to 136 drums per year containing concentrated liquid waste will be stored in the high activity drum stcrage vault. Each drum will contein 30 gallons of concentrate from the Concentrated Waste Tank. The activity inventcry per drum is shown in Table 12.1-46. Up to 17 druros per year containing spent demineralizer resins will be stored. Each drum will contain 17 gallons of spent resins and f rom the Decanting Tank. The activity Inventory per drum is shown in Table 12.1-47. O 12.1-22b Amend. 66 March 1982

12.1.4 Area Radiation Monitoring 12.1.4.1 Design Criteria Area monitors are provided in selected building locations to continuously detect, measure, and Indicate the radiation level and to initiate alarms (audible and visual) for radiation l evel s above preset val ues. In high or varied noise level areas (195db) strobe lights are also provided in addition to the audible alarms. These monitcrs advise plant personneI of extsting radiation levels during nor;nal operation and warn them of pctential radiation hazards that may cause higher exposure level s than expected. l The detector ranges of these rnonitces are chosen to provide continuous monitoring of gamma radiation levels ranging f rom one decado below to three decados above the design background level at each monitor location. l O l l l l l 1 0 1 2.1 -23 Amend. 7 2 Oct. 1982

l l l l The basis for location of tha various personnel protection monitors shall O consider the fof Iowing f actors:

1. The anticipated rcdiation level under operation, shutdown maintenance, and abnormal conditions.

2. The f requency and duration of occupancy, and the ficw of traf fic under normal and accident conditions.

3. The proximity of high radiation scurces.

4 The consequence of an undetected increase in radiation level. In addition to the personnel protection monitoring utilized during normal plant conditions, accidqt area 4monitoring will also be provided. Area monitoring f or range 10 to 10 R/hr will be provided in the following areas:

1. Inside buildings or areas which are in direct contact with primary containment where penetrations and hatches are located.

2. Inside buildings or areas where access is required to service equipment important to saf ety and the threat of radiation contamination extsts.

7 Three high-range monitors of range 1 to 10 R/hr will be provided to monitor the l evel s of ger.ma radi ation in the Contai nment Area. The detectors f or these monitors will be located approximately 120 apart around the Containment O- vessel periphery in the Annulus space so as to alicw a measurement of gamn;a activity being radiated f rom containment. The location of these monitors is in the more benign environment of the Annulus rather than in containment to avoid the severe temperature transient and direct sodium aerosol which may occur during and f ol low ing an aci dent. These monitors are saf ety-rel ated and each is supplled which a separate division of Class IE power. The Accident bbniicrs as identified in Table 12.3-5, will meet the requirements of Section 7.5.11 of the PSAR The locations of the area monitors provided f or the CRDRP are shown on l Figs.12.1-1 to 12.1-19d and are I isted in Tabl e 12.3-5. 12.1.4.2 Monitorina System Descriotion Each area monitoring channel consists of a gamma detector, microprocessor and accessories, local indicators, al arms, and Control Rocm Indication. The gamma detector energy dependence wIlI be fiat wIthin 120% for incident radiation above 100 Kev. L ocal monitor displ ay incl udes Ioss-cf-signal, high and high-high radiation Indicator lights, high and high-high radiation audible alarms and mR/hr rate meter. Al so, an essential feature of each monitoring chennel will be its abil ity to avoid "f ol dover" f ol lowing saturation in high radiation fleids. O , 12.1 -23 a Amend. 72 Oct. 1982

The detector signal is al so displ ayed on redundant Radiation Monitoring System GTs located in the Control Roan and Health Physics Area of the Plant Service Building via their respective Central Processing Units and Mini-Computers (System Control lers). The indicating analog meter ir each local monitor indicates exposure levels on a suitable multi-decade logarithmic scale. The al arm signals are al so permanently recorded by the redundant Radiation Monitoring System Lir.e-Printers located in the Control Roan and Health Physics Area. l Group annunciation is also provided on the Main Control Board. O O 12.1 -23b Amend. 72 Oct. 1982

1 Each area monitor will contain a built-in solenoid actuated shielded check i source which can be actuated f rom the remote process station in the vicinity. All monitor components wil l be modul ar, commercial ly avail able units designed f or rapid repl acement upon f ail ure. Electronic components will be exclusively solid-state, as available; and power will be supplied f rom the instrument AC i l (120V, 60H ) busses f or the non-caf ety monitors. Area monitors perf orming containmenf isolation f unctions (PPS) will be supplied with Class IE power from redundant vital AC busses. The high radiation alarms of all area monitors are transmitted f rom the local monitors to the Remote Data Aquisition Terminal units in the vicinity. The Plant Data Handling and Display system will display and log all high alarms. Figure 12.1-21 shows a f unctional block diagram of an area radiation moni tor. Locations, design dose rates and ranges of sensitivities of the monitors are

 ;                  l provioed in Table 12.3-5.

12.1.4.3 Maintenance and Calibration On completion of the monitoring system installation, each area monitor will be checked for proper operation and cal Ibrated against a radiation chtcr:curce traceable to the National Bureau of Standards or f rom an equally as ceptabl e source. The initial calibration and subsequent calibrations at six month intervals will utilize a minimum of two source strengths to verify the l l inearity of detector output. In addition, each monitor is supplied with a buil t-in check source to provide a rapid f unctional test at periodic i nterval s. 12.1.5 Estimates of Exoosure Peak External Dose Rates and Annual Doses at Unrestricted Locations The peak dose rates and annual doses at the site boundary and control room due to direct plant radiation are low and considered small relative to the natural background radiation. These doses have been estimated and are shown in Table 12.1-49, Parts I, ll, and Ill. i 1 l 4 i 1 2.1 -24 Amend. 72 ! Oct. 1982 I

        . , , , ,,,,-. , . ,--. - - v-          - - -
                                                      ,-,----..,-e           , , ,    .-- - , , - -, , . .r-..- ., , r ,-m,, ,,,_- - , - - - - . , - . -   e u ,    c .

i i i T8BLE 12.1-1 (UNT RADIATION ZONE Q.ASSIFICATION Zoss Dose Rate Design Dose Specification Z,gne Area Tvne Access Rate faren/hr) Imram/hr) Tyne of control

                     -      Unrestricted       Continuous                         -                                                 Uncontrol led Area i     Restricted         Continuous                        0.2                                      10.2        Administrative Control                               !

Area Routinely Occupied

ll Restricted Continuous, 2.0 >C.2 to <5 A4ministrative Controi Area Not Routinely ,

Occupied  ! 111 Radi ation Periodic 10m >5 to 1100 Aeninistrative Control Area Limited Access for Routine

!                                               Tasks

, N j - lW High Radiation Unoccupied

• 100 >100 to <5000 Special work Permits, Limited Access Locked Doors, Signs, 1

7 N Area for Non-routine imporary Barricados, i CD or Infrequent Health Physics asks Survell iance V Extremely unoccupied

• Unilalted >5000 Positive exclusion, High Radiation Locked Doors, Special t Area work Permits, Continuous

 '                                                                                                                                    Heelth fhysics Mont for Ing i

i

• 10CFR20 criteria.

                    "    Approaching background radiation.

3

                   *** 25 mran/hr within HAA 1

1 l ?N i n.3 C j .-a .L i c co u NN e t

TABLE 12.1-2 SHIELD PARAMETERS FOR THE HEAD ACCESS AREA Sg+cific Design Radiation Predominant Sour ce s Consi der a t ions Type cl tkri t ral thickne=s zone __ __ pf_ rad 1Atlp L __ ApflylM.39_SQuf Ces SM eJJ11.nLP gqujted __aLSMfj d___ _ (11 N[# gammas f rm gr imar y Major Source Cemetries: Concrete Walls Along 5' Inlet and outlet coolant (a) 36" Pige Fer ighery of HAA, pi p i rig (b) 24" Pige Atove Suppor i Ledge lli Na gammas f rm In-vessel (a) In-Vessel SodlLm Concrete Suppor t ledge; 6' (serpentine rodium pool and pr imary Fcol Carbon Steel F:eactcr conc r e te ) sodium coolant pirf rg. (b) 36" Pipe vessel Supperi Structure o" (steel) In-vessel neutron leakage (c) 24" Pige through sodium tools (d) Feactcr Cavity "m ex-vessel rieut ron (e) cover Gas Fool

• leakage e

N 111 Same as Above Same as Above S tee t / I r.conr i Pcactcr 53"

• Vessel Closure Head As wrbly til Ex-Vessel Neutron Leakage 14jor Source Gemetries; P4 C Annular Neutron 14" Reactor Cav 1 ty Shield Ring 111 Fadioactive cover gas Fajcr Scurce Gemetries; Penetration Shields gamma scurcesgGanera (a) Annular Gaps CPLH's (15) str caming (Na ); (b) In-wessel sodium EV1M - Nozile por t Neutron streaming Pool IV TM - Noz z le por t (c) Cover Gas Volume Ott.er s : Local Sh adc=

Above Soditm Pool Ul5 Jacks (4) Shields (d) Fencict Cavliy Liquid Level Ports (4) (steel) as Pisers (3) r equi r ed er a

• 3 w .CL CD ch N cT1 8 9

I l TABLE 12.1-47 ( SOLID RADIOACTIVE WASTE RADWASTE DRUM (SPENT RESINS) RADI0lSOTOPE INVENTORY ISOTOPE INVENTORY (CURIES) ISOTOPE INVENTORY (CURIES) Cr-51 4.2(-2) Ce-141 3.02(-1 ) Mn-54 3.02(-1) Ce-143 1.35(-1) Fe-59 2.06(-3) Pr-143 1.35(-1) Co-58 2.8(-1) Ce-144 2.15 (-1 ) Co-60 9.0(-2) Pr-144 2.15(-1) Sr-89 4.62(-1) Nd-147 6.51(-2) Sr-90 3.39(-1) Pm-147 1.22(-1) Y-90 3.39(-1) Pm-149 3.6(-3) Y-91 1.34(-1) Eu-155 1.58(-2) Zr-95 2.54(-1) Eu-156 6.0(-3) Nb-95 2.54(-1) Ta-182 1.33(-2) Mo-99 2.82(-2) Pu-238 2.8(-3) s Ru-103 3.36(-1) Pu-239 7.04(-4) Ru-106 2.72(-1) Pu-240 9.39(-4) Rh-106 2.72 (-1 ) Pu-241 8.28(-2) Ag-111 8.8(-3) Pu-242 1.76(-6) Te-127 9.25(-1) Np-238 5.87(-8) Te-127m 9.25(-1) Np-239 4.09(-4) Te-129m 2.79(0) Am-241 7.54(-4) Te-129 2.79(0) Am-242m 2.93(-5) Te-132 1.98(0) Am-242 1.35(-10) 1-131 1.3(-3) Am-243 1.17(-5) 1-132 1.99(0) Cm-242 1.43(-4) Cs-134 2.6(-3) Cm-243 5.86(-6) Cs-136 2.5(-3) Cm-244 1.64(-4) Cs-137 1.88(-2) Ba-140 1.83(-1) La-140 1.83(-1) TOTAL 16.52 U 12.1-78 Amend. 66 March 1982

O TABLE 12.1-48 HAS BEEN INTENTIONALLY DELETED O O (Next pag }s12.1-806) ,j982

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O O O 1 SYSTEM UHTROLLER IN CONTROL ROOM DETECTOR ASSEMBLY CHECK SOURCE CENTRAL.PROCESSIE m

                                   +

ACTUATOR UNIT /MiHI-COMPUTER thAtw GAMMA Sp

                >       DETECTOR ANWNUATM If                           I f CRT             KEY          LIE II                                                              BOARD         PRINTER M

y 0]ND DATA LINK _

  -                    ACCESSORIES I

[ & TO PLANT DATA

 , L                                                HANDLING AND          SYSTEM CONTROLLERIN HEALTH PHYSICS AREA l              V             y          y            DISPLAY SYSTEM 5

DISPLAY VISUAL AUDIBLE METER ALARM ALARM V V ER KEY LINE SUPPLIES CRT g REMOTE PROCESS STATION (RPS) Ry FIGURE 121-21 FLNCTIONAL BLOCK DIAGRAM 0F AN AREA RADIAll0N MONITOR r9 P e N

Im) v Sections 9.6.1 through 9.6.5 describe the ventilation systems for eacF building and the main control roun. The conceptual design for the RCB provides 14,000 cfm of outside air. This is adequate to meet the design objectives for radiation protection. The conceptual desigt flow rate to each of the 18 IHTS piping cells is 1000 cfm, which is suf ficient to meet the design objective for radiation protection and to satisfy personnel access requi rements . Other plant areas will be designed in accordance with conventional heating and ventilation requirements. Analysis of design requirements for other areas involving potential radioactive release will be undertaken and results incorporated, as necessary, in the heating and ventilation requirements for these areas. 12.2.3 Source Terms The sourges of radioactivity originate from the reactor cover gas leakage and HJ diffusion. The estimated radioactive leakages rates into normally accessible cells are presented in Table 12.2-1. The basis of the table is provided in Section 11.3. 12.2.4 Airborne Radioactivity Monitoring 12.2.4.1 Design Criteria

  /7                        Fixed and mobile continuous air monitors (CAM) will be em-V              ployed in conjunction with portable air sampling equipment to satisfy the requirements of CRBRP General Design Criteria 17 and 56 49l and the relevant sections of 10CFR20; and to verify that radioactive atmospheric contamination within the CRBRP remains normally "as low

49) as reasonably achievable".

The above radioactivity monitoring which is provided for the p' CRBRP reflects a design philosophy which identifies the following l levels of radiation protection (exclusive of the portable personnel nonitoring provisions described in Section 12.3).

1. Continuous monitoring (fixed) performed on i the ventilation which serves the Reactor Containment j

18, Building (RCB) and Reactor Service Building (RSB) oper- ! ating areas. Continuous monitoring is also performed to verify Control Room habitability.

2. Continuous monitoring (mobile) is performed in frequently occupied Nuclear Island operating areas adjacent to potential radioactivity sources. Frequently occupied areas include l

49 l radiation zone I and II (Figures 12.1-1 through 12.1-19:!) l ! cells which house numerous process system control panels. 1 Amend. 49 April 197f 12.2-3

3. Low-vol ume (Integrating) air sampl ing is perf ormed in Inf requently occupied operating areas within the Nuclear Island. Inf req uently occupied areas include radiatloa zone 11 and li t cells where routine tasks are perf ormed on a limited access basis.

4 High-vol ume grab sampi ing is perf ormed (with accompanying Counting Room analysis) prior to personnel entry into Zone IV radiation zones; and whenever a gross determination of short-lived airborne radioactivity in lower radiation zoned areas is desired. Fixed CAM's are provided as of fluent and process monitors (described in Section 11.4) at locations which could concelicably be subject to increases in radioactivity level s during various pl ant evolutions. The precess monitors are used to monitor the ventilation exhaust f rom a particular cell or group of cells. Upon detection of radioactivity above desired levels the radiation monitor will produce an alarm at the process system local panel (in addition to the Control Rom) and some monitors will initiate a signal to autmatically isolate the af fected area. The ef fl uent monitors perf orm surveil lance functions and provide (in the Control Room) Indication of an abnormal occurrence warranting Investigation by Health Physics personnel. Since the ef fluent monitors don't perform initiation of isolation the ranges have been selected to provide monitoring during normal and accident conditions. These monitors are incl uded in Table 11.4-1. Fixed CAMS, except those downstream of HEPA filters will withdraw the samples Isokinoctically in accordance with ANSI N13.1. In addition, the monitors will be located as close as practical to the sample point, and sample line bends are minimized to avoid plate out. Fixed CAM's are also provided to ensure adequate protection against contamination of the Control Room atmosphere due to airborne radioactivity following an accident condition. This monitoring arrangement is described in , Section 11.4. Fixed radiogas monitors (PPS) are also used to initiate Reactor Contai nment isolation as discussed in Section 7.3.1. Mobile CAM's will be provided in select locations throughout the CRBRP to perform the f ol lowing f unctions:

1. Continuously monitor the atmosphere at any specific location where maintenance is perfccmed.

2. Continuously monitor the atmosphere at any specific location where a process system f ail ure is suspected of causing airborne radioactivl 4 leakage.

3. Continuously monitor individual inerted celi purging activities as required by the Heating, Ventilating and Air Conditioning System.

4. Continuously monitor the RG atmosphere following containment i sol atIon, af ter connection to the post-accident contal nment sampi Ing penotrations discussed in Section 11.4.2.2.1.

5. Provide backup support to inoperative stationary airborne radioactive monitors.

O 12.2-3 a Amend. 72 Oct. 1982 l

O The mobile CAM's, will provide local audible and visual alarm Indication of V airborne radioactivity level s which exceed the monitor setpoint(s). and design' parareters of the various mobile airborne activity monitors are Locations l given in Tahid 12.2-3. High and low volume portable air samplers will be employed to obtain representative samples of breathing air at inf requently occupied operating areas of the CRBRP. Samples obtained will be analyzed in the Counting Room for gross activity and radioisotopic identification, as required. The portable air samplers will be supplieri as health physics equipment, and their f requency of use will be governed by the operational procedures of the CRBRP Heal th Physics Program. 12.2.4.2 Monitorino System Descriotion 12.2.4.2.'1 Continuous Air Monitors Continuous air monitors (CAM) are used to provide detection of radiogas, particulate, radiolodine and alpha (Pu) activity as indicated in Table 12.2-3. A combination of single and multichannel Instruments are used to perform the required monitor.Ing functions. The following is a description of each type of monitor provided: Gaseous Radioactivity Monitors , Each radiogas CAM continuously draws gas / air samples through a. particulate p d filter into a shielded 4-PI sample chamber where the gas is viewed by a beta detector, and then returns the gas / air back to the original source. A regulated vacuum pump is used to maintain desired flow rate through the ,, monitor. Samples withdrawn from process or ef fluent flow will be obtained isokinetically from the source stream. Each monitor consists of a radiogas detector, vacuum pump, microprocessor and accessories, local Indycator and al arms. The detector will have a minimum sensitivity of 3 x 10 #CI/cc f or Kr-85, at the 95% conf idence level . Each monitor cabinet will include local loss-of-signal, high and high-high radiation Indicator l ights, gas / air sample flowmeter and count-rate meter. Taps w il l be l provided to allow samples to be withdrawn for analysis in the Counting Rocur. , For stationary monitors, the , detection signal is continuously provided for display on redundant Radiation Monitoring System CRTs located in the Control Room and the Health Physics Area of the Plant Service Building, via their respective Central Processing Units and Mini-Computers (System Control lers). All control signals f rom monitore whicn are transmitted to interf acing systems will originate frort. F 1 Process l Stations which are part of the local monitor cabinet. E M arm signals are permanently recorded by the redundant Radiation Monitoring System Line Printers located in the Control Room and Health Physics Area. i l 12.2-4 Amend. 72 l Oct. 1982 y , 4 .- g-.--

lodine and Gaseous Radioactivity Monitors Radiolodino and radiogas CAM's provide two distinct detection channels within a single monitor housing. A regulated vacuum pump continuously draws a gas / air sample at a measured flow rate into the monitor assembly. The sampled gas / air flows through a fixed lodino f ilter, where a gamma detector observes radiolodine activitQhrough a discriminator window. The minimum radiolodino sinsitivity is 10 ACl/cc for I-131 at the 95% confIdonce ievol. From the lodino filter the air sample passes into a 4-Pi shielded chamber whogo a beta detector observes gaseous activity with a minimum sensitivity of 10 JLCI/cc for Kr-85 at the 95% confidence level. The gas / air sample is then exhausted to the original source. Each monitor contains the detectors, vacuum pump, microprocessor and accessories, and indicators. Display provisions at each monitor cabinet includo (common for each detection channol) loss-of-signal, high and high-high radiation Indicator Ilghts, and (separate for each detection channel) count-rate meters. A sample flow rate gauge is also provided. The detection signal is continuously provided for display on redundant Radiation Monitoring system CRTs located in the Control Room and the Health Physics Area of the Plant Servico Bellding, via their respective Central Processing Units and Mini-Computers (system Controllors). AlI Control signal s f rom monitors which are transmitted to interf acing systems will originato f rom Remote Process Stations which are part of the local monitor cabinet. The alarm signals are permanently recorded by the redundant Radiation Monitoring System Line Printers located in the Control Room and Health Physics Area. Particulate. lodine and Gaseous Radioactivity Monitors Particulate, radiolodine and radiogas CAM's provide three distinct detection channels within a single monitor housing. A regulated vacuum pump continuously draws a gas / air sample at a measured flow rate into the monitor assembly. If process or offluent flow is being monitored, the sample is obtained Isokinetically from the source stream. Particuiates are collected on a filter paper having an ef ficiency of 99.0% for 0.3 micr N0 particle sizes and viewed by a beta detector of minimum sensitivity of 10 A CI/cc for Cs-137 at the 95% confidence level, during an integrating time determined by sampl e f l ow rate. From the particulate filter, the sampled gas / air flows l through a flxed lodino filter, where a gamma detector observes radiolodino activity gough a discriminator window. The minimum radiolodino sensitivity l Is 4 x 10 /4Cl/cc f or 1-131 at the 95% conf Idonce ievel .

12. 2-4a Amend. 72 O

Oct. 1982

From the Iodine filter the air sample passes into a 4-PI shielded chamber where a beja detector observes gaseous activity with a minimum sensitivity of 3 x 10" /ACi/cc for Kr-85 at the 95% confidence level. The gas / air sample is then exhausted to the original source. Each monitor contains the detectors, vacuum pump, microprocesser and accessories, and Indicators. Display provisions at each monitor cabinet include (common for each detection channel) loss-of-signal, high and high-high radiation in.dicator lights, and (separate for each detection channel) count-rate Indicators. Mobile monitors are provided with a multipoint strip-chart recorder and audible and visual alarms for high and high-high radiation conditions. For stationary monitors, the detection signal is continuously provided for display on redundant Radiation Monitoring System CRTs located in the Control Room and the Health Physics Area of the Plant Service BulIding, via their respective Central Processing Units and Mini-Computers (System Controllers). AlI control signal s f rom monitors which are transmitted to interf acing systems will originate f rom Remote Process Stations which are part of the local monitor cabinet. The indicating analog meter in the Remote Process Station w111 Indicate counts per minute on a fIvo decade logarithmic scal e. The alarm signals are permanently recorded by the redundant Radiation Monitoring System Line Printers located in the Control Room and Heal th Physics Area. Gaseous In-Line Monitors Gaseous in-line monitors provided to monitor radioactivity in some process systems incl uding HVAC. The Monitor consists of a shielded section of pipe which is mounted by end flanges in the process line. A penetration through the pipe wall allows a beta scintillation detector to be placed in the process system flow. The detector wilI have a minimum sensitivity of 10~6 ,uCI/cc f or Kr-85, at the 95% conf idence level . Each monitor will have a local microprocessor wIth Iocal Indicator and al arms. The detection signal is continuously provided for display on redundant Radiation Monitoring System CRTs located in the Control Room and the Health Physics Area of the Plant Service Building, via their respective Central Processing Units and Mini-Computers (System Controllers). All control signals from monitors which are transmitted to Interf acing systems will originate f ran Remote Process Stations which are part of the local monitor cabinet. The alarm signal s are permanently recorded by the redundant Radiation Monitoring System Line Printers located in the Control Room and Health Physics Area. Aloha Radioactivity Monitors Each alpha CAM (mobilo units) provided will have the capability to dif ferentiate pl utonium al pha readings f rom the natural radon thoron al pha background through delayed detection techniques. Each al pha CAM continuously draws air samples into a shielded chamber where particulates 3 greater than 0.3 microns are deposited on a filter with an ef ficiency of l 99.0% and viewed by OC detector (s). A regulated vacuum pump will be used 12.2-4b Amend. 72 Oct. 1982

to maintain desired flow rate through the monitor arrangement, and return the air sample back to the original source. Each monitor contains the al pha detector (s), vacuum pump, microprocessor and accessories and indicators. The detector (s) wilI have a minimum sensitivity of 10 -12 ACl/cc f or Pu-239 at the 95% confidence level for a collection time of 8 hours. Display provisions at each monitor cabinet include loss-of-signal, loss-of-sampl e f l ow, high and high-high radiation Indicator lights, sample flow-meter, count-rate meter, strip-chart recorder and audibl e al arms f or high and high-high radiation conditions. These monitors shall have the capability to transmit data to the radiation monitoring consoles in the control room and health physics area when linked to the communication loop at the option of plant operators. Figures 12.2-1 and 12.2-2 show typical block diagrams of the containment exhaust (PPS) and typical fixed (non-PPS) continuous air radiation monitoring channels. The PPS radiogas monitors used f or Containment isolation dif fer f rom the radiogas CAM described previously in the following manner:

1. Each Class IE Monitor is individually wired to a dedicated Display and Control Unit (DCU) in the Control Room.

l 2. An analog output is provided by each monitor to the Plant Protection System (Containment Isolation System) Comparators, Logic and Safety Circuits. l 3. The buf fered output of each monitor is available for display on the Radiation Monitoring System CRTs and logging on Line Printers. AlI CAM components wil I be modul ar, commercial ly avalI able units designed f or rapid roplacoment upon faiiure. Electric components wilI be excl usively sol id-state, as avalI able, and power wilI be suppl led f rom the instrument AC busses (120V, 60Hz), with the exception of Class 1E monitors. These latter CAM's will receive Class IE power (120 Vac, 60Hz) frcm redundant vital Instrument AC busses. Certain design paraneters, as well as locations of the various airborne activity monitors are given in Table 12.2-3. 12.2.4.2.2 Portable Air Samolers Portable air samplers will be used to obtain representative samples of both long and short-lived airborne radioactive contaminants in operating areas of the plant. Their use and placement will be under the direction of the CRBRP site Heal th Physicist. Low Volume Samolqts Each sampling station consists of a regulated air pump and filter arrangement to deposit particulates greater than 0.3 microns in size, atid/or radiolodine, as required. The sampie fIow rate is sot IocalIy and recorded to enable an accurate determination of activity. The f ilterc will be collected af ter a suitable integrating titte interval, and brought to the Counting Room for analysis. The oniy Iocal output f rom the sampier unit is the pump flow signal. The complete pump and filter (s) arrangement are standard, commercially available units designed for ease of maintenance and interchangeability of components. 12.2-4c Amend. 72 Oct. 1982

High Volume Samolers

  ~]

J High voltane samplers will employ high speed air blowers to enable grab samples to be obtained in the 20-35 cfm range. Particulate and/or charcoal filters will be used for sample collection, and analysis in the Counting Room wil l be perf ormed. Thi s ty pe of sampl er w il l be used to determine the airborne radioactivity contribution due to shorter iIved Isotopes. 12.2.4.3 Maintenance and Calibration On completion of the monitoring system installation, each CAM will be checked for proper operation and cal Ibrated against a radiation check source (s) traceable back to the National Bureau of Standards or f rom an equally acceptabl e source. This initial calibration, and subsequent calibration at six month Intervals will verify the electronic operation of both local and Control Room ratemeters and also all annunciation points (loss-of-signal, high radiation, etc.). In addition, each monitor is supplied with a built-in check source to provide rapid functional tests at periodic intervals. 12.2.5 inhalation Doses Inhalation doses to plant personnel will be limited and controlled consistent with 10CFR20 requirements via the heating and ventilation system design. Resulting doses will be kept as low as practicable during operation and maintenance and exposures will be compatible with existing regulations O (10CFR20). U The expected annual inhalation doses to plant personnel in normally accessible cells can be determined f rom the leakage rates given in Table 12.2-1 and the design flow rates f or ventilation air in the Heat Access Area and Intermediate Sodium Piping cells. The concentration in these cells, for the expected leakage rates, is estimated by assuming that thero is a uniform concentration in the cell atmosphere and the ventil ation air stream. Thus, an equilibrium concentration will exist when the curie content discharged per day is equal to the leakage into the cell. The expected concentrations in the accessible cells are given in Table 12.2-4. The doses f rom the expected concentration can be estimated by assuming the ratio of the concentration to MPC occupational Iimits for each l Isotope present and multiplying this by 5 rem, the annual dose which would result fran exposure to the MPC for 40 hours per week for 50 weeks of the year, i O 12.2-5 Amend. 72 Oct. 1982

As shown in Table 12.2-4, the combined expected activity level for the isotopes present is about 0.01 MPC (occupational) In the Head Access Area. l Thus, the corresponding annual dose woul d be about 5mrom/ year. The release to the Intermediate Sodium Piping cells is tritium and the resulting equilibrium concentration is 0.0008 MPC. The resul ting expected j yearly dose would be about 4mrm/ year. Both of the above annual dose estimates are conservative since each assumes occupancy in the cells by an Individual of 40 hours per week for 50 weeks of the year. The expected occupancy is considerably less. The control room will be designed to assure continued occupancy during postul ated accident conditions. The expected radioactivity in the control rom during normal plant operations is background level. Additional discussion is provided in Section 12.1.5. O l l l l l l l l O 12.2-6 Amend. 72 Oct. 1982

i i 1 i TMBLE 12.2-3 LOCATi(N OF CONTINUOUS AIR M)NITORS 1 l LOCATION TYPE OF MONITOR AREA BLOG. ELEV. CELL HO. M)NITOR DES GlPTION BASl$ FOR LOCATION / FUNCTION REMARKS

RW B16 161A Particulate / Radio- Operating Floor Mobile monitor to provide monitoring See Figure 12.2-2, lodi ne/ Gaseous of work areas within containment See Sections 12.2.4.1

                                                                                                                                             & 12.2.4.2.1. This

) locations is the normal storage position of the mobile monitor.

R0B B16 161A Particul ate / Radio- Operating Floor

'                                                                                                     Mobile monitor to provide monitoring See Figure 12.1-2, lodi ne/ Gaseous                                  of work areas snd Inerted cells In     See Sections the containment                        12.2.4.1 &                    '

12.2.4.2.1. This , } location is the normal storage position of the mobile monitor. RG B16 16lA Alpha Operating Floor 3 Mobile monitor to provide monitoring See Figure 12.1-2, of work areas within containment See Sections 12.2.4.1 ] 3

                                                                                                                                            & 12.2.4.2.1. This
       ,                                                                                                                                    location is the e                                                                                                                                        normal storage position of the mobile monitor.

l RG 766 105M Particulate / Radio- Operating Floor Mobile monitor to provide monitoring See Figure 12.2-5, lodi ne/ Gaseous of work areas and inerted cells in See Sections l; the containment 12.2.4.1 & ! i a 12.2.4.2.1.. This location is the normal storage position of the 1 mobile monitoring. 1

         ?? bi

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na . L WO  ! 4 00 sa } DO DO I 4

T ABLE 12.2-3 (Cont 'd ) LOCATION TWE Of MON I TOR AR E A OLDG. ELEV. ULL tJ0. My48 TOR DESCRIPTION DASIS FOR LOCATION / FUNCTION PEMARKS RSS 779 3070 Par t i cu l a t e/ Rad i o- Oper ating Floor Mub l i e moni t or to provide moni toring See Figure 13.1-11, iodi ne/Gase ous of local work areas and post-accident See Sections moni tor ing of containment a tmospher e 12.2.4.1, 12.2.4.2.1

                                                                                                                                          & 11.4.2.2.1. This location is the nor m al s t or age position of the mob i l e moni t or .

RSB 816 308A Airha Oper a t i ng F l oor Mob i l e moni t or to pr ov i de moni t or i ng See F igur e 12.1-9, of l oca l work areas and post-accident See Sections monitor ing of containment atmosphere 12.2.4.1, 12.2.4.2.1,

                                                                                                                                          & 11.4.2.2.1.         This location Is the nor mal s t or age position of the es                                                                                                                                         mob i l e moni t or ,

to QJ RM3 816 308B Par t i cu l a te/ Rad i o- Oper a t i ng F l oor Mob i l e moni t or to pr ov i de moni t or i ng See F igur e 12.1-9, sa l od i ree/ Gaseous of local wor k areas See Sections 12.2.4.1 c) & 12.2.4.2.1. This location is the normal st or age position of the mobi l e moni t or. SGB-lB 816 262 Particulate / Radio- Oper a t i ng F l oor Mobi l e moni t or to pr ov i de moni t or i ng See Figure l odi ne/ Gaseous of SGB-IB local work areas and post- 12.1-19a. See acci den t moni tor ing of containment Sections 12.2.4.1, a tmos pher e12.2.4. 2.1

                                                                                                                                          & 11.4.2.2.1.         This location is the normal stor age position of the mobi l e moni t or .

h! no

= 3 pa .CL 00 ~4 DO hJ O                                                                          O                                                                                  e _-

s TABLE 12.2-3 (Cont'd) LOCATION TYPE OF MONITOR AREA BLDG. EL EV . CELL NO. MONITOR DEsotiPTION BASIS FOR LOCATION / FUNCTION RLMARKS W 816 431 Particulate / Radio- Operating Floor Mobi l e moni tor to provide monitor Ing See Section t odine/ Gaseous of control rom and local work areas 12.2.4.1 & 12.2.4.2.1. This location is the nor mal st or age position of the mobile moni tor. -l N i . N e I w l O C' I i ON nm

• 3 C

H.L CD N NN 1 4 i

TABLE 12.2-4 EXPECTED

• ANNUAL EXPOSURE IN NORMALLY ACCESSIBLE CELLS Head Access Area Expected Concentration MPC+ Expected Isotope (u Ci/ml) (pCi/ml) Concentration : 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-9 5.0 E-06 7.0 E-4 49 6.4 E-ll 2.0 E-06 41 3.2 E-5 H3 9.2 E-15 5.0 E-06 1.8 E-9 491 TOTAL 5.84 E-9 0.0015 Intermediate Sodium Piping Cells 49l H3 4.0 E-09 5.0 E-06 7.9 E-04

    +MPC = MPC for Restricted Areas

• Failed Fuel Fraction = 0.1 percent at 1 year operation Amend. 49 April 1979

DISPLAV AND CONTROL UNIT IN COW 1ROL ROOM Dt:rCOR ASSHELY a m , g(DCU) DETECTOR AND MGONEN SAMPLE CHAMBER AND

                                                                                                                          >   CHECK SOURCE                                                       ACCESSORIES I

Remote Process Station (RPS) I I l l FLOW FLOW DISPLAY VISUAL AUDIBLE k ALARM BUFFER ELEMENT METER ETER ALANS H f3 PUMP

                                                     ."                                                                         FLOW ALARM CONTROL y                                                                             SWITCH          PUMP CLASS IE w

w .m fTO SYSTD4 CONTROLLER To Plant JN MEWH N5KS M Protection NON-CLASS IE MICROPROCESSOR y System Y r AND CEH1RAL PROCESSING ACCESSORIES UNIT AND Mt(RD. >TO M AlH M y -> COMPUTER CONTROL PANFL NN ORS Class 1E Non-class T V To Plant Data k k Handling and DISPLAY VISUAL AUDIBLE Display System I "E METER METER METER CR1 BOARD PRIM 1ER RE Fea {NN , a SYS1EM (ORTROLLER IN LOW 1ROL ROOH FIGURE 122-1 PPS CONTAlHMENT EXilAtisT RADIATION M0HITORING OlANNEl.(.CLASSIE) G G e

O O O DL1ECTOR ASSEMBLY DETECTDR AMD SAMPLC CHAMBER CHEC,K SOURCE + > 10 SYSTEM CONTROLLER IN HEALTH PHYSICS AREA FMER MICR0 PROLE 550R m

                                                        '                                                  CENTRAL PROCEsSLNG UNil           m SUPPUES                 AND A61D MICRO-(OMPti1ER            loLAiN
                    >      KCE550RIE S                                   y To et Auf DAT A                                                 (DNTROL mMEL ltANDUNG AND y                                                                         DtSPLAY SYSTEM ANNUNCIATORS L

prsetAy I vr50AL AUDIBLE METER ALARMS ALARMS REMOTE PROCESS STATION

      -   -      -      ~        -

CRT KEY BOUtD LINE PRINTER FLOW _ m FLOW ELEMENT [ METER J V 4 FLOW ALARM m MP SW11CH CONTROL. R h' F sv5iEm coNin0ttEn in coNin0t R0oM E FIGURE 122-2 NON PPS AIR RADIAT10H MONITOR lHG CHANNEL

TM3tE 12.3-5 FIRSONNEL FROTECTION K)NITOR - AREA K)NITORS LOCAT10N AREA AND/OR ETER OF' ERAT 10NAL 0ASI$ PRO SS MONITOR RANGE BACEROUW MONITOR FOR BLOG. ELEV. ELL K)N ITORED TYPE aNhr (mR/hr) OUTFUT *

• LOCATION
• RG 824' 1&C Cubicle 3 162 Direct Gama 0.01-10 0.2 A 1.

RG 824' 163 14C Cubicle 7 Direct Gamma 0.01-10 0.2 A 1,5 I RG 824' 164 I&C Cubicle Direct G ama 0.01-10 0.2 A 1.,4,3 ] 780' 4 I RG 105U Primary Pil Operating Direct G ama 0.1-10 2.0 A 1.,2 Area RG 766' 105S Operating Floor Direct G ama 0.1-10 4 2.0 A t.,2.,5 RG 780' 4 16tG Operating Floor Direct Gama 0.1-10 2.0 A 1. 2. ! RG 794' 152 Operating Floor 4 Direct Gamma 0.1-10 2.0 A 1.,2. [ - RG 752' 105H Operating Floor Direct Gamma 0.1-10 4 2.0 A 1. 2. ! N PG 766' 105Q Operating Floor Direct G a ma 0.1-10 4 2.0 A 1.,2 4 W b w RG 733' 105A Operating Floor Direct G ama 0.1-10 4 2.0 A 1.,2,5 3 RSD 842'6" 311 Ref uel. Ccam. Center Direct Gamma 0.01-10 0.2 B 1.,3 RG 802' 151 Head Access Area Direct G ame 0.1-107 25.0 A 1.,2.,4 I 816' 308A Operating Floor 3 RSB Direct G ama 0.01-10 0.2 B 1.,2.,3,6 RSB 816' 308A Operating Floor erect Gamma 0.01-10 3 0.2 0 1.,3,6 4 TWB 816' 643 Decontamination Bay Direct G ama 0.1-10 2.0 B 1.,2 RG 794' 105V Operating Floor Direct G ama 0.1-10 4 2.0 A 1.,2,5 779' 4 RSD 307A Ex-Vessel SSP Operating Direct Gamme 0.1-10 2.0 A 1. l Area RG 752' 105K Operating Area Direct G ama 0.1-10 4 2.0 A 1. 4 4 o RSS 755' 306A Ex-Vessel FTl Operating Direct Gama 0.1-10 2.0 A 1. Q Area

                -3
                   ,a 500     836'   27'   SGB(IB) Remote Shutdown Direct Gama         0.01-10 7   Unrestricted      A        1.

j e Panels Area (See NOTE 2) co N NN

TMILE 12.3-5 (Cont'd) LOCATION AREA AND/OR PETER OIT R AT ION AL PASIS PROQ SS FU41 TOR RAtr>E O A& GROUND HON 1IOR FOR B LOG. EL EV . GLL PodlTORED TYPE mR/br (mR/ br ) OUTFtT LOCATION' RW 733' 10$f Make-up Pump Val ve Ofrect G ema 0.1-10 2.0 A 1,5 Oper ating Gallery RW 733' 105D Operating Area Olrect Gamma 0.1-10 # 2.0 A 1. RW 766' 105M Primary SSP Operating Olrect G mma 0.1-10' 2.0 A 1.,2. Area FM) 795' 605C f ALL Distillate Ot reet Gamma 0.1-10 # 2.0 A 1 Storage Tank Area FWD 795' 620 Fil ter Handl ing Rom Direct G mma 0.1-10 2.0 A 1 RSB 733' 3050 Operating Areas Direct G ema 0.1-10 2.0 A 1 RSO 779' 307A Operating Floor Direct Gamma 0.1-10 # 2.0 0 1.,2 8 0 RSO 781' 341 Tuel Handling Cell Direct G mma 0.01-10 2.0x10 0 3. RSO 779' 339A FHC Operating Gallery Direct G ema 0.01-10 0.2 8 1.,2. 1 A RT 749' 336 Spent f uel Cask Corridor Ofrect G ama 0.01-10 0 5.0x10 8 3. and Shaft RT 755' 306AA Operating Areas Direct G mea 0.1-10 # 2.0 A 1.,2. RT 733' 335 SFSC Service Station 01 rect G mma 0.1-10 10.0 0 1. 2.,3 Equirment SGB 816' 26 2 Operating Areas Ot ract Gmme 0.1-10 7 Unr est r i ct ed A t.,4 (See NOTE 2) SGB 794' 253 Emerg. Airlock / Analysis Direct Gema 0.1-10 Unrest r ict ed A 1.,4,6 Operating Area (See NOTE 2) W 816' 431 Control Rom Direct Gamma 0.1-10 7 Unrest r icted A 1. (See NOTE 2) Oe2r 3 PSB 816' 146 Cmbi ned L ab Direct Gmma 0.01-10 3 Unr est r icted A 1. (See NOTE 2) c.

 ~                                                                                                         y C    .%3      775'    605A                          l ALL Distillate Storage  Ofrect G mma      0.01-10          2.0                              A      1.
 $$                                                 Tank Area                                                                                                        ' * "

l O O e

                                                                                                                  \

TM3LE 12.3-5 (Cont'd) LOCATION AREA AND/OR E TER OPERAT IONAL B ASIS FRO &SS O ITOR RANGE BAGGROUND MONITOR FOR BLDG. ELEY. GLL W ITORED TYPE mR/hr (mR/hr) OUTPUTu LOCATION' RG 816' 16tA Equi teent/ Personnel Direct Gamma 0.01-10 3 0.2 A 1. Airlock Area RG 816' 169A RG Annulus Direct Gama 100 -10 7 0.2 A 4 RW 816' 169A Rm Arnutus Direct G ama 10 -10 0.2 A 4 Rm 816' 169A RG Annulus Direct Gamma 10'-107 0.2 A 4 4 RG 794' 161E Primary Pump Drive Direct G ama 10~I-IO 2.0 A 5 RW 794' 161D Primary Pump Drive Direct Gama 10-I-104 2.0 A 5 RG 794' 161C Primary Pump Drive Direct G ama 10~I-104 2.0 A 5 RG 766' 105Y Val we Operating Gallery Direct Gamma 10-I-104 2.0 A 5 RG 733' 111 Stal rwel l Direct Gamma IO'I-104 2.0 A 5 ~ RW 733' 105E Access Area Direct Gamma 10'I-104 10.0 A 5 ro w RW 8258 106 Pofar Crane Operating Direct G ama 10-I-104 0.2 A 5 RW 842' 165 El&C Cubicle Direct G ama 10'I-10 4 0.2 A 5

                                                                        -I      4 RW      842'   167   El&C Cubicle               Direct G ama      10 -10         0.2           A     5 SGB     794'   247   Power Distrib. Panel       Direct Ganes      10'I-10 4   Unr estricted    A     5,6 Area SGB     794'   271   Operating Area             Direct Gamma      10'I-10 4   Unrest ricted    A     5,6 SGB     794'   271   Operating Area             Direct Gama       10'I-10 4   Unrestricted     A     5,6
                                                                        -I      #

Ta 794' 26 2 Operating Area Direct Ganma 10 -10 Unrestricted A 5,6 4 SGB 794' 262 Operating Area Direct G ama 10-I-10 Unrestricted A 5,6 gg SGB 794' 211A Valve Gallery Direct Gamma 10'I-10 4 5xt0 2 A 6 4 4 I$ct SGD 794' 248 tilts Pipe Chase Direct Ganma 10'I-10 Ix10 A 6 $' SGB 794' 251 IHTS Pipe O ase ' Direct Gamma 10-I-10 4 1x10 A 6 500 794' 252 IHTS Pipe Chase Direct G ama 10'I-10 4 lxt0 4 A 6

TN3t E 12.3-3 (Cont'd) LOCATION AREA AND/OR FETER OFIRAT ION AL D AS IS PRORSS MDNITOR RANGE BAO< GROUND MONITOR FOR BtDG. ELEV. GLt F0NITORED TYPE mR/hr (mR/hr) OUTPUT ** LOCATION' RSB 785' 348 Cont. Cleanup Scrubber Direct G ema 10'I-10 0.2 A 6 RSO 785' 349 Cont. Cleanup & HV AC Direct G mma 10'I-10 4 0.2 A 6 Duct Rm 840' 332 FulX 3rd Loop Cell Direct Gmma 10'I-IO 4 0.2 A 5 RSD 864' 395A Annulus Filter Direct Gmma 10'I-10 # 0.2 A 6 4 RSD 733' 350 NAP Storago Vessel Cell Direct G mma 10'I-10 2.0 A 6 RSD 733' 305M Access Area Direct Gmma 10'I-10 2.0 A 6 RSD 733' 305C RSO/SGB Passageway Direct G mma 10'I-10 2.0 A 6 743' 4 2

-  RSO                     311    SDD 82, 85 & 94 Area              Direct Gmma       10'I-IO        1x10             A       6 N

4 RSD 797' 314 SDD 23 Instr u. Area Direct G ama 10'I-10 0.2 A 5 4 Z cr RSB 755' 359 Cont. Cleanup Fil ter Direct G ama 10'I-10 0.2 A 6 Celi RSB 779' 376 RAPS Pipe Gallery Direct G mma 10'I-10 4 5x10 3 A 6 4 R50 775' 3511 EV S Cool Ing Pi pow ay Direct G ema 10'I-10 2x10 A 6 LEGifQ 'f AS I S FOR LOCAT ION f10NITOR OUTPUT RG - Reactor Contalianent Oldg. 1. Provide personnel protection in A. Local and Control Rom: Loss RSD - Reactor Service Oldg. traf ficked area, of signal Indicator light, high SGB - Stem Generator Bldg. 2. Monitor adjecent high radio- level radiation alarm, high-03 - Control B ldg. activity area. high level radiation alarm, PSD - Pl ant Ser v ice B l dg. 3. Monitor refueling operations, exposure meter (mR/hr). RWB - Radweste Area (Bay) 4 High level reactor containment B. Local, Control Room and Ref uel-radiation monitor ( Accident Ing Communication Center: Moni tor ) . (same as above).

5. Monitor areas conntaining saf ety-

@g rel ated equipment ( Accident nm Hon i tor ) .

• 3 6. Monitor areas with hatches or

~O penetrations f rom contalnment y ( Acci dent Moni t or ) . NN NQIES: Unrestr i ct ed: Def ined by 10 CFR 20, Paragraph 20.105. Background specif ied in table is maximum design background value daring operation, based on Na-24 gamma fleid. O O e

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

15.1.4 Effect of Design Changes on Analyses of Accident Events The design of the CRBRP has made significant progress since the consequences of design basis events reported in the renalnder of this chapter were f irst analyzed. A review of approved design changes to determine which may af fect the reported.results and a qualitative evaluation of the ef fecis of these changes has been made. A primary example is the change in core design f rom a homogeneous to a heterogeneous conf igurction. The results of this ef fort are discussed in the following sections. 15.1.4.1 Reactivity insertion Design Events Section 15.2 covers the analyses of reactivity insertion design events. The f ormat progresses f rom anticipated up through f aulted design transients with each accident scenario providing: i o identification of causes and accident description; I o analysis of ef fects and consequences; o conclusions. With regard to accident scenarios, there have been no changes to Section 15.2 since the original PSAR submittal . However, various pieces of design data have changed and have subsequently been incorporated into the appropriate design sections of the PS AR. Modif ications to the nuclear and thermal-hydraulic information af fect the maximum temperatures attained and the O temperature / time traces shown. The purpose of this section is to indicate the ef fect of these various changes to the Section 15.2 results. Reactivity insertion accidents typically result in overpower transients that are characterized by an increase in power such that a proportionately larger increase occurs in f uel temperature than in cladding temperature. This is , opposed to undercool ing design events which have a very small fuel temperature increase as compared to that of the cladding. Worst case overpower conditions commonly have a rapid increase in power which institutes scram of 1he Plant Protection System (PPS). For events having a rapid power burst, the period of the overpower conditions is typically less than one second (see Figure 15.2.3.3-3, f or exampl e) . Although the shutdown occurs quickly, ef fects such as f uel mel ting and the potential for f ool/ cladding interaction are of prime , importance in the f uel pin perf crnance eval uations. To demonstrate temperatures tha+ envelope overpower events with current data appl led, a worst case event was reanalyzed and the results are herein described rel ative to the f ormer val ues. This worst case selected previously was the Seismic Reactivity insertion (SSE) (see Section 15.2.3.3.1 ) with primary control system shutdown (which is an extremely unlike:y event). O

                          ~

15.1-105 Amend. 72 Oct. 1982 _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ . _ _ _ . _._ ___.m.._ _. _

As with the past analyses, the f ollow ing conservative assumptions were made:

1) All f ul l power cases are f or the reactor operating at thermal hydraul ic design conditions w ith a power generation of 975 iMt at three-locp operation. (Power uncertainties are discussed in Section 4.4.3.2.)

2) Since the smallest Doppler coef f icient occurs at the beginning-of-eq u i l i br i um cycl e, the transient reactor power cal cul ation was made f or this particul ar phase in core I if e. This results in the highest possible reactor power changes being calcul ated. For overpower transients, the smal lest Doppler coef f icient of al l core cycles is used (see Section 4.T.2.3) and this val ue is reduced 30%

to account f or 36 uncertainties.

3) The highest cladding and f uel temperature hot rod occurs at the beginning of the first cycle of operation (in F/A #52 and 101).

The conservative reactor power cal cul ation f rom item 2 above was l appl led to this particular rod. With burnup, the power generation l and steady-state temperatures decrease (flows are constant) in the hottest fuel assembl ics, and consequently, the temperatures, due to the transients, woul d decrease af ter beginning of cycle.

4) As described in Chapter 7.0 and Section 4.2, the maximum al lowabl e time del ays f or PPS Iogic and electrical / mechanical del ays have been conservatively enveloped by using a 200 millisecond delay between the instrument channel output going beyond the trip level and the start of control rod i nsertion. *

5) Three sigma (3a) hot channel f actors were used f or al i the analyses and the cladding temperatures shown are the inner surf ace of the hot pin cl adding at the highest temperature position, both axially and circumferentially on the f uel rods. (Position is under the w Ire wrap. )

6) The most rapid fIow decay af ter de-energiz ing the primary pumps was used. (See Figure 5.3-22. )

7) Maximum decay heats were used f or the hot rods considering 3 y uncertainties.

Results f rom FORE-2M analysis are given in Figures 15.1.4-1, 2 and 3 and Tabl e , 15.1.4-1 for a 60d step reactivity insortion occruing at the worst time during I the SSE (see Section 15.2.3.3.1). Comparisons of the heterogeneous core resul ts are made w ith data f or a homogeneous ccre previously reported in th is section. This previously reported data updated earl ier data for the homogeneous core analyzed in Section 15.2.3.3. The figures show the Win this instance the senscr del ay has been encompassed by the 200 msec PPS IogIc and control rod uniatch delay. Th i s I s j ust i f led by ihe sme; i magn fiude of the fl ux sensor del ay which is estimated at less then 10 msecs. O 15.1-106 Amend. 72 Oct. 1982

15.3.1 Anticloated Events 15.3.1.1 l_oss of off-site Electrical Power 15.3.1.1.1 Identification of Causes and Accident Descriotion The of f-site power supply to the 13.8 KV buses is avail able f rom the generating switchyards and the reserve switchyard both of which are powered by outside sources as described in Chapter 8.0. Hence, the postul ated loss of power woul d resul t oniy from simultaneous, multipl e f all ures. The loss of all of f-site power trips all primary and intermediate sodium pumps, commencing a flow coastdown. It also initiates starting of the emergency diesel generators. Action of the Plant Protection System (PPS) trips the control rods thus I imiting core over temperatures f rcen reduced flow. Either emergency diesel provides power to the primary and intermediate sodium pump pony motors and SG# IRS Auxil iary Feedwater Pumps f or decay heat removal. Additionally, a third power supply (250 VDC Diverse Battery and Inverter) provides power to the third loop pony motors. To provide conservatism in the analysis, the most rapid core flow coastdown was assumed by using the minimum pump rotating kinetic energy and the maximum primary system flow resistance specified in the design. The action of the Primary and Secondary Shutdown Systems (SDS) are as follows:

a. Primary trip - Loss of electrical power trip occurring in 0.5 seconds.

The 0.5 second del ay incl udes measurement and trip f unction I ags. O These I ags incl ude bus voltage decay and instrument del ay but not the RSS logic and control rod unlatching del ays.

b. Secondary trip - FI ux-Total FIow trip occurring 2 seconds. af ter ioss of electrleal pumping power. Thi s I ag incl udes time f or the f I ow to coastdown as welI as the measurement Iags.

15.3.1.1.2 Analvsis of Effects and Conseauences s The loss of of f-site electrical power event was analyzed with the DEMO computer code. The overall results of the analysis are summarized in Figures 15.3.1.1 -1 and 15.3.1.1 -2. As shown, the Primary PPS loss of electrical power trip I imits the maximum core hot spot temperature to 1410 F. In the event the primary shutdown system does not operate, Figure 15.3.1.1-1 shows that the secondary shutdown system I imits the worst case cl ad hot spot temperature to 1630 F. While the transient temperature exceeds the design basis emergency transieht envelope temperature by 30 F, the time above the normal operating temperature is only 6 second as compared to 150 seconds f or a 15.3-6 Amend. 72 Oct. 1982

the design basis transient (see Figure 15.3.1.1 -3 ) . Consequently, the cladding cemage due to the transient is less than that due to the design basis transient for which, as shown in Section 4.2, cladding Integrity limits are satIsfled. The capabil ity of the CDF procedure to conservatively predict the results of Fuel Clad Transient Test (FCTT) is demonstrated below. The range of the FCTT temperatures and fluences considered exceeded the data base of the FURFAN CDF computer code. Despite this, the C0F analyses conservatively gredicted the test results with peak ciadding tempegtures n excess of 1900 F, and ciadding fIuence exposures in excess of 3 x 10 n/cm The quantitative critoria in terms of Temperature versus Time for transient events which do not af fect cladding integrity is shown in Figure 4.2-31. The shape of the mergency transient considered in this plot envelopes the loss of of f-site electrical power with scram by the secondary PPS event. The minimum cladding i if etime is determined by the intersection of the peak transient cladding temperature versus time curve and the transient limit curve with maximum design temperatures and maximum uncertainty in properties. Note that the maximum peak cladding temperature occurs at beginning-of-lif e, and the cladding temperature increment due to the transient is assumed constant throughout Iife. Thus, for an emergency transient wIth a maximum peak cladding temperature of 1630 F, the peak clad temperature versus time curve would l ie parallel to and 30 F above the peak clad temperature versus time curve shown in Figure 15.3.1.1-4. The intersection of this curve wIth the minimum transient I imit curve gives a cl adding i ifetime of 450 days or 35 days less than the 1600 F peak cladding temperature transient. In all calculations involved in generating Figure 15.3.1.1-4, cumulative cladding damage is continuously accounted for in the cladding property considerations. It should be noted that the anticipated time temperature curve for the loss of off-site electrical power is considerably less than the time envelope used to develope the transient I imit curves. Theref ore, the above loss of 35 days due to the additional 30 F is belleved to be an overestimate of the transients actual of fect. This not withstanding, the design iIfetime based on the above analysis for the loss of of f site power is still in excess of the 411 day goal Iifetime. As discussed earl ler, the most real Istically severe combination of possibilities allowed in the design specifications were selected to analyze this event. Figure 15.3.1.1-2 shows the ef fects of a possible longer flow l coastdown, enhanced secondary control rod dynamics, and using " minimum j required" instead of " expected" primary control rod shutdown rates. Lower I possible core flow resistances and higher pump rotating kinetic energies, f decrease the core hot spot temperature 10 F for a primary PPS trip and 15 F f or the secondary shutdown system trip. Additionally, increasing the Initial secondary control rod insertion rate to match the primary rates decreases the cl ad temperature 35 F f or the secondary trip.  ; l l 1 15.3-7 Amend. 72 Oct. 1982

Figure 15.3.1.7-1.a Temperatures of Pertinent Parameters as a Function of Time Af ter inadvertent Actuation of the Water / Steam Side of the Sodium / Water Reaction Pressure Rollef System. 100o { soo E I h too -

                  =

o  ! I

> - 1 E

O E g soo - O W S w

soo -

e I e teso me TitIE (SECONDS) O 15.3-28 Amend. 72 Oct. 1982

O 000 800 - C b g 700 - 4 5 0

          $ 600     -

E 8 O 500 - 400 0 1000 2000 TIME (SECONDS) Figure 15.3.I.7-18. Temperature of Core Inlet as a Function of Time After inadvertant Actuation of the Water / Steam Side of the Sodium / Water Reactor Pressure Relief System 7203-17 Amend. 72 15.3-28a Oct. 1982

1500 1400 - 1300 - 1200 - l C g 1100 -

     ?

! E E 1000 - 3 900 - 800 O l 700 - N 600 - 500 0 1000 2000 TIME (SECONOS) Figure 15.3.1.7-1C Temperature of Ilot Spot as a Function of Time Alter inadsertant Actuation of tiie Water / Steam Side ofIlie Sodium > Water Reactor Pressure Relief System 7203-18 Amend. 72 15.3-28b Oct. 1982

15.3.2.3 Small Water-to-Sodium Leaks in Steam Generator Tubes 15.3.2.3.1 Identification of Causes and Accident Descriotion The probabil ity of a tube leak in the steam generators is expected to be quite small as a resul t of caref ul design supported by development and testing of the steam generators. However, the Steam Generator Leak Detection System, described in Section 7.5.5, has been provided to allow operator action to lImit the consaquences of a smalI leak in a steam generator tube. I The water-to-sodium leak detection system is designed to alert the ogerator to the existence of very small leaks, as small as approximately 2 x 10 lb. water /sec. For Inigal very small leaks which can be realistically expected (up to about 5 x 10 lb. water /sec.), the reactor wil l be shut down normally followed by a controlled cooldown and depressurization of the af fected steam generator. The af fected lHTS loop would then be drained to allow repair of  : the steam generator. However in the uniIkely event of a smalI leak exceeding approximately 5 x 10-5 Ib. water /sec, the operator may elect to scram the reactor and Isolate and blowdown all three steam generator modules in the af fected loop. The operator would also drain the af fected lHTS loop, resulting in flow stoppage in that loop. 15.3.2.3.2 Analysis of Effects and Conseauences n It is assumed that the reactor is operating at rated conditions when a leak occurs in a steam generator of such a nature that the operator elects to manually shutdown the reactor, Isolate and depressurize the water side of the af fected loop, and drain the sodium side of that loop. Dynamic analyses have not been completed for this event; however, the primary system response can be conservatively bounded by assuming that alI heat removal capability is Instantaneously lost in the IHX of the af fected Ioop at the time when Intermediate flow stops. The IHX primary outlet temperature increases rapidly to the primary Inlet temperature. Core flow rate and the resulting f uel cladding and core coolant exit temperatures are Identical to those for normal scram until the hot sodium from the af fected lHX reaches the core. This is calculated to occur about 60 seconds af ter reactor scram, assuming a normal fl ow coast down. The hot sodlum from the af fected loop mixes with the sodium f ran the other two loops in the reactor vessel inlet plenum. Assuming perfect mixing in the core inlet region, the core inlet temperature increases about 90 F. If this increase in core temperatures 60 seconds af ter scram is conservatively added to the hot-channel coolant exit temperature for the normal scram, the hot-channel ci ad temperature woul d increase f rom about O 1 O 1 15.3-34 Amend. 72 Oct. 1982

810*F to 900'F. This increase in temperature would be somewhat larger if incomplete mixing occurs in the reactor vessel inlet plenum. However, even for the extreme assumption of zero mixing, the hot-channel coolant temperature could increase a maximum of only 265 F to about 1125 F, still well below the normal steady-state operating value. The core exit temperatures would then decrease as the reactor was cooled by the operable loops. If it is assumed that this event occurs following operation with the maximum undetected intermediate-to-primary sodium leak rate there will be insignificant radiological release. Leakage of primary sodium into the IHTS is prevented by pressurizing the IHTS such that a pressure differential across the IHX (intermediate-to-primary) of at least 10 psi exists during plant operation. This pressure differential could be lost during the sodium dumping process and it is possible that primary sodium could enter the IHTS. Leak rates of approximately 6 gph will be detected during normal operation (Section 7.5.5) and therefore only small amount of primary sodium could , be introduced into the IHTS during the pump coast down. This small amount of primary sodium would mix with the intermediate sodium and either remain in the non drainable sections of the IHTS, steam generators, and IHX, or be drained to the sodium dump tank. Over pressurization of this tank is prevented by either the equalization line or the pressure relief valve, the gases vented through this system will be the inert gas displaced by the sodium entering the dump tank. No sodium will be released in this process and the radiological consequences of this event cre insignificant. 22 15.3.2.3.3 Conclusion Core temperatures following a steam generator tube leak are well within the normal operating temperature range for the fuel and core. Residual heat removal is provided by the operable loops. This event is included in the overall plant duty cycle list tnat provides the basis for the thermal transient design conditions for the reactor and the main heat transport system. O 15.3-35 Amend. 22 June 1976

i O N. VE Av SM ALL INITI AL w 0 LE AK SLOW 2

                 /                                          '.10 -2 g/s 4 L2 10 Iibs/ sect STEP 1    0.110" 4.28cmt                                Leekage oroceolv plugs, or is no negner

• prior to steo 5 betow.

f

                  /P C     STEAM REACTIONIEROSION EROSION GEGINS BRIE 8 INTE AMITTENT '_E AKS
                                       /
                                       /                   Proceoiv plugged for tone periods STEP 2         //
                                           /         /

OEVELOPMENT OP L A AGE CR ATER Leenage patn may open for longer pereoas STEP 3 // O ^ CRATE A NE A AS STE AM $10E

                   /                                        Leamage continuous but varisoie.

still no negner then step 1. STEP 4 ELAPSED TIME FROM STEP 1: HOURS. DAYS TO MONTHS 015" to.38cml A AP'O E AOSiON AT INNE A WALL Rao o iperesse in leakage to 15 gis STEP 5 e (3 a10' IWisci

                             +           &n ELAPSED TIME FROM STEP 4: ONE MINUTE OR LESS (SUPERHEATER CONDITION $1 Figure 15.3.3.3-1. Development of a Large Leak From a Small Steam Leak in 21/4Cr-1Mo Tubing Exposed to Sodium O                                                                                           Amend. 72 15.3-49                                      Oct. 1982

AMENDMENT 72 List of Responses to NRC Questions Received Since the Fall of 1981 and Located Chronologically in PSAR Volumes 25 and 26 QCS220.25 QCS760.13 QCS421.9 QCS760.28 QCS421.17 QCS760.30 QCS421.22 QCS760.36 QCS421.27 QCS760.105 QCS421.30 QCS760.110 QCS421.31 QCS760.116 QCS421.34 QCS760.131 QCS421.36 QCS760.166 QCS421.37 QCS760.172 QCS421.42 QCS760.175 QCS421.47 QCS760.176 QCS421.48 QCS760.177 QCS421.58- QCS760.178 QCS721.1 O l t l Qi l

a / Question 220.25 (3.8.2.2) 7-On page 3.8-1, it is stated that ASME Section 111 Division 1,1974 Edition with Addenda through winter 1974 and ASNE Section lli Division 2,1974 Edition will be used f or the design of the steel contelr. ment and the steel-lined concrete containment f oundation mat, respectively. Indicate what will be the ef fect on the design if the latest editions of the ASME Section t il Divisions 1 and 2 incl uding Code Case N-284 (1980) are used. Resoonsg: The PSAR design was perf ormed to the requirements of the 1974 Code edition specified in the design specifications. The specif ic criteris related to buckl ing are described in the PSAR Appendix 3.8-A. The intent of these criteria is similar to the Code Case N-280 criteria, in that these address buckl ing modes based on classical analysis, provide capacity reduction f actors, f actors of saf ety, and interaction equations f or buckl ing. A significant reanalysis would 'oe required to demonstrate that the Conteinment Vessel meets the requir,ements of the new Code and Code Case N-284. However, the appl icant has compared the PS AR to the ASNE code and N-284. The significant dif ferences have been evaluated and are provided as parts 6.0 and 7.0 of th is response., , , addition the appl Icant has perf ccmed an analysis of the critical buckt ing region just above the operating floor using N-284 criteria and the appropriate l oads. This analysis demonstrated that the design nargir exceeds the p requirements of Code Case N-284. See part 7.0, V i (

                                     \

l t l l l . I O) l QCS220.25-1 Amend. /2 , Oct. 1982 l

                         ,y,,.     - - , ,           +-       -        -     m,w n              -w-   ,v -

r

i INDEX CRBRP CONTAlteENT VESSEL COP @ARlSON AND EVALUATION OF PSAR Am 1980 EDlTlONS OF l ASME BAPV CODE /CCDE CASE N-284 Page No. 1.0 SU MMAR Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.0 - COMPAR ISON OF SECTION l l 1, StBSECT ION NCA . . . . . 4

                .1   NCA-3000 Responsibil ities and Duties .....                                           5 3.0 COMPARISON OF SECTION Ii1, DivIS10N 1 SUB SECT ION NE, CL ASS MC QMPONENTS . . . . . . . . . . . .                               9
                .1   NE-1000      Gener al Req ui rements . . . . . . . . . . . .                         10
                .2   NE-2000 . Nate r i a l s . . . . . . . . . . . . . . . . . . . . . . .               11
                .3 NE-3000       . Design ..........................                                      16 4 NE-3000       'uist of NE-3000 Subarticles Not
                     ,            Ev a l ua te d . . . . . . . . . . . . . . . . . . . . . . .            37
                .5   NE-4000      Fabr i cati on & I nstal l ation. . . . . . .                           38
                .6   NE-5000      Examination .....................                                       45
                .7   NE-6000      Testing .........................                                       47
                .8   NE-7000      Prctection Against Overpressure .                                       52
                .9   NE-0000       Namepl etes, Stemping, & Reports .                                     53 4.0 COMPARISON OF SECTION lli, OlVISION 2 FOUNDATION MAT AND B OTTOM L INER . . . . . . . . . . . . . . .                           54
                 .1  CC-2000       Raterial ........................                                      55
                 .2  CC-3000       Design ..........................                                      57
                 .3  CC-4000       Fabrication and Construction ....                                      59
                 .5  CC-5000       Construction Testing and Examination .....................                                      60 5.0 COMPARISON OF CODE CASE N-284, BUCKLlhG CRITERI A                                            61 6.0 EV ALUATION OF KEY DIFFERENCES 1980 ASME CODE ..                                             71 7.0 EV ALUATION OF KEY DIFFERENCES CODE CASE N-284 ....................................                                          79 O

Q CS220.25-2 Amend. 72 Oct. 1982

 ,        1.0   

SUMMARY

s 1 \- / This section summarizes the dif ferences in the containment design requirements between the PSAR and the 1980 Edition (including Winter 1981 Addenda) of the i ASME B&PV Code Section ill, Division 1 and the 1975 Edition of the ASME B&PV i Code Section lil, Division 2 and the 1980 Edition (including Winter 1981 ' Addenda) of the ASME BAPV Code Section lil, Division 2. Specif ic dif ferences, both design and documentation, are noted in subsequent sections of this attachment. Key dif ferences in the containment design requirements are identif ied bel ow: 1.1 ASME BAPV Code Section Ill. Subsection NCA

1. No impact on the design of the coniainment. Specif ic documentation changes are Identif ied in Section 2.0 of Attachment II.

1.2 ASME BAPV Code Section Ill. Division 1

1. Buckling criteria are added to the design by analysis criterie giving general rules f or buckl Ing conditions not covered by the design by f ormul a criteria.

2. Service Level limits are introduced. By including accident condi-tions under Level s A and B, the Code now requires the eval uation of primary plus secondary stress Intensity range f or accident conditions and provides an allowable stress l Imit. Since this requirement was not in the 1974 Code, calculations of secondary stresses in dif ferent

( areas woul d now be needed.

3. Additional requirements are imposed on nozzles, by changing the classification of stresses at the nozzle piping transition and also rules governing opening and reinf orcement.

4. Additional requirements are imposed on the spacing of areas of primary local manbrane stress intensity at brackets.

1.3 ASME B&PV Code Section Ill. Division 2 ( Af fectine the Foundation M 411

1. The concrete strain corresponding to the maximum allowable primary plus secondary membrane and bending compressive stress of 0.85 f'c has been reduced.

1.4 CONCLUSION

These dif ferences have been reviewed by the appl Icant and compared to the criteria in the PSAR. None of these dif ferences require a change in the present design in the opinion of the appl Icant. The detailed logic f or this conclusion is provided in Section 6.0 of the document. (J~) QCS220.25-3 Amend. 72 Oct. 1982

2.0 COMPARISON OF SECTION lli. SUBSECTION NCA

     .1  NCA-3000  Responsibitifles and Duties 2.1   

SUMMARY

Changes in this section have no impact on the design. O O QCS 220.25-4 Amend. 72 Oct. 1982

t

      'N CONTAlteENT VESSFt - ASK BAPV CODE COWARISON SECTION lil - ARTICLE NCA-3000 RESFONSIBILITIES AND DUTIES PARAGRAPH /TITtE                74 EDITION - W74 ADDE @ A                         80 EDITION - W81 ADDE @A                I FACT ON DESIGN

1. NCA-3125 No provisions f or subcontracted services. Services covered by Section 181 may or.ly be None.

Subcontracted subcontracted to appropriate certif Icate Serv ices holders. An N Type Certif Icate Holder may subcontract to another organization the Survey-ing and auditing f unctions, but must retain the responsibilities for these activities and qual if Ications

2. NCA-3130 Provides that construction includes The term " construction" is no longer def ined. None.

Welding and material s, design, fabrication, examina-Subcontract- tion, testing, inspection and certif ica-Ing durIng tion Constr uct ion

3. NCA-3131 Provides conditions that must be met for Adds the exception f or f urnace braz ing operations None.

Q Wel ding the perf ormance of welding f or shop or as specifled in NCA-3561 (c), NCA-3661 (b), anti m during Con- field work during Code construction. NCA-3761 (b) N struction y O m 4. NCA-3220 Owners responsibilities included eight (8) Owners responsibil ities f or Division 1 increased None, m Categorles of spect f Ic 1tems, to 19 specific items of which 10 are new

  $           the Owners ResponsibilI-responsibilItles. All but 4 can be assigned to the Owner Designee ties

5. Deleted ON sn

= ~3 c3. CO N NN

                               -.                                                     .-                                                            -~

CONTAlfMNT VES set _ - ASME B APV CODE CCNPARISON SECT ION li t - ARTICLE t.CA-3000 PESPONSib lL ITIES AND DUTIES PAPERAPH/T I TI F 74 EDITION - W74 ADDENDA 80 EDITION - WB1 ADDENDA _ IMPACT ON DESfGN

6. NCA-3240 Provices that the Owner shall be respons- The Owner has the additional requirment to None.

Prowlslon of Ible f or adequate structural support and determire allowable bearing pressure or load Aaequate def Inition of boundary interfaces per caisson or pile and f urnish same to designer. Supporting Str uct ure

7. NCA-3252 Provices 8 items to be ir.cl uded in Design Adds 2 additional items which include: None.

Contents of Specification. 1-specify ing operating requirements of a Design component Speci f ication 2-specif y ing ef f ective Code Edition & Addenda

                                                                               & Code Case

8. NCA-3256 (a ) Design Specif ications shall be made avail- Provides that in addition, the applicable None.

Filing of able to the inspector at the construction data for parts, piping assemblies and appurten-Desigt Spec- si te f or all but parts and piping ances shall be made available to the Inspector e ification assembl ies at the fabrication site. m

$      9. NCA-32/0 Owners Data 1-Owner or designee shall prepare Form N-3        An(r or designee shall prepare Form PH3.

2-Sner shall certif y, by signir.g the f orm The other requirments have been deleted None.

]
.          Report and        that each organization was a Hol der of N          Filing            Certif Icate of Authorization T

3-form N-3 shall be f iled in accordance wlth NA-8430

10. NCA-3280 No Requirement Owner shalI be responsible f or designating None.

A ner 's and maintaining records. Responsibli-ity tw Records o> n9 a ca ~ NN

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       \

s CONTAlteENT YESSEt - ASE B&PV CODF COWARISON SECTION lli - ARTICLE NCA-3000 RESPONSIBILITIES AND DUTIES l, PARAGRAPH / TITLE 74 EDITION - W74 ADDEf0A 80 EDITION - W81 ADDEf0A t rACT ON DESIGN 1

11. NCA-3500 Manuf acturer's Responsibil itles Manuf acturer has been changed to read N None.

Responsibil- Certif icate Holder. Some changes in the Ity of an N def Inition wording have been made. Certificate Holder - Division 1

12. NCA-3520 Manuf acturer is charged with nine (9) Certificate holder charged with 15 times of None.

Categories of specific responsibilities responsibilitles that includes 6 additional. the N Certifi-cate Holders ResponsibilI-ties a 13. NCA-3551 This paragraph has been rewritten in its None. Q Design entirety without significant alteration of ro Documents Its content

o. 14. NCA-3561 (c) No provision f or brazing operations per- Permits the N Certificate holder to sub- None.

    @          Scope of       formed by organizations not holding            contract f urnace braz ing to non-Certif icate a         Responsibil- certificate of authorization.                    Holders.

N Ity

15. NCA-3620 HPT Certif Icate Holders responsibit itles None.

Categories of have been added (16 Items) and includes the NPT Cert- scope of the NPT Certificate Hciders ifIcate Hold- Responsibil ity for Quality Assurance. ers Responsi-bilitles kk rao. G' R1M

CONTAINMENT VESSEE - ASME BiPV CODE CCMPARISON SECT ION lli - ARTid.E NCA-3000 RESMNSIBILITIES AND CUTIES PARAGRAPH /TITtF _ 74 EDITION - W74 ADDENDA 80 E0fTION - W81 ADDENDA _IMoACT ON DESIGN

16. FCA-3700 Installers responsibil ities incl udes six NA Certif icate Holders responsibil ities None.

Responsibil- (6) specif ic items incl udes 12 specif ic itms, six (6) of Ities which are additional of an NA Certificate Holder

17. NA-3500 These two Subarticles have been deleted None.

Responsibil- f rom the later code Editions Itles of Inspection Agencies, in-spection Spe-cial Ists and inspector s c O 18. NA-3600 " " " " " " " " " y Engi neer I ng o Organ iz at I on's g Responsibi!- p itles Co

19. NCA-3800 NA-3700 contains prov isIons f or a simii ar NCA-3800 is similar to NA-3700 None.

Metaliic Ma- Quality Systems progra terial Manu-f acturer's and Material Sup-pl ler's Q ual ity System Progrm OF 20. NCA-3900 Not included in the earl ier Code New Qual ity Assurance Provision f or non- None, c+ R$ Nonmetallic Editions metallic material s which is similar to { Material NCA-3800 but not as comprehensive .-.- Manuf acturer's $y and Constituent NN Suppl lers Qual-Ity System Program

      #                                                                                               G                                                          e

i 3.0 COMPAR I SON OF SECT I ON I I I . D l V i S 10N 1 SIESECTION NE. CLASS MC COMONENTS

                                               .1  NE-1000              General Requirernents
                                               .2  NE-2000              Material s
                                               .3  NE-3000              Design 4 NE-3000               List of NE-3000 Subarticles Not Evaluated
                                               .5 NE-5000               Exam ination
                                               .6 NE-6000               Testing
                                               .7 NE-7000               Protection Against Overpressure i

O l l l l l O QCS220.25-9 Amend. 72 Oct. 1932

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

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

O J

                                                                                          ]                                                                 0
               \

i CONTAINE NT VESSFt - ASME B&PV CODE COW /RISON SECTION li t - ARTi(1E NE-1000 GENERAL REQUIREENTS 74 EDITION - W74 ADDEtOA 80 EDITION - WB1 ADDEPOA fWACT ON DESIGN PAPE RAPH/ TITLE , None.

1. tE-1132 Jurisdiction of Subsection NE M ail conform Boundaries of as f ollows:

the Contain- The Design Specif ication shall define the bound-ment System it is intended that the jurisdiction of I this Subsection f or the containment vessel ary of a containment vessel to which piping or j shall conf orm to the prowlslons of (a) and another component is attached. The boundary shall (b ) bel ow, not be closer to the containment vessel thans 1 (a) Connections f or the attachments by (a) the f irst circumferential joint in welded welding of piping or of penetration connections (the connecting weld shall be assembl ies (NA-1262) shall terminate considered part of the piping); i at a circumferential joint exclusive of the connecting weld Iocated at (b) The f ace of the fIrst fIange In bot ted least the greater of the distances connections (the bolts shall be considered normal to the surf ace of the vessel part of the piping); c n (Fig. NE-1132-1) as given in (1) and (c) the f irst threaded JcInt In screwed connections , N (2) below: P (1) the l imits of reinf orcement given N in NE-3334; or -j Y ( 2) the boundary of the connection as j *-* given in the Design Specification j g j (NA-3251) and included in the Manu-f acturer's test (NE-6000) and cert-IfIcation ( E-8000). (b) Connections f or the attachment of locks or hatches shall inclLde all required doors, covers, or other attachmebts required f or the containment f unction. Piping, pumps, c3 2:= S j$ or valves attached to the locks or hatches

         *3                                    shall be classified in the Design SpecIt la l

e

            .'                                 cation (NE-1131),

Co N NN I

CONTAINMENT VESSEt - ASME B APV CODE COMPARISON SECTION 111 - ARTICLE NE-2000 MATERI ALS PAPE RAPH/ TITLE 74 EDITION - W74 ADDENDA 80 EDITION - W81 ADDENDA IMPACT ON DESIGN

1. NE-2110 Scope Thickness def initions added None, of Pr i r.c i pl e Terms Employed

2. NE-2121 Material which need not be tested per NE- 1. List of items not covered by this article None.

(D) & (C) 23 20 expanded. Perm itted Material 2. Permits pressure-retaining material of fer-Specification rltic steel to be quenched and tempered.

3. Corrects ref erence of material which need not be tested to NE-2311.

3. NE-2180 None Requires temperature sur veyed and cal ibrated None.

Procedures furnaces. Q v1 for Heat Treating of N Mater ial s y O k 4. NE-2190 (b ) None Provides f or the repair welding of structural None. (.n Nonpressure- steel rol Ied shapes. 1 Retaining

            ~         Material

5. NE-2223.1 Prov ides test specimen orientation f or Cunbines into one paragraph requirments f or test None.

Location of forgings of 2" maximum thickness and specimen orientation for forging under 2" and Coupons f orgings greater than 2" maximum thick- forgings over 2". ness.

6. NE-2224 Provides Test specimen orientation f cr bars Combires into one paragraph requirments f or test None.

        @y            location of        of 2" maximum thicknass and bars greater        specimen or ientation f or bars of al l th ickne sse s, et g          Coupons            than 2" maximum thickness.

E.

         - -     7. NE-2225               Sane as above                                  Sane as above                                                 None.

8y Tubuler Prod-NN ucts and fittings

            #                                                                     9                                                                          e

                                                          ..          - .. . - . -           . ~                    .-     - -.        -.-.-__.. - .
       \,                                                                        d l                                                        CONTAlteENT VESSEt - AEK BAPV CODE CQWARISQN l

SECTION lil - ARTIQ.E NE-2000 MATERI ALS PARAGRAPH /TITiF 74 EDITION - W74 ADDEPOA BD EDITlON - W81 ADDENDA IWACT ON DESIGN

8. NE-2311 Provides a listing of materials not 1. Provides def Initions of thicknesses f or mat- None.

Material for required to be impact tested eriale, not required to be impact tested. which litpact Testing is 2. Provides Table NE-2311(a)-1 which IIst material Req uired exempt trcn impact testing based on TNDT and lowest service temperature and where LST exceeds 150 F.

3. Requires design specif ication to state LST.

4 Provides exemptions f or drop weight test.

9. NE-23 21.1 Restricts drop weight tests to 5/8" thick No restrictions identif ied None.

Drop Weight and greater and where Charpy V notch test-o Test s Ing is not successf ul, drop weight testing O may not be used as an alternate to the N Charpy V - notch test where the heat-af fected zone of the crack starter weld '

     ?                       Is tougher than the base metal I

N Y 10. NE-2322.1 Location of test specimens shal'l be as impact specimens shall be as f ar f rom the None. [ Location of specif ied in NE-2220 or material specif i- material surf ace as is specif ied f or tenslie Test Speci- cation. The number of test specimens specimens in the material specif ications. For mens shali be per NE-2340. bolting, the impact specimen shalI be Iocated , at 1/2 radius or 1" below surf ace. Fracture plane shall be one diameter f rom the heat , treated end.

11. E-2322.2 Added requirement that drop weight specimens may None, o> Orientation be criented in any direction.

S@ of Test g Specimen

 ~.

c 12. NE-2322.3 & Preparation of test specimens and impact Not specified in this section None.

 $"m           E-2322. 4     test temperature.

CCNTAlhMEf!T VESSEt_ - ASME 0 8.PV CODE COWARISQ SE CT ION 111 - ARTICLE NE-2000 MATERI ALS LASKFfftj/JfTLE 74 E0lTION - w74 A"DE?rA a0 EDITION - w91 AfDE*CA _ IMPACT ON DESIGN

13. NE-2330 Test Rtquirments enganded consicerably to incl ude (a)

Pq u i r erent s Charpy V-notch testing at or telcw the Lowest and Accept- Serv ice kbtal tem;erature. (b) Drop melght test-ance Stand- Ing at (LST-TNDT) (c) Charpy V-notch testing at ards or bel cw l ow est specif ied vessel test tenperature when hydrostatic or pnetrtatic test temperature is i cw or than LST. (d) Charpy V-noth testing f or Nil s Lateral Expansion v al ues (e) Charpy V-notch testing fcr absorbed energy val ues ( f) Drop =eight testing f(r two no-t;reaA specimens. Tests are specif led according to material thickness and other Cr1terIa. a

                       ] 14 NE-2351                      Retest f or Ch ar py V-notch test require        tainitrum requirernents is changed to average three            None.

N Retests fcr that the resul ts meet minimum require- of T abl e NE-2332.1, NE-2332.1-2 or NE-2333.1-1 as N g Mate r i al ments. hhere a test result is tel(w, it appl icabl e. Al so the test resul t bel ow the average other than shall not te l ower th an 10 '-I ts. tel(w. may not te f cwer than 5'-its telcw.

                      $i bolting

[ 15. hE-2552 Retest s f or hot specified Requirments f or retest have teen added None. Dot ting hiertal

16. NE-2420(c) Definitions of tots of covered, fl ux cored, or None.

                              & (f) Re-                                                                   f abricated et tetroce and carbon or icw al ley quired                                                                       steel barrcd electrodes have been expanded con-Tests                                                                       si derabl e p  17 NE-2431.(b)                                                                   1. Provices f cr deposition of meld metal by the

{ if c, & (c) Gene r al Test Ruq u i r e-electrosl ag process and P ments 2. Post mel d heat treat of the electrosl ag None. j process aci d deposi ts NN 9 9 e

s . . - - . .. . - __~ _ _ . - - +-. . _ _ _ . . - - . -- -_ .. - .- - . _ - .- - . . _ . .

                         ~                                                                             r                                                                            \

k

                                                                                                                                                                                -M CONTAINMENT VESSFt - ASME REPV CODE COMPARISON SECTION lil - ARTIQ.E NE-2000 MATERI ALS PARACP>PH/TlT1E                  74 EDITlON - W74 ADDEPOA                                80 EDITlON - WB1 ADDEPOA                  _..lMPACT ON DESIGN

18. NE-2432.1(a) Requires chanical analysis f or A-No. B Requires in addition chemical analysis for any None.

                                  &(c) Test        welding material used with GTM and                     other. welding material used w ith any GTM and Muthods           PAW processes and any other welding                    PM processes.                                                                         ,

i material used with any GMAW process ,

19. NE-2433 Delta ferrite determination shall be 1. Delta ferrite determination shall be per- None.

9 Delta perfctmed by comparison of chemical formed by comparison of chemical analysis to ! Ferrite analysis to Schaef f er diagram. Schaef fler diagram or a Determ ina-tion 2. By magnetic determinatin by Instrument calI-brated to MS-A4.2-74

20. NE-2550 Seamless and wel ded (withc,ut f Iller 1. Wrought seamless and welded (without f iller None.

Exam i na- metal) tubular products and fittings metall pipe and tubing shall comply with re-O tion of shall meet requirements of SA-249, quirments of Class 2 components and SA-655 O N Wrought SA-312, SA-333, and SA-334 pl us one g S w less of the f olicaving weld inspections (a) 2. Similar fittings shall comply with require- ! o-c ;'eided n Ultrasonic, (b) eddy current (c) radio- ments of Class 2 components and SA-652 ! g ( w i th out graphs or (d) magnetic particle or i m Filler Metal) lIquid penetrant. I Tubular Prod-LA ucts and fittings

21. PE-2560 All welds shall be examined radiographic- in addition to material requirements, pipe shall None.

Exam ination ally In addition to material requirements meet SA-655 and f ittings shall meet SA-652 1 and Repair of in accordance w ith NE-5110 l Tubul ar Prod-i (cts and Fittings i ap welded w ith 1 0 5 Filler j f@ct Metal s

        $~                   22. NE-2570           Cast material shall be examined by either              Cast products shell meet all the requirements of              Hone.

CD N Examination radiographic or ul trasonic methods to the SA-613

  '     NN and Repair       acceptance standards of SA-609 of Statical-ly and Centri-f ugally cast products M

i 1 l 1 CONTAINMEffT VESEEL - AEME DiPV CODE CCYPARISON SECTION lil - ARTiaE NE-2000 MATERI ALS i PARACFAPH/ TITLE 74 EDITION - W74 ADDENDA 80 EDITION - h81 ADDEf0A IMPACT ON DESIGN

23. NE-2580 Bol ts, studs and nuts shall be examined in Bolts, studs and nuts shall meet the require- None.

{ Examination accordance w ith requirernents of material rnents of SA-614 of Bol ts, specification j studs and .j Nuts i i .l 1 l, I ) O o Ln N N 1 o

!    N i

I W I W i i 1 O> .I na r+ (D 3 l CL

   ~ .

co N NN i l 4 O O O

J U o b CONTAf tetENT VESSEL - COMPARISON OF 1974 EDITION AfD 198D EDITION Or ASE BAPV CODE SECTICN Ill - DIVISION 1 ARTICLE NE-3000 DESIGN PAP 16PEH/TITl_ E 74 EDITION - W74 ADDEPOA 80 EDITION - W81 ADDE POA 1MPACT ON DE$1GN

1. NE-3111 Changes some load def initions editor tally None.

Loading and adds a f ew ty pical loadings to be Conditions considered in a vessel design.

2. NE-3112 Design Conditions New

Title:

" Design Loadings" None. Design loads redefined

3. NE-3112.1 i) Deletes the f ollowing note: " Maximum contaln- None.

Design ment internal pressure is Intended to include Pressure a margin above the maximum calculated peak internal pressure under which conditions the o containment f unction is required..." n N N ll) " Stability of the vessel shell Deletes the words " Internal concrete". Hence, None. may be provided by Internal shell stability can be provided by internal

o. concrete structures bearing or external structures.

  $:                            directly against the shell."
  "cn                     III) Requirements related to design              Daletes the requirements in NE-3112.1,              None.

pressure are: "The design pres- Similar requirements are given in NE-3131. sure shalI be used in the design formulas of NE-3300 for thickness of components when pressure is the only substantial load and also in the computations made to show compi lance w Ith the stress-Intensity limits of NE-3221, a3 E-3227.1, NE-3227.2, NE-3227.4, n9 NE-3228 and NE-3231." c+ co - a c w.L CO N NN

CONTAINMEffT VESSEL - COMPARISON OF 1974 EDITION AND 1980 EDITION OF ASME B1PV CODE SECT ION lli - Div ISION 1 ARTICll LE-3000 DESIGN fyRACRAPH/T I TLE 74 EDITION - W74 ADDEf0A 80 EDITION - W81 ADDENDA IMPACT ON DESIGN

4. NE-3112. 2 Hydrostatic loads coincicent w ith Design Pressure None.

Design are designated as Design Muchanical Loads Mechanical Loads S a. NE-3112.4 Mai n prov i s ion s ar e: Design a) Use allowable stress-intensity for Revised provisions are stated in a more compre- See items 9 Al l ow abl e design fr m Table 1-10.0, bensive way, as bel ow : through 27. Stress b) Criteria on the maximum allowable v al ues compressive stresses, "The rules f or allowable stresses are given in c) Stress in the wall of a vessel shoul d NE-3200 f or vessel s designed by analy sis where be less than the maximum allowable the allavable stress-Intensity Smc is the Sm stress val ue at the temperature val ue given in Table 1-10.0 of Appendix 1. The a allocat.'e stress S to be used in the equations of n NE-3300 shall be those l isted in Tabl e I-10.0 of

  $N Appendix  l."

O Sb. NE-3112.4 Sm is Design Stress intensity Codes use Smc or Sm1 instead of Sm. None. N NE-3134.6 T ( Nov Code) Smc f rm Tabl e 1-10.0. Usett f or primary q stresses. Sml f rm Tabl e 1-10.0. Used f cr primary pl us secondary stresses. o> O E: et O 3 s-* .CL C CC N NN

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A h I U ( d CONTAINMENT VESSEL _ - COMPARISON OF 1974 EDITION AND 1980 EDITION OF ASME B&PV CODE SECTION lli - DIVISION 1 ARTICLE NE-3000 DESIGN PARAGRAPH /TfTtE 74 EDf TlON - W74 ADDEWA 80 EDITlON - W81 ADDENDA _ f PFACT ON DESICN

6. NE-3113 Requirements are specif ied f or nozzles The requirments f or nozzles are deleted f rom See evaluation in Normal attached to Class 1 piping during Normal this section. The subsection has a new subject 6.0 for item 14 Oper at ing Operating Conditions. " Service Limits". It specifles four different Conditions Service Limits.

f or Nozz les The Code considered only Operating and (1974) Design Conditions. See Table 1 for comparison of requirements. Service L imits OBE was wIthin Design or Operating (1980) Conditions. SSE was within Design Conditions with higher allowable stress intensity. WA was within Design Conditions. O O N

7. NE-3131 General requirements includes design by More emphasis on NE-3200 (Design by Analysis) See evaluation in y General analysis and by formula and deals with than on NE-3300 (Design by Formula). 6.0 for item 21 o Requirements load classification and stress intensity For vessels subjected to compressive stresses or 1imits. Incl udes requirements f or con- external loads, code places emphasis on the new u, figuration having compressive stresses rules NE-3222 (Buckling Stress values). When 1

cn (NE-3133), jet Impingement effects, other pressure loads, appl Icable, NE-3133 may be used, k F "! N

C0f(TAltMNT VESSEL _ - COMPARISON OF 1974 EDITION A?D 1980 E0lTION OF AST P1PV Cf0f( SECTION lit - OlViSION 1 /RTIQ_E NE-3000 DES IGN PARAGPIPH/ TITLE 74 EDITION - W74 ADDEt0A 80 EDITION - W81 A00E POA IMPACT ON DESIGN

8. NE-3133 The rules have been revised. They are more See eval uation Ccrgonent s extensive. More specif ic rul es are prov ided in 6.0 Under f or the ell ipsoidal and the for! spherical heads.

External Rules are al so prov ided f or the cylinders = lth Loading the diameter to the thickness ratio less than 10.

9. NE-3200 Design by Anal y si s Rev i sed Operat ion Service None.

Terms Operational Cycle Service Cycle g Operating Conditions Service Loadings un N 10. NE-3211 il Follow ing requirements are deleted f rom NE-3211: o General Roq ui rment s "The design detail s shat I conf ccm to the rules None, ui f cr Accept- given in NE-3100 (General Design)." ab il ity 5 "In case of conf I Ict between NE-3200 and None. NE-3300, the requirments of hE-3300 shall govern wton considerir,g pressure alone." ll) Requires that the critical buckling Foi t ow ing requirment is added in NE-3211: See eval uation in stress shat I te taken into account 6.0 for i tem 21. f or the conf igurations where ccan- "The buckling stress shall be consicered in pressive stresses occur. accordance w Ith NE-3222. Od Does not g h en speci f ic Specifles new rules f or the buckl ing consicera-r+ ct] saf ety f actors. tion of vessel, w* C Co N NN O O e

3 O O [CNTAINMENT VESSEt_ - COWARISON OF 1974 EDf TION AND 1980 EDITION OF ASK B&PV CODE SECTION Ili - Div ISION 1 ARTICLE NE-3000 DESIGN PARAGRAPH / TITLE 74 EDITION - W74 ADDEPOA 80 EDITION - W81 ADDEPOA I WACT ON DESIGN

11. NE-3213.10 i) Rule for Local Primary Mumbrance Rule is modif led and is stated below: None.

Local Stress is as below: Primary - "A stressed region may be considered local if Numbrane "A stressed region may be considered the distance over which the memt>rane stress Stress local if the distance over which the intensity exceeds 1.1 Smc does not extend in > stress intensity exceed 1.1 Sm does the meridional direction more than 1.0 Rt, not extend in the meridional direc- where R is the minimum midsurf ace radius cf tion more than 0.5 Rt and if it h curvature and t is the minimum thickness in not cicser in the morldlonal direc- the region considered. Regions of local tion than 2.5 Rt to another reofon primary stress Intensity invo!vina where the limits of general primary - axisymmetric membrane stress distributions membrane stress are exceeded, where which exceed 1.1 Smc shall not be closer R is the mean radius of the vessel in the meridional direction than 2.5 Rt. and t is the wall thickness." where R is defined as (R g + R )/2 and t is 2 defined as (t + t )/2, where t and t are Q m j 2 g 2 the minimum inicknesses at each of the regions N y considered, and R, and Ry are the minimum mid-o surf ace radil of curvature at these regions where the membrane stress Intensity exceeds

.n 1.1 Smc."

t N O II) The fcilowing new requirement is added: See evaluation in 6.0

                                                                                    " Discrete regions of local primary membrane stress intensity, such as those resulting f rom concentrated loads acting on brackets, where the membrane stress Intensity exceeds 1.1 Sac shall be spaced so that there is no overlapping of the areas in which the membrane stress intensity oy                                                                                 exceeds I.1 Smc."

oa r+ (D a 3 , CL b-*- CD N NN

CRITAINMENT VESSEL - CO@lSON OF 1974 EDITION AND 1980 EDITf 0N OF ASME PiPV CODE SECTION 111 - Div ISIGN 1 ARTICLE NE-3000 DESIGN 1 74 EDITION - W74 ADDENDA 80 EDITION - WB1 ADDENDA IMPACT ON DESIGN l EtRAGI'ffH/ T I Tt E

12. Tabl e I) Table NE-3217-1 provides cl assi f Ica- Table is mcre extensive f or nozzles. Cl assi f I- See eval uation hE-3217 -1 tion of stress intensity f or nozz l es tion cf stress Intensity is proviced at the in 6.0 fcr item 14.

Cl ass i f i ca- at the f ollowing locations: f ol l ow i ng l ocat ion of ncz z les: l of Stress Intensity o Cross-sect ion perpendicul ar to o Within the limits of reinf orcement For Sune nozzles axis Typical o Outside the l imits of reinf cccement Cases o Noz z l e Wal l o Noz z le Wal l l Details are discussed in item ho. 24.

                                  !!)                                              Additional note to the Teble, as below:                   None.

l "If the bending mment at the edge is required n to maintain the bending stress in the middle to acceptable i imits, the edge bending is classif ied N as P . Otherw i se, it is classified as Q." l b l O l tJ1 e N w 1 1

 - o                                                                                                                                                                      ,

3 ' CL W

• C C; N NN

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b O CONTAllMNT VESSEf. - COW ARISON OF 1974 EDITION AW 1980 EDITION OF ASE BAPV nr SECTION 111 - DIVISION 1 ARTIQ.E NE-3000 DESIGN PARAGRAPH /TITt_E 74 EDITION - W74 ADDFWA 80 EDITION - WB1 ADOF W A frACT ON DESIGN

13. NE-3220 " Stress Limits f or Other Than Bolts" " Stress Intensity and Buckling Stress Values None.

Title f or Other Than Bol ts" Changed

14. Revised Stress Categories and Limits of Stress "Oneratina conditions" changed to " Level s A See evaluation Title for intensity for oneratina conditions and B Service Limits"; and for Level C Service in 6.0 Figure Limits where the structure is not integral and NE-3222-1 continuous."

(Old Code), Figure Accidental pressure and temperatures are NE-3221 -2 wIthin Level A. Primary plus secondary

(New Code) stresses Intensity range must be evaluated, while before only primary stresses were o considered.

O

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O 6 Y' N Rir P

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CONTAINMENT VESSEL - COMPARISON OF 1974 EDITION APO 1980 EDITION OF ASPE B1PV CYOE SECTION lil - DIV ISION l ARTICLE NE-3000 DESIGN PARAGRAPH /TITIF 74 EDITION - W74 ADDEtOA 80 EDITION - WB1 ADDEPOA IMPACT ON DEslGN 15a. Added New Figures NE-3221-3 and 3221-4 added as None. Figures descrit ed below: Figure NE-3221-3 " Stress Categories and Limits of Stress Intensity for Level C Service Limits where the structure is Integral and Continuous, and f or Level D Service L imits where the structure is not integral and continuous and at Partial Penetration Welds." Figure NE-3221-4 " Stress Categories None. and Limits of Stress Intensity for o Level D Service Limits where the Q ro structure is integral and Continuous."

   @    15b. Added New Tabl e Table NE-3221-1 presents stanmory of       See evaluation in Stress Intensity Limits on Design          6.0 fcr item 14
   $                                                                     Linits and Level s A, B, C and D and 4

w Service Limits

16. PE-3221 Gives allowable stress Intensities Gives stress intensity values for See evaluation in for " Design Conditions." " Design Conditions" and Service 6.0 for item 14 Limits A through D. Tabl e NE-3221-1 provides summary of stress Intensity l im it s. See New Flgures NE-3221-1 to NE-3221-4, to cover al l conditions, imposes limits on primary membrane plus See evaluation a3 primary bending stress Intensity when in 6.0 m E3 the local primary membrane stress I$ Intensity exceed 0.67 of yleld D

CD N NN O O e -

CONTAlteENT VESSEt - COWARf SON OF 1974 EDITION APO 1980 EDITION OF AShE B&PV CODF SECTION III - DIV ISION i ARTICLE NE-3000 DESIGN PARAGRAPH /TlTtE 74 EDIT 10N - W74 ADDEPOA 80 EDITION - W81 ADDEPOA IWACT ON DESIGN

17. NE-3222 Gives Iimits for Operating Conditions Gives buckl ing requirement. Does not See eval uation in 6.0, recognize Operating Conditions. Item 14 Instead it provides Service Limits.

18. NE-3221 I) Primary stresses produced only by Deletes the restriction None, and NE-3222 mech anical loads (Figure NE-3221-1)

(01d Code) NE-3221 II) Ref er to Figure NE-3221-1. Expansion Def inition of expansion stresses as See evaluation in 6.0 (New Code) Stresses treated as secondary secondary has been removed. Considered f or NE-3227.5 (item 24) now as primary (see al so NE-3227.5). lii) Primary stresses are produced by The restriction deleted. None, pressure and mechanical loads for O general membrane and by pressure, g m mechanical loads and inertia ef fects y for local membrane and primary o bending (Figure NE-3222-1) m 19. NE-3222.2 The f ollowing note is added to the .None, k (1974 Code) subsect Ion: A NE-3221.4 "The concept of stress dif ferences (1980 Code) discussed in NE-3216 is essential to determination of the maximum range, since algebraic signs must be retained in the computation. Note that this Iimitation on range is appiIcable to the entire history of service conditions, not just to the stresses resulting f rom each individual service condition." kk r8 P $M

C0tITAINMENT VESSEf. - COMPARISON OF 1974 EDITION Af019f 0 EDITION OF ASME D8.PV CODE SECT ION lli - DIV ISICN 1 ARTICLE NE-3000 DESIGN PARACFAPH/ T ITL E 74 EDITION - W74 ADDENDA 80 EDITION - WB1 ADDENDA IMPACT ON DESIGN

20. NE-3 222-4 ( d ) I) Paragraph 2: Paragraph 2:

1974 Code NE-3 221.5 ( d ) E4y Note, adjacent points are def ined By Note, adjacent points are def ined None. 1980 Code as f ol l ow s: as f ollows: Analy sis f or "Adj acent points are def ined as points "a) for surf ace temperature dif ferences: Cy cl ic which are spaced less than the distance Ogeration 2 Rt from each other, where R and t are 1) On surf acas of revolution in the Temper at ure the mean radius and thickness, respect- mericonal direction, adj acent Difference Ively, of the vessel, nozzz le, flange points are def ined as points or other component in which the points which are less than the distance are l ocated. " 2 Rt, where R is the radius measured rormal to the surf ace f rom the axis of rotation of the midwall and 1 is the thickness of the part 43 at the point under consideration. [$ If the product of Rt varies, the pJ average value of the paints shall c3 be used. In 2) On surf aces of revol ution in the cir-y, cumferential d i rect ion, and on flat en parts as f langes and f l at heads, adjacent points are def ined as any two points on the same surf ace. b) For through-th ickness temperature di f f erences, adjacent points are def ined as any two points on a 1 ine normal to any surf ace. " c3 2, il) Paragraph (d)-2 prov ices requironents The stress intensity "Sm" is changed to None. [$3 f or Normal Operat ion Pressure F l uct uat ion, "S,b"w. Al so, " Normal Operat ion Pressure" cr is o cal led " Normal Serv ice Pressure."

p. .

CQ s a to na

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21. NE-3272 Operating Conditions were deleted and replaced See evaluation Operating by buckl ing requirements, in 6.0 Conditions

(1974 Code) For Level A and B; the maximum buck!Ing stress B uck s t r.g val ue shal l be either of . the f ollow..ig

! Stress l V al ues a) One third the.value of critical buckling (1980 Code) stress determinated by one of the methods given below:

1) rigorous analysis which considers the of fects of gross and local buckl Ing, geo-metric imperfections, noniinearities, large g def ormations, and Inertial forces (Dynamic o loads only);

m N

!               N                                                                                                     2) classical (linear) analysis reduced by l

P margins which reflect the dif ference i N between theoretical and actual load

• capacities (knockdown f actors) n N
• 3) tests of physical models under condl-tions of restraint and loading the same as-those to which the configuration is j expected to be subjected; b) the value determined by the appl Icable rules of NE-3133 (This covers only cyiIndrIcal shelIs under uniform exlal compression and cy1IndrIcal and other shells under unif orm external pressure).

Ok For Level C and D, Level A and B buckling f@cL allowables may be Increased by 205 and 505, respectively. 7 e ' CD M NN I i k

cot (TAfNMENT VESSEL - CC+tPARISON OF 1974 EDITION AND 1990 EDITlON OF ASuE E1PV CODE SECTION lil - DIVISICN 1 /RTlaE NE-3000 DESIGN 74 EDITION - W74 ADDENDA 80 EDITION - W81 ADDENDA IMPACT ON DESIGN PARAGRAPH / TITLE

22. NE-3226 Ruquiroments given in NE-6222 and Limits on maximum permissible test pr essure See eval uation Testing NE-6322. are given based on stress l imits f cr th e in 6.0 L im it s vessel.

(1980 Code) i) The average bearing stress under the El iminates the exception "other than f aul ted None.

23. NE-3227.1 Bearing maximum load, experienced as a resul t conditions" and ases " Service Loading" instead Loads of design conditions or of any of the of " Operating Conditions",

operating or testing conditions cther than f aulted conditions is I imited to Sy.

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U1 N N CD N 071 e N N l m

n. 3 C1.

w. CO N NN w

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i O O O CONTAIPeENT VESSEL - COWARISON OF 1974 EDITION Ato 1980 EDITION OF ASE BAPV CODE SECTION 111 - DIVISION 1 ARTICLE NE-3000 DESIGN PARAGRAPH /TITtF 74 EDITION - W74 ADDEPOA 80 EDITION - W81 ADDEPOA IWACT ON DESIGN

24. NE-3227.5 Nozz le Piping Transition a) Within I) Primary mmbrane stress classif ication I) f onal requirements are added for See evaluation the in nozzles applies to those resuUtIng ' ry membrane stress classif ications. In 6.0 Limits from pressure, external load and of mment. . .ses attributed to restrained end Rein- displacement of attached piping should force- ,

be added to stresses due to pressure, ment external loads and a m ents. Al so, states that primary membrane stresses None. g do not include discontinolty stresses, n .

          $                                                                       II) Expands classification of local primary       See evaluation N                                                                           membrane stress intensity due to              in 6.0 P

• discontinuity ef fects resulting fra --

N pressure, external loads and mments to T N include those attributable to restrained f ree end displacements of the attached pipe. i lii) Adds classification of primary local See evaluation plus primary bending plus secondary in 6.0 stress Intensities f or those resulting fra a combination of pressure, external load and aments, and temperatures and restrained f ree end displacements of

attached pipe.

SmE

      -  s b) Beyond the P, + Ph + Q classif ication is appl led tb primary plus secondary stress instead of " Operating Conditions", code requires to consider all pressure, temperature See evaluation in 6.0 gP                  Limits    intensities resulting f rom all ooerattna     external loads and aments. This would include g

ym of Re-inforce-conditions loads, accident conditions, ment l

CONTAINMENT VESSEL - COMPARISON OF 1974 EDITION AND 1980 EDITION OF ASME B&PV.E , SECTION III - Div ISION 1 ARTICLE NE-3000 DESIGN PARACPJPH/ T ITLE 74 EDITION - W74 ADDENDA _ 80 EDITICN - W81 ADDENDA _I W ACr ON DEslGN

25. NE-3228.4 No prov Isions in the Code. This is a new section. New provisions twe.

Irpulsive are added. Loads (1980 Code 1

26. NE-3229 Deleted from the Code. However, parts None.

Design are Ircluded in other subsections, Stress NE-3 ' l 2.4, Design; NE-3134, Material V al ues Properties. (1974 Code) 27 NE-3232 i)

Title:

Operating Conditions New

Title:

" Combined Loads". Nore. Qvs ii) Stress l imits f or bol ting material Stress l imits based on Table 1-10.3 See evaluation based on Tabl e I-1.3. In 6.0 ro o ill) Fatigue analysis based on Sm Stress limits t,ased on Sm1 N m 28. NE-3311 The requiroments f or acceptabil ity of a if the approach of "NE-3300 Design None, r'v Req ui r e- vesse, uesign are given under General by Formula" is used, the rules are o a:ents f or Design Gulet (NE-3130). appilcable to Level A and B Service Accepta- Load!ngs which do not incl ude substantial bitIty - mechaalcal or thermal loads other than (Subsection 1 pressur e. Substantial loads are mechanical of Design or it.prmal loads with cumulative result by Formul a) on stresses that exceed 10% of the primary stresses induced by design pressure.

29. NE-3324.3 There are two equations to determine the in addition to the previous two equations, None, oy Cylindrical minimum thickness (t) of the cylindrical additional equations to determine the QM Shell (Under shells subjected to internal pressure. minimum thickness are provided.

-{ -. Inter nal e Pressure) rv to 9 O e

s

                                                                                                                                             ~

C)

                                                                                                                                                                            %./

s

                     )

CONTAfPefNT VESSFt - COMPARISON OF 1974 EDITlON APO 1980 EDITION OF ASE BLPV rnrF SECTION lli - Div ISION 1 ARTICLE NE-3000 DESIGN 74 EDiTlDN - W74 ADDEPOA 80 EDfTlON - WB1 ADDE 10A IDFACT ON DESfGN PARACRAPHiTiTtE

30. NE-3324.4 There Is only cne equation to determine Several additional ego:ations are provided None.

Spherical the ministan th icknoss "t". The restric- to determine the e inim.an "t". The main Shells tion Imposed on the equation is that dif ference is that the code now provides (under t<0.356R and P <0.6655, a set of equations for the conditions when internal t >0.356R or P >0.6655. Pressu.e)

31. hE-3324.5 There are three requirements. There are six requirements. The following None.  ;

Formed are the additional three: Heads, General i) The Inside crown radius te shich a head Require- is dished shall be not greater than the i l ments outside diameter of the skirt of the c bead. The inside knuckle redlus of a O tortspherical head shall be not less

            $                                                                                       than 65 of the outside diameter of the N                                                                                       skirt of the head but in no case less P                                                                                       than three times the head thickness.

N Y Ill if a tortspherical, ellipsoldal, or hemispherical head Is formed with a y flattened spot or surf ace, the diameter of the fiat spot shalI not exceed that permitted f or flatheads as given by Eq. (l) or (2) of NE-3225.2 using C = 0.25. Ill) Openings in formed heads under Internal pressure shall comply with the require-monts of PE-3330.

       @g r+ ro

• 3 Q.

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CONTAINuENT VESSEL - COMPARf SON OF 1974 EDITION APO 1990 EDITION OF ASME BiPV CODE SECTION lli - DIV ISION 1 ARTIO.E NE-3000 DESIGN PARAGRAPH / TITLE 74 EDITION - W74 ADDEPOA 80 EDITION - WB1 A"DEPCA . IMPACT ON DESIGN

32. NE-3324.6 An additional equation is provided to None.

EllIpsoldal determine "t" f or el i Ipsoldal shelIs Heads of other than a 2: 1 ratio.

33. NE-3324.8 No provisions in the code. A new subsecticn NE-3324.8 is added which None.

(1980 Code) nrov ides new rules f or Tor tspherical TorispherI- Heads. Heads

34. NE-3324.11 The Code sets an additional requirement See eval uation

                                 !1974 Code)                                                    on shear as stated below:                    In 6.0 NE3324.12 (1980 Code)                                                    "The allowable stress value f or shear in C         Nozzles                                                        a nozzle neck shall be 70% of the allowable Q

N tensile stress f or the vessel material". o" NE-3326.2 i)

Title:

" Circular Spherically Dished New

Title:

" Spherical ly Dished Heads w ith None, y Heads w ith Bot ting Fl anges, Concave Bol ting Fl anges" m to Pressure", e d II) Rules of this section are applicable Rules are now appl icable to spherically None. to the spherically dished heads which dished heads either concave cr convex to are concave to pressure, pressure. Scrne revisions were made, but are not applicable to CFORP.

36. PE-33 26.3 "Circul ar Spherically Dished Covers = lth Deleted. None.

Bol ting Fl anges, Convex to Pressure". o> O 3 ("t O Co N

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N CONTAlteENT VESSEL - COWARISON OF 1974 EDITION AM) 1980 EDITIOh OF ASE PAPY CODE SECTION lli - DIVISION 1 NtTICLE NE-3000 DESIGN PARAGRAPH / TITLE 74 EDITION - W74 ADDE W A 80 EDITION - W81 ADDEWA IWACT ON DESlCN

37. NE-3331 Requirenent (b) of NE3331 states as Requirement (b) is modified and Is more See evaluation General f ol low s: restr icted. In 6.0 Req ui re-ments f or "For vessels or parts thereof which are o Rules contained in NE-3330 assure Openings in cycl ic service and do not meet the satisf action for pressure load only.

requirements of NE-3222.4(d) for operating loads so that a f atigue analysis is o An additional following requirement required, the rules contained in NE-3330 is imposed: assure satisf action of the requirements of NE-3221 in the vicinity of openings, "The requirements of NE-3221.4 may al so and no specific analysis showing satis- be considered to be satisfied if, in f action of those stress limits is the vicinity of the nozzle, the stress r eq ui red. " Intensify resulting from external nozzle

       @                                                                                  loads and thermal ef f ects, including m                                                                                 gross but not local structural y                                                                                  discontinuities, is shown by analysis i       o                                                                                 to be less than 1.5 S        In this case, when evaluating the rku.lraments of                                      ,

m NE-3221.5 (e), the peak stress intensities 1 O resulting f rom pressure loadings may be N obtained by appiIcation of the stress index method of NE-3338." 37a. NE-3335.1 Change to NE-3335(e) In the 1980 Code replaces None (1974 Code) the requirements of NE-3335.1 in the 1974 Code. Pad-Ty pe Reinf crcement I n e

     "3 CL O) N NN 1

i f I

CONTAlf# TENT VESSEL - COMPARISON OF 1974 EDIT 10N AND 1980 EDITION OF ASuE BAPV CODE SECTION l l i - DIV IS ION 1 ART I CL E NE-3000 DES IG N PARAGRAPH /TITiE 74 EDITION - W74 ADDEf0A 80 EDITION - W81 ADDENDA . IMPACT ON DESIGN

38. NE3332.2 The f ollowing requirtanent is added: None.

Req uired Area of "At least one-hal f of the required rein-Reinforce- fcecing shall be on each sice of the ment centerl ine of the opening".

39. NE-3334.1 Two requirments on the limit of retn- NE-3332.3 is deleted, however, its intention See eval uation Limit of forcement, in addition to that, req u i re- is carried to NE-3334.1. There is an in 6.0 of Rein- ments are prov iced in NE-3332.3 "Cmpact additional requirement set as descrited force- Reinf orcing in Vessel Wall", below:

ment Along the Vessel "Two-tnirds of the required reinf crement h al I shalI be within a distance on each side of a the axis of the opening equal to the greater o of the f ollowing: m N N 1) r + 0.5 /Rt, where R ls the mean radius P of shell or head, t is the nominal vessel N wall thickness, and r is the radius of the [ w f Inished opening in the corroded condition:

2) the radius of the f inished opening in the corroded condition plus two-thirds the sum of the thickness of the vessel wall and the nozzle wall (new eq ui pment )".

ON n no B Q. w. CO N N r0 0 0 0

s CONTAltMNT VESSEf- - COMPARISON OF 1974 EDITION APO 1980 EDITION & ASW P1PV CODE SECTION lli - DIV ISION I ARTICLE NE-3000 DESIGN fgfCRAPH/ TITLE 74 EDITION - W74 ADDE TA 80 EDITION - W81 ADDE POA IIFACT ON DESIGN

40. NE-3336 " Material la The nozzle wall used f cr Revised prowlslon: See evaluation Strength of reinf orcement shall pref erably be the same in 6.0 Rein- as that of the vessel wall. If material "Materlal used f or reinf orcement shall pre-f crcing with a lower design stress value, S, is ferably be ths same as that of the vessel Material used, the area provided by such material sci l . If the material of the nozzle wall shall be Increased in proportion to the or relnforcement has a lower design stress inverse ratio of the stress values of value $ than the vessel material, the the nozzle and the vessel wall amount of area provided by the nozzle wall material." or reinforcement in satisfying the require-ments of NE-3332 shall be taken as the actual area provided multipt led by the ratio of the nozzle or reinf orcement design stress value to the vessel material m

desip stress val ue."

      $   41. NE-3367                                                     Dimensional standards are updated to the      See evaluation o       Closures                                                    latest appl icable standards.                 In 6.0 6

Ln On Smali Pene-O t> trations raa we 00 N NN l

CONTAINMENT VESEEt - COMPARISON OF 1974 EDITION AND 1980 EDITION OF AsuE BiPV CODE SECTION lli - DIV ISION 1 NiTICLE NE-3000 DESIGN PARAGPNH/ TITL E 74 EDITION - W74 ADDENDA 80 EDITION - WB1 ADDENDA IMPACT ON DESlCN

42. NE-3720 The f ollow irg is an additional req ui rment See eval uation Design imposed: in 6.0 Rules f or Electrical "For closing seams in electrical and and mechanical penetrations rneeting the require-Mechanical ments of NE-4730(c), the closure head Pene- shalI meet the requirments of NE-3325 tration us i ng a f act or C = 0.30. The f illet meld Assembl ies shall be designed using an alic=able stress of 0.55mc."

Design of the pressure retalning portion g of the electrical and mechanical penetration o assembl ies shal l be the same as f or vessel s.

 "m                                                            Ther e i s no exempt ion to t h i s r ul e.

N O hc w I W w ~ea CD N NN

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EMLL 1 iTwis IMTf MKITY 1118111 FOR STIf t fYWTAtans aef t I felemRY STRESS INTierSlists fWenenpy ggus ggtfistmay GElfRAL DE06RAfd[ LOCAL 8(B98WWef OfDE)isc RUS LOC 4 IEleRAset $4155 thfimSitv Nape.t em P Pt*geg L g*Pt 1974 1980 1974 1980 1974 19 4 1974 1980 1974 19a0 LOAD Egde Cods Quig Qude Gods (bee Code Cods Celg CDee lhAT IONS Qade 0+ L

• I' +Pl/Pe/R Design Design $m Sac 1.S$m 1.$1mc I .S $m 8.S$me et/A WA "D+L e l'
• Pl/Pe/R Design Level A $m Sac 4.S $m t.S$me 1.9% I.S$me WA 3.0 ist Level A WA Sac It/A 1.55mc WA 1.55m 5.05= 5.0 $st

       ]      * *Det+ T,            operating N      "0+t e l' +Pt /Pe/

Design Level 8 5e Sac 1.5% 3.S$me t.S$m I.S 5mc 91/A 3.0 5et o R*mE N Oper at Ing level 8 WA $mc II/A t.3 5mc II/A l.S$mc 3.05m 3.0 $ml "D+L

• T,* mE U1 8 When not latogret De L e i'
• Pi/

f4/R+$$e Design towel C and Continuous Se $st 1.55m 1.55mc 4.55m 1.95mc WA sW A and when lavegrat I.25m 1.25mc t.85m 9.t$me 1.85m 8.85mc 0+ L

• f,+ 55E Iq/A st/4 and Continuous or Sy* or sy
• w t.S$y* or 1.55y* or 1.55y* or 1.55,*
• Larger of 2 values.

H see evaluellons in 6.0, flan le under NE-3000 NDTES: (1) PS/fV/R represents consideration of Pf, re er R la the load combinetlan at a time, temover, etI three situations need to be evalmeted.

82) esotellons are es por P54t 3.8.

(3) Level 0 service limit act ovatusted as such goods are not specified f r the vessel design.

n. m ,

a w= CD N t NN

LIST OF ITEMS NOT EVALUATED O The follow ing is a l ist of sections f rom the 1974 Code which have not been evaluated against the 1980 Code because either the items are not appl icable to CR3RP Containment or the containment vessel design does not adopt the techniques or the concepts descrited by the Code Sections. Further ir. the eval uation, consideration was given to only thcse loading conditions which are specif ied f or the design of the vessel.

1) NE-3122 Cl addi ng

2) NE-3131.2 Jet Impingement Effects

3) NE-3133.7 Conical Heads

4) NE-3213.21 Limit Analysis - Collapse Load

5) NE-3213.22 Collapso Load - Lower Bound

6) NE-3216.2 Varying Prir.cipal Stress Direction

7) NE-3222.5 Thermal Stress Ratchet

8) NE-3228 Appi ications of Plastic Analysis

9) NE-3324.7 Hemispherical Heads

10) NE-3324.8 Conical Heads

11) NE-33 24.9 Reducer Sections

12) NE-3325 Flat Heads and Covers

13) NE-3333 Reinf orcement Required f or Operir.gs Ir. Fl at Heads

14) NE-3338.2 Stress Index Methods

15) NE-3352.4, Para. 3( a ), 3(e ), 3( f ) . 3(d) is on Attachment of Nczzles Using Partial Penetration Welds; 3(e) describes Attachment of Fittings w ith Internal Threads; 3(f) is on Aitachment of Tube Connections

16) NE-3356.3 Head Attachments Usirg Corner Jcints

17) NE-3358.4 Fl at Heads wlih Hubs O

QCS220.25-37 Amend. 72 Oct. 1982 _a

p

                                                                                    \                                                                *

(G J 1 CONTA!PMP'T VESSFt - ASPE B&PV CODF COWARISON SECTION lli - ARTICLE NE-4000 FAPRICATION AND INSTALLATION PARACRAPH/ TITLE 74 EDITION - W74 ADDEMA 80 EDITION - W81 ADDE 2A IWACT ON DESIGN

1. NE-4131 No requirement for the time of examination Requires that examination of weld repalr to weld None.

Elimination of neld repair to weld edge preparation edge preparation be'in accordance with NE-5130. and Repair of Defects

2. FE-4213.7(c) Percent strain f x spherical or dished Qianged constant 65 to 75. None.

Procedure surfaces Q ual i f ica-j tion Test 5 strain = 611  % strain = 221 (lepact Testing)

3. NE-4213.2(e) Does not cross-ref erence base material imposes base material Impact test requirements of None.

          @tn impact testing of NE-2300                         E-2300.

N m 4 E-4213.2( f) Determine maximum loss of impact energy Determine maximum lateral expansion or maximum None, c) and maxime POT temperature change change In temperature, pl us, maximum changes M)T m ter.perature. cn E CD

5. NE-4221.2(a) (a) Maximun are length need not be (a) Maximum arc length need not be greater than None.

Maximun 0.300,. Deviation greater than deviation 0.250,true f rom the circlefor determining from True Theoretical Form ior External e'ressure

6. E-4221.2-1 (b) (b) Curves are shif ted siightly fran 1974 None.

Curves f or version. R& r+ i5 Maximum Perm issibl e Deviation,

    - -            from a true
    $y             Circuler Form NN

7. NE-4221.2-2 (c) (c) Curves are shif ted signif icantly to right. None, i Curves for Maximum Arch Length for Determining

Plus or Minus Deviation r

CONTAINMENT VESSEL - ASuE B&PV CODE COMPARISON SECTICN lli - ARTICLE NE-4000 FABRICATION AND INSTALLATION PARAGRAPH / TITLE 74 EDITfON - W74 ADDENDA 80 EDfTlON - W81 ADDENDA .!MPACT ON DESIGN

8. NE-4221.2(c) (d) Provides def inition of length w ithout (d) Clarif les def inition of length to be consis- None.

The Val ue of reference to NE-3133.2 tent alth NE-3133.2. Cy1inder Length (L)

9. NE-4222.1 ( a ) Provides tighter constraints on the devlation of None.

Deviation the head f rm the nominal dimeter of the vessel. for ( f rm) Specifled Shape

10. NE-4222.1(b) Prcvides f or skirts of heads to be within Provides thai hemispherical heads and any spherical None.

1% of nominal diameter portion thereof shall meet NE-4221.2. o Q 11. NE-4222.1(c) Provides that measurements shall be taken on base None. N metal and not on welds. N P 12. NE-4222.2 Forged heads shall conf orm to drawing None. None, N Tolerance w shape as Is practicable e on Forged d Heads

13. NE-4232-1 Provides for fairfr.g of offsets over the 1. Farring of of f set shall be at least a 3:1 taper None.

Fairing w idth of the f inir.shed wel d over the width of the f inished weld. of Of f sets

2. Of f sets greater than those in Table NE-4232-1 are acceptable provided the requirments of NE-3200 are met.

14. NE-4243 Category C welds include the exception that socket None.

Category C welded fIanges are 2" nminal pipe size and iess gg r+ (D Weld Joints and si ip-on flanges may be used. - 3 15. NE-4244(d) Partial Penetration wel ds f or Attachment of None. ~." Category D Nozzics are also limited by additional req ui rment s @y Weld Joints of NE-3359, which prov ides f or suf f icient mel d NN strength and percc..t of allowable stress valves. O G e

A f\ i I O u o CONTAltMf# VESSEL _ - ASE BAPV CODE COMPARISON SECTION 111 - ARTICLE NE-4000 FM3RICAT10N AND INSTALLATION PARAGRAPH /TITIE 74 EDITlON - W74 ADDEPOA 80 EDITION - W81 ADDEM)A _I WACT ON DESIGN-

16. NE-4244(e) (a) Internally threaded f itting and The 2" pipe size l imitation has been corrected None.

Attachment bol ting pads e.ct exceeding 2" to read 3" pipe size. of Fittings pipe size may be attached to With Internal vessel s Threads 16a. NE-4244(g) Nozzles with Integral reinforcing have been None. (1974) deleted.

17. Figure NE- Acceptable types of welded nozzles The reinforced examples are no longer shosn None.

4244(d)-1 using partial penetration welds shows as acceptable types. New acceptable types are examples of reinforced weldnents shown in (e) and (f). (e), (f) and (g)

18. E-4311.1 Stud cross-sectional area is limited to The cross-sectional area is now limited to 1" None, o Stud 1/2" maximum diameter f or stud welding for flat position and 3/4" diameter for all Q Welding other positlons.

m RestrIc-N tions N Capacitor Discharge Welding and Low Energy m 19. E-4311.2 & NE-4311.3 Discharge Welding provisions have been added.

     $         Capacitor Discharger Welding 4

20. E-4322.1 Requires the marking be done with blunt Nose None, identifIca- continuous or Interrupted dot die stamp and tion of provides relaxations where multiple welders Joints by are involved. loentif ication of tack welders gg em Wel der or Welding not required. Deletes ref erence to NE-4122.1.

  *g
  >-a*

Operator W 21. NE-4332 Base material used f or wel d qual If Ica- Base material shalI be In accordance wIth None. gg y NN Base tion shall be same as type and grade the appiIcable requirements of QW-403.4 and Material except that any P- Number 1 material in QW-403.5 of Section IX. to be a Group qualifies f or all P- Ntaber 1 of Empl oyed the same grouping

22. NE-4334.1 Additional requirement that where the postweld None.

Coupons heat treatment temperature exceeds the maximun Representing temperature specif ied in NE-4620, and the test the Weld assembly is cooled at an accelerated rate, the Deposits longitudinal axis of the specimen shall be a I minimim of t f ran the edge of the test assembly. 4 Otherwise the axis of the specimen shall be not

CONTAINMENT VESSEE - ASME B&PV CODE COMPARISON SECT ION t il - ARTICLE NE-4000 FN3RICATt0N AND INSTALLATION PARAGRAPH /TITtE 74 EDITION - W74 ADDEhDA 80 E01TlON - W81 ADDENDA IMPACT ON DESIGN

22. (Cont'd) less than 3/8" f rom the weld surf ace if possible, but not less than 1/4t.

For drop weight specimens, the tension surf ace shall be parallel to the surf ace of the test wel d assembly.

23. NE-4334.2 1. Additionally, def ines axis of the weld None.

Coupons relative to plate or forgings. Repr esent i ng the Heat 2. Provides f or comparison of heat af fect Af f ected zone val ues w ith base material val ues. a Zone m N

24. NE-4335 Impact Specif Ically overrules certain exemptions of impact test ing permitted by NE-2311(a)(8).

None. P Test Require-N ments u, e L

~

25. NE-4335.1 Requires impact testing f or wel ding foquires: None.

impact procedure qual if ication f cr cl assif ica- 1. Impact testing is required f or meld metal for Testing of tion A-Number I weld analysis or any the wel ding procedure f or any wel d repair to hel d Metal other f errific weld analysis base metal that rquires impact testing.

2. Impact test requirements and acceptance standards f or wel ding procedure qual if Ica-tion weld metal shall be the same as specif led in NE-2330 for the base material to be welded or repaired. Dissimilar metals o ;E= shall be impact tested according to requirements S5 f or either metal except where ott erw ise specif ied by NCA-1230 or other parts of Section Ill.

g 3. Impact tests nct required f or austentic and NN nonferrous metal.

4. Welding procedure qualified to lacact testing requirements of Subsection FD or NC may be G accepted as an alternate.

N () e i J V CONTAf rNENT YESSEL - ASME PAPV CODE COWARISON SECTION 111 - ARTICLE NE-4000 FN3RICATl0N AND thSTALLAT10N PARAGRAPH /TfTtE 74 EDITION - W74 ADDENDA 80 EDITION - W81 ADDENDA l W ACT ON DESfGN

26. NE-4335.2(a) Requires impact testing f or base metal 1. Provides exemptions f ran impact testing for None.

impact Tests weld heat-af f ect zone f or mater 8al s af (a) the qual If ication f or melds in P-Ntmiber 1 of Heat P-Number 1 cl assif ication material that Is post meld heat treated and Af f ected made by any process other than electroslag, Zone electrogas, or thermit (b) the qualification of weld deposits on weld cladding on any base material.

27. NE-4335.2(b) Greatly expanded requirements for impact testing None.

Impact Tusts for special case procedure qualification test f or of Heat heat af f ected zone. Affected Zone o O N

28. NE-4335.2(c) Retest of f ailed Impact test results shall be at None.

higher temperature until requirements are met, o

29. NE-4335.2(d) A welding procedure specif Ication qualIf led to None.

u, Impact testing requirements of Subsection P6 or [ NC may be accepted as an alternate. N

30. Figure NE- Figure was redrawn. Back weld added to (C-1) None.

4427-1 sl ip-on f l ange. Removed 2" pipe size 1 Imitation. Fillet and Socket Weld Detalls and Dimensions

31. E-4429 Weld deposited cladding shall be No reference to inspection methods. None.

Welding of examined by a liquid penetrant method

    @5g r+             Clad Parts      In accordance with the requirements d                           of Article 6 of Section V and the acceptance standards of NE-5350 g-c3 -a NN         32. E-4431          Material shall meet requirements of             Materials shall meet requirements of fE-2190 and             None.

Mater ial s NE-4620 and Impact tested to Table impact tested to NE-2300 If exempt f rom post weld f or Perma- l-10.0. Material exempt f rom Impact heat treatment, nent or not post meld heat treated maf not be Struct ural welded closer than 4" or 16 times Attachment thickness f ran weld joint of permanent structural attachment l

CONTAINMENT YESEft - AEME EMPV CODE COMPARISON SECTION 111 - ARTICLE NE-4000 FERICATION AND INSTALLAT!GN PAPAGRAfH/ T I T L E 74 EDITION - W14 ADDEt0A 80 EDITION - W81 ADDE t0 A !MPACT ON DES!CN j 33. NE-4333 Prov ices f or full pe ne tr at i on, f il let, or part iet None. Types of penetr ation continuous or irtermittent melds f u Purmanent attachment cf Permarent Ltr uct ur es. Sir uct ur al Attachment Welds

34. E-4435 At so prov ides that (a) the eldig material is Nore.

Welding of ident if led and cornpatible w ith the mater f als Non-Str uct ural Jof red (b) The welds are post -eld heat treated and Tennporary when required by NE-4620. Attachments

35. NE-4453.1 Walves examination where cef ect removal remov es t.one, Defect essentially the f ull thickness of the meld in a

n Resnoval partial penetration and f li t et mel ds. v1 N N 36. NE-4453.4 Pepair cavities which do not exceed Repair cavities nct exceedin t/3t f or ti tone. P Exam inat ion 3/8" or 10% of wel d th ickness need 1/ 2", 1/41 f or 1/ 2" < 2-1/ 2". '9" or 10$t for N of Repair only magnetic particle or l iquid t > 2-t/2" need only magneti- srticle cr IIquic Y Wel ds penetrant method f or reexamination penetrant method f or reexaminotion, where t= ,

   **                                                                          thickness.

w

37. NE-4622.1 ( a ) No requirement to Ir.cl ude F%HT time 1. FDHT shall be perf ormed in tmperature - None.

Gener al af ter completion of component into surveyed and - cal ibrated f urnaces or w ith Req ui r e- total time aT temperature f or test thermocoupl es in contact w ith the material ments specimen or attached to blocks in contact w ith the material.

2. FWHT time af ter completion of part shall be added to totcl tir.e at ternperature f or test Rg specimen. l rt o

     $  37a. NE-4622.4(c) Frov ides three (3) alternatives f cr Hol di ng       lowering heat treat tmperatures Two al ternatives f or lowerirg heat treat tecpera-tures are cel eted.

tone.

 $y           Times at                                                                                                                                            l NN           Temperature                                                                                                                                         I G                                                                G                                                                               e

CONTAtteENT VESSEL - ASK BAPV CODE EV ALUATION j SECTION III - ARTICLE hE-4000 FERICATION AND INSTALLATION i PARAGRAPH / TIT 1E 74 EDIT lON - W74 ADDEf0A 80 EDlTfON - W81 ADDEPOA 1IFACT ON DES 1GN

38. NE-4622.7( f) Deleted exemption f or IMHT of Type 405 or Type None.

Exemptions to 410 with carbon less than 0.085 and modified i Mandatory some exemption requirements. l Requirements

39. NE-4640 Delete Heat Treatment af ter repair and placed in None.

] ' Heat Treatment Table NE-4622.7(b)-1. j After Repalr by Welding 1

40. IE-4660 Added requirement f or heat treatment of electro- None.

j Heat Treatment siag weld. of Electrost ag

;                                   Welds a      41. NE-4730                                                           Optional method f or closing seam f or the penetra-       None.

O Electrical tion assembly depicted in added Figure NE-4730-1

                        "           and Mechanical                                                   and in accordance ulth listed requirements.

m N Penetration

                        ?

N Assembiles Y 42. E-4714 No requirement Prrivisions f or stud threading has been added. None, g Stud 4 Threading

43. NE-4740 No requirement Qualif ication requirements added. None.

Special Q ual i f Ica-tion Require-ments for i ElectrItal and Mechani-og cal Penetra-Qm3 e tion Assembiles Q w.. CO N NN 4

CONTAINMENT VESSEL _ - COMPARISON OF 1974 EDITION AND 1980 EDITION OF ASME BAPV CODE SECTION lli - Div IS f 0N 1 NtTICLE NE-5000 EXAMINATION PARAGRAPWT I TLE__ 74 EDITION - W74 ADDEtOA 80 EDITION - W91 ADDEPOA IPPACT ON DES f GN

1. tE-5130 No requirement for examination of weld 1. All full penetration weld edge preparation None.

Exm ination edge preparation surf aces surf aces shall be examined by magnetic or of Weld l iquid penetrant method to the acceptanco Edge Pre- standards def ined in NE-5130. paration Surf aces

2. Laminar Indications exceeding 1" in length None, shalI be examined ultrasonically to the acceeptance standards def ined In NE-5130.

3. Weld repairs made to weld edge preparation None, shall be magnetic particle and iIquid pene-o trant examined.

m N 2. NE-5232 No requirements for corner joints Corner joint melds where one plate is more than None. N O Non butt- 1/2" thicker, the cut edges of the plate shall w el d be examined bef ore welding adjacent to the

  @       joint                                                           Intended weld by magnetic particle or liquid g                                                                       penetrant methods. Af ter welding all exposed edges adjacent to weld shall be reexamined.

m

3. NE-5270 Clad plate and applled corrosion layers Weld metal cladding shall be examined by the None.

Speciel be radiographed, liquid penetrant method. 4 hS-5278 No requirment All complete penetration welds made by the None. El ectrosl ag electrost ag process in f errific materials Weld s.halI be ut trasonIcally examined

5. NE-5320 Provides for radiographic acceptance Permits, in addition to radiographic acceptance None, og Radio- standards standards, Internal root meld conditions as acceptable when the density change as Indicated em graphic

{ Acceptance in the radiograph is not abrupt. ~. Standards to Co N NN O O e

                                                                                     \

CONTAt teENT VFSSFl - ASE B&PV CODE CO WARISON SECTION lli - ARTIQ.E NE-5000 Examination PARAGRAPH /TITtE 74 EDITION - W74 ADOFEA 80 EDITION - WB1 ADDE EA I FACT ON DE S t CN l 6. E-5342 Does not provice for a lower limit of Provides that only Indications w ith major di- None. Acceptance relevant Indication mensions greater than 1/16" shall be consloored

Standards relevant l & NE-5352 Acceptance Standards

7. E-5520 Qualification and Certification of QuallfIcation and Certification of personnel None.

Personnel Personnel based upon SNT-TC-1 A description greatly expanded but still based Qual if Ica- upon SNT-TC-1A tion, Certification o and Q ru Verification

                 @      8. E-5600       Defines some limited material requirements Deleted as being redu. ent to NE-2000 require-            None.
                 -          Examination                                                  monts. Noved the requirements for materials l
                 @          of                                                           forming a corner joint to NE-5232 g          Motorial m

r P e IV f9 1

CONTAINMErfT VESSEL - ASME D APV CODE COWARISON SECTION lli - ARTICLE NE-6000 Testing PARAGRAPH / TITLE 74 EDITION - W74 ADDENDA __,80 EDITION - W81 ADDEPOA _ I WACT ON DESIGN

1. tE-6110 Requires all vessels & appurtenances Now rnquires af f oressure retaining vessels None.

Testing constructed or installed per Sub- etc. to be tested. Al so washers have been of HC tect ion NE to be pressure tested gJde.d to l ist of items exempted f rom test ing. Component s

2. NE-6111 Rearranged editorially None.

Pneumatic or hydro-

           *tatic Testing

3. NE-6112 Rearranged editorially hone.

Q m Conditions for Pneumatic y Test i ng o

4. NE-6112.1 Ref ers to " compressed gas" and states NE-6112.2 ref ers to " compressed gaseous None.

m Precautions that precautions shall be taken fluid" and recommends that prorsutions be i f or Pnema- taken N tic Testing

5. NE-6115 Required test to be done prior to Paragraph has been deleted f rom latest None.

Time of stamping a provided instructions to rev ision, however, new paragraph NE-6113 Test inspector for signing Data Report requires all testing done under this Article to be in the presence of the inspector

6. NE-6121 Discusses leaving all mechanical anc Has expanded these requirements to all Joints None.

Exposure welded joints accessible for examination and lef t uninsulated and exoosed for examina-of Joints except as provided in NE-5211 tion o 3> Sk 7. NE-6122 Allows use of temporary supports or Considers stiffening same as support None. - { Tm porary stif fening to support weight of test .-.. Supports IIquid during test e CD N NN G G e

CONTAf tedENT VESSEt - ASME PAPV CODE C000ARISON SECTION lli - ARTICLE NE-6000 TESTING PARAGRAPH / TITLE 74 EDITION - W74 ADDEPOA 80 EDITION - W81 ADDE 2 A IDFACT ON DESIGN

8. NE-6123 Deletes option of Isolating expansion joint None.

Restraint or during the test. ( Note: the title is not isolation of changed) Expansion Joints

9. E-6125 Discusses not testing Flanged joints Allows these same Flanged joints to agt be None.

Flanged with Blanks Installed until af ter the retested af ter removal of blanks Joints Blanks are removed containing Blinds

10. NE-6211 Requires vent at all high points to Reduces discussion to Indicate venting shall None.

g Venting vent air pockets while filling be done to alnimize air pockets n before y Hydrostatic N Test ing o N 11. NE-6212 Limits hydro media to water & requires Expands requirements to allow use of "alterna- None, f Test Nedium testing be done at temperatures above tive IIquid" as permitted by Design g & Tep-erature brittle fracture point Specification

12. NE-6123 & NE-6127 - now contains these relocated require- None.

E-6313 ments Check of Test Equip. (1974 Code)

13. NE-6215 Requirments ere ncw gIwen in NE-6224 None.

Fxam ination ap ior ieakage Q ,E3 (1974 Code)

• 3 s ."

N

CONTAINMENT VESSFi - ASME B&PV CODE COMPARISON SECT ION lli - ARTICLE NE-6000 TESTito PARAGRAPH /TITtE 74 EDITION - W74 ADDENDA 80 EDITION - W81 ADDENDA IMPACT ON DEslGN 14 NE-6 221 Required minimum of 1.35 Times Design Redef ined min. pressure to be not less than None. Min. Test Pressure mul tipl ied by a f actor depend- 1.35 Times design pressure Pressur e ent on stress f or that particular material

15. NE-6222 Provides lengthy discussion of consicer- Requirements are now given by the stress See Evaluation Max. Test ations to determine max. pressure to be limits of NE-3226 when determining in 6.0 for Pressure used max. pressure item 22.

16. NE-6224 Now numbered NE-6223 None.

Hol ding Time (1974 Code) a 17. NE-6224 Requirements given in NE-6215 Now appears as NE-6224 and provides much more None. Q Exam inat ion def initive guidance on pressures to be maintained N for leakage during examination depending on what is being N (1980 Code) exam ined, i t al so prov ides f or al low ances of P certain kinds of leekoges during the test. N 7 18. NE-6510 New Sect ion NE-6320 is prov ided that now incl ude s eve. p Beltows requirements f or test ing bellows. This charige goes in hand with NE-6123 which no longer recog-Expansion Joints nizes isolating expansion joints during hydro (1974 Code) testing & Is tied to thanges made f rm old Section NE-6500

19. NE-6311 Only change is ref erence to NE-3226 for guidance See Evaluation in General on stress Intensities in keeping with change in 6.0 for item 22.

content of NE-6322

20. NE-6312 See comments as Test temperature portion of Test NE-6212 changes Tm perature O E d (D

~ i. CG N NN O G e

Nn

                                                                                                                                                                                          )

v 1 CONTAINENT VESEEL - ASE BAPV CODE COW'ARISON l SECil0N lli - ARTICLE NE-6000 TESTitG 1 PAPJCRAPH/TITI_ E 74 EDITION - W74 ADDENDA _ 80 EDITION - W81 ADDEEA I FACT ON DESIGN

21. NE-6314 NE-6313 now provices this guicance. It no None.

Procedure longer contains the requirement to reduce pressure for the examination. This is now contained in NE-6324

;              22. NE-6315                                                                  Now appears as NE-6324           More al lowance f or                            None, j                     Examination                                                              reducing pressure during examination to the for leakage                                                              greater of Design Pressure or 3/4 Test Pressure vs. 4/5 Test Pressure

23. PE-6321 Required minista of 1.10 times design Redef ined min. pressure to be not less than None.

Min. Test pressure multiplied by a f actor dependent 1.1 times design pressure a Pressure on stress for that particular material n I

       $       24. NE-6322                                                                   Sane comments as Max. Text Pressure of NE-6222 changes.

N Max. Test P Pressure N T 25. NE-5324 Now numbered NE-6323 None. j

       @             Holding Time

26. NE-6411 Reduced reaufrement for using a recording None.

Types & Location gage to a recommandation to be used f or of Gages vessels with large volumetric content j b RE raa H* CO N _ NN .V,I_*

CONTAINMENT VFSSFt - ASME B APV CODE COMPADI SON SECTION lli - ARTICLE NE-6000 TESTito PARAGRAPH /TITt E 74 EDITION - W74 ADDEtoA 80 EDITION - W81 ADDE PO A IMPACT ON DESIGN

27. PE-6500 This incl udes requirements that allow This guicance is now relocated to Section None.

Pressure both hydrostatic or pneumatic testing of NE-6320. New paragraph NE-6320 deletes Testing Expansion (Bellows Type) Joints. direct reference to pneumatic testing of cf Expan- these joints, o O Ln P N U1 a, - nx

• 3 CL ms NN O O e

f

                                                                     %   a
                                                                                                                                        /

Y CONTAf teENT VESSEL _ - ASE BAPV CODE COWARISON SECTION t il - ARTla.E NE-7000 - FROTECTION K,AINST OVERfHESSURE PARAGRAPH /TITtF 74 EDITION - W74 ADDEWA 80 EDITION - W81 ADDEPOA l@ACT ON DESIGN

1. E 7000 General Comment: Article NE-7000 has been See evaluation completely rewritten. A subsection by sub- In 6.0 section comparison is not possible in the strict sen se. Section NE-7000 has been expanded f rom two pages to nino pages providing new guidance in the folIouing areas.

E 7110 Scope NE 7111 General Def initions NE 7120 Integrated overpressure protection NE 7130 Provisions for check operation of pressure rellef devices O NE 7160 Unacceptable relief devices ON NE 7300 Ref leving capacity requirements NE 7400 Set pressure of pressure relief devices g o NE 7500 Operating design requirements for pressure g rellef valves m NE 7700 Certification requirements (specifIcally N Involve certifying capacity of rollef devices) E 7800 Narking, stamping and data reports NE 7141(b) - (d) New provisions f or Installation

2. E 7113 E 7141(a) provides similar guicance only minor None.

Requirments when editorial change Pressure Reilet Dew!ce are Per-manently Installed 3 NE 7114 This allowed f or aveantions as Indicated E 7151 now restricts all construction to Class None. Req ui re- In the article 2 requirenents Rg ments c+ (D for Pressure

• $ Rellef Devices g-NN

CONTAIPNENT VESSEE - ASME B1PV CODF COMPARISON SECT ION lli - ARTICLE NE-7000 - FkOTECTION MAINST OVERFEESSURE PAPAGRAPH/T ITL E 74 EDITION - W74 ADDENDA 80 EDIT!ON - W91 ADDENDA IMPACT ON DESIGN 4 NE-7211 & NE- These two sections are now addressed in None. 7212 Accept- NE-7152. NE-7152 contains only minor atI e ty pes changes to regrouped content of of vacuum previous paragraphs rellef devices

5. NE-7117 Discusses placement of step valves and NE-7142 - New requirement f or means shall be None.

Intervening Interlocks and control s required if they prov ided sJch that the operation of control s Step Val ves are placed such that they m!ght prohibit and interlocks can be verif led proper ret lef protection E, u N o" CONTAltNEt(T VESSEL - ASME P&PV CODE CCMPARISON SECTION lil - ARTICLE NE-8000 - NAKPL ATES, STAMP!PC, & REPORTS u, w PAPACRAPH/ TITLE 74 EDITION - W74 ADDENDA BD EDITION - W81 ADDENDA _ IMPACT ON DESfGN

1. NE-8100 Provides similar guicance; only minor None.

Req uirement s editorial changes nN ~2a w-NN

     #                                                                   #                                                                  G

4.0 COW ARISON OF SECTION 111. DIVISION 2 AFFECTING FOUf0ATION MAT AND BOTTOM LINER

            .1  CC-2000  1laterIals
            .?  CC-3000  Design
            .3  CC-4000  Fabrication and Construction 4 CC-5000   Construction Testing and Examination O

I . O QCS220.25-54 Amend. 72 Oct. 1982

CC-2000 MATERIAL d LIST OF CHANGES CC-2122.3 Reports of tests, treatments, etc. to go to Authorized Irspector (AI). Provisions made for inspections by Al as requirec. - l CC-2131.4 Personnel Qual if Ication. Laboratory personnel perf orming tests required by CC-2000 may be qualified using appropriate industry or l aboratory standards. Appendix V il qual if ication not mandatory for laboratcry tests f or concrete constituents and Concrete. CC-2160 Dimensional Standards. Dimensions of standard items fer pipe, tube, fittings, etc. opdated and enierged Ir Table CC-2160-1. CC-2211 Consideration shall be given f r. The requirements of concrets to minimiz ing the heat of hydration in concrete, strength development w Ith respect to form removal, and construction stresse s. (Note these items are covered in Burns and Roe Specifications). CC-2221 Cement. Added ASTM C595, Type IP, (MS) cr (hH) to other p al Iowabl e types of cement. Eniarged on rcquirements for use of d sul phate resistant cement. CC-2222.1(b) Coarse Aggregate. Added criterie and test for flat and elongated particl es. Test is CRD-C 119, Method of Test f or Fl at and Elongated Particles in Coarse Aggregate. CC-2222.1 ( f) if tangential shear is to be carried by concrete, ' test aggregate loss of weight by ASTM C131, Resistance to Abrasion of Small Coarse Aggregate by Use of Los Angeles machine. Lcss not to exceed 40%. l CC-2224.1 Admixtures for Concrete. Eech admixture shall not contribute more than 5 ppm, by weight, of chlorico Ions to the total conerete constituents. CC-2310(d) Only material listed in Table 1-2.1 may be used f or joinirg reinf orcing bar to 1 Iner plate or structural steel shepes by wel di ng. O QCS220.25-55 Amend. 72 Oct. 1982 l - . -- . - - . -

CC-2332 Bend insts rev ised by mcre expl icit-instructions. Pin diameter changed and I isted. CC-2333 Chcra ical Anal y si s. Requirments f or analysis revised and def ined in more detail . CC-2535 Examination and rcquirements added f or wrought seamless and wel ded tubul er products and f Ittings.

 \

s CC-2537 Examination and repair rcquirements added for statically and contrifugally case products. CC-2600 Welding Materisis. Sone changes in wording, format and tests. CC-2700 Maler f el Manuf acturer's Qual ity System Program. System progran description and ref erences have changed. CONCLUSION. Ncne of these changes impact 1he design. O O QCS220.25-56 Amend. 72 Oct. 1982 l

i 4 s i FOR MAT FOUPOATION DESIGN ! CC-3000 - DESIGN i 1 i OIANGES IMPACT ON DESIGN 1 i Ef fect of changes of ASE Section ill, Division 2 from

1975 Edition to 1980 Edition.

1. Lo d cabination and load f actors (P.172 Table CC-3230-1) See evaluation in 6.0 Constr uction Load Cabination:

U = 1.00 + 1.0L + 1.0F + 1.0To + 1.0W 4 "1.0W" added in the load cabiration !' 8M

2. Allowable Cmpr ession stresses for f actored loads y (Page 174, Table CC-3421-1)

 ,     N P        a. The maximum al1omable pr! mary-pl us-secondary membrane                             2 e evaluation in 6.0 N                                                                     corresponds to

• and bending limiting strain compressive stress of .002 in/In of .85 instead of f'c.033 In/In in

;      m             1975 Edition N

i

b. The membrane portion of the calculated stress shall not None.

exceed the allcweble stress applicable when membrane . stress act alone,

c. The primary portion of the calculated stress shall not None.

4 exceed the allowable stress applicable when primary i stress acts alone. 4

   ,, n 1   . 2 c.
,  w*
!  $N NN J

4

]

FOR MAT FOUt0ATION DESIGN CC-3000 - DESIGN OfAt0ES REMARKS

3. CC-3121 : "L Iner behav f or" changed to "Maximisn strains" No impact.

4 CC-3136.3 thru 313b.5: Classif ication of prirary and No impact. secondary stresses redef ined. Notably the bending stress at a gross structural discontinuity due to external loads is reclassif ied f ran secondary to primary stress. ( Note 3 of Tabl e CC-3136.5-1)

5. CC-3222.3 and Table CC-3230-1: Def ined the internal No impact, f l oodi ng l oa d, Ha, in the load combinations.

6. CC-3410: Add " primary" to f actored loads to clarif y gereeral No impact.

Q yiel d consicerations, m N g 7. CC-3422.1: Requirements f or tension rebars have been No Irpact. O expanded substantial ly. N

  &     8. Table CC-3431-1, subparagraph CC-3432.1 and 2: Changes may                      No lepact.

af f ect the added m ind l oad "1.0W", secondary ef fects and M Co test condition.

9. CC-3521.2.4 (deleted)

10. CC-3 531.1. 2( e ), (h) (3) & (h) (4): Changes in development No impact.

length requirments. O 2" n -s r+ (D - 3 ~ .C2 Co N NN G G e

CC-4000 FABRICATION AND CONSTRUCTION CC-4240(a) Curing. Concrete shall be kept moist and protected through the minimum curing time specif led in construction specif Ications. (Dropped "at least 7 days af ter pl acing"). Note: Burns and Roe specif ications ref er to ACI 301 and 308 which conf orns tc the above curing requirements. CC-4323.4 Fabrication of Reinforcing - Tolerances. Tolerances shown in Figures 4323-2 and 4323-3. CC-4330 Spl Icing of Roinf orcing Bars. Added new types of mechanical spl ices and method of qual if ication of other systems by manuf acturers. CC-4522.1 Liner Shell . Spacirg of measurements f or tolerance conf ornity increased f rom 10 feet to 12 feet. T NQ.US10N. None of these changes impact the design. O O QCS220.25-59 Amend. 72 Oct. 1982

CC-5000 CONSTRUCTION TESTING AND EXAMINATION 9 CC-5221 Cement. Tests Iisted in Tabl e 5220-1 enl arged, but are l ess than those given in Burns and Roe Specification 3066-10-4. CC-5222 Fly Ash. Tests l isted in Tabl e 5220-1 enl arged, but are less than those given ir Specification 3066-10-5. Al so, f requency is less in most cases (1000 tons instead of 200 tons). CC-5224 Aggregates. Frequency changed f rom time basis to vol ume basis. (once daily chenged to each 2000 cubic yard concrete. Depending on production . ate, may be increase or decrease in f requency. CC5225 Water. Soundness test dropped. Frequency decreased f rom monthly to every 6 months. CC-5232 Concrete, Slump. Af ter initial testing, frequency of testing has been decreased f rcrn every 50 cubic yeard to every 100 cubic yards. CC-5232 Air Content. Af ter initial testing, frequency of testing has been decreased f rom every 50 cubic yards to every 100 cubic yards. Temperature: Ditto-Weight /Yiel d. Af ter ir.itial testing, f requency has been decreased f rom every 100 cubic yards to every 200 cubic yards. Compressive Strength. Ditto CC-5500 Examination of Liners. More expl icit instructions added f or testing and examination of wel ds. Ultrasonic testing added as an allcwable method of testing (added to radiographic, magnetic particle and I (quid penetrant methods). CONCL US ION. None of these changes impact the design. O QCS220.25-60 Amend. 72 Oct. 1982

   ~ _ . .         - -      . _ - - _        .     --     .       .              -  --     ._. _           -..       -      .           .- - .         -    - _- . -- .
                 \                                                                             %

1 a i 1 5.0 COMPARISON OF CODE CASE N-284, BUO(LING mlTERI A

l. Key Items f or Comparison The comparison between the PSAR and the Code Case N-284 buckling criteria is based on a qualitative evaluation of both cr i ter i a. Only those items judged to be signifIcant to the design considerations of the CIERP containment vessel are included in the evaluation.

When one of the two criteria is identified as "more conservative" with respect to a particular item, the conservatism Is

;                                     meant to be relative to the other criteria for that particular item only. The evaluation Is also based on loadings and geonetries applicable to the OERP vessel design only.                                                                                  l

Items f or compalson are grouped in accordance with the following 1 Ist

[ '

]

l. Classical buckling analysts f or single stress component

! A. CalcuiatIon by formulas B. Analysis by shell of revolution computer progras l C. Analysis by three-dimensional fintte-element or finite-dlf ference computer prograns 4 i'

11. Critical buckling stress f or single stress component

{

              ,o                               A. Local buckting n

LA

 ,            rv                                      1. Unstif fened cylindrical shells and also cylinders between ring stif feners
              @                                       2. Double curvature shells

3. Thermal stress f $ 4.

5. Membrane stress at structural discontinuity Curved panels between rings and stringers 3 i 4

              ~

l B. Rings and stringers 4 1 4 4 ll 4 r8a i l M;0 1 i i k

ii1. I nier act ion eq uat i ons f or Iocal buckling A. Cy I indr Ical shel i s

1. Axial compression pl us hoop compression

2. Axial or hoop compression plus shear

3. In combination with thermal B. Double curvature shells IV. Factors of safety A. Stringer buckt ing and general Instability B. Local buckling iiI. Description of Numerical Comparison A brief description of the numerical comparison using BOSOR4 computer analyses is given in Section 7.0 following qual itative comparisons of the PSAR and N-284 criterla.

D O

W N N O. U1 s N O no

 = 3
 > CL.

CD N NN

    #                                                                                          9                                 e

O O O CONTAIPMENT VESSFl - BUCFL f NG CRITERI A COWARISON ITEM PSAR CRITERIA CODE CASE N-284 COfGENTS 1 Ciassical buck!Ing analysis f or single stress component lA. Calculation by formulas Conservative Methods of -1712 Formulas for calculating classical buckling (Conservative) values under unif orm loads are well estab-1 I shed. Both criteria assume unIfcrm distrI-bution of the stress component in the circum-ferential and longitudinal (also merldlonal) directions. Generally the two criterla w111 yield similer values. For a longitudinally varying stress com-ponent, a computer analysis f or calculating the unlaxial classical buckling value is o allowed in -1711. m N 38. Analysis by shell of Not detined Methods of -1720 Paragraph -1720 permits use of shell of P revolution computer ( Note: used BOSOR4 (Conservative) revol ut ion computer analyses. Such analyses N

    ,        prograns                        analyses as bases                                         are capable of handling stress components a                                       f or thermal buckl ing                                   which vary In the circumferential and/or
    $                                        criteria)

Reference i longitudinal directions. This method pro-vides ret lef f rom the conservatism Inherent In the methods of -1712 wherein the peak val ue of the stress component is asstmed to apply unif ormly throughout the shell.

 @g          Reference 1:       Response to NRC Question 220.43(a)

I$a G' R!s 1 i \ l

CONTAINMENT VESSEL - BUCKLitC CRITERI A COMPARISON ITEM PSAR CRITERfA CODE CASE N-284 COPHEN1 S IC. Anal y si s by th ree- Not def ined Methods of -1730 Muthods of -1730 permitted by N-284 are dimensional finite- ( Note: eval eat ion (Conservative) similar to methods of -1720 except the f or-element or finite- based on estab- fner methods allow model Ing of material difference computer iIshed engineer- property variations in the circumferential programs ing practices) direction (e.g. large penetration and re-Inf or cement s i n a shel l ) . The 00RP con-tainment vessel f ol l ow s the ASPE area repl acement method f or the penetration reinf orcement design which assumes no re-duction in buckling capacity for a properly reinf orced opening based on establ ished engineering practices. 11. Critical buckling Conservative Conservative The compari son between the PS AR and Code stress f or single Case N-284 criteria 15 based on f ormulas s M stress component given in both criteria which def ine the critical buckl ing stress f or a stress com-N N ponent as the classical buckl ing val ue O mul tipl ied by the corresponding capacity N reduction f actor due to imperf ection knock-f dunn. These formulas given in both criteria c3 are conservatively establ Ished f ran analyses A and tests avail abl e in the l iteratur e. IIA. Local buckling Conservative Conservat Ive N-284 def ines I ocal buckl ing as buckling of shel l pl ate between stI f f eners or bound-ar l es. PSAR considers local buckling to be buckling associated with edge ef fects at a gross structural di scont i nui ty. C 3= 0 3 rt (D -~.8. CO N NN

 #                                                                                                              9                                                                            e

O O O i CONTAINE NT VESSFi - BtKKLIPO CRITERI A COMPARf SON ITEM PSAR CRITERIA _. CODE CASE N-284 COMENTS llAl. Unst i f f ened cy I Indr I-cal shelis and al so cyIinders between ring stiffeners l l Al a. Axial compression More conservative Conservative Both criteria use the same f ormula for calculating the classical buckling values. Hoeever, for cylinders between ring stif f en-ers typical of the QERP vessel construc-tion, the capacity reduction f actors used

!                                                                                                      In the PSAR criteria are more conserva-tive than those used in the N-284 criteria.

a llAlb. O Bending Conservative Methods of -1720 & The PSAR criterion uses higher critical

           $                                                                    -1730                  buckling stresses due to bending f or all N                                                                    (Conservative)          longitudinal compressive stresses which P                                                                                           can be represented by the first or higher

• N Fourier harmonics and are not included in T the axial compression. The methods of

                                                                                                       -1710 treat the peak stress as the uniform
           $                                                                                           stress. However, for a cylinder between a                                                                                                       ring stif feners typical of the OERP vessel construction, the critical buckling i                                                                                                       stress f or bending in the PSAR criteria is I                                                                                                       comparable to the critical buckl Ing stress f or axial compression in the N-284 criteria j

because of the higher critical axial stress allowed in the N-284 criterla, y j O> llAlc. S$ Radial pressure Conservative More conservative Although formulas f or circumferential com-

   -3                                                                                                  pression due to radial pressure are per-4   ~.                                                                                                  altted In the PSAR criteria, they are not

yy used In the actual design. Instead, the f ormulas for hydrostatic pressure are used mm i which are more conservative. For circum-ferential compression alone, the N-284 criteria are more conservative than the PSAR criteria.

4

CCNTAINMENT VESSEL - BUCKLING CRITERI A COfGRISON ITEM PSAR CRs!!R 1 _ CODE CASE N-294 COFNE W S liAld. Hydrostatic pressure Pbre conservative Conservative for hydrostatic pressure alone, the N-284 (external pressure criteria are less conservative than the u lth end pressure PSAR criteria. The use of hydrostatic Included) pressure case for radial pressure case in the PSAR criteria makes the PSAR criteria more conservative overall In dealing wIth the circumf erent ial compression. I l Al e. Shear / torsion Conservativ e More conservative The increased critical buckl ing values f or shear over tcrsion as permitted in the FSAR criteria are based on the recommendation of Reference 2. The critical shear stresses used in the N-284 criteria are mainly based on the torsional consideration, theref ore,

 @                                                                                            they are more conservative than the PSAR criteria.

{ N O il Al f. k Stiffening effect Not used Not defined Although the PSAR criteria include formulas f c3 due to internal pressure to account f or the stif fening ef fects due to Internal pressurization, these formulas O are not used because they could resul t in unconservativa design. The N-284 criteria do not prov ide spect f Ic f armul a f or these st if f ening ef fects, llA2. Double curvature shell s Conserv ative Conservative PSAR criteria are based on criteria given in (For items detined) Welding Research Council (WRC) Bulletin No.

69. f+-284 criteria do not prov ide suf f I-k@ clet guicelines to allow comparisons of f@ct all relevant items, e .

e co N NN Reference 2: "Gulce to Stabil ity Design Criteria f or Metal Structures", 3rd Edition, Structural Statsil ity Research Council, Edited by B.G. Johnston.

     #                                                              9                                                                         e

I v COPITAIPoENT VESSFt - BtKKl_lPC CRITERf A COM)ARISON ITEM PSAR CRITERIA CODE CASE N-284 COM O(TS IIA 3. Thermal stress Conservative Conservative Both sterla treat thermal stresses as r smary buckling stresses which is conserva-tive. The PSAR criteria use material yield-j Ing rather than Imperfect shell capacity as critical buckling stress f or thermal loading. This approach of using yield stress for thermal buckling is based on analyses by BOSOR4 and test data (see Ref erence I under item 18). IIA 4. Membrane stress at Conservative Conservative Both criteria recognize the f act that the 4- st r uct ur al di s- use of peak menbrane stress at structural o continuity discontinuity as uniform buckling stress n is overly conservative. Therefore, both

                                                     $                                                                                                                                                                                           criteria use the stress at /Rt (where N                                                                                                                                                                                           R = shell radius and t = shell thickness)

} P away from discontinuity as the basis for N providing rel lef f ra. the excessive con-

g

;                                                     e                                                                                                                                                                                          servatism.

cn 1IA5. Curved panels between The following evaluations are based on a 4 rir.gs & stringers typical curved panel of the OERP vessel which is located between the upper and lower

}

crane girders and also between two gusset pl ates. I I A5 a. Axlal compression Conservative More c.onservative N-284 criterla allow a lower critical buck-IIng stress than the PSAR criteria.

RR nm

                                                    *3 I

w .c1 l' @ CO N NN i k

CONTAINMENT VESSEL - BL'CKL IPC CRITERI A COMPARf SON ITEM PSAR CRITERfA CODE CASE N-284 COFNENTS llA$b. Circumf erent ial com- More conservative Conserv ative N-284 cr iteria allcw a higher critical pression due to buckl ing stress than the PSAR criteria. hydrostatic pressure liA5c. Shear Conserv atis e More conservative N-284 cr iteria al low a lowe- critical buckl Ing stress than the PSAR criteria. (18 Rings & stringers Conservat ive Conservative Both criteria use f ormulas which are suf f i-ciently conservative to ensure that the shel1 plate or the shel1-stringer combina-tion will buckle bef ore the shell-ring ,l cunbination can buckle, in addition, N-284 c criteria require 20% higher margin for the Q m ring and/or stringer buckling than for the N shell pl ate buckl ing between the sti f f eners g or supports. N cn e Ill. Interaction equations

   @  for local buckling Illa.

Cy l indr ical shel l s The f ollowing discussions on interaction equations f cr cyIIndrical shells are l imited to elastic bucki ing only. BilA1. More conservative Conserv ative The PSAR criteria use a lInear relationship Axi al compression for the combination of axial compression plus hoop com-pression (& al so bending) & hoop compression. The N-284 cr iteria use a l inear rel ationship o3 between the pure hoop compression case and O 9 the hydrostatic pressure case and a square F$ term f or hoop ccrnpression f or the rest of

  ,CL e                                                                                         the axial and hoop combinations, co y NN
  #                                                     G e

l (" O' t i 4 1 CONTAlteENT VESSEL - 9tKXLING CRITERI A COWARISON 4 l ITEM PSAR CRITERIA CODE CASE N-284 CO60ENTS lilA2. Axlal or hoop com- Conserv ative Conserv ative Both criteria use a square term for the pression plus shear shear. j tilA3. In combination w ith Conservative Conservative The PSAR criteria use a square term for ] thermal the thermal while the N-284 criteria treat the thermal stress the same as pressure or i mech anical load Induced stresses. Since the predominant thermal et fect is In hoop com-pression, the N-284 criteria basically i follow the same relationship as the PSAR

!                                                                                                   criterla f or most cases.

1118. Conservative Conservative Both criteria use the approach of WRC 69 m Double curvature shells as the basis for defining interaction N m eq uations. P IV. a i m

;  en              Factors of saf ety                                                               The f actor of safety Is def ined as the
!  E o

ratio of the critical buckling stress over the allowable buckl ing stress. .i IV A. Stringer buckling & Consers ative Pere conservative The PSAR criteria consider the buckling of l

$                   general Instability                                                              un st i f f ened or stringer sti f f ened shel l away fran the edge of fects due to discon-tInulties (supports and ring stiffeners)

) as general buckl Ing. The N-284 criteria classify the stringer and/or ring buckling as general buckling and require 20% higher

'                                                                                                   margin than the saf ety f actor for local o>                                                                                               buckling.

na a ON NN l, l

CONT A I PNENT V E S SEf_ - BUCKL I PC CR I TER I A COMPAR I SON ITEM PEAR CRITERIA CODE CASE N-284 COMMENTS I V B. Lccal buckting Con serv ativ e More conservative The PSAR criteria consider any shell buck-l i ng w Ith i n the eage of f ect s due to di s-continuities as local buckling while N-284 criteria consider the shell plate buckling between stringers and/or discontinuities as Iocal cuckiing, bw N N O b m a N O O> n5 dO 3 Q., W= D CO N NN

• O e

                                   . _ _ _    _ . ~ . . - - -.    .    - . - --      . - - - - .        .-~ . - - - - - . - - - - . .- - . . . . .       . - - -
 \

4 k

 ?

i 5

                          '6.0   EV ALUATION OF KEY DIFFEREN S; PSAR TO 1980 AS E ODDE                                                                            ,

The folicating evaluation pertains to key items identif led in Sections 2.0 through 5.0. . (See table of content Index) 4 j As shown its this section, the 1980 Code criteria appear to be more conservative than the DBRP criteria in some areas and less conservative In other areas. The Intent of the 1980 -i Code has been Implemented by the DBRP criteria given In the PSAR. Evaluation of these dif forences reveals that the existing contalrunent vessel design is adequate and provides a level of safety equivalent to that which would resul t f rom complete appl ication of the 1980 Code and Code Case N-284 1 i

 !                                                                                                                                                                 i i

e i t i O I O LA N N C. 3 1, LD t l ! ~  ; i  ! i i i i  ! e I 1 RE nm g r D CO N } i i ( i  !. ! 1

CONTAltNENT VESSEL - EVALUATION OF DIFFERENCES BEThEEN 1974 EDITION

                                        & 1980 EDITION OF ASME B1PV CODE ITEM NO. & PARACRAPH/ TITLE                                   EVALUATION The f ollow ing items pertain to NE-3000 as l isted in 3.0.3:

8. NE-3133 Component Under External External pressure can only occur f or a shor t Loading period of time late in an accident transient.

Theref ore, the loading combination given in t h e PS AR ( Sect i on 3.8 ) is overly conservative in that it specif les the combination of external pressure w ith an earthquake. The vessel as currently designed satisf les the design f ormulas f or external pressure pl us dead and live loads used in NE-3133 of the o 1980 code for Level A Serv ice L imits. O cn N 11. NE-3213.10 Local 6 -Imary The 1980 Code requires that discrete regions Membrane Stress in the vicinity of brackets, in which the local pr imary membrane stresses exceed 1.1

  $                                                     Smc, shal I not overI ap. Al though not required 8                                                   by the ASE Code, the 043RP containment vendor "m                                                    has f ollowed standard, conservative design pr act ice w h ich includes assuring that the local ef fects (e.g., primary membrane stresses greater than 1.1 Sy ) do not overlap.

Therefore, the exitfing containment vessel design provices a level of saf ety equivalent to that which would resul t f ran cornplete application of the 1980 Code. 14 Figure NE-3222-1 (1974 Code), The 1980 Code requires eval uation of secondary Figure NE-3221-2 (1980 Code) stresses for accidental pressure and op temperature within Level A Service Limits and oB within Level B Service Limits when WE is also f@cL combined. Secondary stresses occur at regions of discontinuity (including non-integral cD N connections). Secondary stresses are small and even if not small, would not af fect the overall saf ety margin f or a single accident event w Ith or w Ithout WE ef fects. O O e

         !                                                                   s                           %M
          \

CONTAltNENT VESSEL - EVALUATION Or DIFFERENCFS BETWEEN 1974 EDITION

                                          & 1980 EDITION OF ASK B&PV CODE ITEM NO. A PARAGRAPH /TfTLE                                EVALUATION

14. (Cont'd) The 1980 Ccde reclassified the operating conditions and required evaluation of primary membrane and bending stresses. Although the evaluation of this combination of stresses is not required in the 1974 Code, the normal practice of containment vendors including G81 is to consider all of these stresses. Such consideration was made for the GBRP vessel.

Theref ore, the current GBRP containment j vessel design provides a level of protection to publIc safety equivalent to that which would result f ran detail implementation of the 1980 Code. O The additional requirements of NE-3221.3(b)(1) Q N

16. NE-3221 Stress intensity val ue and (c)(2) in the 1980 Code impose iimits on l

l N O the primary general or local membrane plus

  -                                                       primary bending s' ess Intensity when the N                                                       primary general or local membrane stress s                                                     exceeds 67% of the ylet d strength at U                                                       temperature. For OBRP, bucki Ing considerations in general govern the design of the containment shell, and the primary general                   i membrane stress does not exceed 67% of the yield strength at' temperature. Theref cre, the existing containment design provides a level of saf ety equivalent to that which would be expected to result f rcm complete appl Ication of the 1980 Code.
            ?1 NE-3222 Operating Conditions               in the 1974 Code, E-3222 is used to specify                      l g              (1974 Code), Buckling Stress              requirteents f or operating conditions. In the                  j n@                                                       1980 Code, operating conditions have been                       I fg                                                       deleted f rom NE-3222 and merged with accident

a. conditions and put into service conditions.

[~ NE-3222 is now used exclusively for specifying co w buckling requirements f or the design by l NN analysis approach. See item 14 above for I evaluation of operating conditions. l

CONTAltofENT VESSEL _ - EV ALUATION OF DIFF EFINCES PEThEEN 1974 EDITION A J9E0 EDITION OF ASME BiPV CODE ITEM NO. & PARAGRAPH /TITtF EV ALUAT I ON

21. (Cont'd) Code Case N-284 is an expansion of the method of linear bif urcation analysis reduced by margins which reflect the dit f erence between theoret ical and actual load capacities (Imperf ectIon and pl astIcIty reduction f actcr s) as permitted by NE-3222.1(a)(2). By folIowing the spect f ic procedures and f actors of saf ety givea in the Code Case, a cylinder under external pressure load will result in similar allowables whether it is designed to the N-284 rules or the NE-3133 rules. For

  'g                                                    eval uation of dif f erences between N-284 and n                                                     the PS AR buckling criteria in addition to the
  $                                                     NE-3133 rules, see Section 7.0.

22. NE-3226 Test ing L imits (1980 Code) Stress l imits f cr testing conditions are N modif ied somewhat in the 1980 Code. This Code T requirement provides protection to the plant N and on-site per sonnel during the containment

   "                                                    test and assures that test perf ormance w il l not et f ect the containment vessel design f unc* lon. The CR3RP containment vessel design is controlled by consicerations other than the internal pressure imposed during containment t e st . The stresses to which the containment is subjected during testing are well within the existing allowables. Tlierefore, current cesign of the containment vessel provides a level of protection to publ le saf ety oy                                                       equivalent to that which would resul t f rom Qy                                                       detail implementation of the 1980 Code.

• 3 w.

C CO N NN

     #                                                                O                                  e

         \                                                               \_.)                             D CONTAlfeENT VFMrf - EVALUATION OF DIFFERENUS BETinFFN 1974 EDITION

, & 1980 EDITION OF ASE BAPV CODF ITEM No. 1 PARAGRAPH /TITIF EVALUATION

24. NE-3227.5 Narz te Piping Transition in the 1980 Code, substantial changes are made

,                                                         to the stress classifications in the nozzle
!                                                         piping transition either within the limits of reinforcement or beyond the limits of i                                                          reinf orcements. Although a quantitative 1

assessment of the of f acts of these changes would require additional detailed ! calculations, a qualitative assessment is provided as f ollows. Consistent with normal containment design practices, extroneIy conservative " enveloping" loads were specified for similar groups of penetrations. The vessel Including the nozzle piping transitions m were designed to accommodate these conservatively specified toeds using N simpi Ifled but conservative design approaches. m O This results in substantial design margin for a large fraction of the penetrations. For j m those penetrations in which design margin Is O W not obvious, It is juJged that the use of detailed analysis approaches would demonstrate that the current design has a capability that approaches, if not meets, the specif ied limits 4 for the load combinations identified in the 1980 Code. Thus, the current design is judged to provide a level of safety equivalent to that which would result f rom full Implementation of the 1980 Code.

27. IE-3232 Operating Conditions in addition to requiring secondary stress 1 (1974 Code), Combined Loads evaluation for accident conditions, there are (1980 Code) also changes to the stress limits for bolts k$

r+rD In the 1980 Code. The only bolts covered by this provision of the 1980 Code would be the 5,, bolts that secure the ref uelIng hatch cover In g* place on the containment shell. The only co w significant operating load is the deadloed NN from the mass of the hatch cover itself. The only signifIcant mechanical load is that

;                                                         resulting from an earthquake. The calculated stresses for the bolts unde. these conditions are well within the specified allowables (1974

CONTAIPMENT VESSFt - EVALUATION OF Of FFERENCES BETWEEN 1974 EDITION

                                         & 1980 EDf TION OF ASPf B&PV CODE fTEM NO. 1 PARAGRAPH / TITLE                              EVALUATION

27. (Cont'd) Code version) for these loads and it is believed that the bolts would readily meet the requirements of the 1980 Code. Therefore, the current design provides the same level of safety as would result f rm implementation of the 1980 Code.

34. NE-3324.11 Nozzle Necks (1974 Code), The requirement of NE 3300 would not be NE-3324.12 Nozzles (1980 Code) mandatory for the GBRP vessel. These requirements are applicable only when the internal pressure loading represents 90% or more of the total loading.

The additional requirement in the 1980 Code, limiting the allowable shear stress to 70% of Q m the allowable tensile stress in the nozzle neck, would require review of design y cal cul ations. Namal containment vendor o practice limits shear stresses to less than 50 6 m percent of the allowable tensile s;ress in the nozzle neck. Therefore, it is judged that the O m current design provides a level of protection to public health and saf ety equivalent to that which would result f rom implenentation of the 1980 Code.

37. PE-3331 General Requirements The requirement of NE 3300 would not be for Openings mandatory for the GBRP vessel. These requirements are applicable only when the Internal pressure loading represents 90% or gd o map of the total loading, c+ co

 . s a                                                     The. changes in the 1980 Code regarding the G*                                                       opening and reinforcement would require review co N                                                     of design calculations. However, there are no NN areas fw the GBRP where the controliIng Iond combinations Ir.clude signif icant contribution from Internal pressure. Therefore, this 1980 Code provision would not be appl icable to GB RP.

G G e

O J O CONTAltNENT VESSFl - EV ALUATICN OF DIFFERENCES BETWEEN 1974 EDITION A 1980 EDITION OF ASME B&PV CODE ITEM NO. 1 PARAGRAPH / TITLE EV ALUAT I ON

39. NE-3334.1 Limit of ReInf orcement See evaluation of item 37 above.

along the vessel wall

40. NE-3336 Strength of Reinf orcing See evaluation of item 37 above.

Material

41. NE-3367 Closures on small Dimensional standards are updated in the 1980 Penetrations Code.. This optional provision applles only to penetrations of 2" diameter or less. The smallest penetration in the OERP containment vessel is da in diameter. Theref ore, this provision is not applicable to OERP.

42. NE-3720 Design Rules The largest, and theref ore the most critical, o mechanical penetrations are designed for 30 psi. The electrical penetrations have been Qto designed with even greater margin and are y.

tested at 50 psi. Theref ore, pressure capability is signif Icantly greater than the 10 psi UERP cesign pressure and tens of times

@ e                                                    greater than the actual accident pressure of d                                                      less than 2 psl. Theref ore the existing design provides a level cf safety equivalent to that which would result f rom complete application of the 1980 code.

o '> r+ a G' l5M

CQtiTAltt!ENT VESSEL - EV ALUATit6 0F DIFFEPEPCES PETWEEN 1974 EDITION M980 EDITION OF ASME B1PV CODE ITEM NO. 1 PARAGRAPH / TITLE __ EVALUATION Tne f ol l ow Ing I tum pertains to NE-7000 as l isted in 3.0.7:

1. tee-7000 Protect ion against Over- Purchase of standard appurtenances only may be impacted. There Pressure (1974 Code), Over pr essure is no ef fect on the containment vessel design.

Protect ion (1980 Code) The f ol low Ing item pertains to CC-3000 as l isted in 4.0.2:

1. CC-3000 Design Tte changes in the 1980 Code impose modif ied roquirements

                                                                            = ith respect to stress and strain in reinf orced concrete design and al so to rebar detail s. The modif ied requirements ao not represent a signif icant departure f rom the design approach in the 1974 edition of the Code. Although the cher:ges generally 43                                                                           result in the 1980 Code teing more conservative than the 1974 Code,

[2 the changes are relatively minor, it is bel ieved that If there hJ were areas of the existing CRBRP containment which have not al ready [$ been demonstrated to accommodate specif ied loads w ithin the revised criteria, the use of r, ore refined analysis techniques tn woul d be expected to demonstrate that the existing design represents i the capabil ity which approaches that dictated by the 1980 Code. oc Theref ore, it is judged that the current CRBRP design embodies a level of saf ety equivalent to that which would result f rom detailed impl ementation of the 1980 Code. c) 32 O 3 a O) %4 h) NO O O e

I 7.0 EV ALUATION OF KEY DIFFERENG; PSAR TO (DDE CASE N-284 l Code Case N-284 criteria appear to be more conwrvative than the PSAR criteria In sme areas and less conservative in other areas. The Appl icant has done an analysis of the critical region using N-284 criteria and appropriate loads. This analysis results in a ! design margin in excess of that required by N-284. i i i l ) C o U3 1 N N i O E N I w i '

.  ~

$ D f l l t i, i em3 C1 i we ' Co N NN i i e

CONTAINMENT VESSEL - EVALUATION DF DIFFERENCES BETWEEN PSAR

                                                                            & CODE CASE N-284 ITEM                                                                          EVALUATION The f oll ow ing item per tains to Code Case N-284 As L isted in 5.0:

1. PS AR Uuckl ing Criteria (Expanded f rom Changing the buck! !ng criteria f rom the PS AR criteria to the N-284 1974 Code), Code Case N-284 (Publ it hed in criteria would require reanalysis of the shell f or buckl ing eval uation.

1980). The most critical region (just above the operating deck) has been analyzed using 80SOR4 computer program and selected critical l oadi ng combinations in order to verif y the design edequacy of the CR3RP containment vessel against the N-284 b ..nl ing cr i teris. Since the 80SOR4 analy ses do not incl ude shear stresses, the ef f ects of shear stresses are manually adjusted using the procedures of N-284 criteria. The results of these analyses show that the design margins of the vessel g) exceed the requirements of N-284. o

 $3                                                                           This analysis was perf orned using the PS AR load combinations f or SSE ha                                                                           and OBE, setting Pe = 0 (External pressure is not postulated to CD exist during a seismic event, see Section 6, item 8) and was no                                                                           eval uated against the AStE Level B and C service l imits. The S"                                                                           B050R4 model is as shown in Figure 7.0-1.       In the analysis f or C3 C3 each l oad combination, pre-buckl ing stresses are determined f rom act ual hoop and axial stress distributions empi lf led by their corresponding imperf ection knockdown f actors. The results of this analysis yield an eigenvalue corresponding to the buckling of the critical reglon just above the operating deck.) The lowest buckl log mode is shown in f igure 7.0-2.

(continued on next page) c3 22 O 9 f' e Co Na hJ r0 9 9 e

             -_. - ~ - . _ __. _                        . -   . _ _ _ . . < . - - . .

_ - _ ~ . . . . _ . . _ . . - O o 4 J 4 I CONTAINENT VESSEt. - EVALUATION 1 0F DIFFEREICES BETWFEN PSAR

'                                                                                             A (11DE CASE M-?84 1

___ ITEM EV AttlAT ION l ____ j The f ollow ing itm pertains to Code Case N-284 As Listed in 5.0: i l 1. PSN1 Duckling Crliert a (Expanded f rom Since BOSOR4 analysis does not include shear, the results are i 1974 Code), Code Case N-284 (Publ Ished in adjusted to determine f actors of saf ety including shear stresses, i 1980). (Contlued) The adjustment is performed in accordance with the following f ormula:

                                                                                                      &# * [F5          "g_   2 ,

{" g "toe th ~ 3 WilERE A = EIGEN VALUE FROM B050R4 ANALYSIS c Q N

                                                                                                                  = Site AR STRESS g

a = KNOCKDOWN FACTOR FOR Site AR (=0.712. EL. 816' TO EL. 839') { *

m " net

                                                                                                           ,    = CLASSICAL BUCKLING VALUE UNDER SHEAR LOAD BASED ON N-284 FORNULA j   f                                                                                                           IS= FACTOR Or SAFETY co                                                                                                                  RESULTING FACTOR OF SAFETY

) - I

!                                                                                                        LOAD (XWeINATION              PSAR           N-284 I

D + L + T' + P, + SSE 1.9 1.67 I D + L + T ' + P, + (B E 2.5 2.0

    ?$

rea 40 TN NN 1

s ] m m m m r r r r-i m m m m

                   -                     u                        m                   y cn-                   e-                       cn-                 m
                                                       -Qw_

2 m.- oe m w-m wm nz M.W "3 m BR t

,   o a   O m                     --                   z A, m

e 1 O C O m :u :o N rr1 C N l  ?  ? 8 i 1

     ~

U1 o I R r 8 w

l y

N i. i1 l i

}

i 276" 204" 228" =

= = = =

70 E SH POINTS 52 ESH POINTS 58 ESH POINTS i RF

  \
    ~. 3 I        a i    Co N NN a

l . i i i i t

,1                                                                                                       h 4

1 I i i

l O l EL 875' EL 856' [ EL 839' l CIRCUMFERENTIAL

                               '                   WAVE N0.=21 l

l El 816' B0SOR4 BUCKLE MODE RADIAL DISPLACEMENT FOR DL + LL + SSE + THERMAL + Pg i FIGURE 7.0-2 Amend. 72 Oct. 1982 QCS220.23-83

cuestion CS421.9 Identify where instrument sensors or transmitters supplying information to more than one protection channel, to both a protection channel and control channel, or to more than one control channel, are located in a common Instrument Iine or connected to a common Instrument tap. The intent of this item is to verify that a single f ailure in a cmmon instrument line or tap (such as break or blockage) cannot defeat required protection system redundancy. Resoonse: Recundant protection channel s, protection channels and control channels, or / more than one control channel instrumentation sensors er transmitters are not located in common instrument I ines or taps. Theref ore, the required protection system redundancy will not be defeated by a blockage or breakage of en instrument i Ine or tap. 3 O l l l l O QCS421.9-1 Amend. 72 Oct. 1982

p OuestIon CS421.17 The inf ormation suppl led f or remote shutdown (PSAR Section 7.4.3) from outside the control room is insufficient. Theref ore, provide f urther discussion to describe the capabil ity of achieving hot or col d shutdown f rom outside the control room. As a minimum, provide the f ol low ing inf ormation: a) A tabl e I isting the control s and displ ay instrumentation required f or hot and col d shutdown f rom outside the control room. Identify the train assignments f or the saf ety-rel ated equipment. b) Design basis f or selection of instrumentation and control equipment on the hot shuidown par.el . c) Location of transfer switches and the remote control station. d) Description of transfer switches and the remote control station. e) Description of isolation, separation and transfer / override provisions. This should incl ude the design basis f or preventing electrical interaction between the control room and remote shutdown equipment. f) Description of control room annunciation of remote control or overricdon status of devices under local control . g) Description of compl iance with the staf f's Remote Shutdown Panel position. v Resoonse: The response to this question is provided in the amended text fcr Section l7.4.4. O QCS421.17-1 Amend. 72 Oct. 1982

,m Ouestion CS421.22 s' The information supplied in PSAR Section 7.5 concentrates on the information and monitoring systems but does not provi je suf ficient information to describe safety-related display instrumentation needed for all operating conditions. Therefore, please expand the PSAR to provide as a minimum additional information on the following:

1. ESF Systems Monitoring

2. ESF Support System Monitoring

3. Reactor Protective System Monitoring

4. Rod Position Indication System

5. Plant Process Display Instrumentation

6. Control Boards and Annunciators

7. Bypass and inoperable Status Indication

8. Control Room Habitability Instr ,entation

9. Residual Heat Removal Instrumentation

Response

This response describes safety-related display information available to the operator in the control room. Display instrumentation provided for ESFs is described below. Section 7.3 will be revised to include this information. The instrumentation for monitoring ESF support systems are described in the Indicated sections of the PSAR: HVAC-7.6.4; Plant Service and Chilled Water Systems-7.6.1; Diesel ('")T ( ,_ Generator-8.3.3; Electric Power Systems-8.3.1.1.2, 8.3.1.1.5 and 8.3.2.1.1. The Reactor Protective Monitoring System is described in Section 7.2. Additional information about the display instrumentation has been inserted into Section 7.2 with this response. A description of the display instrumentation provided in the control room for operators for Rod Position Indication is provided in PSAR Section 7.7.1.3.2. Control Boards and Annunciators are detailed in Section 7.9. The Inoperable Status Monitoring System (including bypass monitoring) is discussed in PSAR Section 7.5.12. Section 7.4.1 discusses instrumentation and controls for the SGAHRS which is a part of the overall Shutdown Heat Removal system. Safetv-Related Disolav information for MSF Svstems Reactor Containment Buildina Annulus Filtration Svstem Monitoring, including Indications and alarms, is provided in the control room for the following parameters for each of the redundant trains:

a. Annulus filter fan discharge flow;

b. Annulus pressure maintenance fan discharge radiation;

c. Annulus filter unit inlet radiation;

d. Annulus filter unit relative humidity;

e. Annulus differential pressure (3 monitors for each train);

(alarm only);

f. Annulus discharge to atmosphere, radiation (2 monitors for each train);

I_s, g. Fan vibration (alarm only);

\~>) h. Filter unit leaving air temperature (alarm only);

QCS421.22-1 Amend. 72 Oct. 1982

1. Individual component differential pressure (alcrm only);

J. Filter unit differential pressure (alarm only). f Status of the following equipment is provided in the control room for each of the redundant trains: Annulus Filter Fan Annulus Pressure Maintenance Fan Annulus Filter Fan Discharge Damper Annulus Pressure Maintenance Fan Discharge Damper Annulus Filter Unit Recirc. Air Damper RSB Filtration System Monitoring including indications and alarms is provided in the control room for the following parameters for each of the redundant trains:

a. RSB cleanup filter fan discharge flow and radiation;

b. RSB cleanup filter train leaving air temperature (2 monitors in each train);

c. RSB cleanup filter unit inlet flow;

d. Fan vibration (alarm only);

e. Individual component differential pressure (alarm only);

f. Filter unit tifferential pressure (alarm only).

Also non-safety-related indications and alarms are provided in the control room for radiation detection in the roof air exhaust discharge. Status of the following equipment is provided Ir, the control room for each of the redundant trains: RSB Cleanup Filter Fan RSB Cleanup Filter Fan Discharge Damper RSB Cleanup ritter Recirc. Air Supply Damper RSB Cleanup Filter Recirc. Discharge Demper RSB Cleanup Filter Normal Exhaust Camper Control Room Habitability System Monitoring inclLding indications and alarms is provided in the control room for the following parameters for each of the redundant trains:

a. Main air intake radiation (control room outsice air);

b. Remote air intake radiation (control room outsico air);

c. Mixed air temperature (2 monitors in each train);

d. Control room A/C unit supply air flow;

e. Control room A/C unit discharge air temperature (2 monitors in each train);

f. Toxic gas in main air intake (elarm enly);

g. Toxic gas in remote air' intake (alarm only);

h. Smoke in main air intake (alcrm only);

i. Smoke in remote air intake (clarn only);

J. Filter unit air flow;

k. Fan vibration (alarm only);

QCS421.22-2 Amend. 72 Oct. 1982 I

1. Filter unit leaving air temperature (alarm only);

(]/ f m. n. Individual filter unit components differential pressure (alarm only); Filter unit differential pressure (alarm only). Status of the following equipment is provided in the control room for each of the redundant trains:

a. Control Room A/C Unit

b. Control Room A/C Unit Discharge Damper

c. Control Room A/C Unit inlet Damper

d. Control Room A/C Unit Supply Air Damper (two for each train)

e. Control Room Filter Unit Supply Fan

f. Control Room FIIter Unit

g. Control Room Filter inlet Damper

h. Control Room Filter Unit Supply Fan Discharge Damper Guard Vessels. Cell Liners and Catch Patui No Instrumentation is required as none is provided for ESF guard vessels (for the reactor and the primary heat transport system), cell liners and catch pans.

Steam Generator Building Aerosol Release Mitigation System Instrumentation and Controls The Steam Generator Building (SGB) Aerosol Release Mitigation System is designed to control the release of sodium aerosols from the Steam Generator v Building in the event of a design basis leak in one of the three IHTS loops. The functional design of this system is discussed in Section 6.2.7. The following instrumentation is provided in the main control room for the SGB Aerosol Release Mitigation System. Main Control Room Instrumentation for the Steam Generator Buildina Aerosol Release Mitigation System

a. Aerosol Detector Alarm Indication

b. SGB Loop #1, #2 and #3 Dampers Position Status Indication

c. RCB Supply And Exhaust Fans Common Alarm

d. RSB Dampers and CB isolation Valves Position Status indication

e. CB Dampers and CB isolation Valves Position Status indication

f. SGB-lB Damper Position Status indication

g. RSB-RWA Supply and Exhaust Fans and the Exhaust Filter Fan Common Alarm

h. ABHX Intake end Exhaust Dampers Position Status indication I. SGB-MB Outside Air Damper Common Alarm J. DGB Inteke Tcrnado Damper and Outsice Air Damper Position Status Indication O

V QCS421.22-3 Amend. 72 Oct. 1982

Ouestion CS421.27 In the PSAR Section 7.3, the statement is made that the Initiation of containment isolation is the only Engineered Safety Feature (ESF) ident! fled which requires a description in this Section. Chapter 6 of the PSAR denotes several systems (Annul us Filtration System, Reactor Service Buil ding Fil tration System, and the Residual Heat Removal System including SGAHRS and OHRS) in addition to the Containment Isolation System as being part of the ESF - Sy stem. Justify why these systems aren't included in Section 7.3 of the PSAR. Al so, the staf f bel leves that the Sodium-Water Reactor Pressure Rel lef System (SWRPS) should be classified as part of the ESF System. Describe the actions to be automatically initiated or to be init!ated by operators to mitigate sodium-water reactions. The discussions should include actions necessary to protect publ ic saf ety or avoid an unanalyzed pl ant upset. Resoonse: Section 7.3 as modif ied by Amendment 71 provides a cross-ref erence to PSAR Section 6.1 which identifies Engineered Saf ety Features (ESFs) and the sections of the PSAR where they are discussed. Additional information is provided in the response to NRC Question CS421.22. The Sodium / Water Reactor Pressure Rellef System's (SWRPRS) safety function is accompl ished by the mechanical actuation of the rupture discs by pressure generated f rom a sodium / water reaction occurring in a steam generator module (ref. PS AR Sections 5.5.2.4 and 5.5.2.6) . Subsequently, the SWRPRS Instrumentation and control has two f unctions. These two electrical f unctions have dif ferent saf ety consequences, and therefore, one is ci assif led as saf ety-rel ated, and the other as non-saf ety-rel ated.

1) Saf ety-rel ated instrument and control function: Actuation of the SWRPRS is detected immediately downstream of the rupture discs (ref. PSAR Section 5.5.2.4). A saf ety-related (Class IE) signal resul ting f rom the sensors, is transmitted to the PPS (ref. PSAR Section 7.2.1.2.2). This initiates a reactor trip and is part of the Plant Protection System. As stated in Section 7.5.6.2 th is compl ies w ith requirements stated in Section 7.1.2 and 7.2.2.

2) Non-saf ety-rel ated instrument and control functions: A buffered signal initiates actions as described in Section 7.5.6.1.2. Since these actions only isolate the loop af fected, the ability of any other loop to ranovo decay heat f rom the reactor is not compromised. Therefore, these f unctions are not considered saf ety-related.

For automatic and operator actions in case of sodium / water reactions, see Sections 5.5.2.8, 7.5.5.3, and 7.5.6. O QCS421.27-1 Amend. 72 Oct. 1982

ouestion CS421.30 To extend our review, the staf f (ICSB & ECM) each require a set of one l ine l&C Draw ings f or the safety related CRBR systems. Drawings should al so be provided that indicate the separation used in the CRBR design. Resoonse: The NRC Staf f in a telecon with the Project on 9/13/82, confirmed that the requested information is currently in their possession. 1 O O QCS421.30-1 Amend. 72 Oct. 1982

Question 421.31

• Address the adequacy of the Reactor Vessel Level gauges with emphasis on the lack of diversity, the level range chosen, the method selected, and the ef fects of temperature on the level accuracy. Provide this same discussion f or the l evel probes in the sodium expansion tank, the sodium dump tank, and the sodium pump tank. Al so, discuss provisions made f or sodium level measurements in the intermediate system.

Resoonse: 1 Mutual inductance type sodium level probes are used for all continuous sodium level measurements Ir. the reactor vessel, sodium expansion tank, sodium dump tank and the sodium pump tenk. This type of level probe has been shown to be superior to other types of level probes during sodium testing of various types of level probes. Other types of level probes which were eval uated ir, this test program include baianced bridge type inductive Ievel probes, di spl acer-fl oat type l evel transducers, del ta P type l evel transducers and time domain reflectemetry transducers. The advantage of using highly relIsble mutual inductance type probes outweighs any advantage that could be obtainea f rcm type diversity. The mutual inductance l evel probe has a primary and secondary inductance coll. Excitation is appi led to the primary coil which develops a signal in the secondary coll. The signal magnitude in the secondary coil is dependent upon the height of the sodium. To compensate for sodium temperature changes a temperature compensation circuit is integral with the signal condition equipment and works on the concept of resistance changing with temperature. The compensation circuit measures the voltage and current in the primary coil and evaluates changes to-determine the resistance change and automatically adjusts the output of the i signal conditioner based on the resistance change. The reactor vessel centains f our narrcw range probes, three of which are used by the Primary Reactor Shutdown System, and two w ide range probes which are designated to the part cf the Accicent Nonitoring (AP) System. The measurement range chosen f or the narrow range probes (30 inches) is based on a range which is wide enough to cover the normal operating ranges of the sodium l evel in the reactor vessel but is narrcw enough that the uncertainty associated wIth the measurement is minimized. The .noasurement range chosen f or the w ide range probes (189 inches) is based on the abil Ity to monitor the sodium level down to the level of the reactor vessel outl et nozzles. Each Primary pump contains two redundant wide range probes (80.5 inches) to monitor sodium Ievel over the f ulI elevation of the pump tank. O O Amend 2 QCS4 21.31 -1

Sodi um l evel measurement is accomplished in the InNrmediate system via the sodium pump and expansion tank, the intermediate sodium pumps have a single w ide range probe (86.9 inches) installed in the pump tank which monitors the f ull range of the sodium level in the pump tank. Two l evel probes are installed in the sodium expansion tank, a wide range probe to measure the f ulI range of anticipated steady state and transient sodium level s in the tank and a narrow range probe f or accuracy during f il l of the system. The wide range level probe in the expansion tank al so provides a signal f or a h igh and l ow l evel al arm. The pump tank l evel probe provides a signal f or a h igh and l ow level al arm, and isolation of lHTS argon cover gas system. Two w ide range l evel probes are installed in each sodium dump tank. These probes are arranged w ith overl ap to prov ide f or monitoring sodium level s during sodium f il l and drain operations of the Intermediate Heat Transport Sy stem. O O QCS421.31 -2 Amend. 72 Oct. 1982

Ouestion CS421.34 PSAR Section 7.5.2.1.2 states in part that a signal is provided to the control room Indicating that the pony motor is running. The staf f requires more information with regard to the CRBR pony motor instrumentation end control sy stem, in particul ar, the initiation signal s f or the pony motors, manual initiation capabit Ity, qual if Ications f or the system, and the design criteria f or the system shoul d be di scussed. PS AR Section 7.5.6.1.1 states in part that the sodium pony motor is tripped upon a large leak detection. Discuss the safety aspects of this trip and provide the staf f information on other signal s that w ilI trip the pony motors. Resoonse: The pony motor runs continuously during all modes of plant operation except during sodium pump or drive system maintenance. Therefore, there is no need f or automatic or manual initiation signal s except f or the start-stop sw itch. Normal pony motor start is through a permissive sequence circuit which starts the external lubricating oil cooling system and high pressure lube oil pump, and when the oil system achieved flow and pressure the pony motor starts. Once started the loss of flow or pressure will not result in a pony motor trip. This method of starting is not cl assif ied as saf ety-related. in the saf ety-rel ated mode, pony motor operation does not require the use of the external lubricating oil cooling system or high pressure lube oil pump. p This f unction is carried out by a start-stop switch on the main control panel ( in the control room. The non-saf ety permissive sequence starting circuit is isolated f rom the saf ety circuit and w ill not prevent the operation of the saf ety f unction. The saf ety circuit w ill be qual if ied per WARD-D-0165 (Ref.13 of PSAR Section 1.6). There is avail abl e in the control room, pony motor speed and current Indications. Pony motor current indication is provided via the PDH&DS. These circuits are non-saf ety rel ated. The only condition which results in an automatic IHTS pony motor trip (the PHTS pony motor is not tripped) is a large sodium / water reaction which results in a rupture dise rupturing. The saf ety aspects of thIs trip are specif Ically addressed in the response to Question CS421.27. O QCS421.34-1 Amend. 72 Oct. 1982

Question CS421.36 Provide a more detailed discussion of the CRBR Leak Detection system and how it meets the provisions contained in the Light Water Reactor Regulatory Guide 1.45. The discussion should include detection methods, detector sensitivity, detector response time, signal correlations and calibration, seismic qual if Ication, testabil ity, and the provisions for technical specif Icaticos. Resoonse: PSAR Section 7.5.5.1.1 has been revised to provide a more detailed discussion 4 of the CRBRP Leak Detection Instrumentation System. A comparison to the provisions of Regulatory Guide 1.45 is contained in Section 5.3 of WARD O-185,

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

Technical Specifications will be developed at the FSAR stage. The Technical Specification wIlI require that the piant wilI be placed in either the hot shutdown or ref ueling condition if there is a confirmed leak in either the primary or Intermediate heat transport system. O d i l 1 O QCS421.36-1 Amend. 72 Oct. 1982

Ouestion CS421.37 Discuss the provisions made for alarming a zero or negative dif ferential pressure (PS AR Section 7.5.5.2.1) as to sensor type, location, setpoints, testabil ity, and annunciation.

Response

The Intermediate loop pressure to primary loop pressure is maintained at presssures greater than 10 psi. When the pressure on the intermediate loop drops to within 10 psi of the primary loop, the operator is alerted by an al arm. The alarm is on a positive pressure dif ferential and not zero or negative pressure dif ferential . Each instrument channel includes provisions for Insertion of a test signal on the sensor side of the signal conditioning electronics. The sensor type, locations, setpoints and annunciation are described in PSAR Section 7.5.2.1.1. PSAR Pages 7.5 -7, 7 .5 -8, 7 .5 -27 have been modified for 4 cl ari f ication. i i j l 4 i I QCS421.37-1 Amend. 72 Oct. 1982

Ouestion 421.42 b V Section 7.1.2 and 7.2.2 of Chapter 7 of the PSAR ref erence the use of IEEE standards. Other sections in Chapter 7 rrake reference to Section 7.1.2 but do not identify specific IEEE standards which were implemented in the system design. Justify why Section 7.3 through 7.7 of the PSAR do not provide enought information to determine whether the IEEE standards are implemented in the design. Resoonse: Chapter 7 has been revised to add specif Ic identification of IEEE standards when appropriate as described below. Compl iance w ith IEEE standards f or non-saf ety related systems is not required and therefore use of IEEE standards f or thcse systems Is not discussed. Section 7.2 - This section is amended to clarify the use of IEEE standards. Section 7.3 - This section is amended to clarify the use of IEEE standards. Section 7.4 - This section is amended to clarify the use of IEEE standards. Section 7.5.1 - The Wide Range and Power Range Flux Monitors discussed in this recticn are saf ety related, the IEEE standards of Table 7.1-3 are appi led to the designs. p Section 7.5.2 - Addresses the types of f unctions and the sensors used in the plant and does not spectfIcally identify these instruments as saf ety rel ated or not. Tabl e 7.5-1 identif ies the variables which are saf ety rel ated as does Section 7.2. Paragraph 7.5.2.2 states that the instruments which are a part of the Protection system comply with the requirements of Section 7.1.2 and 7.2.2 which encompasses the IEEE standards I isted in Tabl e 7.1-3. Section 7.5.3 - The sodlum level probes discussed in this section are IE. The remaining Instrumentation is non-1E. Section 7.5.3.2 states that the sodium level probes are part of the Reactor Shutdown system and will comply with PPS Design Requirements 4 (Sections 7.1.2 and 7.2.2) . The probes, theref ore, wil l comply w ith IEEE standards icentif ied in these sections as appiIcabie to PPS. Section 7.5.4 - The Failed Fuel System is not saf ety related. SectIon 7.5.5 - The ieak detection systems discussed In thIs section are not saf ety rel ated. O QCS421.42-1 Amend. 72 Oct. 1982

Section 7.5.6 - SWRPRS instrumentation and control has two f unctions. One is to initiate a reactor trip, the other is to isolate the af fected loop. The reactor trip f unction is part of the Pl ant Protection system and as stated in 7.5.6.2 compi les w ith Sections 7.1.2 and 7.2.2. Isol ation of the af fected loop is not safety relatcd since it does not comprcmise the abil ity to remove decay heat f ran the unaf fected loops. Sections 7.5.7, 7.5.8 and 7.5.9 - The Instruments discussed in these sections are safety rolated, the IEEE standards of Table 7.1-3 are appl led to the designs. Sections 7.6.1, 7.6.2, 7.6.4 and 7.6.6 - These Sections have been revised to incorporate appl Icable IEEE Standards. Section 7.6.5 - The SGB Flooding Protection System is safety related and section 7.6.5 is amended to clarify the use of IEEE standards. Sections 7.7 and 7.8 - No IEEE standards are appl led in these sections since the systems described therein are non safety related systems. Section 7.9 - This section has been amended to clarify the use of IEEE standards. QCS421.42-2 Amend. 72 O Oct. 1982

Ouestion CS421.47 Discuss the design bases for the ventilation systems used for engineered O safety feature areas including areas containing systems required for safe shutdown. The discussion should cover redundancy, testability, etc. Resoonse: The design bases for the ventilation systems used for engineered safety feature areas are discussed in the PSAR and located in the following sections: (1) Sections 6.3.1 and 9.6.1.1 for the Control Building Control Room HabitabilIty System. (2) Sections 6.2.5 and 9.6.2.1 for the Reactor Conteinment Building Annulus Filtration System. (3) Sections 6.2.6 and 9.6.3.1 for the Reactor Service Building Filtration System. (4) Section 6.2.7 for the Steam Generator Building Aerosol Release Mitigation System. (5) Section 9.6.5 for the Diesel Generator Building. O O QCS421.47-1 Amend. 72 Oct. 1982

Ouestion CS421.48 N Using system schematics, describe the sequence for periodic testing of the: (G a) outlet stem Isolation valves b) main f eedwater control val ves c) main feedwater isolation valves d) auxil lary feedwater syste e) pressure rellef valves at superheater The discussion should include features used to insure the availability of the safety function during test and measures taken to insure that equipment cannot be lef t in a bypassed condition af ter test completion. Resoonse: Periodic testing of those components / system wil l be accompl ished as f ol lows: a) Outlet Steam Isolation Val ves 4 The test mode of the steam and feedwater isolation gate valves, which are opened hydraulical ly and closed pneumatical ly, is as f ol lows: , o A test mode switch must be activated and held in this position by i b,) C the operator. This action simultaneously overrices the pressure switch normally maintaining f ull hydraulic pressure to hold the valve open and de-energize the pneumatic and hydraulic pilot val ves, causing the val ve to being to close pneumatical ly. o The valve closes until the 10%-closed limit switch is activated. This activation energizes the hydraulic pilot valves which blocks j hydraulic flow in the val ve actuator and stops the valve stroke, o Af ter the operator verif les the valve has stroked to approximately 10% closed, he releases the test mode switch. Upon release of this switch the pressure switch controlling hydraulic pressure in the activator is re-energized and f unctions to cycle the val ve open hydraulically. o The operator verifles the valve has returned to f ull open position, thus completing the verification of valve operability. The limiting of the valve stroke to approximately 10% closed will permit normal piant operation durIng valve testing. AlI saf ety functions will remain operable, since trip signals will override the val ve test mode switch. 00S421.48-1 Amend. 72 Oct. 1982

b) Main feedwater control valves. These valves operate at a mid-stroke position which depends on power level and will move whenever power level is changed or whenever there is a system disturbance. Theref ore, no additional testing is required to dononstrate valve operabil ity. c) Main feedwater isolation valves. These valves will be tested in the seme manner as the outlet steam isolation valves. d) ' Auxiliary feedwater systan. The following describes the method of periodically testing the SGAHRS system. This test wil l be perf ormed once every three months to demonstrate the operability of the SGAHRS Auxil iary Feedwater Subsystem. It will be perf ormed during normal plant operation and under conditions that are as close to design as practical and provides for initiation of the complete sequence that brings the AFW Subsystem into operation for a reactor shutdown following a postulated accident. To ensure the availability of the safety function during the test, the logic design provides for the Plant Protection system initiation signal override of the SGAHRS AFW test switch signal. The initial conditions f or the startup of ine SGAHRS for the quarterly test require the plant to be operating at or above 40% power with the SGAHRS f il led, the Plant Protection System in operation and SGAHRS initiation logic in the reset mode. The SGAHRS Instrumentation is operable, the controls are in the automatic mode of operation and the valves are in their normal SGhiRS standby position as shown in PSAR Figure 5.1-5a. The protected air cooled condensers (PACC) are on standby with the f ans of f and louvers closed. All manual interface valves w ith the Steam Generator System (SGS) are open. The pericdic test procedure is as follows: (1) Initiate the Plant Data Handling and Display System (PDH&DS) Procedure for Test Trip Review for recording the following variables:

a. PWST l evel and temperature

b. AFW pump Inlet and discharge pressure

c. AFW pump discharge temperature

d. AFW flow

e. AFW recirculation flow

f. AFW recirculation valve position

g. AFW contrci volve position

h. AFW isolation valve position I. AFW turbine isolation valve position J. AFW turbire pressure control valve position Amend. 72 O

QCS421.48-2 Oct. 1982

n k. Drive turbine steam inlet pressure f 1. Drive turbine exhaust pressure

m. Drive turbine speed

n. Drive motor speed

o. Steam drum pressure

p. Steam drum level (2) Manually start the system for this test with the SGAHRS AFW switch. System startup is entirely automatic upon receipt of the initiation signal. The following cutomatic actions constitute

         . startup of the SGAHRS in the system test mode and occur as a result of manual operator initiation:

a. Drive turbine steam supply valves (52AFW118A, B, C) open and steam is supplled to the AFW pump drive turbine (52AFN001) which in turn drives the full-size AFW pump (52AFP001). The drive turbine pressure control valve (52AFV121) opens to modulate steam pressure at 1000 psig at the drive turbine inlet.

b. AFW pump drive motors (52AFK001A, B) start and drive the haif-size AFW pumps (52AFP002A, B).

c. AFW isolation valves (52AFV103A to F) open.

d. AFW isolation valves (52AFC104A to F) begin control of AFW fIow. In the SGAHRS tost mode these valves wilI close because Os the steam drum level is being maintained at the normal water level (NWL) by the feedwater system. Since the setpoints for the motor-driven pumps are at 4 in, below NWL, and 18 in.

below NWL for the turbine driven pump, no flow from the SGAHRS will be injected into the steam drums.

e. AFW pump recirculation valves (52AFV108A, B, C) begin their recirculation flow function. In the SGAHRS test mode these valves will remain open because no flow is being supplied to the steam drums.

The following automatic f unctions, which normally occur with a , SGAHRS initiation, are suppressed during the system test in order to prevent unwanted loss of steam generator system inventory and tripping of the turbine: o Opening of the superheater vent control valves (52AV116A, B, C) o Open of the steam drum vent control valves (52AFV117A, B, C) o Closure of the superheater outlet isolation valves (53SGV012) i o Closure of the steam drum drain valves (535GV014 and 015) o Opening of the PACC Noncondensible Vent valves O (52ACV1-9A to F) O o Startup of the PACCs QCS421.48-3 Amend. 72 Oct. 1982

(3) Observe and confirm the operation of the ' Auxiliary Feedwater Subsystem af ter the test initiation. The following status represents the normal operation of the AFW Subsystem under test conditions:

a. Motor-driven AFW pumps (52AFP002A, B) running at rated speed.

b. Drive turbine steam supply isolation valves (52AFV118A, B, C) are open.

c. Drive turbine pressure control valve (52AFV121) is open.

d. AFW pump drive turbine (52AFN001) and turbine-driven AFW pump (52AFP001) running at rated speed.

e. AFW flow control valves (52AFV104) are closed.

f. AFW isolation valves (52AFV103) are open.

g. AFW pump recirculation valves (52AFV108) are open.

h. Superheater vent control valves (52AFV116) are closed.

i. Steam drum vent control valves (52AFV117) are closed.

J. Superheater outlet isolation valves (53SGV012A, B, &C) and steam drum drain valves (53SGV014 A, B, C and 105 A, B, C). (4) Shut the AFW Subsystem down af ter 2 minutes and use the plant computer printout to verify all parameters. The subsystem is shut down and returned to standby per Steps (5) to (11) below. (5) Shut down the turbine-driven pump as follows:

a. Transfer the NORMAL /LONG TERM 000LDOWN switch for the drive turbine steam supply isolation valves (52AFV118A, B, C) to the LONG TERM 000LDOWN mode,

b. Close the drive turbine steam supply isolation valves 52AFV118A, B, c.

c. Transfer the drive turbine pressure control valve (52AFV121) to the manual control mode. Close the valve and transfer it back to the automatic mode.

(6) Shut down the motor-driven pump as follows:

a. Transfer each NORMAL /LONG TERM 000LDOWN switch for AFW pumps 52AFP002A and B to the LONG TERM 000LDOWN MODE.

b. Shutdown AFW pump drive motors 52AFK001 A and B.

(7) Reset the NORMAL /LONG TERM 000LDOWN switches for the AFW pumps to the NORMAL rode. (8) Roset SGAHRS test initiation trip logic. When this is done the AFW isolation valves (52AFV103A to F) will automatically close. (9) Switch the AFW flow control valves (52AFV104A to F) to the manual control mode and open the valves. Roset the valves back to the automatic control mode. (10) Confinn that all SGAHRS controls are in the automatic mode of operation. O QCS421.48-4 Amend. 72 Oct. 1982

(11) ConfInn the final conditions of this procedure are identical to the Initial conditions. N (12) Evaluate the system test results recorded by the PDH&DS and perform corrective maintenance on those components requiring it as demonstrated by the test data and repeat test if necessary. The following SGAHRS actuated valves also require periodic test: o Superheater Vent Control (52AFV116A, B, C) o Steam Drum Vent Control (52AFV117A, B, C) At Intervals of three months these valves wl2 l be exercised, one at a time, to the position required to f ul fill their function. The procedure for the exercising test is as foilows: (1) Isolate the approprlate Isolation valvc upstream of the valve to be tested. (2) Transfer control switch of the valve to be tested to the manual mode (skip this step if no such switch is provided for the valve). (3) Open the valve using the start /stop switch and the manual control ler (or open/close switch) . (4) Confirm the necessary valve movanent by exarcising the valve while observing the appropriate control room position Indicator. (5) Close the valve being tested. (6) Open the appropriate upstream isolation valve closed in Step (1). (7) Transfer control switch for the valve back to automatic mode. (8) Repeat Steps (1) through (7) in turn f or each val ve to be tested. (e) Pressure relief valves at the superheater The safety relief valves will be removed and bench tested during plant shutdowns at intervals consistent with ASME code requirements for saf ety val ves. Periodic testing during plant operation is not planned. , l O , Q/ QCS421.46-5 Amend. 72 l Oct. 1982

_ . _ -__ . _ . ._ _ ~ . _ _ _ - .. o Quest Ion CS421.58 Recent rev iew of a plant (Waterf ord) revealed a situation where heaters are to be used to control temperature and humidity within insulated cabinets housing electrical transmitters that provide input signals to the reactor protection systen. These cabinet heaters were found to be unquallfled and a concern was ralsed sInce possible f alIure of the heaters could potentIaliy degrade the tr anstr.itters, etc. Please address the above design as it pertains to CRBR. If cabinet heaters are used then describe as a minimum the design criteria used for the heaters. Resoonse: The only CRBRP IE equipment which use cabinet heaters are the Sodium Pump Drive Systan PPS Breakers. The heaters are Class 1E and are qualified to tunporature and humidity environments of 1250F and 90% relative humidity. When heaters are used in IE cabinets, it is a CRBRP requirement to environmentally qualify them according to IEEE 323, if the heaters are required te enable the equipment in the cabinet to perform its safety function. O O Amend. 72 QCS421.58-1 Oct. 1982

) i Ouestion CS721.1 () The Atomic Safety and Licensing Appeal Board in ALAB-444 determined that the Safety Evaluation Report for each plant should contain an assessment of -each

          - signif icant unresol ved generic safety question, it is the staf f's view that the generic issues identif ied as " Unresolved Safeiy issues" (NUREG-0606) are the substantive safety issues referred to by the Appeal Board. Accordingly, we are requesting that you provide your justification for permitting plant operation in consideration of these issues. This shoul d incl ude a description of any measures in terms of design or operating procedures or investigative i

programs that are being pursued to address these concerns. The Justification should provide an overall summary of your position on each issue in addition to a reference to various sections et the PSAR where related inf ormation is presented. There are currently a total of 27 Unresolved Safety issues. Some of these issues are clearly not appl icable to Cl inch River and need not be addressed. The remaining issues either clearly apply or the general intent of these issues appl les to Cl inch River. Those issues that you should address are identif ied in the fol low ing i ist. '

                                            " UNRESOLVED SAFETY ISSUES" ( APPL ICN3LE TASK NOS. )

Waterhammer - (A-1) i Steam Generator Tube Integrity - ( A-3, A-4, A-5) Anticipated Transients Without Scram - ( A-9) - Resolved * , Fracture Toughness of Steam Generator and Reactor Coolant Pump s Supports - ( A-12) Systems Interaction in Nuclear Power Plant (A-17) Environmental Qual if ication of Safety-Rel ated Electrical Eq u i pment - ( A-24 ) - Resolved

• Residual Heat Removai Requirements - ( A-31) - Resolved
• Control of Heavy Loads Near Spent Fuel - ( A-36) - Resolved
• Seismic Design Criteria - ( A-40)

Shutdown Decay Heat Removal Requirements - ( A-45)

Seismic Qual if ication of Equipment in Operating Pl ants - ( A-46) 1 Safety impl ications of Control Systems - ( A-47)

Hydrogen Control Fkasures and Ef fects of Hydrogen Burns on i Saf ety Equipment - ( A-4G) , in responding to this question for each issue you should address the following gu i del ines: (1) discuss the appl icabil ity of the issue to Cl inch River; (2) If you consider these issues to be resolved for Clinch River provide the basis for this concl usion; and (3) If you consider this issue unresolved as it appi les to Cl inch River provide your basis f or operation and a description of your relevant programs to resolve the issue. I

          *A number of the issues I isted above are technically resolved. Your response

, to this quest ion should address the appl icabil ity of the generic resolution to Clinch River. O I QCS7 21.1 -1 Amend. 72 Oct. 1982 t

Resoonse: CRBRP has considered the "Unresol ved Saf ety issues" identif ied in this ques-tion and has appl led appropriate measures to assure that the pl ant may be per-mitted to operate, given due consideration of these issues. Suitabl e resol u-tions to the Issues which refIect the technology of CRBRP are discussed below. WATERHAMMER A-1 APPL iCAB lL iTY TO CRB RP: Waterhammer and its equivalent, sodium-hammer, are appl icable to the CRBRP plant. Waterhammer events introduce Iarge hydraulic ioads, or pressurc pulses, into a f I u i d sy stem, and are the result of rapid condensation of steam pockets, steam-driven slugs of water, pump startup into volded I ines, and improper (or sudden) val ve cl osures. Where waterhammer has occurred in water l l Ines, the principal damage has been to pipe hangers and snubbers. In none of the waterhammer ir.cidents reported has there been a release of radioactive material or a disabl ing of saf ety systems. RESOLUTION FOR CRBRP: Technical resol ution f or this issue has been ef f ected on CRBRP. The water and steam systems of the CRBRP pl ant [i.e., the Steam Generator System and the Steam Generator Auxil iary Heat Removal System CSGAHRS)] are described in PSAR Sections 5.5 and 5.6.1, respectively. Design resol ution of waterhammer w il l be accompl ished by incl uding f il l and vent holes in the auxil iary feedwater sparger in the steam drum to precl ude waterhammer ef fects resul ting f rom steam-driven sl ugs of SGAHRS water, and by incl uding hydraul Ic dampers in the actuators of the water and steam isol ation vel ves to precl ude waterhammer , ef fects resulting f rom the overly rapid closing of a val ve. The vent holes are described in rev ised PSAR Section 5.5.2.3, and the hydraul ic dampers are discussed in Section 5.5.3.1.5.2. Protection against the ef fects of pipe breaks and waterhammer loads are incorporated in ASkE design codes which require consideration of irrpact leads and dynamic loads in the structural design. The AShE codes are appl led to the sodium systems of CRBRP, i.e., the primary heat transport system, the Intermediate heat transport system (incl uding the steam generator) and the sodium / water reaction pressure rel lef system, as welI as to the water / steam l sy stems. The design of the intermediate heat transport system, described in PSAR Section 5.4, has addressed the occurrence of sonic pul ses, simil er to those produced in waterhammer ir.cidents. Sonic pul ses may occur as a result of a l large sodium / water reaction caused by a postulated steam generator tube I rupture. In addition, the design of the sodium / water reaction pressure rel lef subsy stem, described in PSAR Sections 5.5, 7.5.6 and 15.3.3.3, has considered the ef fects of accelerated sodium slug flows in the component and piping design. O QCS721.1 -2 Amend. 72 Oct. 1982

    ~-          _

i The absence of sodium isolation valves in the IHTS precludes high decelerations of sodium which could cause waterhammer ef fects in sodium. The O high normal boiling point and high heat of vaporization of sodium make vapor-driven sonic pul ses extremely uni Ikely. STEAM GENERATOR TUBE INTEGRlTY (A-3. A-4. A-5) APPL iCAB iL 1TY TO CRBRP: This issue is appl Icable to 013RP. The design uses steam generators in each of the three heat transport system loops f or the transfer of heat f rom the secondary sodium loop to the water systems. The issue concerns 1he capabil Ity of steam generator tubes to maintain their integrity under normal operation

• and accident conditions, should mechanisms exist which can result in tube dcgradation.

RESOLUTION FOR CRBRP: Technical resol ution f or this issue has been ef fected on CRBRP. The CRBRP Steam Generator design has minimized the potential for corrosion / erosion degradation common to pressurized water reactor steam generators. The tubes in the CRBRP Steam Generator are exposed to the water environment only on their inside surface. The waterside consists of smooth wall tubes terminated in spherical plena. This greatly reduces the potential for tube degradation by corrosion induced wastage, cracking and denting. Preferential corrosion product f ormation or deposition is minimized since there are no restrictions, crevices, water l evel s or structure-rel ated concentration-sites O, present. Water side chemistry is maintained by state-of-the-art, all volatile chemistry control which has been modified f rom pressurized water reactor practice and which will incorporate fossil plant experience with 21/4 Cr-1Mo tube material . Full flow domineralizers, a 2:1 full power recirculation ratio I.e., for each 2 parts water flowing into the steam generator, I part is being recirculated and 1 part is f resh feed, and 10% blowdown all contribute to minimiz ing the potential for waterside corrosion-related problems. Steam generator tube integrity has been properly addressed in the CRBRP design through specifyIng that a total of 29% of the 0.109 inch tube wall thickness (Section 5.5.2.3.4 of the PS AR) be allocated f or corrosion, cleaning and wear al l ow ances. The reduced thickness is used for all stress and strain calculations while the full thickness is used for weight and seismic cal cul at t ens. In addition, allowances are provided to canpensate f or material strength degradation by post weld heat treatment, thermal aging and decarbur iz ati on. In spite of these reductions in thickness and material strength conservatively based on end-of-life condition, the tube has a 38% margin over the AShE Class 1 criteria for pressure retention. Erosien of tubes as a result of tube vibration is being addressed in three ways, as discussed in PS AR Section 5.5. First, the design and material selection of the shell (sodium containing) side of the steam generator (SG) provides f or acceptable accommodation of tube vibrations; all known flow induced vibration mechanisms have been evaluated. Tube to spacer plate gaps are consistent with guidel ines used throughout the heat exchanger industry. () Tube spacer plate material (inconel 718) has been chosen since it has a low Q CS721.1 -3 Amend. 72 Oct. 1982

coef f icient of f riction when coupled with the tube material (21/4 Cr-1Mo). Second, to conf irm that al l f l ow induced vibration mechanisms are considered, a f l ow induced vibration program has been impl emented using both a f ull scale model closely representing the prototype unit and a 0.42 scale model. The scal e model fl ow induced vibration tests will assure that mechanisms of unexpected origin in the pl ant unit design do not exist. Third, CRBRP has developed an ultrasonic tube inspection technique which can detect the tube wear wel l bef ore the tube wal l is thinned beyond that specif ied f or the design, which is discussed in PS AR Appendix G. ANTICIPATED TRANSIENTS WITHOUT SCRAM ( A-9) APPL ICAB IL ITY TO CRB RP: This issue is appl Icable to CRBRP. The issue concerns the potential for a common mode f ail ure to reduce the rel iabil ity of protection systems in such a way that the system might not f unction properly in the event of an anticipated transient. RESOLUTION FOR CRBRP: Technical resci ution of th is issue has been achieved f or L ight Water Reactors through publ ication of fiRC staf f position contained in NUREG-0460 Vol . 4.

 " Anticipated Transients w ithout scram f or Light Water Reactors." S peci f ic design f eatures and analyses are prescribed f or LWRs.                                      These prescriptions are not ap propriete f or CRBRP. The issue is resolved on CRBRP as discussed below.

CR3RP incorporates into the design, two independent shutdown systems, either of which has the capabil ity, of itsel f, to terminate reactor transients and to ef fect rapid shutdown of the reactor automatically. Strict attention to diversity of the design f eatures and to the separation of the two shutdown systens reduces the I ikel lhood, of the simul taneous prevention of both systems f rom operating when cal led upon as a resul t of a canmon mode f ail ure, To be incredible. Discussion of the design of the two shutdown systems, and the diversity of their f eatures, is prov ided i n PS AR Sect ions 4.2.3, 4.3.2. , and 7.2. FRACTURE TOUGHNESS OF STEAM GENERATOR SUPPORT ( A-12) i APPL ICAB IL ITY TO CR3RP: This issue is appl icable to CRBRP. This issue concerns the low f ract ure l i toughness and potent ial l amel lar tearing in material s used f or heat transport sy stem component supporis. RESOLUTICN FOR CRBRP: Technical resol ution f or this issue has been ef fected on CR3RP. O Q CS7 21.1 -4 Amend. 72 Oct. 1982 L _ _ - _ _ _ _ _ _ _ _ _

The design of that portion of the CRBRP Steam Generator Support which is in accordance w Ith the ASME Code requires that impact testing, Charpy V-Notch, of O all materials of construction be performed per paragraph NF-2311 of ASME Section Ill. The acceptance standards of NF-2330 must be met at 50 F maximum. Since the lowest operating temperature of the Steam Generator Support is 125 F, there is adequate margin f or protection against non-ductile f ail ures. In addition to the materials f racture toughness requirements, postulated , defects are evaluated using the procedure in Appendix G of ASME Section 111 f or all appi Icable conditions pl us shipping, lif ting and installation. Theref ore, the concern relating to f racture toughness of Steam Generator Supports has been properly and adequately addressed in the CRBRP design. The buil ding structural steel that supports Steam Generators will be designed in accordance with the AISC Code requirements using ASTM A-36 steel and SA-540 bol ting material . Sandia Laboratories' Report SAND 78-2348 ( Appendix C to NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports - Resol ution of Generic l_ Technical Activ ity A-12 for Comment") cl assif ies A-36 as f al l ing w ithin Material Group I I, i . e. , intermediate susceptibil ity to brittle f racture, and identif ies that Group 11 material s have been judged adequate. SAND 78-2348 classifies AS40 bolting material as f alling within Material Group 111, which has al so been judged adequate. The supports f or reactor coolant pumps and intermediate heat exchangers are SS304, connected to ASTM A-36 embedded plate wIth SA-540 bolting material.

CRBRP appi les design criteria to the reactor vessel and steam generator

;          supports to precl ude conditions leading to l amel lar tearing (e.g., material "s)  selection, wel ded joint orientation, and f abrication sequence).

SYSTEMS INTERACTION IN NUCLEAR POWER PLANTS (A-17) APPL ICAB lL ITY TO CRBRP: This issue is appl icable to CR3RP. This issue concerns the suf ficiency of integration of divided responsibil ities f or design, analysis and installation of systems among teams of engineers with f unctional specialties such as civil, el ectrical, mechanical and nuclear, to assure that adverse operational interactions between plant systems are minimized. RESOLUTION FOR CRBRP: Technical resol ution f or this issue has been resolved on CRBRP. CRBRP has implemented a combination of programs and activities directed towards assuring an integrated design which has considered the potential for and provides protection against adverse operational interactions between plant sy stems. These incl ude the CRBRP qual ity assurance program, a comprehensive design control progr am, special ized design reviews, and rel labil ity and 1 probabil istic risk assessment programs. i O i QC07 21.1 -5 Amend. 72 Oct. 1982

The plant has been designed to requirements which support a def ense in depth ph il osophy. These requirements assure physical separation and independence of redundant saf ety systems, diversity of safety features, and protection against I hazards such as sodium leaks, sodium / water reactions, l ine ruptures, missil es, tornadoes, floods, seismic events, f f res, human errors, and acts of sabotage. These requirements are described in PSAR Section 1.1.2 and Chapter 3. To assure that these requirements are properly implemented the CRBRP Quality Assurance Program addresses the design process. This program requires that during the design process emphasis is placed on the control of interfaces between systems. This interf acing is described in PSAR Section 17A.3.1. Independent design reviews, with interdiscipi inary memberships and objectives, are required at various stages of the design process. Requirements f or these independent design reviews are described in PSAR Chapter 17, Appendix G. CRBRP conducted extensive Key Systems Reviews (KSRs) cutting across system boundaries. These. reviews were conducted by multidiscipi ined groups of individual s wIth objectives which incl uded assessments of plant and operator responses during of f normal and accident events. Interactions between systems , were expl icitly considered as part of there reviews. Eval uations of the I results of these reviews addressed the potential for adverse systems l Interactions, incl uding considerations of human, spatial, and f unctional, coupl ing ef fects. A summary report of these reviews (KSRs) was provided in Ref erence QCS271.1-1. The CRBRP saf ety-rel ated rel iabil ity program is described in PSAR Appendix C. The results obtained in this program provide additional confidence that systems designs will minimize the potential for adverse operational interactions. Reference QCS271.1-2 described the CRBRP Probabil istic Risk Assessment (PRA) Program Plan which includes tasks which will demonstrate that the risk of CRBR are acceptably low. The planned methodology will use event trees and f ault trees to identify the component f ailures combinations that could result in a I oss of saf ety f unctlon. The PRA activitles w il I specif Ically eval uate potent ial adverse interactions between pl ant systems.

References:

Q CS7 21.1 -1 Letter Longnecker to Check " Summary Report on the conduct of CRBRP Key Systems Revlews," dated Feb. 19, 1982. QCS721.1 -2 Letter Lonc%er to Check "Probabil istic Risk Assessment (PRA) Program t'l an," dated J une 21, 1982 O QCS721.1 -6 Amend. 72 Oct. 1982

ENVIRONMENTAL OUALIFICATION OF SAFETY REL ATED ELECTRICAL EOUIPMENT ( A-24) APPL 1CABILITY TO CRBRP: This issue is appl icable to CRBRP. CRBRP cesign irclude Class 1E Equipment which must be qual if ied f or the environmental conditions in which it may be required to perf orm. RESOLUTION FOR CRBRP: Technical resol ution of this issue has been achieved f or Light Water Reactors through publ ication of NRC staf f position contained in NUREG-0588 " Interim Staf f Position on Environmental Qual if ication of Saf ety-Rel ated Electrical Eq u i pment. " The issue is resolved on CRBRP through a program for env ironmental ly qual ify ing saf ety-rel ated electrical equipment which is consistent w ith the objectives and requirements conteir.ed in NUREG-0588, Rev. 1 as appl led to CRBRP technology. This program is outl ined in the response to NRC Question CS270.1, and in PSAR Section 3.11. RESIDUAL HEAT REMOVAL REOUIREMENTS ( A-31) APPLICABllITY TO CRBRP: This issue is not appl icable to CRBRP. The issue concerns the capability of RVRs to go f rom hot to col d shutdown w ithout the avail abil ity of of f-site pow er. f} x,j A saf e shutdown condition equivalent to a PWR cold shutdown condition is achieved in CR3RP when the plant is brought down from operating temperature to 600 F using the plant shutdown heat removal sy stems. At the 600 F temperature the plant is in a saf e and stable state, and long term cool ing is in ef fect. There is no subsequent requirement to proceed to another mode or state to ef fect long term shutdown. The normal decay heat removal path is through the use of the main condenser and f eedwater train. Hcwever, as the main condenser and f eedwater train is not available upon loss of of f-site power, the Steam Generator Auxil iary Heat Renovel System (SG AHRS), which is a saf ety-rel ated system, is provided for shutdown heat removal and long term cecay heat removal, and is independent of the avail abil ity of of f-site power. CONTROL OF HEAVY LOADS NEAR SPENT FUEL ( A-36) APPL 1CA8 iL 1TY TO CRBRP: This issue is appl icable to CRBRP. Although the design of CRBRP coes not use spent fuel pool s, this concern is appl icable to the control of heavy loads over the Ex-Vessel Storage Tank Closure Head and Striker Plate, and over ihe Fuel Handling Cell. QCS7 21.1 -7 Amend. 72 Oct. 1982

RESOLUTION FOR CRBRP: Technical resolution f or this issue has been achieved through publication of NRC staf f position contained in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants". The issue is resol ved f or CRBRP by tho appl ication of a single-failure proof crano (in accordance w Ith NUREG-0554, "S ingl e Faii ure Proof Cranos f or Nucl ear Power P! ants") in both the RSB and RG for all critical lifts. The Project appl Ication of NUREG-0612 is presented in respohsc to NRC Ques rion CS410.3. SEISMIC DESIGN CRITERIA (A-40) APPL ICAB IL ITY TO CRBRP: This issue is appl icable to CRBRP. The Issue concerns the conservatism of certain aspects of the overal l seismic design criteria. RESOLUTION FOR CRBRF: Technical resol uticn f cr th i s i ssue has been ef fected on CRBRP. The seismic design bases and the seismic design of CRBRP conf crm to the current NRC criterft. CRBRP se!unic design criteria are described in PSAR Section 3.7. NRC have not establ ished any other bases which woul d render conf ormance to the currert criteria inadequate. SHUTDOWN DECAY HEAT REMOVAL REOUIREMENTS ( A-45) APPLICAB IL ITY TO CRBRP:

• Th Is i ssue i s appi icabl e to CRBRP. This issue concerns the suf fIciency of plant capability to remove decay heat. CRBRP must have a highly rolleble capabil ity to retrove decay heat f rcra the reactor.

RESOLUTICN FOR CRBRP: Fesol ution f or CRBRP f or th is issue has been accompl ished by incorporating Irto the design, multiple, Independent, and highly rol lable heat transport paths, any one of these paths havirg suf f icient capacity to be able to remove the reactor decay heat by itsel f. The various heat removal paths and their operating modes embody substaniial diversity. CRBRP Heat Transport System uses three ir. dependent loops each of which prcv ices a separate path f rom the r eactor vessel to the ul timate heat sinks. The normal heat removal path ir.cl udes the main condenser and f eedwater train which is used f cr normal cperation and some shutdown heat removal conditions. Hcw ever, f or each path an al ternative saf ety-rel ated path is provided, through the Steam Generator Auxil iary Heat Removal System (SGAHRS) which provides its own heat sinks. Thus, it is not necessary to rely upon the main condenser and f eedwater tralr., since SGAHRS is avalI able f or ai i anticipated plant events. The SGAHRS system incl udes the Auxil iary Feedwater Subsystem ( AFWS) and Protected Air Cooled Condensers (PACCs) which serve as alternative heat sinks. O

                                       - QCS7 21.1 -8                  Amend. 72 Oct. 1982

The AFWS provides water make-up to the closed loops between the steam ,ew generators and the PACCs. The AFWS includes two motor driven and one ( ) steam-turbino driven pumps. v The sodlum in the Primary and Intermediate Systems of the HTS loops is always at temperatures well below the flash point. Thus, in the unlikely event of a sodium pipe leak in any loop there wIll not be a loss of heat removal capabil ity due to loss of coolant inventory through flashing. Also, degradation of one locp wIll not af fect heat removal capability in either of the other two loops. Thus, the plant configuration provides multiple independent paths through the Heat Transport System, which contributes to the high reliabil ity of the plant systems f or removing reactor decay heat. These capabil ities are discussed in PS AR Section 5.6 and 5.6.1. CR3RP provides an additional path f or decay heat removal, the Direct Heat Removal Service. This system provides a diverse heat removal path to yet another redundant and diverse set of air cooled heat exchangers. This is described in PS AR Section 5.6.2. SEISMIC OUAllFICATION OF E00!PMENT IN OPERATING PLANTS ( A-46) APPL ICAB IL ITY TO CRBRP: This issue is not applicable to CRBRP. The issue is whether operating plants must be reassessed to assure the adequacy of their seismic qual ification of ( (,,j} eq u i pment. Construction of the Project has not yet commenced and thus, it is not an operating Pl ant. CRBRP resol ution of USI A-40 assures the adequacy of Seismic Design Criieria appl led to it. SAFETY IMPLICATIONS OF CONTROL SYSTEMS ( A-47) APPLICAB lLITY TO CRBRP: This issue is appl icable to CRBRP. CRBRP is dependent upon the proper f unctioning of control systems in order to maintain the plant in a safe condition f or al l normal operations and accidents. This issue concerns the potent ial for transients or accidents being made more severe as a result of control system f ail ures or mal f unctions. These f ailures or mal functions may occur independently or as a result of the accident or transient under consideration. RESOLUTION FOR CRBRP: Technical resolution f or this issue has been ef fected on CRBRP. Design f eatures ensure that control system f ail ures wil l not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe shutdown condition following any anticipated operational occurrence or accident. This has been accomplished by providing independence and physical separation between safety system trains and between saf ety and non-saf ety systems. For the l atter, as a

   ,_   m i nim um, isol ation devices are provided. These devices precl ude the propagation of non-saf ety system equipment f aults to the protection systems.

(v} QCS725 ' -9 Amend. 72 Oct. 1982

l l l Al so, to ensure that the operation of safety system equipment is not impaired, l the singl e-f ailure criterion has been appl led in the pl ant design. PSAR j Section 7.2.2 discusses Plant Protection System (PPS) - Control Sy stem interaction. The 01BRP PPS is conposed of two independent subsystems, either of which is capable of bringing the plant to a safe shutdown condition. Further, these two subsystems employ diverse trip f unctions f or PPS acti vati on. Theref ore, for any Design Basis transient, there is always more than one trip f unction provided by these two totally independent subsystems to activate the PPS and terminate the ensuing transient. Details of this design are described in PS AR Section 7.2, and Table 7.2-2. A wide range of bounding transients and accidents is presently analyzed to ensure that the postulated events would be adequately mitigated by the safe y sy stems, in addition, systematic reviews of saf ety systems have been perf ormed w Ith the goal of ensuring that the control sy stem f ail ures w il l not def eat saf ety system action. The worst conditions f or each given type of transient are assumed in the accident analyses. This inf ormation is provided in PSAR Chapter 15. HYDROGEN CONTROL MEASURES AND EFFECTS OF HYDROGEN BURNS ON SAFETY EOUIPMENT-A-48 APPL ICAB IL ITY TO CRB RP This issue is not applicable to CRBRP. Design basis accidents within the CR3RP containment do not lead to the generation of hydrogen. Accordingly, there is no ef fect of hydrogen burns which coul d impact the capabil ity of saf ety-rel ated equi pment to perf orm iis intended saf ety function. How ever, accidents beyond the design basis involving hypothetical core disruptive accidents may produce hydrogen as a result of sodium-concrete interactions. The control and burning of the hydrogen f rom a hypothetical core disruptive accident is addressed in the CRBRP Thermal Margin Beycnd Design Basis (CRBRP-3, Vol . 2) . In the T1BDB scenario, the hydrogen is ignited in the containment atmosphere by sodium burning wIth the oxygen in containment. CRBRP-3, Vol . 2 al so demonstrates how contai nment integrity is mainteined. O QCS7 21.1 -10 Amend. 72 Oct. 1982

Ouestion CS760.13 \' ') + Section 15.2.2.2 analyzes a 60c radial movement (stick slip) incident. The analysis does not distinguish between primary or secondary scram. (On!y one temperature curve is given). Provide analysis f or this transient, listing the appropriate primary and secondary trip functions. Resoonse For a 60c step reactivity insertion the power increases in almost step f ashion f rom 100% to over 200% as shown by Figure 15.2.3.3-3. Both the primary and secondary high power trip signals are significantly below the increased power level and thus, both trips would occur simultaneously. The table below summarizes results f or the highest cladding temperature hot rod in FA-52 considering both primary and secondary scram (each separately). MAXINllM TEMPERATURES (30) REACTOR POWER CLADDING FUEL COOLANT SHUTDOWN SYSTEM AT TRIP A B C A B C A B C Primary 115% 1491 0.63 1.6 4576 0.53 1.3 1417 0.63 1.6 Secondary 122% 1544 0.83 2.0 4752 0.63 1.7 1467 0.83 2.1 A - maximum 3a hot spot temperature attai ned, OF. B - time to reach .naximum temperature, sec. C - length of tiins temperature is above initial steady state value, sec.

 /"~N  It should be noted that occurrence of a 60c step reactivity insertion combined

(,,) with f ailure of the primary scram would be less probable than an extremely unlikely category event in which case the primary shutdown of a Saf e Shutdowr Earthquake (Section 15.2.3.3) would envelope the consequential core damage. r QCS760.13-1 Amend. 71 Sept. 1982

Ouestion CS760.28 s During the earlier phase of the CRBRP licensing review, it was a regulatory staf f position that no credit would be given f or shutdown (decay and sensible) heat removal by natural circulation through the heat transport trains of .the plant. . This position was based on the lack of a data base to support the - potential natural circulation capability of the CRBR design. Additi onal ly, there are numerous concerns regarding the adequate simulation / prediction of the thermal / hydraulic characterization of the sytem operating under natural ci rcul ation conditions. To adequately portray system operation under natural circulation conditions, there are many inter-related f actors which must be represented. The overall driving force for the natural circulation flow is a balance between competing pressure losses and gains throughout the reactor vessel and primary system. To adequately assess the natural circulation capability of CRBRP, appropriate experimental data, both on a component and system-wide basis, must be provided to validate and improve wherever necessary the presently available analytical tools. Table 15.3-1 summarizes the data needs relative to the adequate assessment of natural circulation capability of CRBR. Incl uded i n th i s Tabl e are the key paraneters and their importance. Please provide data, particularly for those items where the availability is noted as " limited" or "none", so that their inf luence on the natural circulation capability of CRBR can be properly substantiated. O l O QCS760.28-1 Amend. 72 Oct. 1982 . _ - _ _ - - _ _ _ _ . _ .. _ . _. _ ___ _ _ . _ ._ _. _._- ._. . __ _ __= ... _.

(THIS TABLE IS PART OF NRC'S QUESTION CS760.28.) TABLE 15.3-1 LMFBR SYSTEM VAtlDATION NEEDS RELEVANT TO NATURAL CIRCULATION COMPONENT PARAETER/PHENOKNA IPCORTANCE DATA AVAILABILITY

1. REACTOR VESSEL

1. INLET MDDULE PRES $URE DROP HIGH LIMITED MIXING DURING PARTIAL FLOW REVERSAL SEDI UM NONE

2. OUTLET PLENUM MIXING AND STRATIFICATION HIGH LIMITED

3. ASSEWLIES PRESSURE DROP HIGH LIMITED / LIMITED L AMINAR/CR!Tl CAL TRANSITION / SUFF1ClENT TURBULENT HEAT TRANSFER COEFFICIENTS HIGH SUFFICIENT INTRA-ASSEm LY REDISTRIBUTION HIGH LIMITED HEAT / FLOW INTER-ASSEmLY REDISTRIBUTION HIGH LIMITED Q

co HEAT / FLOW LOW-HEAT FLUX BOILING DRYOUT KDIUM LIMITED y CORREL ATIONS Co HIGH E DECAY HEAT  ?

4. STRUCTURAL SHUTDOWN HEAT / HEAT LOSSES KDIUM NONE MATERIAL ll. HEAT TRANSPORT S.YSIf.M

1. PIPING PRESSURE DROPS LOW SUFFICIENT STRATIFICATION EDluM LIMITED

2. PUMPS FRICTIONAL TORQUE HIGH NONE LOCKED ROTOR RESISTANCE HIGH LIMITED

3. OtECK VALVE PRESSURE DROP HIGH LIMITED i

! 4 lHX SHELL SIDE PRESSURE DROP EDIUM LIMITED R@ FLOW MALDISTRIBUTION LOW PO4E

        ~@

P

a 1 i i I J

TMLE 15.3-1 (continued) i

.I

C00F0hENT PARAETER/PHENDENA iWORTMCE DATA AVAILMiLITY 4

fil. STEAM l GENERATOR i iLLM i l

1. S.G. PRESSURE DROPS HIGH SUFFICIENT 4

DRYOUT CDRREL ATIONS EDIUM SUFFICIENT 'I i 1 2. REClRC. HD00LOGOJS PUW QJRVES HIGH LIMITED j PUIF (X)ASTDOWN RATE EDIUM LIMITED  ;

3. HEAT LOSS CDEFFICIENTS EDIUM LIMITED EXCHANGER HEAT TRANSFER HIGH LIMITED FOR

• EVAPORATOR

4. PACC PRESSURE LOSSES Lou LIMITED c HEAT 1RANSFER HIGH NG8E l

, o j m

5. ISOL ATION VALVES FULL OPEN FLOW MEA LOW NONE i o

.l N 6. O*ECX V ALVE LOSS CDEFFICIENTS LOW LIMITED m . w' j I i em i >-a .a. j mN

NN l

4

                                                                                                                                              -_ _                  r-

RESPONSE

!                             1. REACTOR VESSEL 1           gJ                                                                                                                                                                                                        i i

1. Inlet Module The pressure drop data through the lower inlet modules (LIMs) -are given in terms of loss coef f Iclents, K, and ref erence areas, A, as shown in Table 4.4-8 of the PSAR. These val ues were determined f rom the experimental data repor ted in Reference QCS760.28-1, for the bl anket ori f icing test. FulI prototype iniet moduie tests were recent1y performed at ARD. Preliminary evaluations of these tests indicate the measured pressure drop is within the 20.# uncertainty value used in the calculated PSAR values.

1 No flow reversal is expected in assemblies during natural l circulat!7n. Consequently, no such flow condition was contemplated in the LIN experiments. - <

2. Outlet Plenum - Mixina and Stratification Contrary to the bellef that only ilmited experirrental data are available concerning the reactor vessel upper plenum mixing and l stratification phenomena, the problem has been studied quite i extensively both experimentally and analytically. The experimental tests Inelude Argonne National Laboratory (ANL) 1/15 scale modeI water, brine, and sodium tests f or both the FFTF and CRBRP upper pl enum (Ref ereces QCS760.28-3, 4), Battelle-Columbus Laboratory (BCL) 0.55 scale FFTF model water and brire tests (Reference 4

Q CS760. 28-5 ) , and ANL 1/10 scale CRBRP model water tests (References Q CS760.28-6 through 10). The experiments perf ormed with the 1/10 scale CP3RP model incl uded studies of: 1) steady state and transient performance of the CRBRP outlet plenum; 2) the of fects of thermal oscillation; 3) the suppressor plate and shear web; and 4) the influence of the heterogeneous core geometry upon the outlet plenum mixing. In addition to the studies on steady state and

thermal down transient perf ormance, outlet plenum mixing f or i transient overpower conditions had also been studied and the results are documented in Ref erence QCS760.28-11. Finally, the influence of

scale size and fluid thermal properties in simulating LMFBR outlet plenum behavior was also studied and the results are presented in Ref erence QCS760.28-12.

in addition to the above experimental studies, several semi-empirical upper plenum mixing codes have been developed based

primarily on the scale model water test results. A three region outlet plenum mixing code PLENUM-3 was f Irst proposed by j

P. A. Mcward (Ref erence QCS760.28-13) of ANL. Based on addltional { w ater test data P. A. Hcward, et al . , later proposed a two region out l et pl enum model PLENUbb2 (Pof erence QCS760.28-14) . This two ] region model was later revised by the same authors into the

!                                                 PLENUM-2A model (see Ref erence QCS760.28-15).

IO QCS760.28-i-1 Amend. 72 li Oct. 1982

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

Aside f rom these semi-empirical model s, the outlet plenum perf ormance had al so been studied numerically by the two-dimensional V ARR-i l (Ref erence QCS760.28-16) and the three-dimensional TEMPEST (Ref erence QCS760.28-17) codes.

3. Assemblies Engssure Droo The pressure drop in core essemblies consists of the following components: i nl et nozz l e, inlet nozzle orifice-shield, shield, rod bun dl e, and outlet nozzle. There are two f orms of pressure drop, i.e., frictional loss and f orm loss. The f ormer occurs through the rod bundle while the l atter covers all other components of the assembly.

Toe rod bundle f rictional loss data for fuel assemblles for the PSAR were obtained f rom the FFTF Fuel V ibration Tests (Ref erences QCS760.28-18a and 18b). Data for inner and radial blanket rod bundles were cbtained f rom the Radial Blanket Heat Transf er Tests and data f or simil ar bundle geometry (Ref erence QCS760.28-18c) In sodi um and water. These data are shown in Figures QCS760.77-1 and

  -2. (See Question / Response CS760.77.) Both of these figures cover the laminar, transition and turbulent ranges.

Form loss data were obtained f rom the CRBR f uel assembly inl et/outl et nozz le f l ow test and CRBR radial bl anket flow orif icing testing. Tabl es 4.4-6 and 4.4-7 In the PSAR (Chapter 4) cover the component hydraul ic correl ations f or f uel, inner blanket and radial bl anket assembl ies. Additional experimental data for CRBR f uel assembl ies have been obtained as shown in Figure QCS760.77-4 (See Question / Response CS760.77 ). The l atest data on f uel assembly f riction f actor is publ i shed in Ref erence QCS760.28-19. The CRBR control assembly hydraulic tests and blanket assembly flow and vibration tests have been completed and cover a wide range of operating conditions, including natural circulation flow rates. These data is shown in Figure QCS760.77-2 and -3 (See Question / Response CS760.77). Heat Transfer Coefficients Heat transf er coef ficients were selected f rom a wide range data base and are reported in Ref erence QCS760.28-20. It shoul d be noted that for natural circulation conditions, the heat flux if low and the f ilm temperature drop general ly is negl igibl e, which makes the infl uence of the heat transf er coef ficients relatively unimportant f or the natural circul ation event. O Q CS760. 28- 1 -2 knend. 72 Oct. 1982

Intra-Assembly Heat / Flow Redistribution I \ (/ Exter 9ve out-of-pile sodium heat transfer test data over the f ull range. af operating conditions on the 61-rod f uel assembly bundle were obtained in the ORNL Thermal-Hydraul Ic Out-of-Reactor Saf ety (THORS) f acil ity (Ref erence QCS760.28-50). Extensive sodium heat transfer test data on a prototypic 61-rod blanket assembly bundle were obtained over a wide range of operating conditions in the WARD GPL f acil Ity (Ref erences QCS760.28-51 through 60). Both of these data were compared with the COBRA-WC code considering the intra-assembly heat / flow redistribution in Reference QCS760.28-22. In-pile data from the natural circulation experiments conducted in the Experimental Breeder Reactor-lI (EBR-lI) factiIty confirmed such intra-assembly redistributions (Ref erence QCS760.28-21). Specifically, instrumented Subassembly XX08, Test 7A, was used as a verification of the COBRA-WC code in calculating the sodium temperature at different channel locations and time during the transient. The comparison of code calculated temperatures with the experimental data are presented in Reference QCS760.28-22. The recent FFTF natural circulation tests also provide confirmatory experimental data on temperatures measured in the instrumented f uel open test assembiles (FOTAs), involving the offect of intra-assembly heat / flow redistribution. These data are also used to verify the COBRA-WC code predictions (Ref erence QCS760.28-23). Inter-Assembly Heat / Flow Redistribution l I V The EBR-il and FFTF in-pile experimental data 1 isted in the above section also confirm the inter-assembly effects. These factors are al so simul ated in COBRA-WC code simul ation (References QCS760.28-22, 23). The WARD bl anket heat transf er test al so provides inter-assembly heat transfer data. In addition, further inter-assembly heat / flow redistribution experimental data will be obtained f rom the instrumented Inner blanket assembly WBA-45/46 including during natural circulation. Testing of multi-assembiles in sodium in the THORS out-of-pi1e f acil Ity at ORNL wIl l al so provide inter- assembiy flow redistribution experimental data. Low Heat Flux Bol! na Drvout Correlations Experimental data on low flux bcIllng is reported in Reference Q CS760.28-24, which simulates sodium bo!! Ing under low power, low flow conditions. Sodium boil ing in a f ul l length 19-pin simul ated f uel assembly (THORS Bundle 6A) was tested in the ORNL Thermal-Hydraul Ic Out-of-Reactor Saf ety THORS f acil Ity (Reference Q CS760.28-25 ) . An in-pile test was also performed in the Sodium Loop Safety Facility (SLSF) simulating loss of piping integrity accident (Ref erence QCS760.28-26). These experiments provide data on sodium boil ing under low heat fl ux and/or dryout conditions, it shout d be noted that during the CRBR natural circulation event, sodiumboilIngdoesnotoccur. In f act, Ref erence QCS760.28-2 shows that over 150 F margin exists to bolling for the highest temperature O hot rods. QCS760.28-I-3 Amend. 72 Oct. 1982

Decav Heat Decay heat f or an average f uel assembly, average Iniv..- blanket assembly and average radial blanket assembly is p.ovided in Table 3.1 of CRBRP-ARD-0308 (Ref erence QCS760.28-2) for worst case natural ci rcul ation analyses. Corresponding decay heat data for the maximum temperature hot rods In FA-52, iBA-99 and RBA-203 are provided In the response to Question CS760.24. The bases f or these average region and hot rod decay heat values are given in Ref erence QCS760.28-2 (Sections 3.2.2.1 and 4.2.2.3) and Ref erences QCS760.28-38 through 49.

4. Structural Material It is assumed that " shutdown heat / heat losses" ref ers to the inclusion of the reactor vessel and Internal sensible heat and reactor vessel heat losses to the atmosphere in natural circulation analyses. The mass used in DEMO analyses is described in CRBRP-ARD-0005. Heat loss f rom the reactor vessel to the atmosphere, however, has not been included in natural circulation analyses. This is seen as a minor ef fect and neglecting it would yield conservative results.

O O QCS760.28-1-4 Amend. 72 Oct. 1982

II. Heat Transoort System

1. Ploing ,

Pressure Drops - Designated in question as low importance and with suf ficient data available. Stratification - Based on the testing presented in Ref erence QCS760.28-27, a study of the importance of stratification on natural ci rcul ation analyses (Ref erence QCS760.28-28) concludes that "It can be saf ely concluded that for f low and temperature transients seen at the entrance to the piping runs (component exit nozzles) during the transition f rom forced (pumped) flow to natural circulation and

;             operation in that mode, the stratification that would occur in horizontal sections of these piping runs can be ignored. in fact, even under extremely severe strati fication assumptions, i . e. ,

stratification to occur at the inlet to a horizontal pipe run and also at the inlet to a vertical riser, the ef fects of piping stratification on the natural circulation decay heat removal capabil ity are seen to be very smal l."

2. Pumps Frictional Torque - The CRBRP sodium pump f rictional torque corrolations used In the most rocent DEMO natural circulatton analysi s (CRBRP- ARD-0308) account for motcr windage, bearing losses and f riction between the pump shaf t and the surrounding fluid.

h Development of the correlations is based upon experimental data available f rom prototype pump water test results. A trial-and-error procedure using the pump coastdown speed vs. time data, (see Table QCS760.28-i l .1 ) pumping torque correlation and pump Inertia as

;               inputs to the equation of motion for the pump was performed to determine coef ficients of the f rictional torque correlations, i

Locked Rotor Resistance - The CRBRP sodium pump locked rotor rer,lstance correlation used in the most recent DEMO natural ci rculation analysis (CRBRP-ARD-0308) was developed using experimental data available f rom prototype pump water test results shown i n Figure QCS760.28-i l.1.

3. Check Valve i

i Pressure Drop - The pressure drop correlation used in CRBRP natural circulation analyses is based on test data taken on the FFTF 16-Inch val ve and 6-inch model val vo. The detailed testing performed on the FFTF valves and the hydraulic similitude between the FFTF and CRBRP val ves negates the need to test the CRBRP check val ves f urther. Scal ing and analyses accounting f o;- design di f ferences were used to extrapolate thIs FFTF data to the CRBRP data (Figure QCS760.28-Il.2 and .3). O QCS760.28-Il-1 Amend. 72 Oct. 1982

4. .11fX IHX Shell Side Pressure Drop and Flow Maldistribution - The IHX shell side pressure drops used in the DEM0 natural circulation analysis are a combination of vendor experimental data and pressure drop analysis. The IHX pressure drops f rom upstream of the inlet nozzle to downstream of the outlet nozzle are presented in the f ol lowing tabl e.

MASS FLOW AP 0 8620F AP G 7860F AP 8 4000F (T) (osI) (ost) (osi) 100 14.017 13.8586 13.0933 40 2.409 2.38178 2.25025 30 1.46765 1.451 1.37093 10 0.22704 0.22447 0.21208 5 0.07016 0.069367 0.065536 3 0.029877 0.029539 0.027908 1 0.004831 0.004776 0.004512 The fIow Is based on 13.82 X 106 Ibm /hr. FIow maldistribution in the shelI side of the IHX, which may be postulated to be Induced by buoyancy ef fects at low primary flows, show Insignficant impact on natural circulation transients (Ref erence QCS760.28-37) . The ef fect of postul ated f low mal distri-bution was analyzed by varying the of fective heat transf er area f rom

   +33% to -73%. The analysis produced negligible changes In reactor tanperatures.

O QCS760.28-I l-2 Amend. 72 Oct. 1982

l TABLE QCS760htr-!!-! I i

)                                                      Coastdown Run A                                Coestdosn Run B I

TIE RJMP T1T PUMP TIE PUMP TIE PunP Sec SHAFT , Sec SHAFT Sec SHAFT Sec SHAFT RPM * . RPM

• RPM ' RPM
• i i 0 1116 30 120 0 1116 52 -

1 855 32 1@ 1 900 54 - i 2 720 34 1 05 2 738 56 -- 3 618 36  % 3 633 58 - 4 540 38 93 4 552 60 51 ! 5 495 40 87 5 492 62 - ! 6 444 42 75 6 444 64 --- ! 7 405 44 -- 7 405 66 -  ! j 8 360 46 - 8 366 68 -- i 9 336 48 --- 9 336 70 39 l $ 10 312 50 62 10 12 3 09 270 80 90 30 24

[ 11 291 55 57 i es 12 270 60 48 14 237 100 12 I O 13 255 65 45 16 213 110 6

. N 14 240 70 42 18 195 0 i

                 ,         15             228          75                 33                    20     177 i

1

               -           16             213          80                 28                    22     --

7 w 17 204 85 24 24 -- 18 195 90 21 26 -- 19 186 95 - 27 -- 3 20 177 100 18 28 -- ! 22 159 105 - 30 120 I 24 150 120 0 32 - j 26 138 36 -- 1 28 129 38 -- 40 87 l 42 -- 44 -- 46 - + 48 -- < 50 66

• RPM es Cm puted From bearing Proximeter Pulsos i

o> n5 rt O 3 i a w. O ow NN I , I

TABLE QCS760.28-II-2 Ccastdown Run C Coastdown Run D Coastdown Run E Coastdown Run F TIK PUMP TIE PUMP TIE PUW TiK PUMP Sec SHAFT Sec SHAFT Sec SHAFT Sec SHAFT RPM

• RPM
• RPM
• RPM
• 0 795 0 0 1129 0 960 1 7 68 1 840 1 9% 1 --

2 630 2 2 816 2 700 3 540 3 688 3 684 3 -- 4 480 4 522 4 600 4 540 5 435 5 474 5 515 5 -- 6 390 6 421 6 450 6 420 7 360 7 390 7 41 0 7 -- 8 330 8 342 8 378 8 375 9 312 9 312 9 342 9 -- 10 282 sm 10 15 297 210 10 12 315 285 10 12 310 27 0 12 14 252 222 5 20 168 14 258 14 240 16 192 P 25 138 16 220 16 215 18 180 na 30 111 18 198 18 195 20 162

?

35 40 96 81 20 25 185 156 20 25 175 140 25 30 132 108 7 50 63 30 128 30 115 35 90 a 60 47 40 101 35 1 08 40 78 70 35 50 76 40 85 50 60 80 27 60 57 45 72 60 45 90 19 70 45 50 65 70 36 100 13 80 32 60 50 80 24 110 0 90 27 70 40 90 15 100 17 80 25 1 00 12 110 0 90 21 110 0 100 14 110 0

• RPM as Computed From Bearing Proximeter Pulses

   ?$

r2 P e NN 9 9 e -

I , V I .I 4 5 6 / 891 2 3 4 5 6 7 8 91 1000._._-._. _. .. . . . . P- i 00 . l. _.. i..I. - . . - ....: ::. ..lT_ia.9. -

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