ML20028B468

Amend 73 to PSAR

ML20028B468
Person / Time
Site:	Clinch River
Issue date:	11/30/1982
From:	ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:
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ML20028B465	List: ML20028B464 ML20028B468
References
NUDOCS 8211300409
Download: ML20028B468 (351)

Text

O PAGE REPLACEMENT GUIDE FOR AMENDMENT 73 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYSIS REPORT 1

(DOCKET NO. 50-537)

Transmitted herein is Amendment 73 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537. Amendment O

d 73 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 73 pages and instructions for their incorporation into the Preliminary Safety Analysis Report.

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O AMENDMENT 73 PAGE REPLACEMENT GUIDE REMOVE THESE PAGES INSERT THESE PAGES j Chapter 1 1.5-46, 47 1.5-46, 46a, 47 i

Chapter 2 i

2-xiii, xiv 2-xiii, xiv 2-xix thru xxii 2-xix thru xxii 2.4-3, 4 2.4-3, 4 2.4-7, 8 2.4-7, 7a, 8 2.4-9 thru 20 2.4-9 thru 13, 13a, 14, 15, 15a, 16 thru 20 2.4-20a, 21 thru 29 2.4-200, 21 thru 29 2.4-30, 31, 31a, 32 2.4-30, 30a, 31, 31a, 32 2.4-39, 40 2.4-39, 40 2.4-75, 75a, 76 thru 82 2.4-75, 75a, 76 thru 82 i

2.4-117, 118 2.4-117, 118 0 2.4-121 thru 124 2.4-125, 126, 127, 127a, 127b, 128 2.4-121 thru 124 2.4-125, 125a, 126, 127, 127a, 127b, 128 2.4-130a, 131 thru 135, 135a, 136 1.4-130a, 131 thru 135, 135a, 136 2.4-153 thru 156 2.4-153, 154, 155, 155a, 156 2.4-159, 160, 161 2.4-159, 160, 161 2.4-165 2.4-165 Chapter 3 3.7-7, 8 3.7-7, 8 3.7-14, 14a 3.7-14, 14a 3.8-C.1, 2 3.8-C.1, la, 2 3A.1-Sa 3A.1-Sa 3A 8-8, 9, 9a, 9b 3A.8-8, 9, 9a, 9b Chapter 4 i 4.2-162b, 163, 164, 165 4.2-162b, 163, 163a, 164, 165 4.4-89, 90 4.4-89, 90 t

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REMOVE THESE PAGES INSERT THESE PAGES Chapter 5 5.5-8, 9 5.5-8, 9 5.5-11, lla 5.5-11, 11a 5.5-23, 23a 5.5-23, 23a 5.5-44, 45 5.5-44, 45 5.6-8, 8a 5.6-8, 8a 5.6-11, 11a 5.6-11, 11a 5.6-30 thru 33 5.6-30 thru 33 Chapter 7 7.1-9, 10 7.1-9, 10 7.2-1, la, 2, 2a 7.2-1, la, 2, 2a 7.2-11 thru 15, 15a, 16 thru 19 7.2-11 thru 14, 14a, 15, 15a, 16 thru 19 7.2-26, 27 7.2-26, 27 7.4-7, 8 7.4-7, 8 7.5-52 7.5-52 7.7-5 thru 8 7.7-5, 6, 7, 7a, 8 7.7-17, 18 7.7-17, 18 Chapter 9 9-xi, xii 9-xi, xii, xiia 9-xv, xvi 9-xy, xvi, xvia 9-xix, xx 9-xix, xx 9.6-39 thru 42 9.6-39 thru 42 9.6-94 9.6-94 9.7-17, 18 9.7-17, 18 9.9-1 thru 4 9.9-1 thru 4 9.9-7, 8 9.9-7, 8

- 9.9-12a, 12b, 12c

- 9.9-18a, 18b 9.13-13 thru 17, 17a thru 17d 9.13-13 thru 17, 17a thru 17d 9.15-1, 2 9.15-1, 2, 3 9.16-9, 9a 9.16-9, 9a INSERT NEW 9.17 TAB AND PAGE 9.17-1 Chapter 12 12A-5, 6 12A-5, 6 Chapter 15 15.6-1, 2 15.6-1, 2, 2a 15.6-32, 33 15.6-32, 33, 33a O

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l O REMOVE THESE PAGES INSERT THESE PAGES 1

l Chapter 17 REMOVE ENTIRE APPENDIX J INSERT NEW APPENDIX J (Last page is 17J-70)

Appendix A j l

A-113, 114 A-113, 114, 114a i Appendix B B-25, 26 B-25, 26 Appendix J

- INSERT NEW APPENDIX J WITH TAB (Last page is J-27)

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p V AMENDMENT 73 OVESTION/ RESPONSE SUPPLEMENT This Question / Response Supplement contains an Amendmer; 73 tab sheet to be inserted following Qi page Amendment 72, October 1982. Page Qi Amendment 73 is to be inserted following the Amendment 73 tab sheet.

Tais Amendment 73 provides a revised Question / Response page* for NRC Question Received Before The Fall of 1981 plus New and Replacement Pages for NRC Questions Received Since the Fall of 1981.

The following Question / Response pages are to be int rted in numerical order behind the appropriate numbered tabs in the PSAR Question / Response Volumes.

REMOVE THESE PAGES INSER. THESE PAGES

• Q001.245-1 thru 6 0001.245-1 QCS760.60-1 QCS760.60-1, 2 QCS760.77-3, 4, 5, 6 7 QCS760.106-1, 2 QCS760.106-1 (V

With the issue of this PSAR Amendment 73, an additional PSAR Binder (Volume 27) is being provided. In addition, new PSAR Volume 26 and 27 Identification Pages are provided. These I.D. pages should be inserted and retained as the first page in PSAR Volume 26 and Volume 27 respectively.

In order to accommodate the existing volume of New Question / Response -

pages plus the anticipated issue of additional New Question / Response pages in future PSAR Amendments, the shifting of Question / Response pages and their associated numbered tabs currently in Volumes 25 and 26 into Volumes 25, 26, and 27 is reconinended. This page shift should be accomplished so that PSAR Volumes 25, 26 and 27 will contain Question / Response Series pages and tabs as shown below:

VOLUME 25 - SHOULD CONTAIN -

Q/R Series 210 thru 410 VOLUME 26 - SHOULD CONTAIN -

Q/R Series 421 thru 721 VOLUME 27 - SHOULD CONTAIN -

Q/R Series 760 thru 810

• 0ld NRC Question / Response Series U)

D

I 1.5.2.8 Sodlum Fires Test Proaram i

1.5.2.8.1 Purnose The purpose of the sodium fires test program is to verify that plant design f eatures f or accommodat i on of sod i um/ NaK sp i l I s i n a i r-f Il l ed cel I s w 11 I result in acceptable cell pressures and structural concrete temperatures. In addition, this test program will be used to demonstrate that the codes used in sodium fire analyses conservatively predict cell accident conditions.

1.5.2.8.2 Proarams The sodium fire experiments have been or will be performed at the Atomics international test f acil ities in Santa Susana, California. The following small scale tests have been ec'.1pleted:

1) A fast spilI (approximately 15 gal / min) of 1000 F sodium onto the fire suppression deck surface

2) A slow spill (approximately 1.5 gal / min) of 1000 F sodium onto the fire suppression deck surface

3) A spray (approximately 15 gal / min) of 1000 F sodium onto the surf ace of the fire suppression deck

4) A fast spill (approximately 15 gal / min) of 1000 F sodium directly into the catch pan beneath the fire suppression deck O

V 5) A spray (approximately 15 gal / min) of 1000 F sodium, onto the surf ace of the fire suppression deck, through a walk grating above the deck l 6) A spray (approximately 15 gal / min) of 600 F sodium onto the surface of the l

fire suppression deck, ihrough a walk grating above the deck The results of the above small scale tests will be documented as the test reports become avail ab! e. In addition to smalI tests, a Iarge scale test w11I be performed using a large-scale model of the CRBRP catch-pan fire suppression deck system to collect spilled sodium under simulated spill conditions. The test f acil Ity is designed to accommodate a volume gas as large as 6600 gallons of 1000 F sodium with a sodium discharge flowrate of approximately 70 GPM.

1.5.2.8.3 Schedule The small scale tests have been completed. The large scale test is planned to be perf ormed in the l ast quarter of 1982.

O Amend. 73 1.5-46 Nov. 1982 l

l . - . . -. _ _. _

1.5.2.8.4 Success criteria The small scale tests successf ully demonstrated fire suppression deck design f eatures to ensure drainage capabil Ity and f Ire-suppression of fectiveness:

o No blockage of drain pipes during spill.

o Post-spill suppression of sodium burning by control of oxygen ingress to sodium pool via oxide plugging of drain pipes and closure of vent iIds on vent pipes.

o No l eakage of sodlum f ran catch pan.

The success criteria f or the large scale test are that the catch pan shall contain the spilled sodium precluding sodium concrete interactions and that resulting test consequences are enveloped by those calculated with the Proj 9ct's methodol ogy.

1.5.2.8.5 Fallback Position if the ef fectiveness of the fire-suppression deck / catch pan system is not demonstrated, alternative techniques to accommodate design basis lIquid metal spill events will be considered and/or prediction of plant design basis accident consequonces will be made with alternative methods.

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Amend. 73 1.5-46a Nov. 1982

1.5.3 References

1. HEDL-SA-771, " Fuel Pin Transient Behavior Technology Appiled to Safety O- Analysis," Presentation to AEC Regulatory Staf f, 4th Regulatory Briefing on Safety Technology, November 19-20, 1974.

2. Haugen, E. G., "Probabilistic Approaches to Design," Wiley Book Co., New York (1%8).

3. Chang, -. e. C., and C. B. Brown, " Functional RellabilIty of Structures,"

Journal of the FranklIh Institute, September, 1973.

4. Fontana, M. H., et al, "Ef f ect of Partial Blockages in Simulated LMFBR Fuel Assembiles," Proc. Fast Reactor Saf ety Meeting CONF-740401-P3,1139 (1974).

5. Bell, C. R. "TRANSWRAP - A Code for Analyzing the System Effects of Large-Leak Sodium-Water Reactions in LMFBR Steam Generators," Proc. Fast Reactor Safety Meeting, CONF-740401-P1, 124 (1974).

6. Division of Reactor Research and Development, USAEC, RDT Standard M3-7T, "Austenitic Stainless Steel Welded Pipe (ASE SA-358 with Additional Requirements)" November,1974.

7. Marr, W. W., et al, " Subassembly-to-Subassembly Failure Propagation:

Thermal Loading of Adjacent Subassembly," Proc. Fast Reactor Safety Meeting, CONF-740401-P2, 598 (1974).

8. Van Erp, J. B., et al, "An Evaluation of Pin-to-Pin Failure Propagation in LMFBR Fuel Subassembiles," Ibid, 615 (1974).

\ 9. Erc' man, C. A., et al ., " Improvements in ModeIIng Fuel-Coolant interactions and interpretation of the S-11 BEAT Test," Ibid, 955 (1974).

60 34 10. Deleted.

pl II. Letter from R. P. Denise (NRC) to L. W. Caffey (ERDA) May 6, 1976.

12. GEFR-00424, UC-798, "A Physicochemical Model for Predicting Sodium Reaction Swelling in Breached LMFBR Fuel and Breeder elements", R. W. Caputi. M. G.

l 57 Adamson, and S. K. Evans, March,1979.

4 O 1.5-47 Amend. 60 Feb. 1981

LIST OF TABIES (Continued)

('] TABLE NO. PAGE NO.

2.3-12 Annual Joint Frequency of Wind Direction and Wina Speed for all Stability Classes 2.3 -27 2.3-13 Annual Joint Frequency of Wind Direction and Wind Speed for Stability Class A 2.3-28 2.3-14 Annual Joint Frequency of Wind Direction and Wind Speed for Stability Class B 2.3 -29 2.3-15 Annual Joint Frequency of Wind Direction and Wind Speed f or Stabil ity Cl ass C 2.3-30 2.3-16 Annual Joint Frequency of Wind Direction and Wind Speed fcr Stability Class D 2.3-31 2.3-17 Annual Joint Frequency of Wind Direction and Wind Speed for Stability Class E 2.3-32 2.3-18 Annual Joint Frequency of Wind Direction and Wind Speed f or Stabil ity Class F 2.3-33 2.3-19 Annual Joint Frequency of Wind Direction and Wind Speed for Stability Class G 2.3-34 O. ,

2.3-20 Annual Joint Frequency of Wind Direction and Wind Speed f or All Stability Classes 2.3-35 2.3 -21 Monthly Wind Data Summaries 2.3-36 2.3-22 Monthly Average Rel ative Humidity Val ues f or the CRB RP S ite 2.3-37 2.3 - 23 Monthly Average Rel ative Humidity Val ues f or Knoxville Airport, 1961-1973 2.3-38 2.3-24 Precipitation Data Summary Oak Ridge Area Station X-10, 1944-1964 2.3-39 2.3-25 Precipitation Summary for the CRBRP Site 2.3-40 2.3 - 26 Snow or Ice Pellet Summary for Oak Ridge City Of fice 2.3-41 2.3 - 27 Monthly Maan Number of Heavy Fog Days for Knoxville and Oak Ridge City Of fice 2.3-42 O

G T.-xi l l Amend. 71 Sept. 1982

LIST OF TABLES (Continuedl TABLE NO. PAGE NO.

2.3-28 Mean Number of Days with Fog, January 1964 -

October 1970 2.3-43 2.3 -29 Number and Percent Occurrence Pasquili Stabil ity Classes CRBRP Permanent Tower 2.3-44 2.3-30 Design Basis Accident x/Q Values for the Exclusion Area Boundary (EM) and Low Population Zone Diste.nces 2.3-45 2.3-31 Fif tieth Percentile x/Q Values for EM and LPZ Distances 2.3-46 2.3-32 Annual Average x/Qs at Various Downwind Distances for Each Wind Sector Based on Permanent Tower Data 2.3-47

2. 4-1 Clinch River Stream Gage 1.ocations 2.4-61 2.4-2 Average Monthly Discharge in Day-Second-Feet Melton Hili Dam 2.4-62 2.4-3 Watts Bar Reservoir Elevations 1964-1973 2.4-63

2. 4-4 Average Daily Maximum, Minimum and Mean Temperatures f or Each Month (1963-1971) 2.4-64 2.4-5 Cl inch River Water Qual Ity Data 2.4-65, 66 2.4-6 Cl inch River Watershed Municipal Weste Discharges 2.4-67 2.4-7 Cl inch River Watershed industrial 2.4-68, 69 Waste Discharges 70 - 74 2.4-8 Flood Elevetion Summary, CRBRP 2.4-75 2.4-8a Probable Maxime. Precipitation (PMP)

Distribution 2.4-75a 2,4-9 Probable Maximum Storm Rainf all and Precipitation Excess 2.4-76, 77

( 2.4-10 Deleted 2.4-11 Unit Hydrograph Data 2.4-79, 80 0

2-xlv Anend. 73 Nov. 1982

LIST OF FIGURES (Continued)

F1GURE No. PAGE NO.

2.3-13 Topography Surrounding Clinch River Site 2.340 2.3-14 Site Topographic Map 2.341 2.3-15 Topographic Profile Cross Sections f rom Site 2.3-62 2.3-16 Topographical Cross Section including Meteorological Tower and Center of Containment Buil ding 2.3-64 2.4-1 Site Location and CRBRP Layout 2.4-107 2.4-2 Topography of ClIr.ch River Site 2.4-108 2.4-3 Arrangement of Pl ant Structures P1an 2.4-109 2.4-4 Flow Duration Curve 2.4-110 2.4-5 Wastewater Discharges, Clinch River Watershed 2.4-111

2. 4-6 Downstream Profile of the Clinch River 2.4-112 2.4-6a Probable Maximum Precipitation (PMP) Rainf all 0 Time Distribution Adopted Standard Mass Curve 2.4-112a 2.4-6b Plant Area Grading Plan 2.4-112b

2. 4-6c Plant Area Sections 2.4-112c

2. 4-6 d Detail s f or Probable Maximum Precipitation at Saf ety Rel ated Facii Itles 2.4-112d

2. 4-7 Probable Maximum March Isohyets (7980 sq. mi.)

Ist 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (In.) 2.4-113 2.4-8 Probable Maximum Isohyets (4456 sq. mi., Cl inch River Drainage), 1st 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (In.) 2.4-114 2.4-9 Rainf all Time Distribution Adopted Standard Mass Curve 2.4-115 O 2-xix Amend. 71 Sept. 1982

U ST OF FIGURES (Continued)

FIGURE NO.

2.4-10 72-Hour March Probable Maximum Storm Depths (In.)

Tennessee River Watershed Above Watts Bar Dam 2.4-116 2.4-11 CFERP Hydrologic Moriel Sub Areas 2.4-117 2.4-12A 6-Hour Unit Hydrograph 2.4-118 2.4-128 2-Hour Unit Hydrographs 2.4-119 2.4-12C 2-Hour Unit Hydrographs 2.4-120 2.4-12D 6-Hour Unit Hydrograph 2.4-1 21 2.4-12E 6-Hour Unit Hydrographs 2.4-122 2.4-12F 6-Hour Unit Hydrographs 2.4-1 23 2.4-12G 6-Hour Unit Hydrographs 2.4-124 2.4-12H 6-Hour Unit Hydrographs 2.4-125 2.4-1 21 6-Hour Unit Hydrographs 2.4-1 26 2.4-12J 6-Hour Unit Hydrographs 2.4-1 27 2.4-13 Hydrologic Model Verification - 1973 FIood 2.4-127A (Melton Hili) 2.4-14 Hydrologic Model Verification - 1973 Flood (Watts Bar) 2.4-127B 2.4-15 Stage Gages Used for Unsteady Flow Model Verification 2.4-128 l

2.4-16 1967 FI ood-Watts Bar Reservoir Unsteady FI ow Model Verification 2.4-129 2.4-17 1969-70 FIood, Watts Bar Reservoir Unsteady FIow Model Verification 2.4-130 2.4-18 Steady State Model Verif ication Cl inch River Mile 17 2.4-130A 2.4-19 Hydrologic Model Verification-1957 and 1963 Floods 2.4-131 i

2.4-20 Hydrologic Model Verification-1963 and 1967 Floods 2.4-132 2.4-21 Hydrologic Model Verification-1963 and 69-70 Floods 2.4-133 0

2-xx Amend. 73 Nov. 1982

LIST OF FIGURES (Continued) d FIGURE NO. PAGE NO.

2.4-22 Hydrologic Model Verification-1948 and 1%9-70 Floods Emory River at Oakdale 2.4-134 2.4 -23 CRBRPS Probable Maximum Flood Discharge 2.4-135 2.4-23a Norris Probable Maximum Flood Norris Hydrographs and Headwater Elevations 2.4-135a 2.4-24 CRBRPS Probabie Maximum FIcod Elavations 2.4-136 2.4-25 Tennessee ValIey Reglon 2.4-137 2.4-26 A Multiple-Purpose Reservoir Operations -

Watts Bar Project 2.4-138 2.4-26B Multiple-Purpose Reservoir Operations -

Fort Loudoun Project 2.4-139 2.4-26C Multiple-Purpose Reservoir Operations -

Watauga Project 2.4-140 2.4-26D Multiple-Purpose Reservoir Operations -

Fontana Project 2.4-141 2,4-26E Multiple-Purpose Reservoir Operations -

South Holston Project 2.4-142 2.4-26F Multiple-Purpose Reservoir Operations -

Douglas Project 2.4-143 2.4-26G Multiple-Purpose Reservoir Operations -

l Cherokee Project 2.4-144 l

2.4-26H Multiple-Purpose Reservoir Operations -

Boone Project 2.4-145 2.4-261 Multiple-Purpose Reservoir Operations -

NorrIs Project 2.4-146 l 2.4-26J Tellico Project - Multiple - Purpose Reservoir Operations 2.4-146a 2.4-27 Norris Reservoir Areas and Volumes 2.4-147 2.4-28 Norris Dam Plan Elevations and Sections 2.4-148 O 2-xxl Amend. 71 Sept. 1982

LIST OF FIGURES (Continued)_

FIGURE A PAGE NO.

2.4-29 Norris Dam - Spillway + Non-Overflow Results of Analysis f or OBE + 1/2 PMF 2.4-149 2.4-30 Norris Dam - Analysis f or OBE and One Hal f PNF-Assumed Condition of Dam Af ter Fail ure 2.4-150 2.4-31 Norris Dam - SSE + 25 Year Flood Judged Condition of Dam Af ter Fail ure 2.4-151 2.4-32 Melton Hill Dam - General Plan Elevation and Sections 2.4-152 2.4-33 Fort Loudoun Dam - General Pl an Elevation and Sections 2.4-153 2.4-34 Tell ico Dom - General Pl an Elevation and Section 2.4-154 2.4-35 Fort Loudoun Dam Rating Curves 2.4-155 Tellico Dam Rating Curves 2.4-155a l2.4-35a 2.4-36 CR3RP Probable Maximum Flood f or Loudoun Dam Out f l ow 2.4-156 2.4-37 General Plan Elevation and Sections (Main Dam Works)-Watts Bar Dam 2.4-157 2.4-38 Location Plan and Section (West Saddle Dike)

Watts Bar Dam 2.4-158 2.4-39 CRBRP Probabl e Maximum Fl ood Watts Bar Headwater Elevations 2.4-159 2.4-40 Watts Bar Dam Rating Curves 2.4-160 2.4-41 CRBRP Probable Maximum Flood Watts Bar Dam Outfl ow 2.4-161 2.4-42 Headwater Rating Curves, Norris Dam 2.4-162 2.4-43 CRBRP Seismic Analysis Hydrographs 2.4-163 2.4-44 CR3RP W ind Wave Fetch 2.4-164 2.4-45 Watts Bar Reservoir 2.4-165 2.4-46 Combination of Typical Turbine Operations 2.4-166 Amend. 73 2-xxil Nov. 1982

Three minor tributaries which enter the Clinch River between Norris Dam and Melton Hill Dam are Beaver Creek, Bullrun Creek and Hinds Creek. These three streams enter from the south at CRM 39.6, 46.7 and 65.8, respectively.

(Ref.1) Annual average flows and peak flood data for these creeks are not applIcabie to the Site because they enter the Clinch River above Melton Hili Dam. Poplar Creek, a minor tributary below Melton Hill Dam, enters the Cl inch River f rom the north at CRM 12.0. The average annual flow of Poplar Creek is 260 cf s and drainage area at the mouth is 136 square miles.

Several other smalI streams and sioughs enter the Clinch River near the Site; however, they are not considered to be significant tributaries from the standpoint of water flow contribution to the Clinch River. Caney Creek which enters f rom the south at CRM 17 has a drainage area at the mouth of 8.27 square miles and an average flow of 14 cf s. Popl ar Springs Creek at CRM 16.2 has a drainage area of 3.01 square miles at the mouth and an average flow of 5 cfs. Grassy Creek entering from the north at CRM 14.5 has a drainage area of 1.95 square miles and an average flow of 3 cf s. The combined average flows of these creeks total 22 cf s which is only 0.5 percent of the Clinch River fl ow at the S ite.

2.4.1.2.3 Reservoir Water Levels The Site is located on an arm of the Watts Bar Reservoir which extends up the Cl inch River. Thus, the water elevation at the Site is influenced by the operation of Watts Bar Dam. Elevation at the bottom of the Clinch River channel at the Site ranges between 719 and 720 feet above mean sea level (MSL). Water depth at the Site is equal to or greater than the dif ference between the pool elevation at Watts Bar Dam minus this bottom elevation as can be seen in the downstream prof Ile of the Cl inch River, Figure 2.4-6.

l Watts Bar Reservoir is a mul tiple-purpose reservoir providing power generation, navigation aid and flood control. TVA generally maintains a pool elevation between 740 and 741 MSL during the spring and summer months (mid-April through September) and a winter pool elevation of 735 to 737 MSL for the remainder of the year.

During the 32 years of record since the initial filling of Watts Bar Reservoir, TVA has been able to follow closely the above plan of normal operation. Suf ficient inflow has been available each year to raise the reservoir from winter level to summer level on schedule. The water surf ace elevation at the Site is normally one to two feet higher than the level measured at Watts Bar Ban as a result of a backwater. Since 1942, the minimum O

Amend. 73 Nov. 1982 2.4-3

elevation of Watts Bar Reservoir was 733.44 MSL and occurred on March 20, 1945, (Ref. 8). A maximum elevation of 745.40 MSL occurred on March 17, 1973. (Ref. 7) Figure 2.4-26a shows the normal operating level for Watts Bar f

Reservoi r. Table 2.4-3 shows the monthly maximum, minimum and average Watts Bar Reservoir elevations f or the period 1964-1973.

Releases from Norris and Molton Hill Reservoirs, both upstream of the Site on the Clinch River, can be used to regulate flows at the Site. Although Norris Dam is the prime regulator of flow, Melton Hill Dam can influence low flows at the Site. Normal minimum pool stage at Melton Hill Reservoir is 790 MSL.

(Ref. 8)

Norris Reservoir is a multiple-purpose reservoir providing power generation, flood control and low flow argumentation. The normal minimum pool elevation is 960 MSL with storage of approximately 260,000 day-second-feet between elevations 960 and 900 MSL. Although not a primary purpose, stored water below minimum pool elevation 960 is available for low flow augnantation in periods of drought. Release below elevation 960 requires specific approval of the TV A Board of Directors. Power generation at Norris can be naintained to about elevation 900 MSL. Of all the annual maximum elevations . ecorded, the lowest annual maximum elevation of Norris Reservoir was 993.8 MSL and occurred in June 1954. (Ref. 8)

Releases f rom Fort Loudoun Reservoir, located on the Tennessee River 72.4 miles upstream from Watts Bar, can be used to control the Watts Bar pool el ev at i on. Normal .ninimum pool elevation of Fort Loudoun Reservoir is 807 MSL. (Ref. 8) Intlows into Watts Bar Reservoir from the Tennessee River are more than capable of maintaining the minimum pool elevation of Watts Bar Reservoir even under extreme conditions.

Tellico Dan, closed in 1979, is located on the Little Tennessee River at mile 0.3. Tellico Reservoir is connected to Fort Loudoun Reservoir by an uncontrol led canal. Except during large floods, inflows to Tellico Reservoir l are discharged through F mt Loudoun Reservoir.

2.4.1.2.4 Water Flow Stream gages had been me lotained by the U.S. Geological Survey on the Clinch River at the three locat:ols listed in Table 2.4-1. Gages were maintained at these locations f or variou, time periods f rom 1936 through 1968. (Ref.

3,4,5,6). At the present ti.3e, no stream gages are being operated in the vicinity of the Site.

Based upon these stream gage records f rom the three locations, the average flow of the Clinch River was 4,561 cf s (Ref. 3,4,5,6). The maximum discharge during the period was 42,900 cf s and occurred on February 9,1937, (Ref. 5) bef ore the closing of Mol ton Hil l Dam. Based on discharge records f rom Melton Hill Dam since the closing in 1963, (see table 2.4-2) the average annual flow is about 4,600 cfs at the Site. (Ref. 8) The maximum hourly average release was 43,400 cf s and the maximum daily average release was 26,900 cf s; both occurred on March 16, 1973 at Mel ton Hil l Dam. (Ref. 8).

O Amend. 73 2.4-4 Nov. 1982

l 2.4.2.2 Flood Design Considerations i Table 2.4-8 compares the maximum flood level determined for the rain flood and U seismic events specified in Regulatory Guide 1.59, more completely described below. The alternatives evaluated under each category are described in 2.4.3 and 2.4.4.

The computed Probable Maximum Flood (PMF) level at the plant site from an occurrence of the most severe sequence of storms, as defined by the National l Weather Service, is elevation 778.8 at mile 18 and 777.2 at mile 16 excluding the offect of wind waves.

This compares with elevation 777.5 at mile 18 and elevation 776.0 at mile 16 previously given in the PSAR. The dif ferences result from a reevaluation of and refinements in the Tennessee River watershed model and includes Tellico Dam, which was closed in 1979.

A conservatively high velocity of 40 MW wind over land from the most adverse direction, was adopted to associate with the PMF Crest. The probability that this hIgh velocity wind occurs on the same specific day that the PMF would crest is extremely remote. It has been estimated that the probability of the flood g wind occugng on the same day in a given year is on the order of I X 10 to 1 X 10 (Ref. 11).

Waves and runup are applicable to the plant only at MILE 16. At Mile 18, ground levels adequately shield the plant fran coincident winds. For the 40 MW wind from the most critical direction 99.6% of the wind waves were S computed to be less than 2.4 feet high, crest to trough, resulting in a q') maximum water surf ace elevation of 778.80 in the reservoir approaching the plant site. Runup would be about 3.8 feet to elevation 781.0 on a vertical wall and 2.8 feet to elevation 780.0 on a smooth slope of 3:1. The Probable Maximum Precipitation (PMP) and Flood Flow are discussed in Sections 2.4.3.1 and 2.4.3.4.

I The plant site and upstream reservoirs are located in the Southern Appalachian Tectonic Province and, therefore, subject to potential moderate earthquake forces with possible attendant dam f ailure. All upstream dams, including those on the Tennessee River, whose f ailure in a seismic event has the l

potential to cause flood problems at the plantsite were investigated as described in 2.4.4. Studies to determine the potential failure of upstream dams f ran PMF conditions, are also described in the same section.

l The condition producing the maximum flood level at the plant site is the postulated f ailure of Norris Dam under the force of an operational basis earthquake (OBE) coincident with one-hal f of the PM7 with the postulated attendant failure of Melton Hill Dam. This would produce a maximum flood level of 804.3 feet at Mlle 18 and 798.2 at Mlle 16 with a peak discharge of 921,000 cfs. The situations considered are consistent with Regulatory Guide 1.59.

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1 G 2,4-7 Amend. 73 Nov. 1982

cor the analysis of the f ailure of Norris Dam, a debris level was postulated at e!wation 970 feet which is considered the result of a logical mode of fallare. The f ailure mode is discussed in detall in Section 2.4.4.

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9; 2.4-7a Amend. 73 Nov. 1982

When the effects of wind wave and runup are added to still water O

V elevations 804.3 and 798.2, the maximum design water levels will be established to be at elevation 809.2 and 803.1 on a vertical wall and elevation 807.9 and 801.8 on a smooth 3:1 slope, at mile 18 and 16 respectively.

All Seismic Category I structures housing safety-related facilities, systems and equipment are designed for or are protected from the highest flood elevation. The plant grade in the Reactor Containment Building area is established at elevation 815.0 which is well above the maximum flood runup level.

Hydrostatic pressure, buoyancy and dynamic wave effects will be censidered in the design of all Category I structures either completely or partially submerged under the maximum design flood condition. Accesses and penetrations located below the flood level will be reduced to the minimum requirement, and will be designed and constructed as watertight elements.

Specific analysis of Clinch River flood levels resulting from oceanfront surges and tsunamis is not required because of the inland location of the plant. Snow melt and ice jam consider-ations are also not required because of the temperate zone location of the plant. Flood waves from landslides into upstream reservoirs required no specific study, in part because of the absence of major elevation relief in nearby upstream reservoirs '

and because the prevailing thin soils offer small slide volume O potential compared to the available retention space in reservoirs.

2.4.2.3 Effects of Local Intense Precipitation The overall site drainage facilities will be designed for 3.5 inches rainfall in one year. This rainfall corresponds to the maximum rainfall expected during a period of 100 years (Table 2.3-1).

The drainage facilities for safety-related structures will be in-vestigated for local flooding resulting from Probable Maximum Precipitation (PLIP) as specified by the Hydrometeorological Branch of the National lleather Services. (Ref.13) The eight (8) hour PMP depth is 29.5 inches with a maximum one (1) hour depth of 14 inches (Table 2.4-8a). PMP inten-sities beyond 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are less than 1 inch per hour and, therefore, not critical for defining site maximum flood conditions. Time distribution of the 8-hour PMP storm is based upon consideration of the time distribution o." maximum observed storms and the time distribution of design storms adopted by the Corps of Engineers and the Soil Conservation Service (Ref. 47).

The adopted sequence conforms closely to that used by the Corps of Engineers.

No precipitation loss is applied to the 14-inch maximum hourly rainfall. 7 O

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2.4-8 Amend. 7 Nov. 1975

e) The following drainage system will be used in the Access Road and Railroad Area. Pipe culverts will be provided where the drainage channels are interrupted by access roads and railroads. Drainage ditches will be provided along the sides of the road and the rail-road. In high cuts, drainage ditches will be provided at the top or along the slope to intercept surface flow and to prevent excessive erosion of the cut face. These ditches will be led into natural water courses or pipe culverts.

Since the maximum calculated overtopping resulting from PMP is not expected to exceed six inches, there will be no danger of water ponding against safety-related structures.

l Natural drainage will be affected by the plant construction in approxi-mately 100 acres of the 1364 acre Clinch River Site. On high fills, berms will be built along the edge of the fill to control surface flow on the top of the embankment except at points where paved channels will be provided to carry the flow down the embankment. Drainage ditches will be lined when the velocity of flow is abnormally high and at sharp turns, if any.

Settling basins of sufficient capacity will be provided to receive the runoff discharge from the plant drainage systems before discharge into the Clinch River. The purpose of providing the settling basins is to elimi-nate most of the suspended solids in the effluent before discharge, in accor-dance with local, state and federal regulations for effluent discharge. 7

/3 N)

O LJ 2,4-9 Amend. 7 Nov. 1975

2.4.3 Probable Maximum Flood (PMF) on Streams and Rivers The probable maximum fIcod (PMF) would result f rm an occurrence of the probable maximum storm as defined by the National Weather Service. The flood flows and elevations f ran this storm dofine an upper limit of potential flooding at the plant site f rom meteorological conditions.

Occurrence of the PMF as determined and applled in this study is extremely unlikely. The postulated combination of events results in a probability of occurrence which approaches zero. The events combined include a main storm with rainf all volumes which are the physical upper limit that the present climate can produce, an assumed antecedent storm amounting to 40 percent of -

the main storm volume, which exceeds the maximum recorded on the water shed to date, and assumption of an exact centering of the storm to cause that combination of Clinch and Tennessee River flows which produces maximum flood levels at the plant. In applying PMF elevations, further conservatism is introduced by adding the runup due to a 40-mile-per-hour overland wind f rom the most critical direction.

Evaluation of seasonal and areal variations of probable maximum storms l described in 2.4.3.1, and 2.4.3.4, showed that the PMF level at the Clinch River Breeder Reactor Plant (CRBRP) site would be caused by a sequence of two storms occurring in March and centered in the water shed above Watts Bar Dam.

The flood crest at the plant site would be augmented by the f ailure of Fort Loudoun and Tellico Dams, upstream on the Tennessee River, and the nonoverflow section at Melton Hill Dam, upstream on the Clinch River. The estimated maximum discharge at the plant site would be 258,000 cfs. The PMF elevation at the plant site would be 778.8 at Mile 18 and 777.2 at Mlle 16, including the three upstream dam f ailures and f ailure of the earth anbankment at Watts Bar Dam, downstream on the Tennessee River, but excluding any wind wave offects.

O 2.4-10 Anend. 73 Nov. 1982

2.4.3.1 Probable Maximum Precipitation Probable maximum precipitation (PMP) for storms creating

h. maximum flood conditions at the CRBR plant site has been defined L/ for TVA by the Hydrometelorological Branch of the National Weather Service in Hydrometeorological Reports Nos. 41 and 45 (Ref.12,13).

In addition, the Clinch River water shed PMP information contained in Report No. 45 was extended by.the Hydrometeorological Branch to cover the total water shed above Watts Bar Dam and to provide PMP for the 4458-square mile Clinch River watershed (Reference 13a). These 7 define depth-area-duration characteristics of rainfall and their seasonal variations and antecedent storm potentials. Because the water shed lies in the temperate zone, snowmelt is not a factor in generating maximum floods at the plant site. (See page 97 of Report No. 41.)

Four basic storms with five possible isohyetal patterns described in Reports Nos. 41 and 45 and Reference 13a were examined to l7 determine which would produce maximum flood levels at the plant site. I One basic storm would produce PMP depths on the 21,400-square-mile water shed above Chattanooga. Two potential isohyetal patterns are presented in Report No. 41 for this storm. Computations for earlier TVA program studies determined that the downstream centering would be most severe for the CRBRP.

A second basic storm, described in Report No. 41, would produce PMP depths on a 7,980-square mile water shed centered in the Valley below the major tributary dams. The isohyetal pattern I3; for the 7,980-square-mile storm is shown in Figure 2.4-7. The pattern is not orographically fixed and can be moved parallel to the long axis northeast and southwest along the Valley. Critical position centers this storm at Bulls Cap, Tennessee (50 miles northeast of Knoxville ).

The third basic storm would produce PMP depths on the i 4,458-square-mile Clinch River water shed and is described in Reference 13a. The isohyetal pattern for this storm is shown in Figure 2.4-8. This pattern is not orographically fixed and can be moved parallel to the long axis northeast and southwest along the Valley. Centerings both upstream and downstream of Norris Dam were tested.

The fourth basic storm would produce PMP depths on the 2912-square-mile watershed above Norris Dam and is described in Report No. 45. The isohyetal pattern of the storm is shown in figure ?-23 of that report.

The pattern does not cover the full 2912-square-mile watershed. Full coverage was obtained by extending the lowest isohyet uniformly to the watershed boundary. The storm was centered 28 miles northeast of Tazewell, Tennessee, to obtain the maximum watershed rainfall. 7 A 3-day antecedent storm separated by a 3-day dry period from a 3-day main storm is recommended in Reports Nos. 41 and 45.

For the 21,400- and 7,980-square-mile storms the antecedent storm is 40 percent of the main storm with uniform areal distribution pd as recommended in Report No. 41. In the 4,458- and 2912-square-mile Clinch River storms, the antecedent storm is 30 percent of the main storm with uniform areal distribution as recomended in Report No. 45.

Amend. 7 2.4-11 Nov. 1975

A time distribution pattern was adopted for all antecedent and main storms based upon major observed storms transposable to the Tennessee Valley and distributions used by Federal agencies (Ref. 23). The adopted distribution is shown on Figure 2.4-9.

The of fects of seasonal variations in storm anounts and storm centerings on piant sito fIood Ievels were examined in suf fIciont detalI to assure that the situation producing the maximum flood level was found. July prevailed for the 4,458 and 2,912 square-mile Clinch River Basin storm and March prevailed for the others although a May, 7,980-square-mile storm was examined in this process.

The controlling probable maximum storm is the one for 7,980 square miles centered at Bulls Gap, Tennessee, which would follow an antecedent storm commencing on March 15. Over the Watts Bar water shed this results in (1) an l antecedent storm producing an average precipitation of 6.9 inches in 3 days, (2) a 3-day dry period, and (3) a main storm producing an average precipitation of 17.2 inches in 3 days. Figure 2.4-10 is an isohyetal map of the maximum 3-day PMP. Basin rainf all depths by subwater sheds are given in Tabic 2.4-9.

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2.4-12 Amend. 73 Nov. 1982 .

l l

2.4.3.2 PRECIPITATION LOSSES i A multivariable relationship, used in the day-to-day operation of the TVA system, has been applled to determine precipitation excess (P,) directly. The relationships were developed from observed data. They relate precipitation excess to the rainf all, week of the year, geographic location, and antecedent precipitation index (API), in their application P becomes an increasing f raction of rainf all as the storm progresses in tim,e and becomes equal to rainf all when from 7 to 16 inches have f allen.

In the CRBRP application, median API conditions were used at the start of the antecedent storm. This storm is so large, however, that the P, in the main storm is not sensitive to this earller condition.

For review purposes, precipitation losses have been determined by subtracting l P f rom rainf al l. In the controlling probable maximum storm the loss is 33 p$rcent of rainf all in the 3-day antecedent storm and 10 percent of rainf all in the 3-day main storm. Table 2.4-9 displays the API, rain, and Pe f r each of the subwater sheds used in CRBRP water shed model.

2.4.3.3 RUNOFF MODEL The runof f model used to determine Clinch River flood hydrographs at the plant l site is divided into 41 subareas shcwn on Figure 2.4-11. The model comprises the entire 17,310-square-mile Tennessee River water shed above Watts Bar Dam.

This boundary, downstream from the CRBRP site, is appropriate because flood-Induced headwater level at Watts Bar Dam may exert backwater influence upstrean to the site.

O The runof f model used in this amendment to the PSAR dif fers f rom that used previously because of refinements made in some elements of the model during PMF studies for other nuclear plants and those made f rom Information gained f ran the 1973 flood, the largest that has occurred during present reservoir conditions. Changes are identified when appropriate in the text. They include both additional and revised unit hydrographs and additional and revised unsteady flow strean course models.

Unit hydrographs were developed for each unit area f rom maximum flood hydrographs either recorded at stream gaging stations or estimated 'from reservoir headwater elevaticn, inflow, and discharge data. The number of unit areas has been increased from 39 used previously to 41. The differences include:

1. Use of the model developed for the Phipps Bend study which combined the two unit eroas for Watauga River (Sugar Grove and Watauga Iocal) into one unit area and divided the Cherokee to Gate City unit area into two unit areas (Surgoinsville local and Cherokee local below Surgoinsville).

O i

2.4-13 Amend. 73 Nov. 1982

2. Changes to add an unsteady flow model for the Fort Loudoun-Tellico Dam complex which included dividing the lower Little Tennessee River unit area into two unit areas (Fontana to Chilhowee and Chilhowee to Teilsico) and dividing the French Broad River local into two unit areas (Little Pigeon River at Sevierville and French Broad River local).

In addition, 7 of the unit graphs have been revised. Figure 2.4-12, which contains 10 sheets, shows the unit hydrographs. Table 2.4-11 contains essential dimension data for each unit hydrograph and identification of those hydrographs which are new or revised.

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2.4-13a knend. 73 Nov. 1982

The 12,177-square-mile water shed above the Fort Loudoun-Tellico complex comprises 70 percent of the area above Watts Bar Dam. A detailed model is needed for this area, especially to assess the potential for and likely consequence of f ailure of several high tributary dams and of the Fort Loudoun-TelIico comptex itsel f.

Because of the large, assured detention capacity of Norris Reservoir--7 inches, controlled, in March on the 2,912-square-mile water shed plus 4.5 inches, surcharge capacity in the PMF-flood outflows are largely a matter of inflow volume and are not significantly sensitive to inflow peaks and their timing. For this reason the entire upstream water shed may be represented by a single subarea (No.1). Dual-peak performance is a consistent characteristic of the subarea as adequately represented by its unit hydrograph, Figure 2.4-12, Sheet 1.

In contrast, the 431 square miles of water shed, from Norris Dam downstream to Melion Hill Dam was diviaed specifically for this study into quite small subareas (Nos. 2-11) as shown in the Inset on Figure 2.4-11. This was largely to provide accuracy of local inflow location points for the unsteady flow model of Melton Hill Reservoir considered potentially necessary because of the shape variations of contributing subarea units. In retrospect, because of the early timing of noncontrolling local inflow peaks, less detail would have been suitable.

Unit hydrographs are used to compute flows frm the subareas. The subarea flows are combined with appropriate time sequencing or channel routing to compute inflows into the most upstream reservoir. Floods are routed through p the reservoirs using standard techniques. Resulting outflows are combined with additional local inflows and carried downstream using appropriate time sequencing or routing procedures.

Unit hydrographs derived frm observed records were developed from data for the largest floods available using procedures described by Newton and Vinyard (Ref. 14). For subareas 2,3,5-12, and 15 where records were not available, synthetic unit hydrographs were developed using the procedures described by the Corps of Engineers (Ref. 15). Subwater sheds 4 and 13 were gaged upstream f rm the mouth, and unit hydrographs developed f rm these records were O

2.4-14 Anend. 73 Nov. 1982

modified for application at the river mouth. Figure 2.4-12, which contains 10 sheets, shows the unit hydrographs graphically. Table 2.4-11 contains essential dimension data for each unit hydrograph.

Tributary roservoir routings except for Tellico and Melton Hill Reservoirs used the Goorich somigraphical method and flat pool storage. Th!s differs frm the previous submission in that an unsteady flow model has been added for the Fort Loudoun-TellIco compiex.

l Unsteady flow techniques (Ref.16) were used in Fort Loudoun-Tellico, Watts Bar and Molton Hill Reservctrs. Prescribed boundary conditions are inflow hydrographs at ine upstream boundary, local inflows, and headwater discharge relationships at the downstrena boundary based upon standard operating rules, or on rating curves when geometry controlled.

The unsteady flow mathematical model for the 49.9-mile long Fort Loudoun Reservoir was div ided into 24, 2.08-mile reaches. The model was verifled at 3 gaged points within Fort Loudoun Reservoir using 1963 and 1973 flood data.

The uns+eady flow tr.odel was extended upstream on the French Broad and Holston Rivers 17 Douglas and Cherokee Dams, respectively. The French Broad and Holston River unsteady flow models were verified at one gaged point each at mile 7.4 and 5.5, respectively, using 1963 and 1973 flood data.

The Littlo Tennessee River vas modeled f rm Tellico Dam, mile 0.3, through Tellico Rose.'voir to Chilhowee Dam at mile 33.6 and upstream to Fontana Dam at fulle 61.0. The model for Tellico Reservoir to Chilhowee Dam was tested for adequacy by comparing its results with steady-state profiles at 1,000,000 and 2,000,000 cfs computed by the standard-step method. Minor decreases in conveyance in the unsteady flow tr.odel yloided good agreement. The average conveyance correction found necessary in the reach below Chilhowee Dam to make the unsteady flow model agree with the standard-step method was also used in the river reach from Chilhowee to Fontana Dam.

The Fort Loudoun and Tellico unsteady flow models were joined by a canal unsteady flow model. The extel was modeled with five equally-spaced cross-sections at 525-feet intervals for the 2100-foot long canal.

The unsteady flow mathmatical model for Watts Bar Reservoir consists of two units combined in a junction model. These are (1) the Tennessee River from Watts Bar Dam, Mile 529.9, to Fort Loudoun Dam, Mile 602.3, and (2) the Clinch River mbayment f rom its mouth to Molton Hill Dam, Mile 23.1. The model for the 72.4-mile Tennessee River unit was divided into thirty-four 2.13-mile reaches. The model for the 23.1-mile-long Clinch River mbayment was divided into twenty-two 1.05-mile reaches.

The unstead) flow Molton Hill Reservoir model extending to Clinton, about MIIe 59, and upstream on the Clinch River to Norris Dam (Mile 79.8) was divided into twenty-six 2.18-mil e reaches. The model was verified by reproducing the floods of 1967 and 1969-70 at gages at Clinton, below Norris Dam, and Norris Dam tallwater. These verifications are not exhibited because the reservoir model is a part of the total water shed model above Melton Hill Dam which was O

~

i 2.4-15 Amend. 73

' Nov. 1982

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

I, verifled using the 1%9-70 flood and is a part of the total water shed model i

above Watts Bar Dam which was verified using the 1963 and 1%9-70 floods.

These verifications are described in later pages, i

The Watts Bar Reservoir unsteady flow model was verf fled at four points along

, the Tennessee River and at Melton Hill Dam on the Clinch River embayment.

Figure 2.4-15 shows that these locations span the reservoir in a practical l

i f

f 1

O f

1 a

I

\

O 2.4-15a Amend. 73 Nov. 1982

mannor for verification purposes. Floods of March 1967 and December-January 1969-70 woro selected f or this verif ication because they are largo and because of the completeness of observed data for them. Watts Bar outflows were specifled. Figuros 2.4-16 and 2.4-17 for 1967 and 1969-70 respectively show good agremont betwoon computed and observed elevation hydrographs at the fIvo verification points, it is impossible to verify unsteady flow models with actual data approaching the magnitude of the PMF; however, a serios of uniform flows computed by both the unsteady flow model and the standard steady flow method woro compared.

This was dono for both Watts Bar and Melton Hill Reservoirs with good agroment. A ccuparison at Miie 17.0 near the piant site in rating curyo f orm, Figuro 2.4-18, shows agrement within a maximum 1-foot dif f erence. The unsteady flow model produces the highor levels.

The total water shed runof f model was veriflod at five locations critical to the study: the Tonnessoo River at Fort Loudoun and Watts Bar Dams, the Clinch River at Norris and Molton Hill Dams, ar.d the Emory River at Oakdale. The Emory River drainage area of 865 square miles enters the Clinch River at Mile 4.4 and can influence plant flood levels in some flood situations. Emory River flows are not regulated.

Model verification at Fort Loudoun Dam used the largo floods of March 1973 and March 1963. This dif fers f rcm the previous submission in that the 1973 flood was added for verif ication replacing the 1957 flood. Observed voltanes of precipitation excess wero inputs, and reservoir operations were simulated by speci f y ing observed headwater level s. Comparison betwoon observed and computed outflows is shown in Figure 2.4-19.

The medel predicts about 9 porcont high in the 1973 flood and about 10 porcont high I.. the 1963 flood. The model is considorod to be fully adequate.

Norris Dam and Rosorvoir inflow was modelod with a 6-hour unit hydrograph derived f rcm observed data in which timo variant flow is estimated frm observed changes in reservoir volumo and observed project outflows. A best unit hydrograph was determined by the Newton-Vinyard proceduro (Ref.14) using fIoods of 1957 and 1963. VorIfIcatIon using the 1963 fIood and a 1967 fIood, shown on Figuro 2.4-20, is good.

NorrIs Roservolr was modeled by the Goodrich steady fIow method of fIood routing using 3-hour routing intervals. Its verification in the 1963 and 1967 floods, shown on Figuro 2.4-20, speciflod observed headwater levels. This is l

an extrmo test because minuto changes in headwater elevation creato largo changos in volumo of water in the reservoir. Headwater elevations cannot be l observed with total procision. Thus, smal l pl us-and-minus errors in proscribed headwater create magnified plus-and-minus discrepancy in routed outflow to maintain volume continuity in the model. This results in " bouncy" cmputed outflows that are obvious on Figure 2.4-20. Under these i

circumstancos only a general agroment can be expected, and this has been l ach l oved. The verification of Norris outflows used inflows computed f rm procipitation excess.

O 2.4-16 hnend. 73 Nov. 1982

Water shed model verification was carried on down the Clinch River to Melton l HiII using the 1969-70 and 1973 fIoods. The 1973 fIood is the Iargest that Os has occurred since the project was completed. Comparison of observed and computed discharges for the 1973 flood is shown on Figure 2.4-13 and for the 1%9-70 flood on Figure 2.4-21. " Bounce" In the computed outflows is apparent and stems from the same cause as explained for Norris Dam but is less severe because of the small reservoir. Nogative flows are attributable to the same cause. Except for " bounce" the verification is good and predicted a somewhat high peak.

The Emory River runof f model is a 4-hour unit hydrograph. It was prepared f ra the observed records for floods in 1948,1957, and 1963 at Oakdale using the mathematical procedure described by Newton and Vinyard. The Oakdale drainage area, 764 square miles, is nearly 90 percent of the total Emory River water shed. The Emory River runof f model verification at Oakdale is shown in Figure 2.4-22, using the 1948 and December 1969 floods. These were the largest floods with adequate records at the time of this analysis. The model predicts the 1948 hydrograph closely except at the sharp peak, the duration of which is less than one-hal f day. Predicted peak is 20 percent high. The predicted hydrograph is good also in the 1969-70 flood, but is 14.5 percent low in a peak approaching a 1-day duration. Emory River enters the CIinch River about 12 miles downstream from the CRBRP site. Its flow contributions are of short duration and are timed ahead of much broader Clinch River peaks with the only ef fect on plant site flood levels through backwater influence downstream frm Mile 4.4. Hence, Emory River floods have minimal of fect on plant site flood levels. The water shed model is fully adequate in this circumstance.

O V Model verification was continued downstream on the Tennessee to Watts Bar Dam using the 1%3,1%9-70 and 1973 floods. Figure 2.4-21 compares observed and computed discharges at Watts Bar for the 1963 and 1969-70 floods. Figure 2.4-14 compares observed and computed discharges for the 1973 flood. A unique Melton Hill situation existed during the 1963 flood. The partially completed dam modified flows in a way not typical of present, completed dam conditns.

Consequently, observed flows at Melton Hill were used as inflow in the Watts Bar verif ication.

Figures 2.4-14 and 2.4-21 show that the Watts Bar model predicts flood peaks somewhat in excess of observed values in all three cases. It is considered to be conservative as an instrument to estimate larger floods. The negative computed flows in the 1969-70 flood occurred when actual project outflows were zero or minimal and resulted frm specifying observed headwater ievel as the boundary. It is virtually impossible to model a reservoir perfectly enough to verify these severe conditions. Moreover, modification toward doing so would have no influence on model performance during flood peaks.

Studies by others (Ref. 17,18,19) and unpublished work of TVA Indicate linearity of unit hydrographs (capability to predict floods of largely varying magnitude) If they are derived from large, out-of-bank floods produced by major, businwide storms. Total capability of the runoff model has been verified at critical locations (Norris, Melton Hill, Fort Loudoun, and Watts Bar Dams) against the largest available floods. Comparisons revealed that the O

2.4-17 Amend. 73 Nov. 1982

model predicts conservatively high at all four locations. Large volume in upstream reservoirs, particularly in nearby Norris Reservoir, has a strong stabilizing influence on the models. Unsteady flow +echniques, the most advanced state of the art, havo been used in the sognents of the model having principal of fect on flood elevations at the CRBRP site.

Results f rom unsteady flow techniques have been verified by steady flow methods to the extent possible. From these f actors it is concluded that the water shed modeling used in this analysis predicts probablo maximum flood elevations adoquately ar.d safely.

2.4.3.4 Probable Maximum Flood Flow The maximum PMF discharge at the plant site would be a sharp, narrow peak of l 258,000cf s resulting from the f a!Iure of Molton Hill. The discharge hydrograph is shown in Figure 2.4-23. It was computed with the unsteady flow modol.

The flood would result f rom the 7,980-square-mile storm with center at Bulls Gap, Tonnessee, shown in Figure 2.4-10, which also produces near PMP depths on the 17,310-square-mile water shed above Watts Bar Dam. The storm is more completely described in 2.4.3.1.

The flood would overtop and breach the earth embankments at Fort Loudoun and l Tellico Dams upstream and Watts Bar Dam downstream. The concrete nonoverflow section of Melton Hill Dam would also f all. These are the only dams that would f all. The Molton Hill failure and Fort Loudoun-Tellico breach increases while the Watts Bar breach reduces the level of this flood at the plant site.

The analysis of dam f ailures is described in Section 2.4.4.

The influence of the TVA reservoir system on the PMF was computed using operating procedures prescribed for floods.

In addition to spillway flow, these permit turbine and sluice discharge in tributary reservoirs in the antecedent r+crm. Turbine dischargos are not used in the mainstream reservoirs af ter large flows develop because head dif f erential s become too smal l. Normal operating procedures were used in the principal storm except that turbino discharge was not used in either the tributary or main river dams. The previous analysis did include turbine dischargo in tributary reservoirs. All gates were determined to be operable during the flood. Prescribed operating proceduros actually have little influence on maximum flood discharges, however, because spillway capacities and hence uncontrolled conditions are reached early in the main storm flood.

Historic, observed, mid-March reservoir levels were used at the start of the flood f rom the antecedent storm. As a result of the antecedent storm and flood, 51 percent of the reserved system flood detention capacity was occupied at the start of the flood f rom the mala storm.

Norris Dam was examinad f or potential failure during the PMF. The PMF would result f rom a 3-day, July PMP of 18.7 inches with Peof 16.5 inches, preceded by a 3-day dry period and antecedent storm equal to 30 percent of the PMP.

O 2.4-18 Amend. 73 Nov. 1982

The maximum headwater reached would be 1055.5, 5.5 feet below dam top.

Structural analysis confinned that the dam would not f all under this condition. Inflow, outflow, and headwater hydrographs for the Norris PMF are shown in figure 2.4-23a.

Flows and elevations for other candidate storms were not computed at the plantsite because it can be judged from the flood-producing components that l they would produce lower flood levels. Norris Dam outflow and backwater rise in the Watts Bar Reservoir system caused by high flow and Watts Bar headwater elevation are the dominant f r>fluences on mBRP site flood levels. Local inflow from precipitation on the area below Norris Dam peaks early with only the recession of the hydrograph adding to peak dam outflows and hence has a less dominant effect. Precipitation on this area was used to judge flow rates when they were not computed. Emory River inflows at mile 4.4 on the Clinch River contribute somewhat to backwater ef fect.

Backwater elevation at the mouth of the Clinch River is a major influence on l plantsite flood levels (sea Figure 2.4-18), reaching elevation 773.2 in the controlling flood. This flood produces the highest Watt. Bar Reservoir flow and headwater level and, except for the Norris PMF, essentially the highest Norris outflow. The Norris PMF peak outflow of 285,000 cfs would reduce to 230,000 cf s at the plantsite. A site elevation of not more than 772 was estimated for this flood based upon a conservative postulation of elevation

765 at the mouth of the Clinch River.

It is concluded that the March storm on 7,980 square miles produces the probable maximum flood. Any more detailed and definitive proof is not prudent because the con + rolling flood levei at the CRBRP site, resulting from combined O. rainstorm flood and seismic f ailure of Norris Dam as discussed in 2.4.4, overrides the PNF by some 20 feet.

l l

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2.4-19 Amend. 73 l

Nov. 1982 1

2.4.3.5 Water f.evel Determinations l The PNF would produce elevation 7~78.8 at Mile 18 and elevation 777.2 at Mlle

16. Elevation hydrographs for these locations are given in Figure 2.4-24.

Elevations were computed concurrently with the discharges for the site using the unsteady flow model. The influence of Melton Hill Dam f ailure is apparent. The influences of Fort Loudon-Tellico and Watts Bar Dam f ailures are not conspicuous but were an Integral part of the analysis.

2.4.3.6 Coincident Wind Wave Activity Winds are commonly associated with reinstorms. They usually subside, however, when rainf all ends. Flood crests, on the other hand, often occur some time after the end of rainfall. At the CRBRP site this lag f rm total basin runof f is at least 1 day. See Figures 2.4-23 and 2.4-24. Henco, winds integral with a given storm usually have ceased by the time of the flood crest. Yet meteorological conditions conducive to flood-producing storms can repeat themselves. Just such a repetition has been postulated in the PMF analysis which used an antecedent and a larger, main storm. To assign a wind and its wave runup ef fect during the flood crest of the second storm postulates yet a third, repetitive wind-producing event. The probability of such a sequence is extremely remote.

Nevertheless, a conservatively high wind velocity of 40 miles per hour over land f rm the most adverse direction has been applied at the time of the second, main storm flood peak to conform with Regulatory Guide 1.59. This is conservatively high compared to the requirements specified in Revision 2 of Regulatory Guide 1.59 (August 1977).

O 1

0 2.4-20 Amend. 73 Nov. 1982

2.4.3.5 Water Level Determinations, O The PMF would produce elevation 777.5 at liile 18 and elevation 776.0 at Mile 16. Elevation hydrographs for these locations are given in Figure 2.4-24. Elevations were computed concurrently with the discharges for the site using the unsteady flow model. The influence of Melton Hill Dam failure prior to the main storm peak is apparent. The influences of Fort Loudoun and Watts Bar Dam failures are not conspicuous but were an integral part of the analysis.

2.4.3.6 Coincident Wind Wave Activity Winds are comnonly associated with rainstorms. They usually subside, however, when rainfall ends. Flood crests, on the other hand, of ten occur some time after the end of rainfall . At the CRBRP site this lag from total basin ruloff is at least 1 day. See Figures 2.4-23 and 2.4-24. Hence, winds integral with a given storm usually have ceased by the time of the flood crest. Yet meteorological conditions conducive to flood-producing storms can repeat themselves. Just such a repetition has been. postulated in the PMF analysis which used an antecedent and a larger, main storm. To assign a wind and its wave runup effect during the flood crest of the second storm postulates yet a third, repetitive wind-producing event. The probabili ty of such a sequence is extrenely remote.

Nevertheless, a conservatively high wind velocity of 40

(~g Q miles per hour over land from the most adverse direction has been applied at the time of the second, main storm flood peak to conform with Regulatory Guide 1.59.

l l

Amend. 7 O 2.4-20a Nov. 1975

A 40-mile-per-hour wind of suf ficient duration to produce maximuin wave runup at the plant is, in itsel f, a rare event for March, the month of the PMF. As judged f rm 30 years of record at Chattanooga, Tennessee m ed winds on 930 March days show a probability in the order of 1 X 10'3 for a 40-mil e-per-hour wind on a specific March day. It is f urther postulated that, as a third, sequential meteorological event,yt is Independent of the PMF f rom two prior meteorological events. A 1 X 10 probability for a PMP storm and resulting PMF from a single meteorological event coincident with a high velocity wind f rm a second meteorological event has been receiving professional acceptance.

Ccrnbining a 40-mile-per-hour wind frm yet a thig independent event extends the probability of the combination to the 1 X 10 probabliIty range. Newton and Cripe (Ref. 11) Ingternative approaches applying to the Tennessee Valley estimate in the 1 X 10 range.

Wind waves were computed using procedures of the Corps of Engineers (Ref. 20).

Waves and runup are applicable to the plant only at Mile 16. At Mile 18 ground Ievels adequately shield the plant in the PMF wIth coincidont wind.

(See Figure 2.4-44 which applies specifically to higher flood levels from combined hydrologic and seismic causes.) At Mlle 16 the critical direction is f rm the southeast with an ef fective fetch of 0.5 mile. For a 40-mil e-per-hour overland wind, 99.6 percent of the waves would be less than 2.4 feet high f rm crest to trough, resulting in maximum water elevation 778.8. Runup above still reservoir levels would be 2.8 feet to elevation 780.0 on a smooth 3:1 slope and 3.8 feet to elevation 781.0 on a vertical wal1.

2.4.4 Potential Dam Failures (Seismically and Otherwise; induced)

The plant site and upstrean reservoirs are located in the Southern Appalachian Tectonic Province and, therefore, subject to moderate earthquake forces. All l upstrean dams, including those on the Tennessee River, whose f ailure in a seismic event has the potential to cause probims at the plant when combined with appropriate floods were Investigated as described in 2.4.4.2.1. Studies to determine the potential f ailure of upstream dams f rm hydrologic conditions are described in 2.4.4.2.2.

An operational basis earthquake (OBE), imposed concurrently with the one-hal f PMF resulting in postulated Norris f ailure, would be the controlling f ailure situation. Maximum water surf ace elevation would be 804.3 and 798.2 at Mlle 18.0 and Mile 16.0 respectively, excluding any wind wave of facts. As shown in Table 2.4-8 this condition produced the maximum plant flood level frm any of the PMF cr seismic conditions stipulated by Regulatory Guide 1.59.

O 2.4-21 knend. 73 Nov. 1982

This information is presented solely to confirm that the CRBRP can w!thstand 7- postulated floods caused f rom probable maximum rainf all, from seismic dam f ailures, and f ran combinations of rainf all and seismic dam f ailure. TV A is (3v) of the strong opinion that the iIkelIhood of events occurring concurrently which produce controlling flood levels for plant design is extremely remote.

By furnishing this Information TVA does not imply or concede that its dams are inadequate to withstand great floods and/or ecrthquakes that may be reasonably expected to occur in the TVA region under consideration. TVA has a program of inspection and maintenance carried out on a regular schedule to keep its dams safe. Instrumentation of the dams to help keep check on their behavior was Installed in many of the dams during original construction. Other instrumentation has been added since and is still being added as the need may appear or as new techniques become available. In short, TVA has confidence that its dams are safe against catastrophic destruction by any natural forces that could be expticted to occur.

2.4.4.1 Dam and Reservoir Descriotion Characteristics of TVA dams and reservoirs are contained in Table 2.4-13.

Their location with respect to the plant site is shown in Figure 2.4-25.

There are nine dams upstream in the Tennessee River System which influence flood levels at the plant site and Norris and Melton Hill Dams upstream on the Clinch River. Elevation-storage relationships and seasonally varying storage allocation in the major projects are shown on the nine sheets of Figure 2.4-26. No guide is provided for Melton Hill because the reservoir is held et f ull pool elevation 795 throughout the year with only minor fluctuations. An area-volume curve for Norris Reservoir is given in Figure 2.4-27.

There is essentially no likelihood that future dams and reservoirs could adversely affect flood levels at the CRBRP. Tha Clinch River already is f ully developed essentially to Norris Dam, and additions upstream of Norris Reservoir are not needed. There is smalI chanco of future dams on other Tennessee River Tributaries upstream from the mouth of the Clinch River and on the Emory River. Even if these forecasts are incorrect, any new dams would be designed and bulli to withstand floods and seismic forces that otherwise could O

2.4-22 knend. 73 Nov. 1982

endanger by flooding not only the CRBRP but erher nuclear power generating plants as wel l. Hence, the not influence of future dams, however unlikely, would be f avorable rather than adverse.

Most of the dams upstream frm the plant and Watts Bar Dam were designed before the hydrmeteorolcgical approach to spillway design gained its current level of acceptance, and spillway capacity is probably less than would be prov ided today. Arbitrary freeboard provided at these dams, however, permits many of the to meet today's criteria. Those dams whose falIure in the PMF have a potential to InfIuence plant fIood Ievels were examined, as discussed In 2.4.4.2.2.

2.4.4.2 Dam Fallure Permutations The discussion of dam f ailure permutations has been separated into twe sections--Seismic Failure Analysis (2.4.4.2.1) and Hydrologic Failure Analysis (2.4.4.2.2).

2.4.4.2.1 Solsmic Failure Analysis l There are 11 major dams that can infIuence plant site fIood Ievels--two on the Clinch River and nine on the Tennessee River System upstream of Watts Bar Dam.

All 11 were examined Individually and in groups to determine if postulated seismic f ailure combined with appropriate flood conditions would produce a controlling flood condition at the plant site. Two basic conditions were oeamined, a safe shutdown earthquake (SSE) during a 25-year flood with full reservoirs, and an operational basis earthquake (OBE) during the one-hal f PMF with full reservoirs. The latter combination produced controlling flood l evel s.

The FSAR for Sequoyah Nuclear Plant (Reference 20a) describes the investigation of potential single and multiple f ailures of Watts Bar and all 11 dams upstream during the two postulated seismic-flood combinations. All events ref erred to in that report were reexamined using flood conditions specifically applicable to the CRBRP. In the OBE the seismic dam f ailure combinations with a potential to create maximum plantsite flood levels are Norris Dam singly and Cherokee and Douglas Dams concurrently. In the SSE the candidate situations include failure of Norris Dam singly and concurrent f alI ure of Norris-Cherokee-DougI as and NorrIs-DougI as-Fort Loudon-Tel lIco. In the situations involving Norris Dam f ailure, Molton Hill Dam was postulated to f all when the flood wave reached headwater elevation 804. Watts Bar Dam would be overtopped and the mbankment would be breached f rm postulated Norris Dam fofIuro in the OBE. However, f alIure would occur af ter the Clinch River fIood peak had passed the plantsite and hence have no lowering ef fect. Because of this, plantsito fIcod ievels were computed as though Watts Bar mbankment did not f all.

Flows and elevations for the potentially critical situations are strnmarized in Table 2.4-12. The single, Norris Dam OBE f ailure combined with the one-hal f PMF was controliIng.

O 2.4-23 Anend. 73 Nov. 1982

Regulatory Grade 1.59 recommends use of Appendix A to 10 CFR 100 for estimates k of seismically Induced flood levels. As described above, both a saf e shutdown earthquake with a 25 year flood and a 1/2 safe shutdown earthquake together with 1/2 PMF are considered in accordance with 1his guide. For this site the 1/2 SSE corresponds to a horizontal acceleration of 0.125 g at rock foundation.

The basic procedures and the specific analysis to determine Norris Dam stabil ity are described below.

A standard method of computing stabil Ity is used. The maximum base compressive stress, average base shear stress, the f actor of safety against overturning, and the shee. strength required for a shear-friction f actor of saf ety of one are determined. To find the shear strength required to provide a safety factor of one, a coef ficient of friction of 0.65 is assigned at the elevation of the base under consideration.

The analyses of earthquakes are based on the static analysis method as given by Hinds (Ref. 21) with increased hydrodynamic pressures determined by the method developed by Bustamante and Flores (Ref. 23). These analyses include applying masonry inertia forces and increased water pressure to the structure resulting from the acceleration of the structure Horizontally in the upstream O

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Amend. 73 2.4-24 Nov. 1982

. . . - . 1

direction and simultaneously in a downward direction. The masonry inertia forces are determined by a dynamic analysis of the structure which takes into account ampi If Ication of the accelerations above the f oundation route.

No reduction of hydrostatic or hydrodynamic forces due to the decrease of the unit weight of water f ran the downward acceleration of the reservoir bottm is included in this analysis.

Waves created at the f ree surf ace of the reservoir by an earthquake are considered of no importance. Based upon studies by Chopra (Ref. 24) and Z ienkiew icz (Ref. 25), it is TVA's judgment that before waves of any significant height have time to develop, the earthquake will be over. The duration of earthquakes used in this analysis is in the range of 20 to 30 seconds.

Although accumulated slit on the reservoir bottom would dampen vertically travel ing waves, the ef fect of slit on structures is not considered. There is only a small amount of silt now present and the accumulation rate is slow, as measured by TVA for many years (Ref. 26).

Figure 2.4-28 Is a general plan of Norris Dam showing elevations and sections.

Results of Norris Dam stability analysis in the WE for a typical spillway block and a typical nonoverflow section of maximum height are shown in Figure 2.4-29. Because only a small percentage of the spillway beso is in compression, this structure is judged to f all. The high nonoverflow section wIth a renalI percentage of the base in compression and wIth high compressive and sheai Ing stresses is al so judged to f all. Based on stability analysis the iower nonoverfIow blocks remaining in piace are judged able to withstand the W E.

Blocks 34-33 (665 feet of length) are judged to f all by overturning at the base foundation because the resultant of all forces f alls very near the downstream toe which results in high compressive and shearing stresses.

Supporting this judgment is a statement by Hinds, Creager, and Justin (Ref. 21): "As the resultant approaches the f ace the compression stress increases rapidly, hence overturning would be preceded and accelerated by a compression f ail ure." Stabil Ity analyses indicate f ailure by overturning at a plano in the concrete above the foundation of the structure is less I !kely, principally because the height of the dem above such a plane is decreased and because a drainage system for upl If t rol lef is provided in the structure above the inspection and drainage gallery.

O 2.4-25 Amend. 73 Nov. 1982

l The dam is located on the dolomite series of rock which belongs to the lower O part of the Copper Ridge formation and is in turn the lowest part of the Knox group. The etructures are well entrenched into rock which dips siIghtly downstream. Tests made during the original design of Norris Dam Indicate ihe rock has high shear strength and, therefore, is judged able to resist s1Iding of the dam due to the additional earthquake forces.

Figure 2.4-30 shows the 665-foot-long part of the dam Judged to f all under OBE conditions and the judged location and height ' elevation 970).of the debris of the failed portion. The location of the debris is not based on any calculated procedure of f ailure because it is belleved that this is not possible. It is TVA's judgment, however, that the f ailure mode shown is one logical assemption and although there may be many other logical assumptions the amount of channel obstruction would probably be about the same.

Under SSE conditions, blocks 31 through 45 (833 feet of length) are judged to f all . The resulting debris downstream would occupy a greater span of the valley cross section than would the debris f rom OBE f ailure but with the same top level, elevation 970. Figure 2.4-31 shows the prt of the dam Judged to f all and the location and height of the resulting debris.

2.4.4.2.2 Hydroloalc Failure Analvsis All upstream and downstream dams which are close enough to have a sign 8' : ant influence on flood levels at the GBRP were examined for potential fa ire during the PMF. Concrete sect 7ons were examined for overturning and for Q

V horizontal shear f ailure with a resultant siIding of the structures.

and lock gates were examined for stability at potentially critical water Spillway levels, and against f ailure from being struck by waterborne objects. Concrete lock structures were examined for stability, and earth embankments were examined for erosion due to overtopping, it was concluded that the only potential failures during the PMF would be of l the earth embankments at Fort Loudon-TelIico and Watts Bar Dams due to erosion f rom overtopping and all the concrete nonoverflow portion of Melton Hill Dam to the lef t (looking downstream) of station 19+54 and above elevation 774.5.

Concrete Section Analysis For concrete dam sections, comparisons were made between the original design headwater and tallwater levels and those that would prevall in the PMF. If the overturning moments and horizontal forces were not increased by more than O

V Amend. 73 2.4-26 982

20 percent, the structures were considered saf e against f ail ure. All upstream dams passed this test except Mel ton Hil l, Dougl as, and Fort Loudon. Original design showed that the spillway sections of Fort Loudon and Douglas Dams to be most vul nerabl e. These were examined in f urther detail and Judged to be stabl e. The nonoverflow portion of Melton Hill Dam lef t of station 19+54 and above elevation 774.5 was judged to f all by overturning If headwaters reached elevation 804. Figure 2.4-32 is a general plan of Molton Hill Dam showing elevations and sections.

Solliway Gate Failures Consideration was given to the potential ef fect at the CRBRP of the f ail ure of spillway gates at Watts Bar and upstream dams in the PMF. The analysis f or the Sequoyah FSAR show that at Fort Loudon and Watts Bar Dams the gates would remain intact except possibly when struck by waterborne objects. These dams woul d be overtopped by the PMF. Gate f ailure would only make relative small changes in the timing of such failure. Because of this it was concl uded that gate f ailures are not important to this analysis and were dropped from further consideration. Gates were assumed operable and not to f all in all routings.

Lock Gates The lock gates at Fort Loudon, Watts Bar, and Molton Hill Dams were examined with the conclusion that no potential for f ailure exists because the gates are designed for a of f ferential hydrostatic head greater than that which would exist during the PMF.

Embankment Breachina Earth embankments at Fort Loudon-Tellico and Watts Bar Dams would be overtopped and subsequently brooched by the PMF. The Fort Loudon-TelIico breach would add to PMF elevations and the Watts Bar breach would reduce flood level s at the pl ants'te. These situations will be described in some detail.

The adopted relationship to compute the rate of erosion in an earth dam f ailure is that developed and used by the Bureau of Roclamation in connection w i th i t s saf ety of dams program ( Ref . 27 ) . The expression relates the volume of eroded fili material to the volume of wator fIowing through the breach.

The equation is:

l l

Amend. 73 O

Nov. 1982

O O soil 4 Owater

, g*

where:

Q,g; = Volume of soll eroded in each time period Q water = Y Iume of. water discharged each time period K = Constant of proportionalliy,1 for the soll and discharge relationships in this study e = Base of natural logarithm system X =

tan (d b = Base length of overflow channel at any given time H = Hydraul Ic heat at any given time d = Developed angle of friction soll material A conservative value of 13 degrees was adopted for materials in the dams investigated.

Solving the equation, which was computerized, involves a trial-and-error procedure over short depth and time increments. In the program, depth changes of 0.1 foot or less are used to keep time increments to less than one second during rapid f ailure and up to about 350 seconds prior to breaching.

The solution of an earth embankment breach begins by solving the erosion equation using a headwater elevation hydrograph assuming no f ailure. Erosion is postulated to occur across the entire earth section and to start at the downstream edge when headwater elevations reached a selected depth above the dam top el evation. Subsequently, when erosion reaches the upstream edge of the embankment, breach ing commences. Thereaf ter, computations incl ude headwater adjustments f or increased reservoir outflow resulting from the breach. Breaching proceeds relatively slowly for a short period; then, typically, breaching proceeds rapidly and the embankment is washed away in m inutes. For purposes of routing, complete f ailure was assumed to occur at the beginning of rapid f all ure.

O Amend. 73 Nov. 1982 j

2.4-28

During the hour of f ailure the peak discharge was determined based upon the headwater and tallwater depth at that time. Unsteady flow routing techniques were used to define the rest of the outflow hydrograph.

Scrne verification for the breaching computational procedures Illustrated above was obtained by comparison with actual failures reported in the literature and in Informal discussion with hydrologic engineers. These reports show that overtopped earth embankments do not necessarily f all . Earth embankments have sustained overtopping of several feet for several hours before f ailure  !

occurred. An extreme example is Oros earth dem in Braz il (Ref. 28) which was overtopped to a depth of approximately 2.6 feet along a 2,000-foot length for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before breaching began. Once an earth embankment is breached, f ail ure tends to progress rapidly, however. How rapidly depends upon the material and headwater depth during f ailure. Complete f ail ure computed in this and other studies has varied from about one-half to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> af ter initial breaching. This is consistent with actual f ail ures.

Fort Loudoun-Tellico Embankment Fallure Figure 2.4-33 is a general plan of Fort Loudoun Dam showing elevations and sections. Figure 2.4-34 is a general plan of Tellico Dam showing elevations and sections. Fail ure calcul ations were made f or the earth embankments at Tel i Ico and Fort Loudoun. TelIico would f all about 1-1/2 hours earller than Fort Loudoun but the rel ief af forded woul d not prevent .f ail ure of Fort Loudoun. To conservatively determine a maximum plant site flood level and to f acil Itate computations, compl ete, Instantaneous disappearance of the Fort Loudoun-Tellico complex was assumed at the earlier of the two calculated f ail ure times. Figure 2.4-35 shows the headwater and tallwater discharge rolationships for Fort Loudoun. Figure 2.4-35a shows the headwater and tallwater rating curves f or Tell ico. Figure 2.4-36 shows the computed outflow hydrograph for the CRBRP PMF immediately below the f ailed Fort Loudoun-Tell ico compl ex.

Watts Bar Embankment Failure Figure 2.4-37 is a general plan of Watts Bar Dam showing elevations and sections. Figure 2.4-38 is a general pl an and section of the west saddle dike. Fail ure calcul ations were made for the 750 feet of earth anbankment shown on Figure 2.4-37 which was assumed to erode down to average ground l elevation 700. The computed rate of f ail ure of the embankment saction is l shown on Figure 2.4-39.

( The west saddle dike was examined and al so f ound subject to f ail ure f rom l overtopping. This f ail ure would be a complete washout and would occur some 8-1/2 hours before that of the main embankment. The relief af forded would not prevent f ail ure of the main embankment and, theref ore, was ignored.

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O Amend. 73 Nov. 1982

, 2.4-29 l

1

A Figure 2.4-40 shows the headwater discharge relationships for Watts Bar Dam, one bef ore f ail ure and one af ter f ail ure of the 750-foot earth embankment sect i on. The tallwater rating curve is also shown for comparison. The tallwater curve dif fers f rom that originally provided and results from changes made in the Chickamauga Reservoir hydraulic model based upon March 1973 flood data. The headwater discharge relationships also dif fer as a result of the revised tallwater and improved definition of flow at high levels where the spil lway acts as a submerged orif ice. Figure 2.4-41 is the computed outflow hydrograph fran Watts Bar Dam for the CRBRP PMF. Corresponding headwater l levels are shown on Figure 2.4-39.

2.4.4.3 Unsteadv Flow Analysis of Potential Dam Failures An unsteady flow model of Norris Reservoir was developed in suf ficient detail to define the manner in which the reservoir would supply and sustain outflow at postul ated seismically f ail ured Norris Dam. The 61-mile reach of reservoir upstream to Clinch River Mile 141 was divided into twenty-eight 2.2 mile reaches. The model was verified by comparing its routed headwater levels in the one-hal f PMF wIth those using simpl ifled routing techniques. Headwater levels agreed within a foot, and the model was considered adequate for the purpose.

Discharge rating curves for Norris Dam for both the postul ated OBE and SSE f ail ure conditions are shown on Figure 2.4-42. These rating curves were developed from 1:150 scale hydraulic model studies at TVA's Engineering S Laboratory and verif led closely by hydraul Ic analysis. Outfl ow for f ail ed 1 conditions is controlled by the degree to which the vailey cross section downstream from the dam is obstructed by debris. This debris, not the dam breach, forms the discharge control section. Debris resulting from the SSE f ail ure is more extensive than from the CBE f ail ure, as shown by figures 2.4-30 and 2.4-31. Thus, discharge under WE conditions wIth the shorter f ailed section but less downstream debris is greater at a given headwater than for SSE conditions with wider dam breach but greater downstream debris, as shown by the rating curves, figure 2.4-42.

In addition to the postulated OBE f ailure condition for Norris Dam shown in Figure 2.4-30, four other f ail ure conditions were arbitrarily assumed. There is no engineering basis for these conditions which were assumed solely for sensitivity analysis. These are (1) overturning of blocks 33-44 (665-foot width) with 945 debris level, (2) overturning of seven blocks, 37-43 (370-foot width) with 925 debris level, (3) vanishment of the three tallest middle bl ocks, 38-40 (168-foot width) to ground level, and (4) instant vanishment of entire dam. Discharge rating curves f or the first two conditions were developed f rom 1:150 scale hydraul Ic model studies at TVA's Engineering Laboratory. The discharge rating for the assumed three-block f ail ure condition was developed analytically using hydraulic relationships for contracted openings. The outflow for the instant vanishment of the dam was defined by the unsteady flow model s of Norris Reservoir and Melton Hil l Reservoir, which were coupled together for this condition.

Amend. 73 Nov. 1982 2.4-30

Unsteady flow routing was used in Melton Hill and Watts Bar (including the Cl inch River embayment) Reservoirs to provide the accuracy needed to account for rapid flow and elevation changes at the plant site resulting from both upstream and downstream dam f ailures during the various postulated flooding conditions.

For Mel ton Hil I f all ure in the PMF, headwater and tallwater curves approprlate to the overturned nonoverflow section were used as boundaries for the models.

In the Norris Dam seismic f ailure flood wave, Melton Hill Dam was conservatively assumed to f all completely and Instantaneously with no debris interf erence at which time the unsteady flow models upstream and downstream were coupl ed together. This allowed computation of wave propogation both upstream and downstream in one continuous analysis.

Routings of seismic dam f ailure surges upstream of Watts Bar Reservoir were made using short interval storage routing procedures. These def ine Watts Bar lake inflows with suf ficient accuracy to demonstrate that Norris Dam WE f all ure Is the control lIng situation.

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1 Amend. 73 0

Nov. 1982 2.4-30s

2.4.4.4 water Level at Plant Site The unsteady flow analyses discussed in the previous section yield flow and elevation hydrographs in one operation. Resul ts f or PMF conditions are given in 2.4.3. These hydrographs f or floods f rom the controlling combined seismic dem f ailure and precipitation flood are shown on Figure 2.4-43.

Peak flow at the plant f or the controlling, OBE-one-hal f PMF Norris f ailure situation would be 921,000 cfs. Crest still reservoir levels would be elevation 804.3 at Mlle 18 and elevation 798.2 at Mile 16.

Plantsite flood elevations were also determired f or the arbitrarily assumed Norris f ailure conditions discussed in the previous section. These f ailure situations were combined with the one-hal f PMF and were determined only for comparative purposes. The tabulation below provides computed elevations f or these specified arbitrary conditions and f or the adopted level.

Still Location Reservoir Fallure Mode (Mile) Elevation Adooted Condiffon Blocks 33-44 overturned 18 804.3 (665-f oot w idth) 16 798.2 970 debris level O Arbitrarv Conditions Blocks 33-44 overturned 18 808.9 (665-foot width 16 80 9.6 945 debris level Blocks 37-43 overturned 18 811.9 (370-f oot w idth) 16 805.3 925 debris level Vanishment of blocks 38-40 18 808.4 (168-foot width) to 16 80 2.2 ground l evel Instant vanishment of 18 81 8.0 entire dam to ground 16 811.0 level O

V Amend. 73 2.4-31 Nov. 1982

The only f ail ure condition that would create flood levels above pl ant grade elevation 815 is the Instant vanishment of the entire dam, an unrealistic assumption. TVA concludes that f ailure of Norris Dam coincident with a large fIcod w IlI not endanger the pl ant.

A coincidental 40-mil e-per-hour overland wind was appl led f or the fetch radial a and directions shown on Figure 2.4-44. Critical direction is from the northeast for Mlle 18 and from the southwest for Mlle 16, both with an ef fective f etch l ength of 0.8 mil e. For these conditions 99.6 percent of the waves would be less than 3.0 feet from crest to trough. Runup would be 3.6 feet on a 3:1 smooth siope and 4.9 foot on a vertical wal1. Resulting elevations for the adopted condition are as follows:

Elevation Stili Maximum Runup Location Reservoir Water Surface Smooth 3 1 Sloce Vertical Wall Miie 18 804.3 806.3 807.9 809.2 Mlle 16 798.2 800.2 801.8 803.1 l Windwaves were not computed for the arbitrarily assumed f ailure conditions.

2.4.7 tce Floodino Because of the location in a temperate climate, significant amounts of Ice do not f orm on the lakes or rivers in the area and Ice jms seldom occur and are not a source of major flooding. There are no records of frazil or anchor Ice on the Cl Inch River in the vicinity of the plantsite.

The potential for ice formation at the site is less today than in the past because (1) daily water level fluctuations f rom operating Watts Bar (closed 1942) and Mel ton Hil I (closed 1963) Reservoirs would break up surf ace Ice before significant thickness can be formed, (2) increased water depths due to Watts Bar Reservoir result in a greater mass needing to be cooled by radiation compared to prereservoir conditions, (3) Cl inch River flows are warmed by release f rom near the bottom of Melton Hili Reservoir, and (4) Melton Hili Lake waters, in turn, are warmed by releases f rom near the bottom of Norris Reservoir (cl osed 1936).

Since Molton Hill was closed in May 1963, daily winter variation in tallwater l evel, mil e 23.1, has ranged f rom 0.2 foot to 7.4 feet and f rom 0.02 foot to 4.1 feet at the USGS strem gage near Oak Ridge, mile 14.4. FIuctuation at j the pl antsite woul d be somewhero in-between.

Minimum average water depths encompassing the 2-mile plantsite reach have been increased f ran 7 feet to 19 feet due to Watts Bar Reservoir. The lowest observed temperature in Melton Hill Lake was 40.4 in January 1964 at 1

O 1 Amend. 73 2.4-31a Nov. 1982

{

l

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2.4.9 Channel Diversions

! Channel diversion is not a potential problem for the

plant. There are now no channel diversions upstmam of the CRBR plant that would cause diverting or rerouting of the soun:e

of plant cooling water, and none are anticipated in the future.

The floodplain is such that large floods do not produce major channel meanders or cutoffs. Carbon 14 dating of material at the high terrace levels shows that the Clinch River has essentially i maintained its present alignment for over 2,000 years. The l topography is such that only an unimaginable catastrophic event could result in any flow diversion above the plant. l

2.4.10 Flooding Protection Requirements All Category I Structures, housing safety-related facilities, '

systems and components, and on-site power supply, will be designed and

! constructed for protection against all possible flooding con-di tions. These Category I Structures, capable of surviving the design flood l conditions, include the Reactor Containment Building, Reactor '

Service Building, Steam Generator Building. Intennediate Building, g Diesel Generator Building and the Control Building.

With the maximum flood level established at elevation j 809.2, structures which are either completely or partially j located at elevations below this level will be analyzed for the effects of the following forces

a. Hydrostatic pressures

b. Buoyancy

c. Wave action.

Hydrostatic pressures and dynamic wave effects (where applicable) will be conbined with other loads in the design of the Category I structure or components.

Stability against floatation will also be provided.

Protection against buoyant effects will be provided for Category I structures by resistance from dead loads or mechanical anchors to bedrock.

j All safety-related systems and equipment will be either located on floors above the maximum flood level, or will be I protected by the following neasures:

I 1

!O 2 4-32 2" die!!

Q D

dard conversion of units).

Releases from Norris Reservoir, located on the Clinch River 56.7 miles upstream of Melton Hill, flow into Melton Hill Reservoir and subsequently by the site. Norris Reservoir is a multiple-purpose reservoir providing power generation and flood control. The nomal minimum pool elevation is 960 (See Figure 2.4-58). Power generation at Norris can be main-tained to about elevation 900. Although not a primary purpose, stored water between elevations 960 and 900 is available for low flow augmentation in periods of drought. However, minimtsn levels will not be 9161ated without specific TVA Board of Directors' action. The total volume of storage in Norris Reservoir between elevations 960 and 900 is 260,650 sfd. This volume of water represents an average discharge of 714 cps for a perioo of one year. It is possible to lower Norris Reservoir to about elevation 860 by the use of slide gates. The total storage volume in Norris Reservoir between elevations 900 and 860 is 46,940 sfd (see Figure 2.4-59).

Releases from Fort Loudoun Reservoir, located on the Tennessee River 72.4 miles upstmam from Watts Bar, can be used to control the Watts Bar pool elevation. The normal minimum pool elevation for Fort Loudoun is 807 (See Figure 2.4-60

. The minimum pool elevation of record is 805.54, l 27 and Reference on January 18,19537 )4.

It is possible to lower Fort Loudoun n

Reservoir to about elevation 783 by the use of the spillway. The (di total volume of storage in the reservoir between elevations 807 and 783 is 97,500 sfd (See Figure 2.4-61).

Inflows into Watts Bar Reservoir from the Tennessee River are large, even during periods of low flow. Observed low flows at Loudoun (gaging station Nunber 3-5200, 10 3/4 miles do'.mstmam from Fort Loudoun Dam and 9 3/4 miles down-stream from the mouth of the Little Tennessee River) during the period from 1923 through 1954 are 1,820 cfs for one day and 2,790 cfs for 30 days. Observed low flows at this location since the filling of Fort Loudoun Reservoir are 1.820 cfs for one day and 9,020 cfs for 30 days. Thus, the 30-day low-flow volume for the period of record is equal to 83,700 sfd. The 30-day low-flow since the filling of Fort Loudoun Reservoir represents a volume of more than 270,000 sfd. An appraisal of the significance of these flows may be obtained by noting that the storage capacity of Watts Bar Reservoir at elevation 735 (minimum pool elevation) is about 15,000 sfd per foot (See Figure 2.4-62). Thus, the Tennessee River is more than capable of maintaining the minimum pool elevation of Watts Bar Reservoir even under extreme conditions.

Amend. 27 2.4-39 Oct. 1976

t 2.4.11.4 Future Control No plans for new structures on the Clinch River are known which might result in f uture low flows at the site significantly dif ferent f rom those observed in the past. The extended periods of no release f rcrn Molton Hill Reservoir in the past have been the result of special operations either upstream or downstream of Melton Hill Dan which are unrelated to either power generation or navigation. These extended periods of no release will be avoided in the f uture thru appropriate reservoir operations should plant requirements so dictate.

Flows at the site can be augmented f rom storage in Norris and Molton Hill Reservoirs. Inflow into Watts Bar Reservoir ca ' also be augmented f rom l storage in Fort Loudon and Tellico Reservoirs in the Tennessee River.

Characteristics of these reservoirs are described in Section 2.4.11.3.

2.4.11.5 Plant Reautrements 2.4.11.5.1 River Water Service System This system incorporates a non-Seismic Class l Intake structure designed to withstand a flood level of 750'0". The system supplies all plant make-up water from the Clinch River to the Emergency Cooling Tower Basin and the Main Cooling Tower Basin. This system also provides the Plant Water Treatment Facility with a source of water to meet alI demands for potable and process water.

The River Water Pump House is designed such that make-up water supply wilI not be Interrupted during periods when river level drops to minimum water elevation of 735 feet. Additional description of the river water system is prov ided i n Section 9.9.5.

2.4.11.5.2 Circulatina Water System The circulating water system is a closed cycle utilizing mechanical draf t i

cooling system. This system rolles upon the river only fcr make-up supply.

l The River Water Service System is designed to provide this water for river I stege levels down to minimum water level of 735 feet. River flow conditions I will not ef fect the perf ormance of the system as long as the river stage is at

, or above 735' . TVA operating procedures are such that Watts Bar Resenoir is I

maintained at or above the level at all times. The circulating water system is described in Section 10.4.5.

l l

l 2.4-40 Amend. 73 Nov. 1982

i TABLE 2.4-8 FLOOD ELEVATION SUP44ARY, CRBRP l

Clinch River Flood Elevations Wave Runuo Elevations Mlle Still Reservoir Wave Too 3:1 Slone Vertical Wall Norris Failure In nR7 With One-half PMF 16 798.2 800.2 801.8 803.1 18 804.3 806 .3 807.9 809.2 Probable Maximum Flood 16 777.2 778.8 780.0 7 81 .0 18 778.8 780.4 7 81 .6 7 82.6 O

1 l

l O

2.4-75 Amend. 73 Nov. 1982

1 O

TABLE 2.4-8a PROBABLE MAXIMUM PRECIPITATION (PMP) DISTRIBUTION Time Period Rainfall Rainfall (Hours) (Inches) Accumulation (Inches) 1 0.9 0.9 2 1.1 2.0 3 2.3 4.3 4 5.0 9.3 5 14.0 23.3 6 3.0 26.3 7 1.7 28.0 8 1.5 29.5 The above +1bulated time distribution of the PMP is depicted in 7 Figure 2.4-6.:.

Amend. 7 2.4-75a Nov. 1975

TABLE 2.4-9 m

PROBABLE MAXIMUM STORM RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Subwatershed Rain Pe,* Rain Pe,**

Hg. Location inches Inches Inches Inches 1 Norris 6.16 4.58 16.71 15.49 2 Coal Creek 6.16 4.25 16.00 14.59 3 Hinds Creek 6.16 4.25 17.70 16.29 4 Bullrun Creek 6.16 4.41 18.50 17.09 5 Beaver Creek 6.16 4.25 19.10 17.69 6 Clinch River local above M71.3 6.16 4.25 16.90 15.49 7 Clinch River local above M55.2 6.16 4.25 17.10 15.69 8 Clinch River local above M41.0 6.16 4.25 17.10 15.69 9 Clinch River local above M35.4 6.16 4.25 17.10 15.69 10 Clinch River local above M28.0 6.16 4.25 17.10 15.69 11 Clinch River local above M25.5 6.16 4.25 17.10 15.69 12 Clinch River local above M16 5.16 4.25 16.90 15.49 13 Poplar Creek 6.16 4.25 16.70 15.29 14 Emory River at mouth 6.16 4.25 14.60 13.19 15 Clinch River local at mouth 6.16 4.25 16.00 14.59 16 Watts Bar local below Clinch Rv. 6.16 4.25 13.30 11.89 17 Watts Bar local above Clinch Rv. 6.16 3.79 16.20 14.21 18 Little Tennessee River local, Fontana-Chi l howee 6.16 2.71 15.40 12.72 18a Little Tennessee River local, Os Chil howee-Tel l ico 6.16 3.79 16.10 14.11 19 Fontana local 6.16 2.71 14.40 11.72 20 Tuckasegee River at Bryson City 6.16 2.71 12.80 10.12 21 Nantahala 6.16 2.71 11.20 8.52 22 Little Tennessee River at Needmore 6.16 2.71 11.20 8.52 23 Fort Leudon local 7.48 4.99 19.90 17.91 24 Holston River local 7.48 5.52 21.90 20.30 7.48 / 23.3 0 25 French Broad River local 5.17 21.51 25a Little Pigeon River at Sev ierv il l e 7.48 4.99 20.00 18.01 26 Little River at mouth 7.48 4.99 19.80 17.81 27 Douglas local 7.48 5.88 27.00 25.78 28 Pigeon River at Newport 7.48 4.99 15.80 13.81 29 French Broad River, Newport to Asheville 7.48 4.99 16.60 14.61 30 French Broad River at Asheville 7.48 4.03 10.80 8.12 0

2.4-76 Amend. 73 Nov. 1982

TABLE 2.4-9 (Continued)

PROBABLE MAXIWM STORM RAINFALL AND PRECIPITATION EXCESS Antecedent Storm Main Storm Subwatershed Rain Pe,* Rain Pe,**

h Location inches Inches Inches _ Inches 31 NotIchucky local 7.48 4.99 21.50 19.51 32 NolIchucky River at Embreev il l e 7.48 4.99 16.30 14.31 33 Surgoinsville local 7.48 5.88 22.80 21.58 33A Cherokee local below Surgoinsvllle 7.48 5.88 24.00 22.78 34 North Fork Holston River near Gate City 7.48 5.88 '7.40 16.18 35 Fort Patrick Henry 7.48 5.88 23.80 22.58 36 Boone local 7.48 4.99 19.80 17.81 37 South Holston 7.48 5.52 17.00 15.40 38 Watauga 7.48 4.99 16.70 14.71 Average above Watts Bar Dan 6.9 4.6 17.2 15.4 O

1. Adopted API prior to antecedent storm,1.0 inch, based on median observed conditions.

l ** Computed API prior to main storm, 3.65 inches.

l l

l l

l 1

1 1

0 2.4-77 Amend. 73 Nov. 1982

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

e

/

I i

I i

\ w TABLE 2.4-10 INTENTIONALLY DELETED i l

t 4

[

,A 2.4-78

's Amend. 73 Nov. 1982

.- - , - - v_ _ -

o . -

TABLE 2.4-11 UNIT HYDR 0 GRAPH DATA Drainage Subwatershed Area, h Location So. Miles b b b b b b Duration 1 Norri s 2912 43,300 0.07 6 15 8 118 6 2 Coal Creek 36.6 2,150 0.64 8 9 5 40 2 3 Hinds Creek 66.4 3,620 0.68 9 7 5 54 2 4 BulIrun Creek 104 2,400 0.47 14 21 14 84 2 5 Beaver Creek 90.5 2,600 0.58 14 14 10 88 2 6-11 Clinch River local 22.25 1,350 0.10 2 8 5 34 2 12 Clinch River local above M16 37 4,490 0.95 6 4 3 46 2

' 13 Popier Creek 136 2,800 0.61 20 25 13 88 2 14 Emory River 8 mouth 865 34,000 0.37 9 13 8 8: 6 l

15 Clinch River Iocal at mouth 32 3,870 0.95 6 3 2 46 2 16 Watts Bar local below Clinch Rv. 427 16,300 0.36 9 9 7 84 6 17 Watts Bar local above Clinch Rv. 293 11,300 0.30 8 9 7 84 6 18 Little Tenn. River local,Fogtana- 84 6 Chi l howee 406 16,900 0.58 12 9 5 18a Little Tenn. River local, Chilhowoo-TelIIco 650 17,000 0.61 18 21 11 72 6 19 Fontana local 389 16,350 0.46 10 9 5 94 6 20 Tuckasegee River at Bryson City 655 26,000 0.43 10 12 7 58 6 21 Nantahala 91 3,770 0.45 10 12 7 70 6 22 Little Tennessee River at Neednore a 436 9,130 0.49 18 23 12 126 6 23 Fort Loudoun Iocal 323 20,000 0.29 6 10 6 36 6 24 Holston River local" 289 6,800 0.55 18 22 15 96 6 25 Frencg Broad River local 207 7,500 0.51 12 11 8 60 6 25a Little Pigeon River at Sevierville 353 15,600 0.62 12 10 6 102 6 l

l 1 O

. 2.4-79 Amend. 73 Nov. 1982

TABLE 2.4-11 (Continued)

UNIT HYDR 0 GRAPH DATA Drainage Subwatershed Area h Location Sa. Miles b b b b b b Duration 26 Little River at e 8 96 4 mouth 379 11,730 0.68 16 14 27 Douglas Local a 832 47,930 0.27 6 8 6 60 6 28 Pigeon River at Newport 666 26,600 0.56 12 11 6 78 6 29 French Broad River, Newport to AsheviiIe 913 35,000 0.53 12 12 7 108 6 30 French Broad River at Asheville 945 15,000 0.27 14 35 12 166 6 31 NolIchucky local 378 10,600 0.40 12 16 9 87 6 32 Nolichucky River at Embreev il le 805 27,300 0.58 14 14 9 82 6 b

33 Surgoinsville local 299 10,280 0.48 12 13 9 66 6 33a Cherokeelocalgelow Surgoinsville 554 18,750 0.48 12 14 7 66 6 34 North Fork Holston River near Gate City a 672 12,260 0.60 24 33 25 108 6 35 Fort PatrickaHenry 63 3,200 0.40 8 8 6 64 6 O 36 37 Boone Iocal SouthHglston 669 703 22,890 0.16 6 13 16,000 0.53 18 24 8

17 90 100 6

6 38 Watauga 46 8 17,700 0.53 12 13 7 84 6

a. Rev ised

b. New Definition of Svmbols Q = Peak discharge in cfs p

Cp = Snyder coeffIciont T = Time in hours from beginning of precipitation excess to peak of unit E

hydrograph W50 = Width in hours at 50 percent of peak discharge W75 = Width in hours at 75 percent of peak discharge TB = Base length in hours of unit hydrograph Dur = Duration in hours of unit hydrograph 2.4-80 Amend. 73 Nov. 1982

TABLE 2.4-12 FLOODS FROM POSTULATED SEISMIC FAILURE OF UPSTREAM DAMS CRBRP OBE Failures Headwater Peak Flow, Elevation With One-half PMF Watts Bara Norris CFS Mile 16 Mile 18 Norris 765.8 1035 921,000 798.2 804.3 35,000 c c Cherokee-Douglas 765.0 -

765.4 765.6 7

SEE Failures With 25-Year Flood Norris b 754.5 1027 d

744,000 790.5 796.3 Norris-Cherokee-Douglas 754.5 1024.3 770,000 d 791.6 797.7 Norris-Douglas-Fort Loudoun- d d Tellicob 764.5 1024.3 754's000 791.6 797.7 7

a. Lake level at mouth of Clinch River concurrent with peak CRBRP stage.

b. Taken from recent analysss for Sequoyah Nuclear Plant.

c. Estimated by steady flow backwater with starting elevation 765 at mouth of Clinch River from unsteady flow analysis,

d. Difference in Norris headwater elevations and peak site flows from that for Norris single failure results from use of the Sequoyah watershed 25-year flood. 7 l

l Amend. 7 2.4-81 Nov. 1975

I FACTS ABOUT MAJOR T i

1 M4 N H.. e - ,4 ' e cou nt y N e a rb, T. pe can, rete r.rth M.s. i ench D,..nare isnath Me.. A rea HI% D H C,, , of sa d.m. .nd /ar +)eets ..e. aheve of lake W id t h of I. eke B snot a T.s p.m H ,k n e.6 h,t d.m e miles s of I.she et Feu (gp so,h pe rh , A.e ten . stoee,t tog. mi.) (moles) roei f ra. , de.) ru. ,ds. ( .,re s ,

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W ,i...., o , T.. s i. t.....

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Ilfilti T 4 H Y l'Ht d l L15

1. . . ..,4 ,a T . , n. r .nu.n w. . e.te, ex. . . n. 3.n ..... n. i.n. n, n -

i.....

A . . i. . . . n . . .. .e e s. c. ne eiee ,c, r.. e, cc n,.... . n. i .u. . .. n ... .3 i.i..

n,. .~e u . a .. .e , s . c. ne,esee u.rpi , - ....u. . nr i.u. ,o n ... .....

c . ..e n.. .me s. c. ci., n.,eee.ne E is.n. 2.3. . .. . . i.. . .s, . i., i3 ... ....

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i....

o~e s . . in o,..ee Tenn. rem ne nt.n arr . . 3. .i. ii. - - -

/

( o~e s. o,eee Ten.. rem o. u... cc n.n. n.... n. .n o. , .3 .n

.u.e n.d.e n, Te~e. c r. .n oi.e d.e e _ i........ i., ..... .3 i. - 3.2, .

senei, s e,,,i, c .. c, .en n i. i,.. .ne E&= i i .2 3. i .i s,.3.. iu r. .. ii. i. i.i ..i..

wei,e., n,n ci. . Tenn. ";d; ,;" n. . . .d .e cc n . .. .. . in i.n. 3.3 n u ... ..

s e. ..e nn T en n. ^y;;;;el,u u n.....ie ccE i..n.n. ni.n. ns ..... i., a n i ., u u.

T e ll... Lettle T. Tenn. i uden Lenee, cit, ccE  ?. 0 0 2.%).3 0 129 3.23. 2. 27 33.2 .3 is.aao r .. . O',','t s. c. c'ef. n'a n . .. .ne cc 2.. n.3.. n.m. c. n.3 n i.in n ... i ..u .

n . . i.. 5,',';ll Tenn. se,ie, se. .e. , ,ne ccE s s..n. nr.... in i.ns ..in u.i i .s n.o.

n e,.6 ee n.i.t.n T, n. 30;;;u Je e,,.en c. , - n.a. 33..n. ns . .n . 3.n . i, os um.

s r...

cc

r. t r.t ,,,, n e n,, noi.ien Tenn. son...n x.n, -n nu. . .3 ni i. n i. 3 ..n .n S Suil.een a b, no.ne H a,Fe.

eten k Tenn. w ..h nsten Jahnson hing.p.c,i,., ccE i , 9. . .. f i . .... i.9 i.332 ..... it.3 .5 .....

M. B ,i..... V a..

Sog.h neletan y, F,,,,, k, Tenn. Sailiv e n y,... E&R ,7.5.. 5. .,7,. . . 295 ..... 73 2. 3 .3 7.5..

W at.u s e Wat.es. Tenn. certer Fiis sheihten a ...... 3. . ,7. n .. Si, ... ... i ..? ... . . 3.

l c,eet F.H.

4 en cumbe...d) nd C **7 Tenn. Wenen (H Rock ietsad cc - ~ ,8 3.. .. 75 22 -

2.1..

l gsn,y) F e.h w hite Tet et. .34.353 i'i M P) D SToM AG.: Tenn. Tenn. M a rian ( heit.ne s e EAR i t;.... .2...... 23. ...n. - - - 52.

He coon MountainO)

.. Foundshen to ope, sting deck.

b. River i. event, line.

l i c. re.erhouse i. In relk count,. Tenn

d. Orir.nsi construction er a,guisitie e.bsequent addeteens. retirements.

et p,ejecer under constreetsen.

e. constructeen das,s.ntinued e.ri, in i1 I- 3," d',*I TENNESSEE V ALLEY AUTuGR TY. KNOXVILLE. TENN. Sn.2 , , ,, ,",,','"ll,**"j'*,'j,,Fe,,b,r,e.

g. Abbr cc.cenerete g r.* i

( Revised September i.f. e m b.eei.

nk me ntion t s. s : E.Larth an. EAR I Filli.R b. Unit 2 6. a rever.ibie pomp-terb6saw

\

l m

6 VA DAMS AND RESERVOIRS i.ahe Fle atee.n i'*e f el Can- ( lesete Cn ee r. Cost of Ultimate I ek imek MAIN Sher.s lo w fleet shove see Betell

~i she Veleme t erre. feet i Centrailed strettian Com. Planten Gene r-J ' s g Nee Man. RIVE R se rsti

- - ~ ~ - ' ~

Ordinner Tap et Store Started pleied Se Capacity t Feen 84f t PROJECTS Peel Ordenary f ull M enemum f.e t e s ( Ac . set.) .T ice 4d: KW and No.

(Int enet 4 3.rv. (reen c a.eles) Minsenum T.ap

.ateset reel (m) Llev aten I' lee steen en hae) (mathens) of Unite ( )

2.31. n. 373  %, u t i e.. 4.12,.e.. ......... 733. 4 3... . ..i.... S i i ... 17..... i s t it. 6.. 11 Kenterby 4,4 ... 41. .10 4 4,*. . .. i . I . 3.e. . 417,... 3.a.35 p.3. 42M6 43.1 216.... ( 6 ) ii.s4.. 43 Piche ech Land.ns ll.s6.. I Wileen til 154 54.5 3 7.. . 37.5 362.... $ s t.o.. 5,.... ..l . la e.] . 2. ..it 25 i .M 42,....n t )

43, 6

s2 3 lH 6 12

i. 63 55. 556.3 3 *.6 72..... t ..; 1.4a. 3 31.a.. i t.21.n u.3-3 6 i t .. .a *?.3 %4.4..n t ) ii 4s6 .. 32 Whnin 4 36

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o u. un u,. i o.. n o.. no.. n. .n m u.on 7. .n u n.mm eine R d.e m i.6 u, . un 1.n, u.a. n o.. n u.. 7.n.n ne.n n..u e. . n...m wei,ei, m m m m ,u.. n o.. 3 u.. ..u. n.n un u. n...m u.... u Meit.n nin t

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m m .n .n m.. ., n o.. nuo 3. i s.67 112. ,, <o In , .o to Teiiin m un ui. u.. m.... u . u.. i.n u a un u.7.u n..n u.. nuam r.niana m m un u.. . u .. u u.. u.u.. n.n n .n ni.o n, nuam o.e. iaa

... m on un . u .. u. u.. u n.o. ..u. n.5.n .. int u. 6 no m Ci,e e ee n un un un tu.. nua o.. n ..n i..n.n n.s.n in u.... m r., r.irea nen,,

n. un un un . u.. n o .. n o . . .. noe inen no3 n. . n.mm a.e-in un un un nu.. nuo n u.. o.7m u.u-u nui 3u n.... m se. .i n.isten in un un un S u .. n u.. n u.. 7.n.ui f t in.o -a o n5 u. .m W.t....

Crest ralla (Il it. 78. k. 5. 3. . 5 3. ..6.. S i .6.. 3 7.... 35 it.. 16 .it s.2 3i. 4.(1) ten Cumberland V alley) i t ..'i . n.42 8.. . . n.73 2.35. I S.11..p e. 3.iss. .t l.5) Totals

- i.13. _ l.673 3.... 3 7.e .. 35.n.. 7..- 7 .--7. l- .74 ISS.. i .3 3..... ( n Pt %tPED MTOR ACE Herreen MeenemenO)

.. Tellice presert has ne f.ek er p .eehem e. Sereamae. threesh navigable thennes se f ort 14edoe n Hese rs m all increase average annual energy threesh vert taudoun pomerhouse by 3.o.irsnilleen bok.

ease. 3 i nder constreeten; cost and eventity data estesmated.

i esse, inel* ding s=6tchterd, as adjested br k. .%aajack Dam replaced the old neles Bar Dese 6 m6ies apetreaan, and esthesitesteens. includes eetsmated emets i. A egeered: W.I.en by tren fer from i'. S. Cerpe of Ensineere in i,33: oceee S i. seemee Se. 2. Rime Redse. and Great rolle by perchase fresa TEP Co. i t4; per.sned in March i.3.. i ,e.,

3 . seb equent to sco.... tion. TVA heightened and enstalled addit 6enai ana.n e at W61aea.

7 it. i. 23 tasuporarHF daseentaneed te um. g , g 6tled. % here eterage space se s*e ntable above thee levei, addellenal 41Hng may p dams. CGE Concrete steesty with earth be snade as needed for Seed sentrol.

nrth and tsch $11. RFT.Rech.eued tasaber, m Cn.tr een e t Neaejuk main Iwk hmited to andnestu pwuens for esse-pletese later.

TABLE 2.4-13 2.4-82

O O O

g......

\ 3  %

s ____,__

s

'>- /~'

s& K ,-

\\

!~~ ', / /~ ,/ ~

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50 40 -

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O I 2 3 4 5 6 TIME - DAYS LEGEND:

AREA 1, NORRIS DAM , 2912 SQ. MI.

Note: Two peaks result from different contributing times for two principal tributaries making up this long, narrow watershed.

Figure 2.4-12a. 6-ilour Unit flydrograph 6650-6 2.4-118

O 40 30 E

o 8

9 1 20

<r Oi O 10 3 0

0 1 2 3 4 5 6 TIME - DAYS LEGEND:

t REA 14, EMORY RIVER AT MOUTH,865 SQ. MI.

FIGURE 2.4-12d. 6-Hour Unit Hydrograph Amend. 73 2.4-121 Nov. 1982

1 l

l 25 -

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---

ARE A 16, WATTS BAR LOCAL BELOW CLINCH R., 427 SQ. MI.

---

AREA 17, WATTS B AR LOCAL ABOVE CLINCH R., 293 SQ.MI.

AREA 18, LITTLE TENNESSEE RIVER LOCAL, F ON T ANA - CHILHOWEE , 4 0 6 SQ . Ml.

ARCA 180, LITTLE TENNESS EE RIVER LOC A L, CHILHOWEE -TELLICO D AM,650 SQ.MI.

FIGURE 2.4 -12e, 6-Hour Unit Hydrographs Amend. 73 2 4-122 Nov. 1982

O 3 O r-

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ARE A 19, FONTANA LOC AL,389 SQ. MI.

ARE A 20, T!JCKASEGEE R. AT BRYSON CITY, 655 SQ. MI.

ARE A 21. N ANTAHALA,91 SQ. MI.

- AREA 22, LITTLE TENNESSEE R. AT NEEDMORE,436 SQ.MI.

FIGURE 2.4-12f. 6-Hour Unit Hydrographs 2.4-123 Amend. 73 Nov. 1982

O 20 0

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AREA 23, FT. LOUDOUN LOCAL , 323 SQ. Mt.

l

- - - - - - AREA 24, HOLSTON RIVER LOC AL, 28 9 SQ. MI.

AREA 25, FRENCH BROAD R. LOCAL,207 SQ.Mt.

-- AREA 26, LITTLE RIVER AT MOUTH, 379 SQ.MI.

--- AREA 25a, LITTLE PIGEON R. AT SEVIERVILLE,353 SQ. MI.

l

[

FIGURE 2.4-12g. 6-Hour Unit Hydrographs Amend. 73 2.4-124 Nov. 1982

O 50

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i ARE A 27, DOUGL AS LOC AL ,832 SQ.MI.

I FIGURE 2.4 -126 '-Hour Unit Hydrograph O 2.4-125 Amend. 73 Nov. 1982 1

35 g, II II II II I

so tIg I I l I ip i li 25 -3 8-$

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l -

AREA 30, FRENCH BROAD R. AT ASHEVILLE,94S SQ. MI.

AREA 31, NOLICHUCKY LOCAL,378 SQ MI.

l ---- ARE A 3 2, NOLIC HUCKY R. AT EMBREVILLE ,805 SQ. MI.

FIGURE 2.4-12ha. 6-Hour Unit Hydrographs Amend. 73 2.4-125a Nov. 1982

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AREA 3 3, SURGOINSVLLE LOCAL,299 SO.MI.

AREA 34, N. FORK HOLSTON R. N R. CATE CITY, 672 SQ. Mt.

---

AREA 3 5, FT. PATRICK H ENRY, 63 SQ. M l.

-- A R E A 33 o, CHEROKEE LOCAL BELOW SUR40lMSVILLE, 5 54 SQ. MI.

FIGURE 2.4-12i. 6-Hour Unit Hydrographs 2.4-126 Amend. 73 Nov. 1982

25

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ARE A 3 6,000NE LOC AL , 669 S C. MI.

ARE A 37, SOUTH HOLSTON ,703 SQ. M I.

A RE A 3 8, WATAUG A , 44s SQ. NI.

Note: Two peaks for area 37 result from different contri-buting times for the two large streams making up this long watershed.

FIGURE 2.4-12j. 6-Hour Unit Hydrographs 2.4-127 Nov. 1982

O FIGURE 2.4-13 HYDROLOGIC MODEL VERIFICATION - 1973 FLOOD 60

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Figure 2.4-22 Hydrologic Model Verification - 1948 & 1969-70 Floods Faory River at Oakdale 6650-25 2.4-134 l

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

FIGURE 2.4-23 CRBRR'S PROBABLE MAXIMUM FLOOD DISCHARGE DISCHARGE- 1000 CFS o E b b o 5i J  :

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, Amend. 73 l 2.4-159 Nov. 1982 1

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47 3.7.2.1.2 Seismic Category I Systems and Components The analysis of Seismic Category I systems and components is detemined by a detailed dynamic analysis using either the response spectrum method or the time history method. The analysis is performed on a multi-mass O 3.7-7 Amend. 47 Nov. 1978

mathematical representation of the system or components. A sufficient number of masses with their appropriate degrees of freedom are used in trie model to adequately describe the behavior of the structural system, and to insure an accurate determination of the dynamic response. Significant non-linearities, such as gaps or elearances between PCRS components, are inct uded in the mathematical model. In this case, a nonlinear time history analysis is perf ormed, which considers the impact f orces generated at the gap locations.

Non-symmetr ical features of gemetry, mass, and stif fness, are modeled to include their torsional ef fects in the analysis. Hydrodynamic ef fects of partially filled tanks will be evaluated wherever they are significant in magnitude. Descriptior.s of a prel iminary reactor system I inear model and a prol iminary PCRS non-1 Ir.9ar model are given in Section 3.7.3.15.

The methods of response spectra analysis and time history analysis are described in a number of publications. A description of these analyses techniques is provided in Appendix 3.7-A.

The system or component is analyzed with the seismic input (floor response spectra or time histories) derived at the particular points of support on the str ucture. AlI signifIcant modes of the mathematical model are included in the analysis. The signifIcant, dynamic response modes are those predominant modes which contribute to the total, combines modal response of the system.

Other modes, whose inclusion in the square root of the sum of the squares modal summation have anegl Igible ef fect on the total response would not necessarily be used. With this procedure the number of modes included will be such that inclusion of additional modes will not result in more than a 10%

increase in responses. Where the response spectrum method is used, the individual modal responses are combined by the squars root of the sum of the squares, except for closely spaced modes (frequencies less than about 10%

apart) where the modal responses are combined by the absolute sum. The analysis is perf ormed independently in each of the two horizontal directions, and the vertical direction. Simil ar of fects obtained for each of the three directions are combined by the square root of the sum of the squares. This is consistent w Ith Regul atory Guide 1.92.

A simplified analysis based on a single mass model or an equivalent static load method may be used when it can be demonstrated that the simplified analysis provides adequate conservatism. For the simpl ified analysis, the equival ent static f orce, F , is distributed proportional to the mass of the component, and is calculat3d by the following equation:

Fs= 1.5 W A s where W is the total weight of the component, and A is the maximum peak acceleration of the response spectra, which apply af the points of support of the component. Ccmponents whose f undamental frequencies are greater than 33 Hz in any direction, are assumed to be rigid in that direction and may be designed for at least the maximum acceleration at their supports.

O 3.7 -8 Amend. 73 Nov. 1982

3.7.3.13 Interaction of other Pfofna with Seismic Cateaorv i Ploina O' For Category 1 oiping have non-Category 1 piping systems connected, the analysis of the Category 1 piping will include, as a minimum, the section of the piping system to the first anchor point beyond the cl assification boundary or suf ficient non-Category 1 piping and seismic restraints to assure decoupl ing between the Category 1 piping and the remaining non-Category 1 piping. This will assure that the dynamic coupi Ing ef fects at the interf ace between piping systems has been considered.

in any given fluid system, a valve will serve as the seismic Category I and i non-Category I boundary. The valve capabil Ity to maintain a pressure boundary in the event of a seismic event is to be assured by designing piping on the non-Category I side through the first anchor beyond the valve for that same seismic event or through suf ficient seismic restraints to capture the dynamic ef fects of the dif ferent seismic category piping systems at the interf ace.

For the seismic restraints, the piping system analysis includes the structure or building Interaction by considering the appropriate stif fness values in the analytleal model s. The structure /bullding mass is usually not considered since its dynamic response is negligible. For the anchors, the piping system is modeled to the anchor with the appropriate stif fnass values considered.

The resultant anchor loads are summed to form the design loads for the anchor.

3.7.3.14 Field Location of Suncorts and Restraints For the analysis of multiple supported subsystems, the ef fects of relative p displ acements between piping and support points at dif ferent elevations on the Q supporting system are considered as discussed in Section 3.7.2.7. The response spectra for the dif ferent elevations were superimposed to yield an envelope response spectrum to be used in the response spectrum analysis of mul tipl e supported subsystems.

3.7.3.15 Seismic Analyses for Fuel Elements. Control Rod Ass-blies and Control Rod Drives The seismic analyses that wilI be used to establish the seismic design adequacy of the reactor Internal s, assembl les, control rod drives, etc., is discussed in Section 3.7.2.1.2. For components such as the assembl les and control rod drives where clearances exist between adjacent members, a non-1Inear time history analysis has been performed, see Section 4.2.3.3.1.4. The j mathematical model consists of the whole reactor system. Preliminary models i

I for Iinear analysis are discussed below.

l l

l l

l b

V 3.7-14 Amend. 73 Nm. 1982  ;

1 1

3.7.3.15.1 Reactor System Structural Arrangement Simplified sketches of the reactor configuration as modeled for seismic analysis are shown in Figures 3.7-17 A, B and C. Figure 3.7-17A shows the reactor and reactor enclosure system as idealized for normal operation. Figures 3.7-178 and 3.7-17C show the additional head mounted equipment during refueling and preparation for refueling respectively.

In Figure 3.7-17A the reactor vessel flange is attached to the support ledge in the reactor cavity through a bolt and support pad system. The outer plug riser is bolted directly to the vessel flange.

Therefore, both the vessel and riser are assumed cantilevered from the flange which is attached (with an appropriate stiffness) to the support ledge.

The head is comprised of three separate plugs (large, inter-madiate and small). The rim of each plug is suspended within the penetration of the mating plug (or flange in the case of the large plug) by bearings mounted on concentric cylindrical risers. Both primary and secondary CRDM nozzles as well as the surrounding shield and seismic support structure are cantilevered from the iatermediate plug. The upper internals columns are attached to the same plug through the jacking mechanisms. The upper internals structure is assumed laterally restrained by the core barrel in the operating and preparation for refueling cases. The core barrel is rigidly attached to the core support plate which is, in turn, attached to the vessel through the support cone. The lower end of the thermal liner is also directly attached to the vessel wall.

~

The fuel blanket, control and radial shield assemblies are all piloted into the inlet modules at their lower ends and laterally sup-ported through adjacent assemblies to the core former rings attachad to the core barrel at two elevations. Tolerances, twist, and bow of the assemblies as well as the sodium between assemblies tend to prevent relative lateral motion of the assemblies. Therefore, inter-assembly gaps and clearances within the core barrel are of relatively minor importance to the overall system. The assemblies and core barrel are assumed to be effectively coupled together in the lateral direction at the load pad-former ring elevations.

The primary and secondary control absorbers and drivelines are each effectively connected vertically to the CRDM on the head and laterally to the CRDM and core at several CRDM bushing and absorber wear pad elevations. The drivelines are free at other elevations where the ciearances are larger. Section 3.7.3.15.3 gives a discussion of the nonlinear control rod and driveline model used to determine the scram retarding impact forces during a seismic event.

The reactor vessel is partially filled with sod'ium. The normal level of the sodium is about 36 inches above the suppressor plate, or 46 about 12 inches below the bottom reflector plate of the closure head. The 3.7-14a Amend. 46 August 1978

1. SCOPE This appendix establishes the baseline requirements of the design and analysis of the steel catch pans and fire suppresion decks for the Clinch River Breeder Reactor Piant.

2. APPLICABLE DOCUMENTS The edition and addenda of the following publications are part of this document and are appl icable to the extent specif led herein.

2.1 American Society of Mechanical Encineers (ASME) 2.1.1 Boller and Pressure Vessel Code,1977 Edition incl uding Addenda through the summer 1977.

(a)Section II, Material Specifications (b) Section iII, Division 1, Nuciear Power Piant Components (c)Section V, Nondestructive Examination (d)Section IX, Welding and Braz ing Qual if Ications 2.1.2 Boller and Pressure Vessel Code, Section lil, Division 2, Code for Concrete Reactor Vessels and Containments,1977 Edition I.ncluding Addenda through Summer 1977.

s 2.1. 3 Boller and Pressure Vessel Code, Section Vill Division 1,1977 Edition I inciuding Addenda through Summer 1977.

2.2 American Institute of Steel Construction (AISC)

Specifications for the Design, Fabrication and Erection of Structural Steel for Buil dings. (1969 incl uding Supplements 1 (11/70), 2 (12/71), and 3, (10/75).)

2.3 Westinahouse Electric Corocration. Advanced Reactor Divison (WARD)

WARD Document No. WARD-D-0037, Seismic Design Criteria for Cl inch River Breeder Reactor Pi ant (Rev 1,1977), (PSAR Appendix 3.7-A).

3.0 TECHNICAL REQUIREMENTS 3.1 Design Reautrements Catch pans and fire suppression decks are located in non-radioactive Na and NaK cells in order to prevent a chemical reaction between Na or NaK and concrete following a accidental spill and to protect the structural integrity of cell structures f or the preservation of the capital investment.

Catch pans, fire suppression deck and supports shall be designed as Seismic Category I components.

O Amend. 73 3 3.8-C.1 Nov. 1982

l The design requirements and the associated criteria used to satisfy each of the requirements for catch pans, fire suppression decks, penetration assembl les, brackets and attachments, and seismic equipment and othe.-

structural supports, are described as f ollows:

O O

Amend. 73 3.8-C.1a Nov. 1982

3.1.1 Catch Pan Reculrements

1. The catch pan plate shall be designed to contain a large sodium /NaK spill (faulted condition) with temperatures as per Attachment A

" Design Parameters".

Criterion There wIll be no catch pan faliure under a Na/NaK spiil such that Na/NaK penetrates the catch pan plate and Interacts with the structural concrete. This is ensured by the strain Iimits under load combination C per Table 3.8-C-1 not being exceeded.

2. The catch pan plate shall bg designed for maximum long term operating conditions of 120 F.

Criterion Strain Iimits under Load Combinations A and B per Table 3.8-C-1 shalI not be exceeded.

3. The equipment supports in the catch pan area shalI be designed l Independently of catch pan plate. i Criterion The equipment will not be supported on the plate but on local structural supports Independent of catch pan plate. During h maintenance, timber dunnage will be placed on the catch pan plate to facilitate equipment handilrj. Stresses under this condition shal I not exceed those specl* ;ed in Table 3.8-C-1.

4. The catch pan plate shall be designed to insure an essentially elastic response under normal operating conditions.

Criterion Strain limits under Load Combination A shall not exceed 0.002 In/

In. strain.

5. Catch pan plate surface shalI be protected to fact lItate

! decontamination after a sodium spill.

Criterion The hot-rolled natural finish surf ace condition is considered adequate. A protective coating will be applied during construction I to prevent corrosion.

l l 6. The catch pan plate shall be designed for corrosion allowances l commensurate with environmental conditions for a 30 year plant design life.

I O Anend. 64 3.8-C.2

l Cell Liners and Liner Support System Carbon Steel l Piping Carbon Steel & Stainless Steel Pipe Insulation and Canning Material Note 1 Pipe Supports and Auxillary Steel Carbon Steel Conduit Carbon Steel Embedmonts Carbon Steel Penetration Seal s Piping Wel ded Hatches and Doors Silastic Rubber Compression Gav.iets*

Electrical TBD

• Some hatches, such as the piping cell hatches (Cells 1010, D, and E) may be seal-wel ded.

Note 1: Material requirements for piping insulation and piping are appiIcable to the components and piping in the inner cells. These are discussed in Chapter 9 for individual systems.

1 O O .

3 A.1 -Se Amer:*. 73 Nov. 1982

3A.8.4 TESTING AND INSPECTION 3A.8.4.1 Development Testing Programs A series of development testing programs have been developed to support the cell liner design. These programs provide materials data to support the objective of designing the cell liners to accomodate large sodium spills without failure, demonstrate through qualification testing that integrity of the liner is maintained under sodium spill conditions, and provide test materials data on sodium-concrete reactions to assess the consequences of cell liner failure.

Five individual testing programs have been completed or are ongoing in support of the cell liner design. These development programs are:

(a) Comprehensive Testing Program for Concrete at Elevated Temperatures (b) Sodium-Concrete Reaction Tests (c) Sodium Spill Design Qualification Tests (d) Cell Penetration Sealant Tests (e) Base Material Tests for Liner Steels The tests included in the development programs listed above are modeled to minimize the difference between small scale tests results and the actual mass concrete response at elevated temperatures. The development programs indicated above are directed toward the goal of designing and testing a cell liner system which will not fail, even 59 under the unlikely event of a large sodium spill.

! Comprehensive Testing Program for Concrete at Elevated Temperatures 45l This ongoing experimental program will define the variation with temperature of various physical and thermal properties of prototypic-CRBRP limestone aggregate concrete and lightweight insulating concrete.

The properties include, but are not limited to, compressive strength, l'

modulus of elasticity, shear strength, bond strength, thermal conductivity, specific heat, and coefficient of thermal expansion. The series of experiments will be carried out at various temperatures including those 37 representative of accident conditions.

O Amend. 59

• C' 3A.8-8

The results of this testing program can be directly appl led to the analysis of the building structures supporting the cell liners. The testing program is nearing completion and the results will be included in an ORNL/CRBRP report f ol l ow ing compl etion.

Since the blaxial and triaxial testing of concrete at elevated temperatures will yield a greater compressive strength than unlaxial testing due to the infl uence of the lateral confining stress, the concrete tests perf ormed on specimens in the uniaxial state of stress will yield a more conservative value of strength. Theref ore the consequences of blaxxlal and triaxial loading can be disregarded.

Sodium-Concrete Reaction Tests The objective of this ongoing program is to determine the rate and extent of penetration due to sodium-concrete reaction. The ef fect of reaction product accumulation and gas release on the sodium-concrete reaction rates will be determined to allow upgrading of analytical capabil Ity. Additionally, intentionally def ected I Iner tests w il l be perf ormed to assess the response of the I iner to a sodl um-concrete reaction. Results of these tests will be documented as they become avail able.

The dimensions of the test articles have been selected to ensure that results representative of 1he actual mass concrete structure can be obtained.

Sodium Solli Design Oualification Tests A l arge scal e model of a CRBRP cell l iner has been perf ormance tested to demonstrate the abil Ity of the cell liner system to maintain l Iner integrity, mitigate consequences of a l arge sodium spil ;, and prevent sodium-concrete reactions. A total of 3500 pounds of liquid sodium at 1100 F was spilled against a CRBRP cell liner wall forming a 50 Inch deep sodium pool above the CR3RP l iner fl oor in the test article. The sodium pool was then heated, using electric heaters, to temperatures ranging between 1460 F and 1580 F and maintained until six days af ter the spill. The 1100 F sodium spill simulated a Design Basis Accident sodium spill event and the subsequent heat up to approximately 1600 F simulated the fission decay heat of a sodium pool under TNBDB Accident conditions.

The test data and post test examination revealed no f ailures or liner defects and minimal def ormation of the I iner system under the DBA and TbBDB spill conditions. The resul ts of this testing program are incl uded in the HEDL f inal report (Ref erence 5) .

Cell Penetration Sealant Tests The objective of this progran was to determine the ef fects of temperature, sodium and radiation on various candidate sealant materials f or cell penetrations. This series of experiments enables selection of the most suitable sealant material for use in the CRBRP. Following selections of the prime sealant material, prototypic electrical cable penetration assembly i

perf ormance test ing were conducted. The results of this testing program were I publ ished in Ref erence (4).

Amend. 73 O

3A.8-9 Nov. 1982

O Base Material Tests for Liner Steels The objective of this completed testing program was to determine the response of the cell Iiner plate matertal (SA-516 Grap 55) and its associated weldment material to elevated temperatures up to 7100 F. The base liner steel will be tested f or residual tensil e strength (incl uding stress-strain response),

stress-rupture (Creep) and thermal expansion. The weldment material was tested f or residual tensile strength (Including stress-strein response) and stress-rupture (Creep). Both longitudinal and transverse welds were invastigated. The results of the base Iiner steel and weldnent material tests have been published in Reference 6.

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

O v

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

References:

1. McAf ee, W.J., Sartory, W.K. , " Eval uation of the Stru.tural Integrity of O

LMFBR cel l l Iners - Results of Prol iminary Investigations", ORNL-TM-5145, January, 1976.

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

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

4. Humphrey, L.H., Horton, P.H., Al-D0E-13227 " Selection of a Sodium 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 i Test (LT-1)", December, 1980.

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

Febr uary , 1980.

O O

3A.8-9B Amend. 72 Oct. 1982

! The core support structure is welded Type 304 stainless

' steel structure which includes the core support plate and the core barrel. The core support plate contains module liners which serve as receptacles for the lower inlet modules. The core support structure carries the weight of the other portions of the lower internals structure, the reactor removable assemblies (fuel, blanket, control and radial shield assemblies) and the core former structure. The core support structure provides the upper boundary of the vessel inlet 58l plenum and distributes the coolant to the lower inlet and bypass flow modules.

The core support structure transmits the dead weight hydrostatic pressure ard seismic loads to the reactor vessel.

The core support structure concept is based upon the FFTF core support structure, however, the FFTF manufacturing experience has been utilized to reduce the complexity of the core basket. The FFTF core basket was a core diameter size structure containing receptacles so that each reactor assembly could be " plugged" into the core basket. This single large core basket has been simplified by designing mini baskets (lower inlet modules). Each inlet module receives seven reactor assemblies.

Each module in turn plugs into liners which are integral to the core support plate. The concept of these liners is shown in Figures 4.2-38 and 4.2-39. Each liner is a Type 304 stainless steel tube inserted into the support plate seated to the bottom of the plate by a flange and clamped to the support plate by a cap at the top of the liner. The q cap complies with the ASME Code requirements for the use of the non-Q 51 integral joints. The liner is sealed near the lower surface of the l

l Amend. 58 4.2-162b O

V l

plate to permit hydeaul Ic bat ance of the iowor Iniet modules. The 1iner has an alignment feature mating wIth the sup' port plate and an alignment feature for the iower Inlet modules. These two al ignment f eatures assure that the lower Inlet modules are positioned correctly. The reactor assembly discrimination feature precludes placing an assembly in an improper location.

Auxil iary flow ports and debris barriers, as shown in Figures 4.2-38 and 4.2-39 have been provided in each module i Iner to precl ude the possibil Ity of large debris of any type f rm blocking all flow to one or more of the inlet modules. The auxil lary fIow. ports are Iocated immediately below the CSS plate in a secondary Inlet plenum forrred by the ~ hexagonal debris barriers, which separate the auxillary flow ports f rm the primary flow ports and the radial ribs on the peripheral lIners. The primary flow ports are designed to prevent Iarge debris f rm entering the module l Iner stem and blocking the auxil lary ports f rom the inside and the peripheral ribs prevents debris f rom working its way in f rom the side of the array. In the event that one or more of the primary flow ports become blocked, the af fected l iner woul d then draw cool ing sodium via the auxil iary flow ports f rom the secondary plenum. Sodium feed to this secondary plenum is by (1) the auxli lary flow ports in the unblocked l Iners and (2) the array of 2 Inch diameter holes in the hexagonal debris barrier array.

Lower Inlet modules support and position the reactor assembl les on the core support pl ate. These modules, as shown in Figure 4.2-40, distribute the coolant to the various reactor components: f uel assembi les, bl anket assembl ies, removable shiel d assembl ies and control rod assembl ies. Each module fits into a l iner integral to the.' support plate and receives seven reactor assembi les and provides orifIcing that is unique to specific reactor assembly locations as shown in Figure 4,2-41.

Each of the LIMs f eature one al ignment pin, and two shorter discriminator pins.

Proper al Ignment of each LIM is assured through the mating of the al Ignment pin to the module I iner hole. Each LIM group has two uniquely machined discriminator pins that mate with two uniquely drilled holes on each of the module Iiners. DurIng Instal Iation, the al ignment pin wll l properly align the LIM. However, compieto InstalIation w!!I be prevented If the two diseriminator pins do not lIne up with module iIner holes.

Suf ficient clearance exists between the LlM and the module l iner, as well as pin / hole dimlnsions, to allow thermal expansion. The module l iner has an Interf erence f it with the Core Support Plate and it maintains a f ixed position w ith the pl ate. Both the l iners and the< Core Support Plate experience simil ar steady state temperatures and are made f rom the same material, theref ore, thermal expansion variations between the two are minimum.

Mechanical discriminating features are designed into each raodule to assure pimment of the reactor assembi les into the proper region (i.e., fuel, bl anket, and control) so that assembi les cannot be undercooled. Furth ermore, mechanical discrimination assures the proper core lattice positions fcr fuel assembl ies. Angul ar al ignment to the module f or the correct l attice position is assured by an al ignment pin between the I iner and the core support plate.

The modules are shielded by the lower shield within the reactor assemblies so that the loss of ductil ity limit is not exceeded during the plant l ife. The modules are a welded 304 stainless steel structure and all 61 modules have the Amend. 73 O

4.2-163 Nov. 1982

l l

same envelope dimensions. However, there are several distinct configurations due to the dif fering flow requirements of the reactor assemblies.

t i

]

4.2-163a Amend. 73 Nov. 1982

Loads from weight, hydraulic pressure drop and seismic acce-1eration are transmitted by the support plate to the reactor vessel.

Sizing analysis for internal pressure, flow blockage, control rod drop, and seismic loads indicate that under normal operating loads with flow blockage the inlet module meets the ASME Section III cri-teria for primary stresses.

Six bypass flow modules, surrounding the lower inlet modules, distribute low pressure coolant received from the lower inlet modules to the ,emovable radial shield assemblies. The bypass flow modules provide receptacles to accept the removable radial shield assemblies that are not positioned in the lower inlet modules.

51 The details of the FRS are provided in Section 4.2.2.2.1.4.

O 54 Amend. 54 4.2-164 May 1980

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t

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x The general design rule of 5.0% minimum residual ductility insures that non-ductile fracture will not occur during short term loadings in reactor internal structures. This criterion is based upon the minimua residual total elongation of 10.0% and the established relationship between total and uniform residual elongation of et = Cu + 5% as noted in Table 4.2-53. This relationship is based upon the end-of-life tensile test data in Tables 4.2-54 through 4.2-57 and data from References 178, 179 and 180. It is conservatively based upon a data set showing the least uniform elongation for a total elongation of 10.0%. An evaluation of all current data indicates that when the degradation on ductility is greate3t at a particular fluence level the uniform elongation tends to be a greater .

fraction of the total than this relationship indicates. Since thistlimit is based upon uniaxial test data a correction for the multiaxial state of stress for actual reactor component conditions is required. This correction -

can be performed using scientific paper 67-1D0-CODES-P1, " Applied Mechanics in the Nuclear Industry Applications of Stress Analysis". For a typical thermal stress conditions which causes an equibiaxial stress state the 5.0% .

would be reduced to 0.9%. The elongation available to insure ductile behavier can be determined by considering the factor of safety, consistent with the s ASME Code Section III factor of safety protecting against ultimate failure. ' -

The use of the factor of safety of 3.0 would reduce the elongation for a equibiaxial state of stress to 0.30%.

The applied strain considered relevant to this elongation limit is the maximum value of the three principle strains and represents an accumulation of elastic plus plastic strain at the end of life. These limits would apply (m at a minimum to membrane plus bending strains regardless of whether the s loading is primary or secondary. Thermal transient strains in reactor in-ternal components are less than the 0.30% membrane plus bending. Therefore, from the tensile data base that is presently available, the ductility required

, at the end-of-life in reactor internal components is sufficient to insure their integrity when 10% residual total elongation is available and the criteria described is applied. In locations where significant fatigue damage occurs in the low cycle regime, which is also affected by the ductility of the material, corrections to the fatigue design curves are applied using

accepted theories of fatigue design curve construction which are based upon l reduction in area.

A test program is presently in place which will experimentally characterize the fracture toughness of reactor component materials when subjected to a fast-neutron irradiation environment. This program includes tests of smooth, notched and welded specimens. The establishment of the fracture toughness and fatigue crack propagation characteristics will provide 57 a basis for confirmation of the described criteria or the substitution of a more refined criteria.

4.2.2.2.1.2 Lower Inlet Module Sixty-one inlet modules support and position the reactor assemblies on the core support plate. These modules distribute the coolant to the following reactor components: fuel assemblies, 51 blanket assemblies, removable shield assemblies, control rod Amerd. 57 Nov. 1980 4.2-165

/

T.9 LE 4.4-3 00lM. ANT LIMITim TE85 FRAT 15ES FOR TEl_T cAa ria ATIONS (TEWERAREE5 iN %)

STEACY STATE TE W.

HETEROGENEOUS ,(DRRESR)NDilG TO STEADY STATE TEW.

CORE MXilt)N HEMENEOUS (X)RE (DR(SPONDilG TO TYPICAL WORST CASE TRANSIENT TEW. , MXIMJM TRMSIENT 15507 MAXIMJM T FOR ASSEb6LY TYPE (F6AE-2M CALQX.ATED) TE W. (F8RE-2M) TRMSIEhr TEW. M Fuel Assembly 1571 1338 1316 1252 First Core 1261 Second Core inner Blanket Assembly 1498 124T 1282 1198 First Coro 1207 Second Core Radlet Blanket Assembly 1580 1331 1310 1252 i / s ,

V v l

O Temperatures at THDF, 3F, 750'F Inlet Teeperatiires f or g MOV, 2a-

! oE 58 P

$d l

TABLE 4.4-4 CORE ORIFICIf4G ZONES FLOW ALLOCATION FLOW (Ib/hr)

NO. ASSYS/ CYC:.ES CYCLE CYCLES ZONE 1,3,5,.. 2 4,6,8,..

ZONE TYPE 1 Fuel 39 169,990 (201,900) 188,520 (200,340) 187,050 (198,780)

Fuel 54 176,750 (187,870) 175,420 (186,420) 174,060 (184,970) 2 (174,620)

Fuel 21 166,900 (177,360) 165,610 (175,990) 164,320 3

Fuel 18 153,400 (163,020) 152,220 (161,760) 151,030 (160,500) 4 149,480 (158,850) 148,330 (157,630) 147,170 (156,400) 5 Fuel 24 Fuel 0,3 or 6 179,W M 89,M ) U7,M M88, M 6

Inner Blanket 6,3 or 0 68,790 (73,100) 69,330 (73,680) 7 inner Blanket 57 88,790 (94,360) 88,110 (93,630) 87,420 (92,900) 19 76.030 (82,920) 77,420 (82,270) 76,810 (81,620) 8 inner Blanket Radial Blanket 12 62,MO (66,210) 61,820 (65,700) 61,340 (65,190) 9 Radial Blanket 36 48,300 (51,330) 47,930 (50,930) 47,550 (50,530) 10 Radial Blanket 48 35,090 (37,290) 34,820 (37,000) 34,540 (36,710) 11 Radial Blanket 30 25,740 (27,350) 25,540 (27,140) 25,330 (26,920) 12 g

P NQIE: Flows are for THDV (PEOC) conditions.

4 5

o CORE REGION FLOW FRACTIONS CYCLES CYCLE CYCLES REGION 1,3,5... 2 4,6,8...

Fuel 0.65 0.66 0.66 Inner Blanket 0.17 0.16 0.16 Radial Blanket 0.12 0.12 0.12 Total 0.94 0.94 0.94 S$

?8

,_. P

$2 9 6 e

i 4 Piping shall be designed with suitable access to permit in-service  ;

testing and Inspection.

5. All " horizontal" piping shall be sloped. Steam traps and drain valves shall be located at the low points to permit complete draining of the piping. l

6. Piping sizes shall be chosen such that average fluid velocities at the 100% plant power condition will not exceed the following values:

a. water 25 fps l

b. water-steam mixture 50 fps

c. saturated steam 125 fps

d. superheated steam 175 fps System Descriotion All Steam Generation System piplng is shown in Figure 5.1-4. The design characteristics and ASE Code classifications are presented in Table 5.5-7.

The only field run piping planned for the steam generator system is non-safety class piping. The Internal diameter of the piping will be 2 inches or less and is used for drain lines f rom steam traps. The design pressure would not exceed 100 psia and the design temperature would be less than 3000F.

The Seismic Category I design requirements are placed on the Steam Generation l System's steam-water piping. Superheater and evaporator modules and the steam

drum are provided with quick acting isolation valves. Design pressures of alI l piping are nominally 110% of the operating pressure at rated power.

l The use and location of rigid-type supports, varleble or constant spring-type supports, and anchors or guides will be determined by flexibility and stress analysis. Piping support elements wilI be as recommended by the manuf acturers and will meet applicable code requirements. Direct weldment to thin wall piping will be avolded where possible.

Attachment and penetrations shalI be designed and f abricated according to the ASME Code requirements.

l l

l O

5.5-8 Amend. 71 Sept. 1982

Design loading used f or flexibil Ity and seismic analysis for the determination of adequate piping supports w Il l include all expected transient loading conditions. Spring-type supports will be provided f or the initial dead weight loading during hydrostatic testing of steam systems to prevent damage to piping supports.

Test and insoection in-service inspection is considered in the design of the main steamwater and f eedwater supply piping. This consideration assures adequate working space and access f or the Inspection of selected pipe segments.

Af ter completion of the Installation of a support system, all hanger elements will be visually examined to assure that they are correctly adjusted to their col d setting position. Upon hot start-up operations, thermal growth wil l be observed to confirm that spring-type hangers are functioning properly. Final adjustment capabil Ity will be provided for all hanger or support types.

5.5.2.3.4 Steam Generator Module The steam generator module shown in Figure 5.5-2 is a shell and tube heat exchanger with fixed tubesheets. Fl ow is counter-current, with sodium on the shell side and water / steam on the tube side. The evaporator modules transf er heat f rom the sodium and generate 50 percent qual ity steam from the subcooled recirculation water. The steam-water mixture exiting f rom the evaporator is separated into saturated water and saturated steam in a' steam drum. The superheater modules transfer heat from the sodium to superheat the saturated steam to the temperature required for admission to the turbine.

The Atomics international - Modular Steam Generator (MSG) was a 32.1 Mnt maximum power, hockey stick designed unit used as the basis for the CRBRP Steam Generator design. The sal lent features of the MSG unit are as follows:

. Ebximum Power 32.1 Mwt

. Temperature 930 F

. Pressure 2550 psig

. Startup/ Shutdown 37 Cycles

. Tube Design 158 Tubes 5/8 In. 0.D. x 109 mil . wal l

. Length 66 ft

. Material 100% Ferritic Steel - 21/4 Cr-1 Mo For f urther detail s see Reference 4.

Evaporator and superheater modules are identical in all respects except for the inlet orifices that may be added to the evaporator tubes at the lower tubesheet to increase the evaporator water flow stability margin. Each module consists of a 531/2 inch 0.D. shell containing a tube bundle with locations l f or 739 5/8 in. 0.D. x 0.109-Inch wal l tubes. The design anploys O

5.5-9 Amend. 73 Nov. 1982

[

p)

A.

An upper header thermal l iner and an inlet nozzle thermal liner are provided to mitigate the of fects of system sodium transients.

c. Shell Arranaement (1) Maior Comoonents of Shell The shell connects to an upper and lower tubesheet, and consists of two reducers, an el bow, an inlet header " tee" section, an outlet header " cross" section, a main support section and a main shell section. These components have been sized structurally to contain postulated maximum large leak SWR conditions es welI as meet design operating conditions.

( 2) Shell Penetrations Each superheater and evaporator module is f Itted with one inlet sodium nozzle and two outlet sodium nozzles. Present Ir.termediate sodium loop arrangement drawings show both superheater outlet nozzles being used, while only one of the two outlet nozzles is used on each of the two evaporator units. The spare evaporator exit nozzles are capped. The inlet sodium nozzle is a 30-inch nozzle that attaches to the 41/4-Inch thick Inlet sodium header in the direction of the hockey stick. The 30-Inch nozzle is reduced to a 26-inch, I-Inch thick wall pipe, which will be mated to the loop piping. 0 The two outlet sodlum nozzles are 22-inch nozzles that attach at 90 to the direction of the hockey stick to the 41/4-inch thick outlet sodium header. The 22-inch nozzles reduce to 18-Inch, schedule-60 pipes, which will be mated to the loop piping. The purpose of the oversized nozzles in regard to the piping size is U to provide space in the nozzles for thermal liners and to reduce flow velocities in the inlet / outlet regions.

Two 8-Inch sweepolets are attached to the reducers located at both tubesheets.

These serve as ports to inspect the final closure welds. Al so, one of the ports on the lower reducer is attached to a 6-inch schedule-80 pipe by a transition section to provide for rapid drainago of the lower stagnant end of the modules, should it be required. Again, the purpose of the transition section is to provide for possible Iining of the nozzles. A one-incF. drain is al so provided through the lower tubesheet to drain the lower thermal baf fle region. A three-Inch sodium bleed vent is provided in the hockey stick end of the moduls to provide for: 1) venting during initial filling of the chell sida, and 2) a small sampi ing flow to a hydrogen detector to allow deteciicn of any small leak in that region during operation.

(3) Steam / Water Heads The steam / water heads are integrally welded to the tubesheets. The steam piping is in turn welded to the steam heads. An Integral steam head provides an enhanced maintenance capability since 1) the heads are not removed for in-service inspections, 2) drainage of the module is not required since the integral steam head w ll l serve as the tank to contain the water medium and 3) the alr/ water exposure of the steam tubes wIlI be minimizod. The wel ded steam head al so signif icantly reduces potential steam water leakage by axchanging a p large diameter steam head seal for a smaller diameter manway seal which is h relatively insensitive to distortion and leakage during normal transients.

5.5-11 Amend. 73 Nov. 1982

Steam Generator Insoection Access to the heat transfer tubes of the steam generator is readily obtained by removal of the manway nuts and renoval of the manway cover. The steamhead is basically a 32 inch radius sphere which provides larger stress margin than the al ternate bolted design. The manway is a standard 16 inch diameter port.

The 57 inch ID sperical head provides adequate space and headroom for inspections and maintenance and tube plugging as required. The upper steamhead also sorves as the water tank for in-service inspection (ISI).

The inner diameter of the heat transfer tube is readily available for inspaction by ultrasonics, eddy current and/or other suitable means which will be determined acceptable at the conclusion of a development program (now in progress). The outer surf ace of the heat transfer tubes cannot be readily inspected since the shelI of the steam generator is a fut ly welded assembly.

How ever, it is expected that the above tube inspection techniques will give suf f Iclent information on the condition of the tubes to provide assurance of integrity of the sodium / water boundary.

5.5.2.3.5 Steam Drum The steam drum, shown in Figure 5.5 4, is a horizontally mounted 82 inch 0.D.,

35 ft. long cyl inder w ith hemispherical heads (42 f t. overall length). Most of the major nozzles are located in a vertical plane through the steam drum centerlino. These consists of one 12 Inch steam outlet nozzle located at vescel midpoint and directed vertically upward, two 16 inch riser nozzles (evaporator return) located at approximately cylinder quarter points and directed downward, four 10 inch downcomer nozzles (recirculation pump suction) spaced evenly along the cylinder and directed downward, one 6 inch continuous drain nozzle Iocated in one head and directed downward normal to the head at a 45 angle to the vertical, and one 10 inch feedwater inlet nozzle located in the opposite head and directed downward normal to the head at a 45 angle to the vertical. The only nozzle that is not coplanar with the vessol centerline is the auxil iary feodwater nozzle. This is a 4 inch nozzle located on the same head as the main feedwater inlet nozzle in a vertical plane rotated 45 f ran the vessel centerline; the nozzle is directed downward normal to the head at a 45 angl e to the vertical .

O l

5.5-11 a Amend. 73 I Nov. 1982

Safety / power rel ief valves are installed on the outlet l ine of the evaporator i I

i units, on the steam drum and on the outlet Iine f rom the supacheater. These valves all meet the requirements of Section t il of the A?lE Boiler and Pressure Vessel Code for protection against overpressure. Table 5.5-8 Indicated design pressures and valve settings for the steam generator l l saf ety/rel lef val vos. Additional valve data is provided in Table 5.5-8A.

5.5.3.5 Steam Generator Module Characteristics Each evaporator module will' produce 1.11 x 10 6 lb/hr of 50% qualfty steam from 4

subcool ed water. Each superheater module wilI produce 1.11 x 10 ib/hr of superheated steam from saturated steam. The thermal hydraul Ic normal design operating conditions are given in Table 5.5-9.

j The steam generator modules will supply the turbine with steam at design

, conditions over a 40% to 100% thermal power operating range for both : lean and i

f ouled conditions. The steam generator modules are al so capable of removing reactor decay heat wIth the natural convection in both the intermediate sodium loop and the recirculaton water loop.

This hockey stick unit is of the same basic design as that of the Atomics

! International-Modular Steam Generator (Al-MSG) unit which was tested in a test l program carried out at the Sodium Component Test Installation. The Al-MSG employed a 158-tube module with an overall length of 66 feet, as compared to l the 739-tube CRBRP Steam Generator which has an overalI length of 65 feet.

The Al-MSG heat exchanger was operated for a total of 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> incl uding operation both as an evaporator (siIghtly superheated steam out) and as a once O through evaporator-superheater (from sub-cooled liquid to completely superheated steam).

The Al-MSG served as a proof test of the Al prototype hockey-stick steam generator design. The unit was operated for 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under steaming conditions; alI of these 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, the unit was at the same temperature i

level at which the prototype will operate, with a steam pressure equal to or i

greater than prototype conditions. Tabl e 5.5-9A compares various design operating conditions for the CRBRP Units to the Al-MSG, and iIsts the number of hours which the Al-MSG operated under respective conditions. The Al-MSG operated at steam pressures equal to or greater than the CRBRP Units for essentially the whole 4,000 hrs., and at CRBRP superheater inlet temperature for 750 hrs.

Since the Al-MSG unit was operated in the once-through mod, simultaneous simulation of both inlet and outlet CRBRP conditions for the separate CRBRP evaporator and superheater units was not achieved, but operation over the l QBRP temperature and pressure range was achieved on both the sodium and steam conditions for significant portions of the test. 4

)

i I'

O 5.5-23 Amend. 73 l Nov. 1982

Safety Evaluation The steam generators are essential to remove reactor decay heat.

However, since there are three independent loops with each loop containing two evaporator modules and one superheater module, the loss of one loop would not preclude removal of reactor decay power. The steam generators 41 [ are Safety Class 2, but shall be constructed to Class 1 rules.

Design transients for normal, upset, emergency and faulted conditions are discussed in Section 5.7.3 and Appendix B.

Methods for detecting internal leakage between sodium and the

! water or steam, the margin in tube walls for thinning and time dependence of tube wastage to effect adjacent tubes are discussed under Steam Generator System Leakage Detection System, Section 7.5.5.

The rationale for the selection of any given number of failed tubes to establish an overpressure design for the IHTS is discussed under Evaluation of Steam Generator Leaks, Section 5.5.3,6.

O 1

l 5.5-23a Amend. 41 Oct. 1977

TE LE 5.5-5 SGS PUMP AND VALVE DESCRIPTION ACTUATING PUMPS ACTIVE INACTIVE SIGNAL RectrcuiatIon Pump X N/A VALVES Pump Suction isolation X Manual (Remote)

Evaporator inlet isolation X SWRPRS Evaporator inlet Water Dump X SWRPRS Evaporator Outlet Rei lef X SWRPRS**, High Pressure Evaporator (Steam)

Steam Dr'un Reilef X High Pressure - Steam Drum Superheater inlet Isolation X SWRPRS Superheater Rellef X SWRPRS**,High Pressure Superheater (Steam) l Superheater Outlet isol ation X SWRPRS**, OSIS/SGAHRS or Low Super-heater Outlet Pressure l Superheater Bypass Valve X SWRPRS**, OS IS/SGAHRS, or Low Super-heater OutIet Pressure Steam to SGAHRS HX Manual (L.O.)*

O Water f rom SGAHRS HX X

X Manual (L.O.)*

Steam to SGAHRS Auxil lary FW Pump X Manual Feedwater f rom SGAHRS X Manual (L.O.)*

Main Feedwater SGB isol ation X SWRPRS**, High Steam Drum Level, Low Steam Drum Pressure, Cell Temp and Humidity Main Feedwater Drum Isol ation X High Steam Drum Level Main Feedwater Check Valve X Simple Check Main Feedwater Control X High Steam Drum Level, Cell Temp and Humidity Startup Feedwater Control X High Steam Drum Level, Cell Temp and Humidity Evaporator Outlet Check Valve X Check Val ve Superheater Outlet Check Valve X Check Val ve Steam Drum Drain isol ation X SWRPRS**, SGAHRS Initiation, Low Steam Drum Pressure

• L.O. - Locked open

• • This function is not safety active O

Amend. 73 5.5-44 Nov. 1982

TABLE 5.5-5 (Continued)

ACTUATING O

Valves ACTIVE INACTIVE SIGNAL-SWRPRS Stack Check Valve X Check Valve SWRPRS Atmospheric Seal Bypass X Manual Sodium Dump Tank Pressure 41 Reilef X High Sodium Dump Tank Pressure 59 Evaporator Water Dump Tank Drain X Manual 42 O

O 5.5-45 Amend. 59 Dec. 1980

s The pump discharge lines contain check valves to prevent back flow through inoperable pumps. The motor driven pump discharge lines also contain a manually operated, locked open isolation valve downstream of the check valve. All three Class 3 discharge lines also have a 2 inch pump recirculation line containing an electrically-operated, normally closed isolation valve, branching off and running back to the PWST.

e. Auxillarv Feedwater Sunolv Lines The six auxillary feedwater supply lines from both the turbine and motor driven pump discharge headers are 4 inch diameter and contain (in order and in direction of flow) a manually operated, locked open isolation valve; a normally open electro-hydraulle control valve; a normally closed, electric operated Isolation valve; and a manually operated, locked open iso'ation valve. After the final isolation valve, the turbine and motor driven pump supply lines are joined. The resulting 4 Inch carbon steel line, which contains two check valves and a manual isolation valvo, is then routed to the steam drum.

Routing of the auxiliary feedwater supply lines is such that high pressure lines (high pressuring during normal plant operation) are not located in cells containing the PWST, auxiliary feedwater pumps or other SGAHRS equipment whose f ailure could cause a loss of SGAHRS saf ety function.

O f. AFW h Test Loon U Downstream of the tee where the motor-driven and turbine-driven pump supply lines join at the loop #1 valve station, an AFW pump test line returns flow to the protected water storage tank during periodic testing. This line contains redundant automatic valves for Isolating the AFW supply from the PWST should SGAHRS be Initiated during testing.

g. Steam Sucolv Line From Steam Drum to AFP Drive Turbine There are three 4 inch steam supply lines, one from each steam drum.

Each of these lines contains a locked open, manual isolation valve, an electrically operated, normally closed Isolation valve, a check valve, and another locked open, manual isolation valve. Downstream of the final isolation valves, the three lines are headered together. The resulting 4 inch line then passes through a normally closed, electro-hydraulic operated pressure control valve before entering the drive turbine.

Routing of the turbine steam supply lines is such that they do not pass through the PWST cell. When the turbine lines pass through adjacent cells, protection is provided from missiles and jet Impingement.

\s.

5.6-8 Amend. 65 Feb. 1982

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

h. Steam Drum to Protected Air Cooled Condenser (PACC) i This is a high temperature, high pressure insulated 8 inch diameter O' carbon steel line. There are three parallel lines, one to each of the 1 PACCs, which are separated by the Steam Generator Building Containment w al l s. Each line, which suppl les steam from the steam drum to the PACC, has two l ocked open, manual ly operated Isol ation val ves. Before l entering the PACC, each 8 Inch line tees into the 6 inch lines, each of which leads to one of the PACC's two half size tube bundles.

During normal plant operation, these lines remain hot due to the PACC l heat losses and natural circulation flow.

l 1. Protected Air Cooled Condenser to Steam Drum Recirculation Lines l

Condensate from each of the half size PACC tube bundles will be piped in a separate 8-inch insulated line down to an elevation 3 feet below normal water l evel in the steam drum (See Figure 5.6-7). These separate lines assure that each half size PACC bundle is isolated from the other by a water seal . The isolation allows one hal f-size PACC bundle to be started and operated independently of the other. At an elevation 3 feet below the normal water level the 8-inch hal f PACC returns join to a single 6-inch line which continues down to the l recirculation header 19 feet below the normal water level. This I common condensate return l ine contains two locked open manual l Isolation valves and a ventur! flowmeter. Above the water seal elevation, condensate flow will be a vertical annular or stratified two phase gravity fl ow pattern. A l arge l ine si'z e (8-inches) is used l to assure the two phase gravity flow remains stable and does not i resul t in entrainment over the PACC operating range. (See Section 5.6.1.3.2.3) The l ines f rom each PACC to its steam drum are separated f rom 1he l ines f or other PACCs by the Steam Generator Buil ding walls.

J. Steam Drum cr_d Suoerheater Steam Vent Lines These two Iines, one branching from the steem drum to sunsrheater piping and the other branching from the superheater to main turbine l Ine, contain a l ocked open, manual isol ation val ve and a normally closed electro-hydraulic operated pressure control valve. Both l ines are used to vent steam from the system to release heat from the plant and maintain the steam drum at a pressure below the design head of the auxil iary feedwater pumps. The superheater vent valve and vent line i

are made of 1 1/4 CR - 1/2 Mo steel; the steam drum vent valve and vent l ine are carbon steel . Following the plant trip and the initial pressure reducing transient, these valves will normally be used as the only means f or venting steam during SGAHRS operation. Power rel ief val ves l ocated at the superheater outlet wil l serve as a backup shoul d both the SGAHRS superheater and steam drum vent valves be unavailable.

These steam generator system valves will be set to open at a higher pressure. The advantage of separate SGAHRS vent valves is a controlled steam drum pressure by venting through valves designed for low erosion rather than the on/of f operation of the safety valves.

O Amend. 73 Nov. 1982 5.6-8a

I 5.6.1.3.1.5 Analvtical Method for Comoonent Suonorts (Vessels. PloInc. Pu=ns.

O and Valves)

In accordance with the ASE Code, component supports will have the same code classification as the components they support. Design of each component support will comply with the ASME Section lli design rules corresponding to the component support classification. In order to provide assurance that the component support stresses comply with limits specified in Section 5.6.1.1, i

analysis of each component support will be perf ormed. The applicable

, analytical techniques and applicable computer codes discussed in Section 5.3.3.1.5 wil l al so apply to detailed analysis of support components. The classification of components within the SGAHRS is included in Section 5.6.1.1.1.4. Allowable stress limits and pressure limits are specified in Tables 3.9-3 and 3.9-4.

5.6.1.3.2 Thermal Hydraulic Design Analysis 5.6.1.3.2.1 Natural Circulation The SGNiRS auxillary feedwater supply subsystem draws its driving force f rom the Auxiliary Feedwater Pumps. The Protected Air Cooled Condensers (PACC) operate on natural circulation on the steam and water side. The relative elevations are shown in Figure 5.1-6.

Since the relative densities of water and steam are 10:1, there will be no dif ficulty in ansuring steam supply to the PACC. The condenser design will

, O permit adequate circulation within the condenser tubing.

be verified by analyses and by proof testing af ter installation.

The PACC design wil l The PACC closed loop schematic is shown on Figure 5.6-7. The steam / water side natural circulation is comprised of two parts as follows:

(1) Steam flow from the steam drum superheater supply piping, through the steam inlet piping, into the tube bundle.

(2) Condensate flow from the tube bundle through the condensate return piping, to the recirculation pump header located below the steam drwn.

The tube bundles during normal plant operation are filled with saturated steam at steam drum conditions and kept on hot standby (i.e., Isolation from ambient by air side isolation louvers). Assuming 3% heat loss through the insulated Isolation louvers (design goal) dJring standby, condensate is formed at the rate of 2974 lbm/hr. The condensate outflow from the tube bundle during its period is due to gravity.

Upon SGAHRS Initiation signal, the isolation louvers are opened, the f an is turned on, and steam condensation increases. Condensation causes a volume collapse Inside the finned tube bundle. This volume collapse causes the bundle pressure to drop below the steam drum pressure as makeup flow from the drum is establ ished. The return piping connected to the recirculation pump header is supplied with water from the steam drum. Because this line contains

,~ ,f relatively high density water (43.2 lbm/ft3 for water as compared to 3.36 lbm/ft3 for steam) the low pressure in the bundle causes the IIquid level in 5.6-11 Amend. 71 Sept. 1982

the return piping to rise above the steam drum liquid level while steam flows into the tube bundle through the supply line. The units are designed to l condense 89,000 lbm/hr of saturated steam from the steam drum. The combined pressure drops associated with flow of steam through the inlet piping, steam /

water mixture through the tube bundle, and water through the return piping is cal cul ated to be 4 psi. This causes the liquid level in the condensate return pipe to rise 16 f t. above the steam drum l iquid l evel . This height is 11 ft.

below the low point of the tube bundle (i.e., the tube bundle exit header nozzle). This 11 f t. margin is enough that the tube bundle pressure drop coul d be as high as 4.6 psi without drawing water into the tube bundle. The tube bundle pressure drop is not expected to be more than the 2 psi allowed by the PACC Equipment Specif Icatlon.

The condensate outflow from the tube bundle is caused by two f actors as f ol l ows:

(1) Shear forces resul ting f rom flow of steam over the condensate formed in the tubes. These forces are directly proportional to the velocity dif ferential between the steam and the condensate as predicted by the relation:

T =

p(f"y)y=6 where T= he shearing stress at steam / condensate interf ace u = Steam viscosity u = Steam vel ocity 6 = Location of the steam / condensate interf ace (2) Gravitational forces causing the condensate to flow to the low point of the tube bundie.

The tube bundle l ength may be divided in three parts. The condensate flow through the first region is primarily due to shear forces as described above.

In the second region the steam velocity is greatly reduced and both gravitational and shear forces cause condensate to flow towards the tube bundl e exit header. The governing f orces in the third region are gravitational, shear, and pressure gradient induced. These f orces cause the condensate to flow into the tube bundle exit header where it is returned to the recircul ation header. The steam inlet nozzle location (high point of the tube bundle) with respect to the condensate return nozzle (low point of the tube bundle) also serves to insure flow of all condensate steam towards the condensate return pipe.

O Amend. 73 Nov. 1982 5.6-11 a

TABLE 5.6-2 v CLASSIFICATION OF SGAHRS COMP 0l4EllTS QUALITY SAFETY NATIONAL Q'JALITY _ ASSURANCE.:

COMP 0NEtlT CLASS CODES STAT lDARDS* ASME Protected SC-2 ASME III/2 Group B flA-4000 Wate r Store.ge Tank (PWST)

PWST SC-2 ASME III/2 Group B I4A-4000 l Piping PWST SC-2 ASME III/2 Group B fA-4000 Valves Protected SC-3 ASME III/3 Group C NA-4000 Air Cooled Ccndenser (PACC)

PACC )iping SC-3 ASME III/3 Group C NA-4000 d

Auxiliary SC-3 ASME III/3 Group C NA-4000 Feedwater System (AFS)

Piping AFS Pumps SC-3 ASME III/3 Group C NA-4000 AFS Valves SC-3 ASME III/3 Group C flA-4000

• NRC Regulatory Guide 1.26 " Quality Group Classifications and Standards,"

17 March 23, 1973.

A N

Amend. 17 5.6-30 Apr. 1976

_ _ _ ~ _ . _ _ . _ _ _ . _ ._ _

l TABLE 5.6-3 i SGAHRS EQUIPENT LIST AND MATERI AL SPECIFICATIONS ASE DES IGN DES IGN SECT 10N III TEMP PRESSURE SGAHRS COMPONENT _ CODE CLASS MATERIAL * (OF) ( PS IG) air Cooled Condenser Bundie 3 CS 650 2200 Air Cool ed Condenser Fan, Motor, l Louvers - --

100 ---

l l Auxil lary Feedwater Pump 3 CS 200 2200 l Pump Motor Drive - -

104 ---

Pump Turbire Drive Downstream of Admission Val ues -

CS 600 1250 Upstream of Admission Values -

CS 650 2200 1

Water Storage Tank 2 CS 200 15 l

. SGAHRS Piping:

l FWST to First isol ation 2 CS 200 15 Val ve First isol ation Val ve to AFW Pumps 3 CS 200 100 AFW Pumps to AFW Headers 3 CS 200 2200 AFW Headers 3 C!. 200 2200 AFW Headers to Electrically Operated Isolation Valve 3 CS 2Cf 2200 AFW Pump Test Loop to and between Isol ation Val ves 3 CS 650 2200 AFN Pump Test Loop From isolation Valves to FWST Fill Line 3 CS 200 100 I sol ation V al ve to Main FW Line 3 CS 650 2200 Superheater Iniet LIne to PACC 3 CS 650 2200 PACC to Evaporator Recirc Line 3 CS 650 2200 AFW Pump Recirc to Orif ice 3 CS 200 2200 Orif ice to FWST-Recirc 3 CS 200 250 Superheater Vent Line (Upstream of Vilve) 3 1 1/4 Cr-1/2 Mo 935 1900 Steam Drum Vent L ine (Upstream of Valve) 3 CS 650 2200 Superheater Vent L Ine (Down-stream of Val ve) 3 1 1/4 Cr-1/2 Mo 850 250 Steam Drum Vent LIne (Down-stream of Valve) 3 CS 400 250

, Steam Supply Line to Drive l Turbine 3 CS 650 2200 l

PACC Vent Line Upstream 3 CS 650 2200 l

of Vent Orifices) l

• CS - Carbon Steel l Amend. 73 Nov. 1982 5.6-31

TELE 5.6-3 (Cont'd)

ASE DES IGN DES IGN SECT 10N IIl TEMP PRESSURE SGAHRS COMPONENT CODE CLASS MATERIAL * ( F) (PSIG)

PACC V7/.. LIne (Downstream 3 CS 400 250 of Vent Orifices)

FWST Fil l L ine 3 CS 200 100 AFW Pump Alternate Supply Line 3 CS 200 100 Drive Turbine Exhaust 3 CS 340 100 SGAHRS Val ves:

Al ternate AFW Supply 3 CS 200 100 FWST FiiI 3 CS 200 100 W ST Drain 2 CS 200 15 FWST Level Indicator 2 CS 200 15 AFW Pump inl et (Manual) 2 CS 200 15 AFW Pump Inl et (El ectrIcal ) 2 CS 200 100 Alternate AFW Pump Inlet 3 CS 200 100 Pump Recirculation 3 CS 200 2200 Pump Recircul ation c/v 3 CS 200 2200 Pump Discharge C/V 3 CS 200 2200 Pump Discharge isolation 3 CS 200 2200 AFW Supply isolation (Manual) 3 CS 200 2200

_ AFW Supply Control 3 CS 200 2200 AFW Supply isolation (Electrical) 3 CS 650 2200 (s AFW Supply C/V 3 CS 650 2200 AFW Supply isolation (Manual) 3 CS 650 2200 AFW Supply C/V 3 CS 650 2200 AFW Pump Test Loop Isol ation 3 CS 650 2200 Superheater Vent Control 3 21/4CR-1 Mo 935 1900 Steam Drum Vent Control 3 CS 650 2200 Drive Turbine Steam Supply I sol ation (El ect. ) 3 CS 650 2200 Drive Turbine Steam Supply C/V 3 CS 650 2200 Drive Turbine Steam Supply isol ation (Manual) 3 CS 650 2200 Drive Turbine Steam Supply Pressure Control 3 CS 650 2200 PACC Steam Supply 3 CS 650 2200 PACC Steam Supply Bypass 3 CS 650 2200 PACC Condensate Return 3 CS 650 2200 PACC Noncondensibl e Vent 3 CS 650 2200 PACC Noncondensible Vent i sol ation 3 CS 650 2200 Pressure Instrument (Pump Inlet) 3 CS 200 100 Pressure instrument (Pump Discharge) 3 CS 200 2200 Pressure Instrument (Turbine inlet) 3 CS 650 2200 Chil led Water Isol ation 3 CS 200 100

• CS - Carbon Steel Amend. 73 Nov. 1982 5.6-32

TABLE 5.6-4 SGAHRS WELD FILLER METAL SPECIFICATIONS BASE MATERIAL ASME SECTION II SPECIFICATION Carbon Steel SFA-5.1 Specification for Mild Steel Covered ARC--

Welding Electrodes 1 1/2 Cr-1/2 Mo SFA-5.5 Specification for low-alloy steel covered arc-welding 26 electrodes.

O Amend. 26 Aug. 1976 5.6-33

TE LE 7.1 -3 LIST OF lEEE STANDARDS APPLICABLE TO SAFETY RELATED INSTRUENTATION AND CONTROL SYSTEMS lEEE-279-1971 IEEE Standard: Criteria for Profection Systems for Nuclear Power Generating Stations lEEE-308-1974 Criteria for Class IE Power Systems for Nuciear Power Generating Stations lEEE-317-1976 Einctric Penetration Assembl les in Containment Structures for Nuclear Power Generating Stations I EEE-323-1974 Qualifying Class IE Electric Equipment for Nuclear Power Generating Stations I EEE-323-A-1975 Supplement to the Foreword of IEEE 323-1974 lEEE-336-1971 IEEE Standard: Installation, inspection, and Testing Requirements for instrumentation and Electric Equipment During Construction of Nuclear Power Generating Stations I EEE-338-1977 Criteria for the Periodic Testing of Nuclear Power Generating Station Saf ety Systems p I EEE-344-1975 IEEE Std. 344-1975, IEEE Recommended Practices for Seismic Qual if ication of Class 1 Equipment for Nuclear Power

\

Generating Stations I EEE-352-1975 General Principles for Reilability Analysis of Nuclear Power Generating Station Protection Systems I EEE-379-1972 IEEE Trial-Use Guide for the Appl Ication of the Single-Fail ure Criterion to Nuclear Power Generating Station Protection Systems I EEE-383-1974 Standard for Type Test of Class IE Electric Cables, Field Spl Ices, and Connections f or Nuclear Power Generating Station.

IEEE-384-1974 IEEE Trial Use Standard Criteria for Separation of Class lE Equipment and Circuits l EEE-420-1973 Trial-Use Guide f or Cl ass lE Controf Switchboards f or Nuciear Power Generating Stations I EEE-494-1974 IEE~ Standard Method f or Identif ication of Documents Related to Class 1E Equipment and Systems for Nuclear Power Generating Station I

o 7.1 -9 Amend. 72 Oct. 1982

TABLE 7.1-4 RSS DIVERSITY Primary Secondarv Logic: Local Coincidence General Coincidence Sensors: Inlet PIenum Pr 9ssure Primary Loop FIow Primary Pump Speen Primary Loop FIow Intermediate Pump Speed Intermediate Loop Flow HTS Bus Frequency HTS Bus Voltage Steam FIow 1 Steam Drum Level 4

Feedwater Flow j Reaction Products Flow IHX Primary Outlet Evaporator Outlet Temperature Sodium Temperature Logic isolation: Photo Coup!Ing Direct Coupied Equipment:

o Circuitry Integrated Circuits Discrete Components o Power Supplles Separate vendors utilized o Potentimeters Separate vendors util ized o Buffers Light Coupling Magnetic CoupiIng o Control Rod Circuit Breakers in Solenoid Operated Rel ease 2/3 Logic Arrangement Pneumatic Valve in a 2/3 Logic Arrangment O

7.1-10 Amend. 73 Nov. 1982

7.2 REACTOR SHUTDOWN SYSTEM 7.2.1 Description 7.2.1.1 Reactor Shutdown System Description The Reactor Shutdown System (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 unlikely events can be terminated without exceeding the specified limits 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 unlikely events without exceeding speci-fied limits even if the most reactive control rod in the system cannot be inserted. To assure adequate independence of the shutdown systems, mecha-nical 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 con-figuration while the Secondary RSS uses a general coincidence. Sufficient redundancy is included in each system to prevent single random failure 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 n

v 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 amplified and converted for transmission to the trip comparator in the control room. Three physically separate redundant instrument channels are used. When necessary, calculational units derive additional variables from 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 specified limit 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 coin-

, cidence logic arrangement.

l 1 Operating bypasses are necessary to allow RSS functions to be bypassed during main sodium coolant pump startup, ascent to power, and two loop operation. Operating bypasses are accomplished in the instrument channels. For bypasses associated with normal three loop operation, the bypass cannot be instated unless certain permissive conditions exist which assure that adequate protection will be maintained while these protective functions are bypassed. Permissive comparators are used to determine when bypass conditions are satisfied. When permissive conditions are within the 57 allowable range, the operator may manually instate the bypass. If the G Amend. 57 7.2-1 Nov. 1980

out of the allowable range, the protective f unction is automatically rei nstated. The trip f unction will remain reinstated until the permissive conditions are again satisfled and the operator agair, manually initiates the by pass. Operator manual bypass control is not of fective unless the bypass comparator indicates that permissive conditions are satisf led. A functional diagram of the Primary and Secondary bypass permissive logic is shown in Figure 7.2-2AA.

Two loop bypasses are established under administrative control by changing the hardware conf Iguration wlthin the locked comparator cabinets. These bypasses are also under permissive control such that the plant must be shutdown to estabi Ish two Ioop operation and if the shutdown Ioop if activated the bypass Is autmatical ly removed.

Bypass features included within the Primary and Secondary RSS hardware for two loop operation will be deactivated during all three loop operating modes so that the three loop operating configuration can not be af fected by these bypass f eatures either by operator action or by two loop hardware f ail ure.

Bypass permissives are part of the Reactor Shutdown System (RSS), and are designed according to the RSS requirements detailed elsewhere in this section of the PS AR.

Continuous local and remote indication of bypassed instrument channels will be provided in conf ormance w ith Regul atory Guida 1.47, ' Bypassed and inoperable Status Indication f or Nucl ear Power Pl ant Saf ety Systems".

O l

l l

O 7.2-la Amend. 73 Nov. 1982

t v Figure 7.2-2B is a logic diagram of the Primary RSS logic trains.

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

In the Secondary RSS, the sensed variables are signal conditioned and compared to specified limits by equipment which is different from the Primary RSS equipment. The secondary logic is configured in general rather than local coincidence to provide additional protection against comon mode failure. Each instrument channel comparator outputs its trip or reset signal to a 1 of 16 logic module. The 3 redundant secondary instrument channels from each subsystem feed 3 redundant logic trains, which are coupled to the secondary scram actuators. Figure 7.2-2D is a logic diagram for the Secondary RSS logic.

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

However, since a measure of nuclear flux is necessary in both the Primary and Secondary RSS, nuclear flux is sensed with compensated ionization chambers

[] in the primary while fission chambers are used in the secondary. The Primary v RSS monitors primary and intermediate pump speed while the Secondary RSS monitors primary and intermediate coolant flow. Similarly, the steam flow to feedwater flow ratio is used in the Primary RSS while the steam drum level is sensed for the Secondary RSS.

Figure 7.2-2C is a typical Secondary RSS instrument channel logic diagram. Each protective subsystem has 3 redundani; sensors to monitor a physical parameter. The output signal from each sensor is conditioned for transmission to the trip comparator located in the control room. Redundant instrument channels are used. When necessary, calculational units are placed in front of the comparators to derive additional variables. The output of the comparators are input to redundant logic trains in a general coin-cidence arrangement.

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

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

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

b 1

Amend. 57 7.2-2 Nov. 1980

T l

Figure 7.2-2 shows the RSS Interf ace wIth the Heat Transport System (HTS) pump breaker control. Two HTS pump breakers are connected in series f or each HTS p um p. Each HTS breakers receives input f rom the Primary RSS and Secondary RSS pump tr i p l ogic, tipon receipt of a reactor trip signal f rom either Primary or Secondary RSS, the HTS pump breakers open to remove power f rom the primary and intermediate pumps.

Provisions are made to allow testing of the HTS breaker actuation f unction during reactor operation. A test breaker is used to bypass the main HTS breaker during a test condition. Test signal s are then Inserted through the Primary or Secondary RSS pump trip logic to open the main HTS breaker.

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

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

l Where signal s are extracted f rom the Reactor Shutdown System, buffers are prov ided. These buf fers are designed to meet the requirements of IEEE-279-1971. The buf fers prevent the ef fects of f ail ures on the non-lE output side f rom af fecting the perf ormance of the RSS equipment. The buffers are considered part of the RSS and meet all RSS criteria.

System Testabilltv Both Reactor Shutdown Systems are designed to provide on-line testing

a pab i l I ty. For the Primary RSS, overlapping testing is used. The sensors are checked by comparison with redundant sensor outputs and related measurements. Eoch Instrument channel includes provisions for Insertion of a signal on the sensor side of the signal conditioning electronics and test points to measure the perf ormance at the comparator (or calculational unit)

Input. Where disconnection of the sensor is unavoidable f or test purposes, the comparator is tripped when disconnected. The instrument channel electronics including trip comparators and bypass permissive comparators are tested f or abil Ity to change val ue to beyond the trip point and provide a trip input to the logic. The coatparators and logic are tested by the PPS Monitor.

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

O i

7.2-2a knend. 73 Nov. 1982

Evaoorator Outlet Sodium Temnerature

" 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 era three of these subsystems, one per loop. These subsystems detect a large class of events which Impair the heat removal capabil Ity of the steam generators. These subsystems are never byoassed.

Sodlum Water Reaction The Sodium Water Reaction Subsystems (Figure 7.2-10) detect the occurrence of a sodium water reaction w Ithin a superheater or evaporator module. There are three of these subsystems, one per loop. Each subsystem receives nine signals f rcm the sensors in the reaction products vent l Ines of a steam generator.

These subsystems are never bypassed.

7 . 2.1. 2. 3 Essential Performance Reoufrements in order to implement the required protective f unctions within the appropriate I imits, RSS equipment must meet several essentiel performance requirements.

These essential perf ormance requirements and the RSS equipment to which they apply are summarized below.

l The RSS Instrumentation will meet the essential performance requirements of Table 7.2-3. This table defines the minimum accuracy and time constants which l will result in acceptable performance of the RSS.

l Analysis of worst case RSS functional performance is based on the values given in Tabl e 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. This time incl udes the del ays due to the calcul ational units, comparators, logic, 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 second.

This time also includes the delays due to the logic and HTS scram breakers.

The RSS is designed to meet these essential performance requirements over a wide rangs of environmental conditions and credible single events to assure that environmental ef fects do not dograde the performance of the PPS. The environmental extremes are documented in Reference 13 of PSAR Section 1.6.

Provisions are incorporated within the PPS which provide a defense against the folIowing incidents:

O 7.2-11 Amend. 73 Nov. 1982

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

Measures have been teken 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 ied f or al l the electronic equipment considered here is greater than is expected to occur during normal or abnormal conditions. Electronics do not f all catastrophical ly when these I imits are exceeded even though this is the assumed f ail ure mode. The detailed design of the circuit boards, board mounting and racks incl udes f ree ventilation to minimize hot spots. Ventilation is a result of natural convection alr iiow.

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

Vibration and shock are potential causes of f ailure in electronic components. Design measures, incl uding the prudent location of equipment, minimize the vibration and shock experienced by RSS el ectron i cs. The equipment is qual if led to shock and vibration specif Ications which exceed al I normal and of f-normal occurrences.

The RSS comparators and protective logic are designed to operate over a power source voltage range of 108 to 132 VAC and a power source f requency range of 57 to 63 HZ. The maximum variation of the source voltage is expected to be t10%. More extreme variations in the power source may result in the af fected channel comparator or logic train outputting a trip signal. in addition, testing and monitoring of RSS equipment is used, where appropriate, to warn of Impending equipment l degradation. Therefore, 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 simul taneous f ail ure of al l instrument channel s or logic trains.

The majority of the RSS electronics is located in the control l building, and is not subjefred to a radioactive environment. Any PPS equipment located in the radioactive areas (such as the head access area) will be designed t'o withstand the level of activliy to which it w il l be subjected, if its f unction is required.

O 7.2-12 Amend. 72 Oct. 1982

o Tornado i

The RSS is protected f rom the of fects of the design basis tornado by locating the equipment within tornado hardened structures.

o Local Fires All RSS equipment, including sensors, actuators, signal conditioning l equipment, wiring, scram breakers, and cabinets housing this equipment is redundant and separated. These characteristics make any credible f Ire of no consequence to the safety of the pl ant. The separation of the redundant components increases the time required for fire to cause extensive damage and also allows time for the fire to be brought to the attention of the operator such that corrective action may be initiated. Fire protection systems are also provided as discussed in Section 9.13.

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

o Earthauakes All RSS equipment, including sensors, actuators, signal conditioning l equipment, wiring, scram breakers and structures (e.g., cabinets) housing such equipment, is classed as Seismic Category 1. As such, all RSS equipment is designed to remain f unctional under CBE and SSE l conditions. The characteristics of the CBE and SSE used for the eval uation of the RSS are f ound in Section 3.7. l 7.2.2 Analvsis The Reactor Shutdown System meets the safety related channel porformance and l rel labil Ity requirements of the NRC General Design Criteria, IEEE Standard Z19-1971, applicable NRC Regulatory Guides and other appropriate criteria and standards.

The RSS Logic is designed to conform to the IEEE Standards lIsted in Table 7.2-4.

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

O 7.2-13 Amend. 72 Oct. 1982

Sinole I-allure l No singt e f all ure wIthin the Reactor Shutdown System nor removal from service O of any component or channel will prevent protective action when required.

Two independent, diverse reactor shutdown systems are provided, either of which is capable of terminating ali excursions wIthout alIcw!ng plant paraneters to exceed specif led i Imits. Each system uses three redundant instrument channel s and logic trains. The Primary RSS is configured using local coincidence logic while the Secondary RSS uses general coincidence logic. To provide f urther assurance against potential c.:egradation of protection due to credible single events, f unctional and/or equipment diversity are included in the hardware design.

Bvoasses Bypasses for normal operation require manual In stat i ng. Bypasses wIlI be autcrnatically removed whenever the subsystem is needed to provide protection.

l The equipment used to provide this action is part of the RSS. Adm ini strative procedures are used to assure correct use of bypasses f or infrequent oporations such as two Icop operation. If the protective action of some part of the system has been bypassed or del Iberately rendered inoperative, this f act w il l be continuously Indicated in the control rocrn.

Multiole Setooints 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 l operating conditions, the RSS design will provide automatic means of assuring that the more restrictive setpoint is used. Administrative procedures assure proper setpoints i. Inf requent operations.

For CFBRP, power operation on two-loops wil l be an inf requent occurrence, and will only be initiated f rom a shutdown condition. While the reactor is l shutdown, the RSS equipment w il l be al igned f or two-loop operation which will incl ude set down of the appropriate trip points. Sufficient trip point set l down Is being designed into the RSS equipment to adequately cover the possible range (conceptually from 2% to 100%) of trip point adjustment required. In addition, administrative procedures (specifically the pre-critical checkof f) l will be invoked during startup to ensure that the proper RSS trip points have been set.

The analysis of pl ant perf ormance during two-loop operation has not been compl eted to date. Theref ore, the exact trip point settings f or two-loop operation cannot be specifled at this time. However, the range of trip point -

settings Indicated above is adequate to ensure that trip points appropriate for the anticipated lowest two-loop operating power can be achieved.

l In summary, the design of the RSS equipment trip point adjustments and other f eatures f or two-loop operation coupled w ith the anticipated two-loop operating power level and administrative procedures assure f ul l compl iance wIth Branch Technical Position EICSB 12 and satisfy Section 4.15 of IEEE std 279-1971.

7.2-14 Anend. 73 Nov. 1982 I

Comoletion of Protective Action i The Reactor Shutdown Systems are designed so that, once initiated, a protective action at the system level must go to completion. Return to normal operation requires eanual reset by the operator because the Primary RSS scram breakers or Secondary scrm Iatch circuitry must be manually elosed fof Iowing trip. Trip signals must be cleared prior to closure of scram breakers.

Manual Initiation l The Reactor Shutdown System tacludes means for manual initiation of each protective action at the system level with no single f ailure preventing Initiation of the protective action. Manual Initiation depends upon the operat!on of a minimum of equipment because the manual trip directly operates the scram breakers of the solenoid scram valve power supply.

O

[

\

l l

l l

l t

' 7.2-14a Amend. 73 Nov. 1982

Access Admiristrative control of access to all setpoint adjustments, module O cal Ibration adjustments, test pohts and the means for establishing a bypass permissive condition is provided by locking cabinets and other access design f eatures of the control room and the equipment racks.

InformatIon Read-Out Indicators and alarms are provided as an opr. ating aid and to keep the plant operator informed of the status of the RSS. Except for the IHX primary outiet temperature analog Indicators which are part of the accident monitoring sy stem, al l indicators and al arms are not saf ety-rel ated. The f ol l ow ing items are located on the Main Control Panel for operator information.

Analoa Indication A. Secondary Wide Range Log MSV Power Level B. Secondary Wide Range Linear Power Level C. Primary Power Range Power Level D. Reactor Vessel Level E. HTS Pump Speeds F. HTS Loop FIows G. Reactor Iniet Pressure H. lHX Primary Outlet Temperature

1. Evaporator Outlet Temperature J. Stoam Flows K. Feedwater FIows L. Steam Drum Level Indicatino 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 Te-t Status Annunciators A. Instrument Channel Trip / Reset inf ormation is provided f or each f unction i Isted in Tabl e 7.2-1 B. Logic Train Power Supply Fall ure C. Two Loop Bypasses instated Most information is also available to the operator via the Plant Data Handling and Displ ay System.

lAnnunciatorforRSSChannelTrios l A visual and audible Indication of all channel trip conditions within the RSS will be provided in the control room. These al arm conditions incl ude any l tripped RSS comparators in the Primary RSS or Secondary RSS. The Plant Data 7.2-15 Amend. 73 Nov. 1982

cm Handl ing and Displ ay system al erts the operator to signif icant deviations T" ) between redundant RSS analog Instrumentation used to monitor a reactor or pl ant parameter for the RSS.

Control and Protection System Interaction l The Reactor Shutdown System and the Piant Control System have been designed to assure stable reactor plant operation and to protect the reactor plant in the

[ event of worst case postul ated Pl ant Control System f ail ures. The RSS is designed to protect the plant regardless of control system action or lack of action. Isol ation devices w ill be used between protection and control funetions. Where thIs is done, all equlpment common to both the protection l and control function is classified as part of the RSS. Equipment sharing between protection and control is minimized. Where practical, separate equipment (sensors, signal conditioning, cabi Ing 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 l Pi ant Control System mal f unction and prevents the appropriate RSS acton.

Periodic Testing l The Reactor Shutdown System is designed to permit periodic testing of its f unctioning incl uding 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 verifying 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.

q,) Because of the time response of the undervoltage relay coil s of the scram breakers and very short duration of the test signal, the reactor does not trip, in the Secondary RSS, an instrument channel can be tested frcrn sensor l

O 7.2-15a Anend. 73 Nov. 1982

to scram actuator by inserting a single test signal because of the general coincidence conf iguration of the 3 redundant channel s. 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 j and control rod drop wIlI be tested and maintained when the plant is shutdown 1 (See Section 7.1-2). Whenever the abil Ity of a protective channel to respond '

to an accident signe! Is byparsed such as for testing or maintenance, the channel being testeu is placad in the tripped state and lts tripped condition is automatically indicated in the control rom.

Failure Modes and Effects Analysis A Failure Modes and Ef fects Analysis (FMEA) has been conducted to identify, analyze and document the possible f ail ure modes within the Reactor Shutdown System and the ef fects of such f ail ures on system perf ormance (see Appendix C, Suppl ement 1). Cmponents of the RSS analyzed are:

o Reactor lessel Sodium Level input +

l o RSS Sodium Flow Input o Pump Electric Power Sensor o Compensated Ion Chamber Nuclear input o Fission Chamber Nuclear input o Primary Loop Inlet Plenum Pressure input .

l o Sodium Pump Speed (Primary and Intermediate) o Steam Mass Flow Rate input o Feedwater Mass Flow Rate input o Steam Drum Level input o Primary Ccrnparator o Secondary Comparator o Primary Logic Train l

o Secondary Logic Train o Primary Calcuiatlonal Unit I o Secondary Calcui ational Unit l

l l

9 7.2-16 Amend. 73 Nov. 1982 l

> 1

(

i

o Scram Actuator Logic o Heat Transport System (HTS) Shutdown Logic o Control Rod Drive P4echanism (CRDM) Power Train l o RSS isolation Buf fer Figures 7.2-3 and 7.2-4 provide assistance in locating the above system level components w Ithin the overal i RSS.

The probabil ity of occurrence of each f ail ure mode is listed in the tables of Appendix C, Supplement 1, in the Probabil it/ Col umn. The ef fects of each potential f ailure mode have also been categorized in the tables in -ihe Critical ity Col umn. Even though the f all Jre of an Individual element may result in the inabil ity to initiate chan iel trip, the provision of redundant Independent instrument channels and loc,Ic trains assures that single random f ail ures cannot cause loss of either '.he Primary or Secondary RSS thereby meeting the design requirements of IEEE 279-1971. The high rellabilIty of components, redundant configuration, provision for on-1Ine. monitoring and on-line periodic testing f urther assure that random failures will not accumulate to the point that trip initiation by either Primary or Secondary RSS is prevented. All failure ef fects are therefore categorized as not causing any degradation or f ail ure of a system saf ety function. The majority of the identif ied f ailure modes can be el iminated f rom consideration based on their low probabil Ity of occurrence and the insignificance of their critical Ity.

They are included in the FMEA, however, to document their consideration.

O O

7.2-17 Amend. 73 Nov. 1982

TM3LE 7.2-1 REACTOR SHUTDOWN SYSTEM PROTECTIVE FUNCTIONS Primarv Reactor Shutdown Svstem Number of Inouts I o FI ux-Oel ayed FI ux (Positive and Negative) 2 o Fl ux-Pressure 1 o High Flux 1 o Primary to Intermediate Speed Mismatch 3 o HTS Pump Frequency 1 o Pump Electrics 1 o Reactor Vessel Level 1 o Steam-Feedwater Flow Mismatch 3 o lHX Primary Outlet Temperature 3 Secondarv Reactor Shutdown System Number of Inouts o Modified Nuclear Rate (Positive and Negative) 2 o FI ux-Total FI ow 1 o Startup Nuclear Flux 1 o Primary to Intermediate Flow Mismaten 2 o Steam Drum Level 3 o Evaporator Outlet Sodium Tanperature 3 o HTS Pump Vol tage 1 o Sodium Water Reactio- 3

' The Primary RSS can accept a t'otal of 24 Inputs and the Secondary RSS can accept 16 Inputs. There are 9 spare Primary inputs.

O 7.2-18 .4nend, 73 Nov. 1982

~

TABLE 7.2-2 RSS DESIGN BASIS FAULT EVENTS Primarv R w tor Secondarv Reactor Fault Events Shutdown Svstem Shutdown System

1. Anticinated Faults A. ReactIyIty Disturbances W Positive Ramps g d/sec ar.C Steps 110 Startup Flux-Oelayed Flux or Startup Nuclear j Fl ux- Pressure 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate or y Fl ux- Pressure Flux-Total Flow I

40-1005 Power Flux- Pressure Flux-Total Flow w

e Fuli Powor High FIux F1ux-lotal F1or

Negative Reps and Steps Flux-Delayed Flux Modified Nuclear Rate l B. SodIass FIow Disturbances Coastdown of a Single Primary or Primary-intermediate Primary-intermediate t intermediate Pump Speed MImmatch Flow Ratio l Loss of I HTS Loop Fl ux-Pressure Primary-intermediate 1 Flow Ratio l

Loss of 3 HTS Loops HTS Pump Frequency Flux-Total Flow EF 58 F

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7 .4 . 2.1. 2 Eautoment Design 6  ; A algh steam flow-to-feedwator flow ratio is Indicative of a main steam supply V leak downstream from the flow meter or insuf ficient feedwater flow. The superheater steam outlet _ valves and superheater bypass valves shall be closed with the appropriate signal suppl led by the heat transport Instrumentation sy stem (Section - 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 initiatina Circuits l The OSIS Is initiated by the SGAHRS Initiation 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 f rom the main superheated steam system by closure of the superheater outlet Isolation valves and '

superheater bypass val ves.

7.4. 2.1. 4 Bvoasses and interlocks Control interlocks and operator overrides associated with the operation of the superheater outiet isol ation val ves have not been completely def ined.

Bypass of OSIS may be required to allow use of the main steam bypass and condenser f or 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 outiet isolation valves.

7.4.2.1.5 Redundancv and Diversity Redundancy is provided within the initiating circuits of OSIS. The primary trip f unction takes place when a high steam-tc f edwater flow ratio is sensed by two of three redundant subsystems on an/ one SGS loop. The low steam drum level sensed by two of three b

o 7.4-i Amend. 73 Nov. 1982

redundant channel s in any one loop provides a backup trip f unction.

Additional redundance is provided by three independnt SGS steam supply loops serving one common turbine header. Any major break !n the high pressure steam system external frcrn the Individual loop check valves will be sensed as a steam feedwater flow ratio trip signal in al l three loops.

7.4.2.1.6 Actuated Device The superheater outlet isolallon and superheater bypass valves utilize a high rol labil Ity el ectro-hydraul ic actuato' . These valves are designed to f all ciosed upon Ioss of eloctrIcal supply to the control solenold.

7.4.2.1.7 Seoaration The OSIS Instrumentation and Control System, as part of the Decay He.?t Removal System is designed to maintain required Isolation and separation between redundant channel s (see Section 7.1.2).

7.4.2.1.8 Ooerator Information Indication of the superheater outlet isolation valve position is supplied to the control room. Indicator lamps are used for open-close position Indication to the pl ant operator.

7.4.2.2 Design Analvsls 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.

As a saf ety related system, the instrumentation and controls critical to OSIS operation are subject to the safety criteria identified in Section 7.1.2.

Redundant monitoring and control equipment will be provided to ensure that a single f ailure will not impair the capability of the OSIS Instrumentation and Control System to perf orm its Intended saf ety f unction. The system wil l be designed for f all safe operation and control equipment, where practical, will assure a f ailed position consistent with its intended saf ety tunction.

7.4.3 Pony Motors and Controls There are six pony motors, one in each primary and Internediate heat transport l oop to prov ide sodi um f l ow for decay heat removal . These motors through the use of a gear box are capable of providing fIvo to ten percent sodium flow in f Ivo discrete steps by gear changes. Sec1 Ion 5.6 describes the Interaction of l

the primary and intermediate heat transport loops with the SGAHRS to provide l decay heat removal.

7.4.3.1 Design Descriotion The pony motors are 75 horsepower, 480 VAC, 3 phase, 60 Hz, totally enclosed f an cooled Cl an IE motors. These motors are mounted on top of the sodium pump vertical drive motor. They are 1800 rpm motors which del iver power to the sodium pump via a reducing gear, an overrunning cl utch, and the vertical motor shaf t.

7.4-8 Amend. 72 Oct. 1982

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J U Amend. 73 Nov. 1982

,oLATION VALVES CONTROL LOGIC DIAGRAM 7.5-52

i p

V Each CRDM controller reouires control power to operate the interface circuitry, programmer, gate drives, internal interlocks and display equipment. As shown on Figure 7.7-4, redundant AC power sources 57 ener,gize redundant DC logic power supplies whose outputs are 'auctioneered.

This design, prevents failure of a power supply from causing a rod to drop.

The power supplies are sized to provide sufficient capacity for all of the CRDM controllers in the primary group. Transformer isolation, including grounded Faraday shields, is used to prevent failures from propagating into the controller electronics.

CRDM Motor Controller The CRDM Motor _ requires DC energization of coils in the pro-per sequence to develop the required setpoint motion. The sequence '

of coil energization for rod motion is in a two coil-three coil 57 sequence. Thus a forward step is produced each tire a leading coil is energized and also when a trailing coil is de-energized. To reverse the motion, the sequence is reversed.

The CRDM Controller uses six SCR's for each stator coil to half wave rectify the 6 phase AC input power and supply DC output to a stator coil. All six SCR's for a stator coil are turned on by one gate drive unit. The Controller incorporates the logic necessary to correctly sequence the gate drive units on and off, thereby O~ sequencing the coils in appropriate order. Separate controllers are provided for each individual mechanism. Holds are provided when input or output logic errars are detected.

571 In Single Rod Control Mode, the input circuitry to each controller accepts on-off inpu.ts for IN, 0UT,, an.d HOLD commands and, provides the sequencer with an IN pulse train, OUT pulse train, or HOLD DC output. The IN command steps a single rod down in the core at a predetermined rate. The OUT command steps a single rod up out of the core at a predetermined rate (not i

necessarily the same as the IN rate) and the HOLD command maintains the rod in its present position (no motion). The input circuitry also incorporates adjustable speed settings for the IN, OUT, and LATCH modes of CRDM operation and assures that an IN command takes precedence over an OUT command. In addition to the adjustable speed settings, the controller provides an independent speed limitation

, which has a separate clock and power supply from that used by the l input circuitry. If the input circuitry called for a speed greater 57l than 10% above 9 inches per minute due to a postulated failure, the speed limiter circuit will place the rod in the Hold Mode.

Amend. 57 7.7-5 Nov. 1980

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

in any automatic control mode, or in Group Manual modo, the mechanism controllers are operated in sequence one step at a time to keep the rod bank in required alignment. The sequence rate and direction are determined respectively by analog and digital signals f rom the reactor control system. If the selector sequence rate is higher than a predetermined trip point, an overspeed detector will alarm and place the controllers in HOLD. A functional block diagram of the control is shown in Figure 7.7-5.

Hold Bus A Hold Bus Power Supply and transfer select circuitry are provid6d to allow any controller to be replaced without a plant shutdown, in the event of a controller failure, the mechanism controller in question can be switched out and transferred to a Hold Bus. Power to the Hold Bus Power Supply is provided downstream from the scram breakers. This ensures that if a scram is initiated, a rod on the Hold Bus will also scram.

7.7.1.3.2 Primary Rod Position Indication System l

Two independent Rod Position indicating Systems are provided for each primary control rod: An Absolute Position Indication System (ARPI) and a Relative l

Position Indication System (RRPI). These syr>tems assure that the plant operators can continuously determine the position of the control rods.

The ARPI provides a direct measuranent of rod position at any time and, unlike the RRPl, does not require re-zeroing after a scrw or temporary loss of power. The system is solid state, utilizing ultrasonics and magnetics to provide a D.C. output Indicative of rod position.

The sensor for this system consists of a tube extending down f rom the top of the motor tube and into the inside diameter of the PCRDM lead screw. A nickel-cadmium wire is stretched axially through the tube. As the l u d screw translates, the flux from a torroidal magnet mounted on top of the lead screw intersects the wire at a point indicative of the rod position. Electrical pulsos sent down the wire generate magnetic fields which, when they intersect the flux of the lead screw magnet, causes a torsional strain creating a sonic pulse which travels from the point of flux Intersection upward. The sonic pulse is detected at the top of the wire, and the time of propogation is measured electronically. This propagation time is converted to a D.C. signal which is analagous to rod position.

This signal is read out on the main control panel by rod top and rod botton Indicator lights and a vertical bar graph indicator, it is al so used to operate the rod out of alignment alarm, the rod misalignment rod block system and rod control interlocks.

O 7.7-6 Amend. 73 Nov. 1982

n The Relative Rod Pcsitian Indication System provides a digital rod position Indication on a (RT at the Main Control Board. Two pairs of magnetic coil (L} pick-ups are mounted within each stator jacket above the stator and on oppo-site sides. A 6 pole magnetic section is attached to the mechanism rotor and rotates in the plane of the pick-up coll .. Voltage pulses caused by the move-ment of the poles in the proximity of the pick-up coils are sent to a digital to analog converter. The D/A converter produces an analog signal which is a measure of rod position. This analog signal is sent to the PDH&DS and the rod misalignment rod block system., The resolution of this signal is 10.1 inch.

Unlike the Absolute Position Indication System, this system must be reset after each scram and in the event of a power failure reset after power is restored. The puises are also counted by an odometer type readout in the rod control equipment room.

7.7.1.3.3 Fod Misalignment Rod Block System The rod misalignment rod block system ensures that a row 7 control rod cannot be withdrawn more than a set distance above the average position of the six row 7 control rods when the plant is operating. As shown in Figure 7.7-6, rod position signals f rom the Relative Rod Position Indication (RRPI) and Absolute Rod Position Indication (ARPI) systems are used by two redundant trains of rod blocking logic. Each logic train outputs a rod block signal when the position of one of the six row 7 control rods is more than a set distance above the average position of all the six row 7 rods comprising the operating bank. A rod block signal from either of the two redundant logic trains results in all controllers f or the six rods of the operating bank switching to the HOLD mode.

A Signals are also provided to the unit load controller of the supervisory

(-) control system to ensure that a plant loading or unloading is stopped upon the occurrence of a rod block. This prevents a reactor trip due to power / flow mismatches which may occur if sodium flow is allowed to change without a corresponding change in reactor power. In addition to the redundant logic trains, the rod block system includes:

1) Circuitry necessary to convert the pulses of the RRPI signal conditioners to en analog signal.

2) Deviation alares which continually compare the RRPI signal and ARPl signal from each rod and f rom the rod position overage circuit and provide a position f ault alarm to the Plant Annunicator System when the two signals dif f er by a set amount.

3) A Low Power Bypass in each logic train which may be manually instated at low power to disable the rod block system. This bypass is provided to allow for control rod movement which is necessary to perf orm low power physics and startup testing. This bypass is automatically removed during the ascent to power.

4) A momentary manual override f eature to allow the removal of the rod block so that the operating bank may be realigned if a misalignment occurs.

When the manual override f eature is engaged, the operator may manually insert control rods to realign the operating bank. Withdrawal of control rods *lle the manual override feature is engaged is automatically prohibited.

V 7.7-7 Amend. 73 Nov. 1982

5) Testing and bypass features to allow for the testing and maintenanco of the RRPl, ARPI or one train of the rod block system during plant operation.

6) System alarm outputs which provido signals to the Plant Annunciator System when either train is bypassed or upon the occurrence of a rod block.

7.7.1.4 Sodlum Flow Control System The Sodlum Flcw Control System consists of six controllers used to drive the three primary and three intermediate sodium pumps. Each controller consists of a cascado system with an Inner loop using speed as the feedback signal and en outer loop based on a flow feedback signal. The flow control range is 30 to ~00% of rated flow. The flow setpoints are generated either manually or by the Supervisory Control .

Figure 7.7-7 Is a block diagram of the flow / speed control loop which is typical of the six controllers in the system. The Speed Control System is an inner loop and used pump speed, which is sensed via a pump shaft mounted tachometer, as the feedback variable. The Speed Control System is lImited internally by the torque limit circuit which sets both the accelerating and decelerating torque of the variable speed pump drive.

The demand to the Speed Controller is set by the FLOW / SPEED Mode Select Switch, in the Speed Mode, pump speed is set by a manually adjusted potentiomotor; In the FIow Mode, pump speed Is set by the FIow ControfIer.

The Flow Controller uses the filtered, median select signal of three availablo redundant flow meter buffered PPS outputs as the feedback signal. This signal, along with the flow demand, is used to generate the error signal which is compensated through the Control Compensation Network and then limited by the High Spood Limit Circuit prior to being used as the speed demand signal.

The demand to the Flow Controller is set by the MAN / AUTO Select Switch, in the automatic mode, the demand comes f rom the supervisory control, while in the Manual Modo, the demand comes f rom a manually adjusted potentiometer on the control panel.

O 7 . 7-7 a Amend. 73 Nov. 1982

7.7.1.5 Steam Genera 1DC,_ Steam Drum Level Control Syltem

()

The steam' drum level control system regulates the feedwater flow to the steam drum to maintain a constant water level in the steam drum during plant operation. ,

The control system consists of a three element (steam flow, f eedwater f l ow and steam drum water level) controller and a median select module. Each of the input elements have three redundant measurement channels. The median select module selects the median signal of the three channels as the input to the controller.

l Independent Class 1E high steam drum level trip logic trains are provided at 8 Inches and 12 Inches above steam drum normal water level. Each logic train also uses three redundant inputs and a median select module.

The steam drum level control signal, the 8 Inch high level signal and the 12 inch high level signal, have separate buf fered signals provided f rom the PPS instrument channels f or isolation and independence.

The control logic is shown in Figure 7.7-1.

7.7.1.5.1 Egedwater Flow Control Valve Control The startup f endwater control valve conntrols flow in the range of 0 to 15% of rated f low. The control loop f or this valve is a single element controller, using drum water level to control valve position. The main f eedwater control

/'~} valve is closed during this operation. When the flow rate Increases to

( _) approximately 155, the control system will automatically open the main f eedwater control valve and close the startup control valve. A deadband is provided f or this switchboard point to prevent cycling f rom one valve to the other.

The control loop for the main valve is a three element controller, using drum normal water level, steam flow, and f eedwater flow, to control the valve position. Drum drain flow rate, which remains essentially constant at all power level s, is a manual input to the controller. The controller compares steam f l ow to f eedwater f l ow, and the resulting net flow error signal is canbined with the drum water level error signal, to control the valve position. Drum water level is controlled within 12 inches of the normal water level. Three redundant buf fered signals are provided from the PPS for steam flow, feedwater flow and steam drum level. The median signal of each element is provided to the steam drum level controller. Manual control of the startup I

and main feedwater control valves is provided in the control room.

Instrumentation required by this control system is obtained as follows:

o Steam Drum Level - Water level is measured by a dif ferential pressure transmitter which senses the dif f r ence between the pressure resulting from a constant ref erence column of water and the pressure resulting f rom the varlable height of water in the steam drum. The measurement is density compensated.

m U 7.7-8 i Anend 71 Sept. 1982

- initiates a transient requiring Protection System action and could con-currently degrade the performance of one shutdown system. The consequences v of this potential failure will be mitigated by diverse instrumentation in the second Reactor Shutdown System which, being independent, is unaffected by the sensor failures.

Postulated failures for the Plant Control System, their actuators, and sensors and the features included to mitigate results of these failures are described below.

7.7.2.1 Supervisory Control System _

The function of the Supervisory Control Systm shown in Figure 7.7.2, is to relate the plant load demand to the second level (subloop) control system demands and to provide trim of the subloop controls to achieve the desired temperature or pressure operating conditions. Failures of this control system could result in either a combination of misdirected subloop control system demands, or a consistent, but erroneous, set of subloop control system demands.

The first case may be caused by a failure of at least one section (e.g. , one or more programmers, one or more sensors, etc.) of the supervi-sory control. This will result in some of the subloop controllers beinn directed away from their desired profiles, while others would be controlling normally. An obvious result of this failure mode is a mismatch of some key plant variables. For instance, if the intermediate flow of a single HTS loop is at its desired flow and the primary flow is directed otherwise, the

( intermediate flow to primary flow ratio would be incorrect. The Plant Pro-tection System would then trip the plant based on this erroneous ratio. In general, failures of this type, would result in activation of those Plant Pro-tection subsystems which are based upon a ratio or mismatch of plant variables.

In the second type of failure mode, it is assumed that all plant variables are maneuvered in such a way that no mismatch occurs but that the general direction or rate of the control demands are wrong. This would result from a misinterpretation of the plant load demand or a gross failure of the entire Superviyory Control System. In general, those Plant Protection sub-systems based .on single variables (i.e., high flux, flux-delayed flux) would be activated under these conditions.

The Supervisory Control design uses multiple sensors and average /

reject, auctioneer, or medial select circuitry to minimize the possibility that single sensor failure will result in inappropriate control system action.

Failures in the electronic controllers can only affect the plant at the rate of change of the actuators (pump drives, control rod drive mechanisms, etc.).

As shown in Chapter 15, the Plant Protection System acceptably terminates the results of all incidents involving incorrect actuator response. There-fore, the Supervisory Control System is inherently incapable of initiating a transient which is more severe than the PPS design basis.

O 7.7-17

7.7.2.2 Reactor Control System The Reactor Control System shown in Figure 7.7-3 contains an Outer Core Exit Tanperature Controller with an inner loop based on flux feedback. Failure in this system could result in erronecas movement of the control rods. This could result f rom f ailure in the sensors or feedback signal conditioning (i.e., flux or temperature), f ailures in the controller electronics, or a failure in the CRDM controller. The Reactor Control System has redundant sensors and average / reject or , median select circuitry to prevent single sensor failures from initiating reactivity transients. Even though it is highly improbable, simul taneous multiple f ailures in PPS compensated lonization chamber instrumentation could cause the loss of flux Instrumentation channel s in both the Plant Protection and Control Systems, the consequences of this potential f ailure will be mitigated by diverse fission chamber instrumentation in the secondary reactor shutdown system. Rod withdrawal block circuitry and a rod misalignment rod block system are provided which are independent of the normal control to prevent control electronic f ail ures f rom causing reactor trip. Withdrawal blocks are initiated f or both high power and power-to-flow ratio. These withdrawal blocks operate directly on the Control Rod Drive Mechani sm Controllers to stop outward rod motion of all primary rods.

Withdrawal and insertion blocks are Initiated by the rod misalignment rod block system to prevent severe misalignment of the control rod bank. The Control Rod Drive Mechani sm Controller and Rod Sequencer include overspeed detector and block circultry to provide assured limitation of rod withdrawal speed even if reactor control f ailures and f ailures of the rod block or overspeed circuitry are postulated. The PPS acceptably terminates the results as shown in Chapter 15.

It is also considered that a f ailure of the Reactor Control System could resul t in improper banking of the control rods which is not severe enough to require action by the rod misalignment rod block system. Under these condi-l tions the reactor operator would have to readjust the out of bank rods manually. To aid the operation, the main control board is equipped with rod position Indications f or each rod and also an alarm if the rods deviate f rom the proper banking requirements.

7.7.2.3 M jum Flow Control Systsm A block diagram of the Sodium Flow Control System is given in Figure 7.7-7 which is typical of the six HTS flow loops. The controller contains an outer flow loop w i th an inner loop based on pump speed. A f ailure of any of the six flow controllers would resul t in improper pump speed and, consequently.

undesired sodi um f low. Power to flow or primary to intermediate flow mismatch would occur resul ting in a plant t, rip. Even though it is highly improbable, mul ti pl e f ai l ures i n PPS f l ow instrumentation could cause the loss of flow instrumentation channels such that the secondary RRS f all s to trip; the consequences of this potential failure to initiate control system action which requires Protection System action will be mitigated by the primary reactor shutdown system. Pump speed instrumentation; this is independent of the flow Instrumentation and is, therefore, not af f ected by these f ail ures.

O 7.7-18 Amend. 73 Nov. 1982

TABLE OF CONTENTS (cont'd.)

Page 9.8.5 Instrumentation Requirements 9.8-8 9.8.5.1 Sodium Sampling and Monitoring 9.8-8 9.8.5.1.1 Plugging Temperature Indicators 9.8-8 9.8.5.1.2 Sodium Sampling Packages 9.8-9 49 45 9.8.5.2 Cover Gas Sampling and Monitoring 9.8-9

' 9.9 SERVICE WATER SYSTEMS 9.9-1 15 9.9.1 Normal Plant Service Water System 9. 9-1 34l 9.9.1.1 Design Basis 9.9-1 9.9.1.2 System Description 9.9-1 9.9.1.3 Safety Evaluation 9. 9-1 S.9.1.4 Tests and Inspections 9.9-1 9.9.1.5 Instrumentation Application 9.9-2 9.9.2 Emergency Plant Service Water System 9.9-2 9.9.2.1 Design Basis 9.9-2 i

9.9.2.2 System Description 9.9-3 9.9.2.3 Safety Evaluation 9.9-3 9.9.2.4 Tests and Inspections 9.9-3 9.9.2.5 Instrumentation Application 9.9-4 1

9.9.3 Secondary Service Closed Cooling Water System 9.9-4 9.9.3.1 Design Basis 9.9-4 9.9.3.2 System Description 9.9-4 9.9.3.3 Safety Evaluation 9.9-5 15 9.9.3.4 Tests and Inspections 9.9-5 IA U

9-xi Amend. 49 Apr.1979

TABLE OF CONTENTS (cont'd. )

P_ER 9.9.3.5 instrumentation Application 9.9-5 9.9.4 Emergency Cool ing Towers and Emergency Cool Ing Tower Basin 9.9-5 9.9.4.1 Design Basis 9.9-5 9.9.4.2 Design Description 9.9-6 9.9.4.3 Saf ety Eval ution 9.9-8 9.9.4.4 Test and Inspection 9.9-10 9.9.4.5 Instrumentation Appl Ication 9.9-11 9.9.5 River Water Service System 9.9-11 9.9.5.1 Design Basis 9.9-11 9.9.5.2 System Description 9.9-11 9.9.5.3 Saf ety Eval uation 9.9-12 9.9.5.4 Tests and Inspections 9.9-12 9.9.5.5 Instrumentation Appl Ication 9.9-12 9.9.6 Potable Water System 9.9-12a 9.9.6.1 Design Basis 9.9-12a 9.9.6.2 System Description 9.9-12a 9.9.6.3 Saf ety Eval uation 9.9-12a 9.9.6.4 Tests and inspections 9.9-12a 9.9.6.5 Instrumentation Appl Ications 9.9-12a 9.9.7 Make-up Water Treatment System 9.9-12b 9.9.7.1 Design Basis 9.9-12b 9.9.7.2 System Description 9.9-12b 9.9.7.3 Saf ety Eval uation 9.9-12b 9.9.7.4 Tests and Inspections 9.9-12b O

Amend. 73 Nov. 1982 9-xil

I TABLE OF CONTENTS (cont'd.)

eaa.

9.9.7.5 Instrumentation Appl Ications 9.9-12b 9.9.8 Demineral ized Water System 9.9-12c 9.9.8.1 Design Basis 9.9-12c 9.9.8.2 System Description 9.9-12c 9.9.8.3 Saf ety Eval uation 9.9-12c 9.9.8.4 Tests and inspections 9.9-12c 9.9.8.5 Instrumentation Appl Ications 9.9-12c 9.10 COMPRESSED AIR SYSTEM 9.10-1 9.10.1 Service Air and Instrument Air System 9.10-1 9.10.1.1 Design Basis 9.10-1 9.10.1.2 System Description 9.10-2 9.10.1.3 Design Evaluntion 9.10-2 9.10.1.3 Tests and Inspections 9.10-3 9.10.1.5 Instrumentation Appl Ication 9.10-3 9.10.2 Hydrogen System 9.10-3 a 9.10.2.1 Design Basis 9.10-3 a i

O Amend. 73 Nov. 1982

TABLE OF CONTENTS (cont'd.)

Page 9.14 DIESEL GENERATOR AUXILIARY SYSTEM 9.14-1 9.14.1 Fuel Oil Storage and Transfer System 9.14-1 34 9.14.1.1 Design Basis 9.14-1 9.14.1.2 System Description 9.14-1 9.14.1.3 Safety Evaluation 9.14-2 9.14.2 Cooling Water System 9.14-2 34 9.14.2.1 Design Basis 9.14-2 9.14.2.2 System Description 9.14-2 9.14.2.3 Safety Evaluation 9.14-3 9.14.3 Starting Air System 9.14-4 9.14.3.1 Design Bases 9.14-4

9.14.3.2 System Description 9.14-4 9.14.3.3 Safety Evaluation 9.14-4 9.14.4 Lubrication System 9.14-5 9.14.4.1 Design Bases 9.14-5 9.14.4.2 System Description 9.14-5 9.14.4.3 System Evaluation 9.14-5 9.15 EQUIPMENT AND FLOOR DRAINAGE SYSTEM 9.15-1 9.15.1 Design Bases 9.15-1 9.15.2 System Description 9.15-1 9.15.3 Safety Evaluation 9.15-2 9.15.4 Tests and Inspections 9.15-2

o 9-xv Amend. 49 Apr. 1979

TABLE OF COMTENTS (cont'd. )

=

9.15.5 Instrumentation Appi ication 9.15-2 9.16 Recircul ating Gas Cool ing System 9.16-1 9.16.1 Design Basis 9.16-1 9.16.2 System Description 9.16-1 9.16.2.1 Primary Heat Transport Systems (PHTS) -

Subsystem PA, FB, PC 9.16-2 9.16.2.2 Control Rod Drive Mechanism (CRDM) - Subsystem CR 9.16-2 9.16.2.3 Sodium Makeup Pump and Vessel - Subsystem CR 9.16-2 9.16.2.4 Sodium Makeup Pump, and Pipeways - Subsystem M3 9.16-3 9.16.2.5 Col d Trap, NaK Cool ing System, PTI, SSP -

Subsystem CT 9.16-4 9.16.2.6 Reactor Cavity - Subsystem RC 9.16-4 9.16.2.7 EVS Loop 1 - Subsystem EA 9.16-5 9.16.2.8 EVS Loop 2 - Subsystem EB 9.16-5 9.16.2.9 EVS Loop 3 - Subsystem EC 9.16-6 9.16.2.10 Ex-Vessel Storage Tank (EVST) Cavity -

Subsystem ET 9.16-6 9.16.2.11 Fuel Handl ing Celi (FHC) - Subsystem FH 9.16-6 9.16.3 Saf ety Eval uation 9.16-6 9.16.4 Tests and Inspection 9.16-7 9.16.5 Instrumentation and Controf 9.16-7 9.17 Sewage Disposal System 9.17-1 9.17.1 Design Basis 9.17-1 9.17.2 System Description 9.17-1 O

Amend. 73 g, ,

Nov. 1982 l

l

} l TABLE OF COMTENTS (cont'd. )

1 j b l j 9.17.3 Safety Evaluation 9.17-1 i

9.17.4 Tests and inspections 9.17-1

j i

l 9.17.5 Instrumentation Appl Ications 9.17-1 l l

4 I '

i j

i i

i 1

j i

1 I

i I

l 1

l I l I

i i

,1 1

i t

j Amend. 73 Nov. 1982 i 9-xvi-a

q LIST OF TABLES (cont'd.)

(V6 TABLE NO. PAGE NO.

9.9 -3 Components Served by Emergency Pl ant Service Water System 9.9-15 9.9-4 Emergency Plont Service Water System Major Ccuponents 9.9-16 9.9 -5 Equipment Cooled by the Secondary Service Closed Cool ing Water System 9.9-17 9.9-6 Single Fail ure Analysis-Emergency Pl ant Service Water System 9.9-18 9.9 -7 Buil dings and Services Suppl led by Potable Water System 9.9-18a 9.9 -8 Systems and Services Suppl led by the Domineral Ized Water System 9.9-18b 9.13-1 Non-Sodium Firo Protection System Areas of Coverage and Fire Barriers 9.13-18 9.13-2 Design Basis Fire Hazards 9.13-T7 9.13-3 Non-Sodium Fire Protection System Design Basis / Features 9.13-28 9.13-4 Fire Protection System Design Description 9.13-38 9.13-5 Fire Protection System - Fail ure Mode and '

Ef fects Analysis 9 .13 -41 9.13-6 Celi Designation and Sodium Fire Protection . .

Requirements (SFPS) 9.13-44 9.16-1 Safety and Seismic Cl assif Ication 9;16-8 9.16-2 System Parameters 9.16-9

\;

s O 1 s

Amend. 73 g_ Nov. 1982

LIST OF FIGURES Figure No. Page No.

9.1-1 General frrangement Plan of Fuel Storage /

Handling Equipment and Facilities 9.1-73 9.1-2 Arrangement of Fuel Storage / Handling Equipment and Facilities 9.1-74 9.1-3 New Fuel Unloading Station 9.1-75 9.1-4 Deleted in Amendment 44 9.1-76 9.1-5 New Fuel Shipping Container Concept 9.1-77 9.1-6 Ex-Vessel Storage Tank (EVST) 9.1-78 9.1 -7 , Fuel Handling Cell 9.1-79 9.1-8 Fuel Handling Cell Spent Fuel Transfer Station 9.1-80 9.1-9 Schematic of Crane Handled Gas Cooling Grapple in FHC 9.1-81 7.1-10 Sodium /NaK Flow During EVST Normal Cooling Operation 9.1-82 9.1-11 Sodium /NaK Flow During EVST Plus Reactor Decay Heat Removal 9.1-83 24 9.1-12 Sodium /NaK Flow During Operation of the EVST Backup Cooling Circuit 9.1-84 9.1-13 Ex-Vessel Traasfer Machine on Gantry 9.1-85 9.1-14 Ex-Vessel Transfer Machine (EVTM) 9.1 86 9.1-14a Deleted Amendment 44 9.1-87 9.1-14b Deleted Amendment 44 9.1-87 20 9.1-15 EVTM Cold Wall Cooling Concept 9.1-88 9.1-16 In-Vessel Transfer Machine (IVTM) 9.1-89 9.1-16A IVTM Grapple in Core Assembly Socket 9.1-90 34 49 9.1-16B Schematic of IVTM Interlock Over Core Positions 9.1-91 0

~**

Amend. 49 Apr. 1979

Main supply ducts leaving the air handling units run in the Operating Floor E1. 862'-0" parallel to column i Ines TC and TF to serve the west and east zones respectively. Branches f rom the main ducts are routed to the lower elevations to distribute the supply air to the various areas of the building.

l Eleven (11) roof exhaust f ans are provided, with a total capacity equal to ine total air supply, for the building. The three (3) exhaust f ans serving the i Chemical Storage Area, Lube Oil Storage Area and Operating Floor are sized to '

handle the minimum outside air and run continuously while the number of remaining exhaust f ans that are operating is determined by the percentage of outside air in the total air supply for the building.

The Roof Exhaust Fans are located as follows:

2 Fans at Roof E1. 861 '-0" l 2 Fan at Roof E1. 878'-0" 5 Fans at Roof E1. 910'-6" 2 Fans at Roof E1. 921 '-0" A rel ief hood, located above the deserator area, will relieve the air from the building and maintain the air balance during the various steps of exhaust f ans operation.

Two (2) unit coolers are provided to serve the condensate pumps and L.P.

Feedwater Heaters to supplement the main HVAC system and conserve energy during part load operation.

A separate Heating and Ventilating Unit at E1. 816'-0" along with a roof exhaust fan at Roof E1. 861'-0" serve the Ammonia Storage Room and operate continuously. The supply and exhaust air quantitles are balanced to maintain sl ightly negative pressure inside the room.

The sampl Ing room is served by a branch duct f rom one air handl ing unit.

Constant temperature and humidity are maintained inside the room hrough a heating coil and a Steam Humidif fer.

The caustic and acid storage room is served by a branch duct from the main air handling system with a reheat coil that maintains the required Indoor design conditions.

The Radiation Monitoring System will provide the equipment necessary to sample l

and analyze tritium in the exhaust air released f rom the building to meet the l requirnments of 10CFR20.

9.6.4.3 Safetv Evaluation l This section describes the design and operation of the TGB HVAC Systen during single f ail ure of the TGB HVAC System Components.

I l

Amend. 73 9.6-39 Nov. 1982

The TG8 FNAC System consists of activo and passivo compononts.

This system design has no provisions f or f ail ures of passivo components.

Activo components in the TGB INAC System which are susceptible to f ailure are as f ollows:

Supply fans Exhaust Fans Autcrnatic Rol 1-Typo Fil tors Outsido Air Dampors Return Air Dampers Exhaust Air Dampors Unit Cooler Supply Fans l The TGB IN AC Sy stem is prov ided w ith f lvo (5) supply f ans and twel vo (12) oxhaust fans. Failure of any one supply or exhaust f an would not increaso tho l

average temperaturo in the af f acted area abovo 120 F.

l The TGB Air Handling Units are provided with automatic roll-typo fil tors with an autcznatic advanco mechanism that advances the f Ilfor medium on the basis of sensed dif ferential pressuro across the f il ter. The f ail ure of the advance mechanism resul ts in increased pressure across the f iltors. A sensing dovice is provided for each f ilter with an alarm sotpoint to Indicate higher than normal dif forontial pressure across the f il ter. The alarm sotpoint is solocted on the basis that af ter Initiation of the alarm, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> are available to correct the f ailure without signif icantly deviating f rom the systom design paramotors.

Tho steam generator f ood pump aroa, the condensate pump area, the L.P.

Foodwater Heator Area, are provided with two (2) unit coolers. Each unit cooler has ono (1) 100% capacity contrif ugal fan. Fail ure of any one unit cooler f an would not increase the spaco temperature in the immodlato area abovo 120 F.

9.6.4.4 lostina and Insocction Reautrements AlI compononts are testod and Inspected as separato components and as Integrated systems. Velcrnoter readings are taken to ensure that all systems are balanced to del Ivor and exhaust the required air quantitles. All water colI s are hydeaul leal ly tested f or ieakago pr f or to being placod in servIco.

Capacity and performance of the f ans are tested according to the Air Moving and Conditioning Association requirements prior to operation of the plant.

9.6.5 Diosol Generator Building HVAC System 9.6.5.1 Design Basis 9.6.5.1.1 Diesel Generator Rooms HVAC System The Diosol Generator Rooms HVAC System is a safety related system designed to provIdo vontllatIon to the Diosol Genorator Rooms under alI conditions.

O Amend. 73 9.6-40 Nov. 1982

l i

I g- S The system provides the required environment to permit personnel access during i

'! normal plant operation and to ensure operability of the equipment under all conditions. The HVAC system serving the Diesel Generator Rooms is designed to:

a) Limit maximum temperatures in the Diesel Generator Rooms to 120 F.

b) Operate from the Cl ass IE AC power supply during loss of of f-site pow er.

c) Provide air movement through the Diesel Generator Rooms to the final exhaust points during normal plant operation and when the Diesel Generators are in operation, d) Provide heating during the winter months to the Diesel Generator Rooms during normal plant operation and when the Diesel Generators are in operation.

9.6.5.1.2 Diesel Generator Buildino Class IE Switchaear Room HVAC System The design basis f or the Diesel Generator Building Class lE Switchgear Rooms HV AC System is provided in Section 9.6.1.1.2.

9.6.5.1.3 Diesel Generator Building Non-Class IE Switchgear Rooms and Motor Generator Set Rooms HVAC System

~'N The design basis f or the Diesel Generator Buil ding Non-Class IE Switchgear (d Rooms and Motor Generator Set Rooms HVAC System is provided in Section 9.6.1.1.3.

9.6.5.1.4 Diesel Generator Building Motor Generator Sets Unit Cooler Svstem The design basis f or the Diesel Generator Building Motor Generator Sets Unit Cooler System is provided in Section 9.6.1.1.3.

9.6.5.2 System Descriotion 9.6.5.2.1 Diesel Generator Rooms HVAC System The Diesel Generator Rooms HVAC System P&lD is shown on Figure 9.6-11. The classification of the Diesel Generator Rooms HVAC System components and their primary parameters are Indicated in Table 9.6-6.

One (1) 100% capacity Heating, Ventilating Unit is provided for each Diesel Generator Room to satisfy the Ventil ation Requirements during normal operation, in addition, two (2) 50% capacity A

Amend. 73 9 .6 -41 Nov- 1982 9

Emergency Supply Fans are provided for each Diesel Generator Room. The operation of the supply fans are in conjunction with the operation of  ;

the Diesel Generator which they serve. The temperature in each cell is controlled by modulating the outs'de air and return air dampers. The air is relieved from the Diesel Generator Rooms through an exhaust damper connected to the DGB exhaust structure.

The day tank cell in each of the two (2) Diesel Generator Rooms at El. 816'-0" is ventilated by an Exhaust Fan using infiltrated air from the Diesel Generator Rooms in which they are located. These Exhaust Fans also exhaust the air from the fuel oil transfer pump cells.

One of the 100% capacity Heating, Ventilating Units is located at El. 829'-0" south of the missile protected air intake structure serving Cell No. 511. The other unit is located at El. 829'-0" south of the missile protected air intake structure serving Cell No. 512. The units are connected by a plenum to their respective air intake structurcs.

The suction side of each unit is connected to the plenum by an automatic damper and a flexible connection. The discharge side of each fan is provided with a flexible connection and supply ductwork to distribute the air to the cell. The fans are V-belt driven centrifugal fans.

Two of the 50% capacity supply fans are located at El. 837'-0" south of the missile protected air intake structure serving Cell No. 511.

The other two 50% capacity supply fans are located at El. 837'-0" south of the missile protected air intake structure serving Cell No. 512.

Each pair of supply fans are connected by a plenum to their respective air intake structure. Two (2) return air openings with automatic dampers are provided in each plenum for each pair of supply fans. The suction side of each fan is connected to the plenum by a flexible connection and an inlet bell. The discharge side of each fan is provided with a flexible connection, duct transformation section, and an automatic damper. The fans are direct driven vaneaxial fans.

A missile protected exhaust structure is located on the roof a t El . 847'-3". Gravity dampers are provided which connect the exhaust structure with each Diesel Generator Room.

An Exhaust Fan is provided for exhausting each day tank cell at El . 816'-0", and each fuel oil transfer pump cell at El. 808'-0".

The fans are located at El. 816'-0" and are connected by ductwork to the DGB missile protected exhaust structure. The fans are direct driven vaneaxial fans with flexible connections at the inlet and outlet of each fan.

49

/ mend. 49 Aoril 1979 9.6-42

l O O O TABLE 9.7-1 COMP 0NENTS SERVED BY THE NORMAL CHILLED WATER SYSTEM LOCATION E_0UIPMENT TITLE _

BLDG. CELL ELEVATION MG Sets A/H Unit CB 412 847'-3" Loop #1 A/H Unit SGB 244 852'-6" Loop #2 A/H Unit SGB 245 852'-6" Loop #3 A/H Unit SGB 246 852'-6" SGB-IB A/H Unit SGB 262 816'-0" Maintenance Bay A/C Unit SGB 261 816'-0" i Primary Na Tank Unit Cooler SGB 211 733'-0" Below Operating Floor A/C Unit RCB 105I 752'-8" Below Operating Floor A/C Unit RCB 105K 752'-8" Operating Floor Unit Cooler RCB 161A 857'-11" Operating Floor Unit Cooler RCB 161A 857'-11" Operating Floor Unit Cooler RCB 161A 857'-11" LCCV Unit Cooler RCB 125 733'-0" RCB A/H Unit SGB 271 836'-0" RSB A/H Unit RSB 305H 733'-0" P RWA A/H Unit RSB 660 867'-0"

? Comunication Center A/C Unit RSB 328 865'-0" O Air Handling Unit TGB -

892'-0" Air Handling Unit TGB -

862'-0" Unit Cooler TGB -

816'-0" Unit Cooler TGB -

838'-0" Air Conditioning Unit PSB 105 816'-0" Air Conditioning Unit PSB 151 816'-0" Air Conditioning Unit PSB 151 816'-0" Air Conditioning Unit PSB 151 816'-0" Air Conditioning Unit WB 212 828'-0" Air Conditioning Unit WB 210 828'-0"

, Air Conditioning Unit WB 210 828'-0" CRDM RCB 152 794'-0" CRDM RCB 152 794'-0" Cold Trap, NaK Cooling, etc. RCB 105V 794'-0" RSB 306B 755'-0" Q =

59lEVST Cavity 44

.-. ." 47 I

TELE 9.7-1 (continued)

LOCATION EQUIPKNT TITLE BloG, GLL ELEVATION HITS, il RG 1501 752'-8" PHTS, #2 RG 105J 752'-8" RITS, #3 RG 105K 752'-8" Reactor Cavity RG I05E 733'-0" Cool er RG 125 746'-0" SGB 235 7 82'-0" l Intermediate System Cooler & Condenser 125 746'-9" Vapor Condenser RG Intermediate System Condenser SGB 235 782'-0" l Autociave Sparge Gas Condenser RSB 640 8168-0" Prleary Cold Trap NaK Cooler RG 131 790'-4 7/8" CAPS Conpressor Cooler RSB 365 7558-0" e RSb 366 755'-0"

. CAPS Cmpressor Cooler N RAPS Cmpressor Cooler RG 1050D 733'-0"

' RAPS Canpressor Cooler RG 105BE 733'-0" os SGB/IB Air Handling Unit SGB 271 836'-0" Constant Temperature Bath TGB 838'-0" Third Loop RSB 324 8168-0" HAA Unit Cooler RG 152 800'-9" RAPS & CAPS Unit Cooler RSB 365 755'-0*

N$

5a J

$d O O O

I p

d 9.9 SERVICE WATER SYSTEMS 9.9.1 Normal Plant Service Water System 9.9.1.1 Design Basis The Normal Plant Service Water System is a non-safety related system designed to provide cooling water for the Normal Chillod Water Sysiem chiller condensers, the Secondary Service Closed Cool Ing Water System and other equipment i Isted in TaH e 9.9-1 during normal pl ant operation and pl anned outages. The system will be designed according to the ASE Section Vill / ANSI B31.1 requirements.

9.9.1.2 System Descriotion The Normal Piant Service Water System is shown in Figure 9.9-1. The system consists of two (approximately 26,600 GPM) 100 percent capacity electric motor driven vertical, wet-pit, circulating water pumps and the required piping, val ves and instrumentation. The Normal Pl ant Service Water is pumped f rom the basin of the Circulating Water System cooling tower to the equipment to be cooled, and is returned to the cool ing tower return header. The pumps are Iocated in the Circulating Water Pumphouse. Normally, one pump is operating with the second pump in an auto-standby mode.

p The components served by the Normal Plant Service Water System are l isted in Q Tabl e 9.9-1. Design data for the major system components are I isted in Table 9 .9 -2.

9.9.1.3 Safetv Evaluation

( The Normal Plant Service Water System is a Seismic Category lli and a nonsafoty cl ass system.

Pipe break analysis for this moderate energy fluid system will be provided in the FSAR.

9.9.1.4 Tests and insoections The Normal Plant Service Water pumps are tested at the manuf acturer's f acil Ity and rotested in the system prior to continuous plant operation. The operation of the pumps wIlI be rotated to equal ize wear.

9.9.1.5 Instrumentation Aoolication Indication of the Normal Plant Service Water header pressure is provided in the Control Room. Normal Plant Service Water low discharge header pressure U

Amend. 73 Nw. 82 9.9-1

is annunciated in the Control Rocq. A logic circuit is available to automatically start the standby pump when the operating pump motor trips or is inadvertently stopped.

9.9.2 Emergency Plant Service Water System 9.9 . 2.1 Design Basis The Emergency Plant Service Water System is designed to provide suf ficient cool ing water to permit the safe shutdown and the maintenance of the saf e shutdown condition of the plant in the event of an accident resulting in the loss of the Normal Plant Service Water System or the loss of the plant AC power supply and all of fsite AC power suppl ies. The Emergency Pl ant Service Water System is not used during normal pl ant operation. The system provides the Emergency Chilled Water System chiller condensers and the Standby Diesel Generators w ith cool ing water. Additionally, this system provides f ire f Ighting water f or the solsmical ly qual if led f Ire pumps of the nonsodium f Ire protection system. The Emergency Pl ant Service Water System includes the Emergency Cool Ing Towers and Emergency Cool ing Tcwer Basin, as described in Section 9.9.4.

The Emergency Plant Service Water System is designed to Seismic Category I requirements as def ined in Section 3.2. Pumps, valving and piping required for the saf e shutdown of the plant are designed to ASE Section 111, Class 3 requirements, as def ined in Section 3.9.2. All electric motors serving the system are connected to the Class 1E onsite power supply, in case of loss of plant and of f site power, these motors are switched autcmatically to the Standby Diesel Generator. The piping and equipment for each redundant loop of the system is physically separated or protected with a barrier to conform to common mode f ail ure criterion. System piping is below ground between the Seismic Category l Emergency Cool ing Tower and Diesel Generator Bull ding. The Emergency Cool ing Tower structure is tornado missile hardened as described in Section 9.9.4.1.

9.9.2.2 System Descriotion The Emergency Plant Service Water System (EPSW) consists of two 100 percent capacity f ul ly redundant cool ing loops. Each cool ing loop incl udes one circul ating pump, one make-up pump, one emergency cool ing tower and associated piping, val ves, instrumentation and control s. Figure 9.9-2 shows the various equipments and represents the system component conf Iguration and rel ationship.

The components served by the Emergency Plant Service Water System are i Isted in Tabl e 9.9-3. Design data on the major system components is I isted in Table 9.9-4.

Upon loss of Normal Chilled Water or upon start of the Standby Diesel Generators, the EPSW pumps, EPSW makeup pumps, and Cool ing Tower Fans wIl I autanatically start and provide cool Ing water at 90 F maximum to the O

l 9.9-2 Amend. 73 Nov. 1982 L

(N Emergency Chiller Condensers in the SGB and the Standby Diesel Generators in h the DGB. The EPSW pumps take suction f rom the Emergency Cool ing Tower operating basins which are located adjacent to the Emergency Cool ing Tower.

During system operation the EPSW makeup pumps will transfer water trom the common storage basin to the redundant operating basins to compensate for evaporative and drift lccses from the towers.

Cooled water from the Emergency Cooling Tower operating basins is pumped via underground supply mains to the emergency loads in the DGB and SGB. After cool Ir.g the emergency chillers and the standby diesel generators, warm water is returned, al so through underground mains, to the Emergency Cool ing Towers.

To account for seasonal temperature variations, temperature control valves served by electro-hydraulic operators bypass a portion of the returning water back to the pump suction. A temperature indicator controller automatically adjusts the valves as required to maintain supply temperature above 55 F, the minimum required for chiller operation.

in addition to cool ing the Emergency Chilled Water chillers and the standby Diesel Generators, each loop of the EPSW System provides a connection to supply water to the Non-Sodium Fire Prctection System. The EPSW pumps and the Emergency Cooling Tower Basin are designed to allow fire protection operation while maintaining the capabil Ity for supplying 100 percent cool ing to the unorgency loads. The fire protection pumps are provided with Instrumentation that will automatically terminate operation when a r. escribed amount of water has been used (see Section 9.13). This ensures that the guaranteed 30 day s

supply of water for EPSW system operation wilI not be cocoromised. In T addition, this system is connected to the EPSW loops in such a manner as to preciude a single f all ure from compromising the capabil Ity of the EPSW system to perf orm its required f unction.

9.9.2.3 Safetv Evaluation The EPSW system is a Seismic Category 1, safety related system designed to have 100% redundancy in both active and passive components. The system is provided w ith AC power f rom the Cl ass 1E power sources. EPSW Loop "A" is suppl ied from Cl ass 1E Division 1 and Loop "B" is suppl ied from Cl ass 1E Division 2. This arrangement assures that 100 percent cool ing capabil ity will be available even if one of the Standby Diesel Generators or one of the EPSW loops shoul d f all .

The EPSW system is a f ully automatic system, normally controlled from the Main Control Panel in the Control Room. Redundant control s have been provided that wilI alIow fulI operation of the system from a controf panel in the Diesel Generator Bull ding.

~

Pipe break analysis f or this moderate energy fluid system will be provided in the FSAR.

l l

1 9.9-3 Amend. 73 Nov. 1982

During the initial phase of recovery from an accident, one O

Emergency Plant Service Water loop satisfies the cooling of the Standby Diesel Generators and the Emergency Chilled Water Chiller Condensers.

The Emergency Plant Service Water System is capable of accommo-dating any single component failure without affecting the overall system capability of providing cooling water to achieve a safe shutdown con-di ti on . A single failure analysis of the Emergency Plant Service Water 59 System is given in Table 9.9-6.

15 9.9.2.4 Tests and Inspections The system components will be tested at the manufacturer's facili-ties, and a complete system test will be accomplished prior to plant operation.

The EPSW System does not operate during normal plant operations. However, the system, including all active components will be operated periodically during the year in conjunction with the Standby Diesel Generator testing program as outlined in USNRC Regulatory Guide 1.108. The system can be proven operable at any time by manual initiation. Inservice inspections will be conducted according to ASME Section XI, as described in Section 9.7.2.1.g. In addition, isolation valves and pressure test connections on the supply and return headers in the pumphouses and the DGB pennit 50 inservice inspection of the buried piping by hydrostatic testing.

9.9.2.5 Instrumentation Application Instrumentation will be provided for local and/or remote (Control Room) indication of the following parameters as indicated:

pump discharge pressure (local / remote) 59l - diesel generator / emergency chilled water chillers supply temperature (local / remote) 50 - storage basio level (local / remote)

- diesel generator and emergency chiller flow rate (remote) diesel generator and emergency chiller supply temperature (local)

- diesel generator and emergency chiller return temperature (local / remote)

- diesel generator and emergency chiller supply and return pressure (local)

- operating basin level (local / remote)

- makeup water flow (local / remote - alarm on low)

A flow switch, located in the return line from each diesel generator and emergency chiller will detect an abnormal low flew condition 43 33 and energize an annunciator in the Control Room.

1519.9.3 Secondary Service Closed Cooling Water System The objective of the Secondary Service Closed Cooling Water (SSCCW) System is to provide cooling to auxiliary equipment located in the turbine building.

Amend. 59 Dec. 1980 9.9-4

l l

(3 The Emergency Cool ing Towers pumphouses, operating basins and storage basir.

V are designed to withstand the most severo natural phenomena (e.g., Safe Shutdown Earthquake, tornado, tornado missiles, wind, Probable Maximum FloM or drought). The design has the necessary redundancy of components.

Electrical power f or The Emergency Cool ing Tower f ans, pumps, and control equipment is provided f rom the Class 1E AC power supply. One loop is provideo with electrical power f rom System Class 1E Division 1 and the other from System Class IE Division 2.

9.9.4.2 Design Descriotion The Emergency Cool ing Tower Structure consists two of pumphouses (containing the pumps and piping of the EPSW System, Section 9.9.2) located directly above the operating water storage basin. The cool ing towers, pumphouses and operating basins are 100% redundant Seismic Category 1, Tornado protected structures. The common storage basin is a Seismic Category I, flood and tornado protected structure. The storage basin has suf f Iclent storage capacity for 30 days of operation, incl uding 30,000 gallons of water storage f or the non-sodium Fire Protection System pl us adequate al lowance for drif t and evaporation losses.

Each cooling tower is designed to achieve jhe required heat dissipation rate at any time, approximately 2.36 x 10 BTU /HR at the maximum Emergency Plant Service Water Flow of approximately 3600 gpm.

The change in water chemistry due to the absence of blow-down f rom the cool Ing g towers has minimal ef fect on operation of the Emergency Plant Service Water

[Uj Sy stem. Proper selection of the Emergency Plant Service Water components, appl led blocide additives, and maintainence of proper water chemistry will provide compensation for the increased tube foul ing. The maximum makeup water required af ter 30 days of operation is approximately 100,000 gallons per day.

In case the make-up water is not available af ter 30 days, make-up water can be suppl led by either truck, rail or temporary piping f rom the Cl Inch River or purchased under agreements wIth the Department of Energy, Oak Ridge Operations.

The top elevation of the Emergency Cool ing Tower Basin is 818 ft, which is 9 l f t. above the probabl e maximum flood l evel . The entire basin and the cooling tower supports are founded on siltstone. The basin is a below grade reinf orced concrete structure. For f urther detail s on the basin, ref t to Section 3.8.4.1.5.

Each Emergency Cool ing Tower consists of a single cell, provided with an induced draf t f an system. Each cool ing tower is enclosed in a Seismic Category 1, tornado missile protected structure. The water intake and 9.9-7 Amend. 73 Nov. 1982

discharge piping are located within the tower or safely below the ground for tornado missile protection. The water intake and discharge piping and the Internal distribution piping are Seismic Category I, ASE Section Ill, Class 3 design. Each Emergency Cool ing Tower has a design flow rate of 3600 GPM.

The Emergency Cool ing Towers are of a counter-flow, wer-type, mechanically induced draft design. The Internal distribution piping distributes the intake water evenly over the f ill area so that suf ficient water area is exposed to the counter air flow to provide evaporation for the required heat removal.

The counter air flow is provided by the induced draft fans.

Drif t el iminators are located above the Internal water distribution piping and below the induced draf t f ans. The drif t ol iminators are a z igzag pattern of channel s which prevent water carryover through the f an stack.

The Emergency Cool ing Towers are supported by the reinforced concrete storage basin. The top of the cool ing towers is approximately 44 ft. above the maximum water level of the storage basin.

The Emergency Cool ing Tower Basin is f illed with potable grade water which is treated for bacteria control. The qual ity of the stored' water is analyzed at regular interval s and the required blocido additive is injected manually in quantitles required to control seasonal variations of the bacteria growth.

The Emergency Cool ing Towers and Emergency Cool ing Tower Basin wilI be seismically analyzed as described in Section 3.7.

9.9.4.3 Safetv Evaluation l The Emergency Cool ing Tower structure consists of two 100 percent capacity l cool ing towers pumphouses, and operating basins and one 100 percent capacity below grade cool ing water storage basin. The entire structure is Seismic Category 1, tornado, and fl ood protected. Piping, associated with the Emergency Cool ing Tower is designed to ASE Section lil, Cl ass 3 requirements.

The structure can withstand the most severe natural phenomena expected, and other site related events, such that the Emergency Cool Ing Tower cool ing capabil Ity is assured under required conditions. The method of analysis is similar to that used for other Seismic Category I structures. The entire structure is designed to w Ithstand the Safe Shutdown Earthquake. The fili, l drif t el iminators, motors, mechanical drives, piping, electrical conduit, l cabi es and supports w ll l be seismically analyzed in accordance wIth the procedures discussed in Section 3.7.

O 9.9-8 Amend. 73 Nov. 1982

1 9.9.6 Potable Water System 1

9.9.6.1 Design Basis The non-saf ety related Potabl e Water System receives drinking qual Ity water from the Bear Creek Road Filtration Plant and supplles it to the Fire Protection Storage Tank and to the Potable Water Storage Tank, which feeds the Potable Water Supply pumps and the Make-Up Water Treatment System. The supply pumps distribute water to the various plumbing fixtures and other services throughout the pl ant buidings. All piping and components shall be designed, fabricated, inspected and erected in accordance with the Standard Plumbing Code.

9.9.6.2 Svstem Descriotion The Potable Water System consists of a transmission l Ine, Potable Water l Storage Tank, two (2) 100 percent Potable Water Supply Pumps, (one (1) acting as a spare), distribution piping, valves, instrumentation and control s. The Potable Water System provides potable water to the buildings and services i Isted in Tabl e 9.9-7. The Potable Water Supply Pumps take suction from the Potable Water Storage Tank and deliver water to the distribution header.

Demand incl udes; supply for sinks, toll ets, showers, eyewashes, water fountains, service hose connections and other services in BOP and Reactor Support Bulldings; Circulating Water Pump seal; and Hypochiorito Generating Plant supply. The storage tank is sized to maintain a one (1) day reserve for p

b all normal services based on the maximum estimated short term demand, except f or suppl ler to the Hypochi orIte Generating Pl ant. A recirculation orifice is installed in a bypass line running f rom each pump discharge, back to the Potable Water Storage Tank through a common header. Backflow preventers are installed in the supply lines to the Hypochlorite Generating Plant, the Circulating Water Pump seal s, the Fire Protection Storage Tanks and to the Makeup Water Treatment System Clearweli Pumps to prevent possible contamination of the Potable Water System.

! 9.9.6.3 Safetv Evaluation The Potable Water System is a Seismic Category lil and non-safety class sy stem.

9.9.6.4 Tests and insoections Potable Water Supply Pumps are tested at the manuf acturer's f acil Ity and again prior to normal pl ant operation.

I 9.9.6.5 Instrumentation Aoolications Each pump is provided with a storage tank low-low level Interlock and alarm to stop the pumps. The spare pump is on standby and starts autanatically upon f ail ure of the operating pump. Inlet flow to the Potable Water Storage Tank is automatically controlled by a level control valve.

O i

Amend. 73 I Nov. 1982 9.9-12a

9.9.7 Make-Uo Water Treatment System 9.9.7.1 Deslan Basis The non-saf ety related Make-Up Water Treatment System receives potable water f rm the Potable Water Storage Tank and provides high purity demineralized water for the Dmineral Ized Water and the Condensate Systems.

All pip'ng and components are designed, fabricated, inspected and erected in accordance w Ith ANSI B31.1, Power Piping.

9.9.7.2 System Descriotion The Make-Up Treatment System consists of clearwell pumps, granular activated carbon units, two (2) 100 percent capacity domineralIzer trains, chemical injection equipment, instrumentation, pumps, val ves and piping. During startup operation, both trains can be put into service while normal operation requires only one (1) train to be in operation producing 100 percent flow while the remaining train is in standby or regeneration mode. The demineral Izer system portion of the Make-Up Water Treatment Syster sploys two (2) parallel domineral izer treins each consisting of one (1) cation, one (1) anion and one (1) mixed bed Ion exchanger, piping, valves, and control s. A resin trap is installed in thu outlet of each domineralizer train to prevent resin f rom entering the Demineral Ized Water System on f ailure of a domineral Izer underdrain screen. The regeneration system consists of bulk acid and caustic storage tanks with respectivt. Injection pumps, a hot water tank w ith tmperature control and means of dil ution f or chemical s. Means of transferring regenerant wastes, rinses and backwash water frm activated carbon f ilters and domineral izers to the Waste Water System is provided.

9.9.7.3 Safety Evaluation The Make-Up Water Treatment System is a Seismic Category lli and non-safety cl ass system.

9.9.7.4 Tests and insoections Vessel s are ASE Section V ill except for the Hot Water Tank which is designed to Section IV. They are tested in accordance wIth Code requirements.

9.9.7.5 Instrumentation Aoolications Mixed bed demineral izer ef fl uent is checked f or pH, sil ica, and sodium with Indicator / recorders tied to al arms. Conductivity of dil ute acid and caustic is measured wIth a Indicator / recorder coupled to high/ low alarms. Anton and mixed bed demineral izer ef fl uent is measured for conductivity on an indicator /

recorder w ith high al arms. Flow to the cation and mixed bed demineralizer is measured util Iz ing an Indicator / recorder / total izer wIth total fIow al arm.

Carbon f il ters and resin trap strainers are f itted with high pressure dif f erenti al al arms. Tmperature is monitored on the hot water tank and the hot water mixing valve outl et w ith a high al arm.

O Amend. 73 Nov. 1982 g

(m) 9.9.8 Demineral17ed Water System 9.9.8.1 Deslan Basis The non-safety related Demineralized Water System receives demineralized water f rom the Make-Up Water Treatment System and pumps the water as required to the various systems and services.

All piping and components shall be designed, f abricated, inspected and erected in accordance w ith ANSI B31.1. Power Piping.

9.9.8.2 System Descriotion The Domineral Ized Water System consists of a Demineral Ized Water Storage Tank, three Demineral ized Water Pumps (which incl ude a Domineral ized Water Jockey Pump and two (2) 100 percent Demineralized Water Transfer Pumps, one acting as a spare), distribution piping, valves, instrumentation and control s, the domineral ized water provided by the Make-Up Water Treatment System is pumped to the Domineralized Water Storage Tank by the clearwell pumps. The Domineral ized Water Jockey Pump and Domineral ized Water Transf er Pumps take suction from the Domineralized Water Storage Tank and deliver water to the di stribution header. The Demineral ized Water Jockey Pump ' operates continuously to maintain pressure in the system and to supply water demand rates up to approximately 40 GPM. One (1) of the two (2) 100 percent

~

Danineralized Water Transfer Pumps is on standby and starts and stops automatically to maintain system pressure under varying system demands. A

(_,}/ recirculation orifice is installed in a bypass line running from each pump discharge back to the Domineral ized Water Storage Tank through a common h ea der. The Domineral Ized Water System provides water for the systems and serv ices l isted in Tabl e 9.9-8.

9.9.8.3 Safetv Evaluation The Domineral Ized Water System is a Seismic Category 111 and non-safety class sy stem.

9.9.8.4 Tests and insoections Dominer al ized Water Transfer Pumps and Jockey Punp are tested at the manuf acturer's f acil Ity and again prior to r >rmal pl ant operation.

9.9.8.5 _ Instrumentation Aoollcations One (1) of the Domineralized Water Transfer Pumps is on standby and will start and stop automatically under the control of a low pressure svitch. Each pump i s prov ided w ith a ntorage tank l ow-l ow l evel Interlock and al arm to stop the pumps. The l evel in the storage tank is controlled automatically by a level control val ve.

()

Amend. 73 n v. 2 9.9-12c i

1

T/8LE 9.9-7 BUILDING AND SERVICES SUPPLlED BY POT /BLE WATER SYSTEM

a. Turbine Generator Bullding

b. Steam Generator Bullding

c. Maintenance Shop and Warehouse Bullding

d. Control Bull' ding

e. Plant Service Bullding i

f. Reactor Service Building (Radwaste Area)  :

g. Gate House

h. Circulating Water Pump seal water during pump start-up I. Hypochlorite Generating Plant J. Fire Protection Storage Tanks

~

k. Make-up Water Treatment System l

l l

l i

O Amend. 73 Nov. 1982 9.9-18a

TELE 9.9-8 SYSTEMS AND SERV ICES SUPPLIED BY THE DEMINERAllZED WATER SYSTEM

a. Backwash and regeneration for the condensate pol Ishers in the Feedwater and Condensate System.

b. Domineral Ized water for Initial fill and raake-up water for Diesel-Generator water Jacket coolers.

c. Domineral ized water for initial fill and make-up water for the Secondary Services Closed Cool ing Water System with provision for chemical

! conditioning.

l l d. Domineral Ized Water for Initial fill and make-up water for the Hot Water Heating System l

e. Domineral ized Water for initial fil l and make-up f or Normal and Emergency Chil led Water Systems.

f. Domineral ized Water f or initial fill and make-up water supply for the stator cool Ing system,

g. De:alneral Ized service water f or decontam; nation f acil ities in the Radwaste area of the RSB and in the BOP Regulated shop complex fee equipment and personnel, and rinse water to the Intermediate Componen'. Cleaning System.

h. Backwash and regeneration of the domineralIzer trains of the Make-Up Wpter Treatment System and source of make-up water for the Feedwater and Condensate System.

I. Equipment washdown and decontamination hose connections in Radwaste area.

l J. Chemical dil ution water for chemical feed units for miscellaneous closed water cool ing systems.

l

k. Personnel decontamination f acilities in the Plant Service Building and Combined Laboratory services.

I. Emergency Cool Ing Tower f Il I and make-up.

l O

Amend. 73 Nov. 1982 9.9-18b

i 9.13.2 Sodlum Fire Protection System (SFPS)

The'SFPS provides the means of detecting, locating, alarming, containing and extinguishing sodium and/or NaK f f res. The system consists of fire detection and alarm instrumentation, aerosol release limiting Instrumentation, a catch pan system, portable f ire extinguishers, and personnel protective clothing and equi pment.

The steel catch pans, insul ation between the catch pan and the structural concrete, and steel fire suppression decks, herein referred to as the catch pan system, and the aerosol release Iimiting instrumentation comprise the Engineered Safety Features of the SFPS. This equipment is Installed in the air-filled cells of the plant in which sodium-NaK piping and other equipment containing sodium-NaK are located. In the event of a liquid metal spill, the catch pan system functions to: (1) Iimit burning and the production and spread of combustion product aerosols, and (2) limit the temperature imposed on the structural concrete. The aerosol release I imiting instrumentation provides an initiating signal to interf acing systems for actions to limit the release of aerosols to the outside atmosphere. Other equipment that must perf orm a safety function to lImit the release of aerosols in the event of a design basis leak in the intermediate Heat Transport sodium piping include:

1) fire dampers in the HVAC outside openings in the SGB loop cells;

2) ciosure devices in the SGB vent stack

3) smoke detectors at the PACC's airside Inlets.

O The operation of this equipment is described in Section 6.2.7. Cel i s containing primary and EVST sodium systems and piping are equipped with steel lIners and inert (nitrogen) atmospheres. These features are described in Sections 3, 8 and 9.5.

9.13.2.1 Desian Bases The catch pan system which is an Engineered Safety Feature is designed to mitigate the consequences of a design basis sodium or NaK spilI in an airf illed cell. The design basis spill is based on leakage from a sharp edged circular orifice whose area is equal to one quarter of the pipe wall thickness multiplled by the pipe inside diameter. These spilIs are ciassifled as Extremely Unilkely Events and are analyzed as f aulted events.

The f unctional design and evaluation of the catch pan system is based on the sodium-NaK l eak rates and spil l vol umes l isted in Tabl e 9.13-9. The relevant Engineered Safety Features for these accidents is the catch pan system. In all cases, with the exception of cell 211 A (which contains the ex-contcInment storage tanks), the spilI volumes are predicated on the assumption that no action is taken to terminate the leak. In cell 211 A, action is required to l imit the spil l vol ume to 3400 gal . This is ensured through the operating procedures governing the transfer of sodium to the ex-containment storage tanks.

/' The specific f unctional requirements imposed on the catch pan system as an Engineered Saf ety Feature are:

Amend. 73 9.13-13 Nov. 1982

1. The system shall be designed to contain the entire spillable volume f rcm a full-flow piping leak in a leak-tight manner to preclude chemical reaction between the i Iquid pool and the structural concrete.

2. The system shall be designed to lImit the temperature imposed on the structural concrete, in the event of a design basis leak, to a level suf fIclent to ensure the structural Integrity of the bullding.

The f unctional requirement imposed on the aeroso! release l Imiting instrumentation, as an Engineered Saf ety Feature, is to provide an initiating signal to the HVAC system within 10 seconds of the timeyhe combustion product aerosol concentration in the SGB exhaust air reaches 10 gm/cc.

The fire detection and alarm instrumentation, the portable fire extinguishers, and the perscnnel protective clothing and equipment portions of the SFPS are designed to the requirements of appl Icable National Fire Protection Association (NFPA) codes. These features of the SFPS are not safety related.

9.13.2.2 System Descriotion 9.13.2.2.1 Catch Pan System Catch pan system features are provided in alI air-fIlIed celIs of the Steam Generator Building (SGB) and the Reactor Service Building (RSB) which contain nonradioactive i Iquid metal systems and piping. These cells contain the nonradioactive sodium piping and components of the intermediate Heat Transport System, the Auxil lary Liquid Metal System, the impurity Monitoring and Analysis System, and portions of the non-radioactive NaK piping and components of the Auxil iary Liquid Metal System.

Catch pan system features are al so provided in SGB Cells 211 and 211 A. These cells contain the ex-containment primary sodium storage tanks and associated piping of the Auxil iary Liquid Metal System, and are inerted when radioactive sodium is inf requently present in the storage tanks.

The catch pan system consists of four basic features: (1) catch pans, which contain spilled i Iquid metal; (2) insul ation between the catch pan pl ate and the surrounding structural concrete; (3) fire suppression decks that cover the catch pan open area; and (4) interconnections between adjacent catch pan cells that allow drainage of liquid metal from one cell to another. A steel grating above the fire suppression deck will be provided, where required, to serve as a walkway and to provide equipment and personnel access. The steel grating also acts to prevent damage to the fire suppression deck. A typical catch pan-fire suppression deck arrangement is shown in Figure 9.13-2. The plant arrangement of catch pan system features is summarized in Table 9.13-10.

Catch Pans The catch pan consists of a carbon steel plate assembly which covers the enttre fIoor surf ace of the celi and extends vertically up the walI to a l minimum height of one foot above the maximum sodium level in the catch pan to prevent spilied i Iquid metal from fIowing over the edge of the plate into the area between the plate and the walI.

Amend. 73 Nov. 1982 9.13-14

A continuous lIp plate is provided at the top of the catch pan side wall to prevent sodium or NaK f rom running down the structural concrete cell walls into the region behind the catch pan plate sidewalls. The catch pan is free floating and is supported above the concrete floor of the cell by a continuous Iayer of Insulating material and by steel beams. In the event of a l Iquid metal spilI, the catch pan contains the iIquid metal and prevents contact between the 1 Iquid metal and the concrete structure. Open catch pans (without fire suppression decks) are used in those cells where the postulated spill voi umes are smalI and open pool burning does not result in concrete temperatures that degrade structural concrete or release unacceptably high quantitles of aerosol s that af fect safety-related equipment in adjacent loops.

Open catch pans cover,the concrete fIoor surfaces of SGB CelIs 244, 245, and 246, to prevent sodium concrete reactions during a spilI event. The steel plates of the open catch pans are sloped toward the existing floor openings such that sodium will not be contained at this elevation. Sodium leaked onto the plates wilI spill into CelIs 224, 225, and 226, respectively, and drain f ran there into Cel I s 207, 208, and 209, respectively, where the sodium is contained wIthin a catch pan equipped wIth a fire suppression deck, insulation insulation is provided between the catch pan plate and the concrete floor, and behind the wall sections of the catch pans.

p The insulation behind the wall sections of the catch pans is provided in the form of aluminum silicate, blanket type, and is attached to the structural V

concrete wal I s. An air gap between the Insulation and the vertical catch pan plate sidewall provides additional insul ation, allows for rel ative movement l between the insulation and the catch pan plate, and vents hot gases f rom behind the catch pan plate to the cell atmosphere to prevent pressure buildup behInd the pl ate.

Insulation in the floor consists of M 0gaggregate and extends to the bottom of the catch pan, in the event of a lIquid metal spil1, the insulation and air gap act as a thermal barrier between the hot iIquid metal pool and the bullding structural Concrete.

Fire Sucoression Decks in cells where l iquid metal spills are contained and where open pool burning may pose a chtllenge to the structural integrity of the building and/or to saf ety-rel atet equipment, the catch pans are provided with fire suppression decks. The deck is supported above the catch pan plate surf ace by a structural (steel) framework supported at the edges of the cell by embedments in the structural coner The stub col umn base ates pl,ete, areand in the Interior anchored directlyofinto the the cell structural by stub colconcrete umns.

f l oor. Around +he stub col umns, as wel l as around penetrations through the s concrete floor, a vertical plate is provided to form an enclosure to allow for the catch-pan f ree-expansion and to prevent leakage of spilled sodium from the

' catch pan. The fire suppression deck is connected to the support f raning, with all edges sealed to form an essentially airleak-tight cover.

Amend. 73 Nov. M82 9.13-15

Carbon steel drain pipes (downcomers) are welded to the deck and extend l downward to a point 1/2 inch above the catch pan plate. The pipes are un I f ormi y spaced to f orm a un i f orm array over the cei I fI oor crea. Vent pipes arn welded to the fire suppression deck and extend slightly below and above the deck. The vent pipes are provided to vent hot gases f rom the region below the deck to the celi atmosphere to prevent pressure bulldup underneath the deck.

In the event of a linuid metal spill, the liquid metal flows f rom the surf ace of the fire suppression deck through the drain pipes into the catch pan. As the l Iquid metal drains into the catch pans, the drain pipes become partially fillod, and the ef fective burning surf ace of the resulting IIquid metal pool is I imited to the cross-sectional area of the vent and drain pipes. Af ter the sodium has drained into the catch pan, burning is terminated when the pipes become pl ugged with combustion products and air is prevented from reaching the i Iquid metal surface.

Cell Interconnections in certain cells, where the postulated spill volumes are large compared to the floor area of the cell such that consideration of cell penetrations and buil ding structural loading make it impractical to contain the entire vol ume, open catch pans equipped with drains are provided. The catch pan plates are pitched toward the drains. The minimum slope toward drains is 1/8"-1/4"/f t, except for cells 244, 245 and 246 where the slope is approximately 1/10"/f t.

The drains are sized to accommodate the maximum spill rates f rom postulated design basis accidents.

The drains are in the form of carbon steel pipes passing through the structural concrete of the cel l . Horizontal pipes Interconnect cells on the same I evel; vertical pipes interconnect cel Is on dif ferent Ievel s.

In the event of a iIquid metal spilI, the catch pan prevents contact between the i Iquid metal pool and the structural concrete. The 1 Iquid metal is drained into a celi which has the capabil Ity for fire suppression (catch pan w ith fire suppression deck). This concept has been extended to incl ude dralning iarge upper ievel celIs into Iower elevation celIs, I.e., CelIs 224, 225, and 226 draining into cells 207,3208and2g9. Net vol umg of catch pans in cel I s 207, 208 and 209 are 5180 f t , 4207 f t , and 5763 ft , respectively.

Svstem ConffouratIon l The catch pan fire suppression deck arrangement for Loops 1 and 3 of the l Intermediate Heat Transport System, in the Intermediate Bay (IB) of the SGB, is shown schematical ly in Figure 9.13-3.

The catch pan f Ire suppression deck arrangement for Loops 1 and 3 in the steam generator bay of the SGB is shown schematically in Figure 9.13-4. The catch pan f Ire suppression deck arrangement for Loop 2 in the 18 and the SG3 is sImilar to Ioops 1 and 3, wIth appropriate adjustments for the modifled Ioop I ayout.

O Amend. 73 9.13-16 Nov. 1982

The catch pan arrangement for SGB Cells 211 A and 211 is shown schematically in Figure 9.13-5.

(V]

Cells 352A, 353A, and 332 in the RSB are equipped with open catch pans.

Cells 354, 355, and 350 of the RSB are equipped with catch pans with fire suppression decks.

Aerosol Release Limitina Instrumentation Sets of saf ety-related aerosol detectors are Installed in the HVAC exhausts at each Steam Generator Cell (Cells 244, 245, 246). Each detector set consists of three detectors which are provided with power from the three 1E battery power sources. These detectors trip when the so aerosol concentration in the exhaust reaches 10"pium combustion gm/cc. product signal An initiating is generated when any two of the three detectors in a set trip.

9.13.2.2.2 Fire Extinaulshers Portable sodium carbonate (NaX) fire extinguishers, hand-held and wheeled-cart types, are provided and stored in locations convenient to spaces in which there is sodium and NaK equipment in the RSB, RG, SGB, and the sodium and NaK receiving station. The extinguishers and their storage locations are compatible with building space allocations and passageways. Distribution and Iabel ing of these commercially avail abt e extinguishers are in accordance with NFPA 10.

9.13.2.2.3 Instrumentation and Control The SFPS instrumentation is designed to:

1. Detect the presence of and location of incipient and existing sodium and NaK f ires, and provide th is inf ormation to the pl ant operator.

2. Detect inoperative detectors, and provide Inoperative detector inf ormation to the pl ant operator.

Sodium Fire Detection The fire detection and alarm channel arrangements are as shown in Figure 9.13-6.

l Fire detectors are used to detect the presence of an incipient or f ully developed sodium fire. The detectors are permanently Installed at the l ocations I isted in Tabl e 9.13-11.

Receptacles, connected to the appropriate area panel s, are provided both inside and outside of the cells l Isted in Table 9.13-10 for which the normal atmosphere is inert (nitrogen or argon). The ex-cell receptacle is located in proximity to the cell exhaust connection to the Heating, Ventilating and Air Conditioning System deinerting apparatus. This deinerting apparatus includes a smoke detector in the exhaust iIne. This detects is temporarily connected to the receptacle during the deinerting procedure.

9.13-17 Amend. 64 Jan. G82

l 1

Detectors are pl ugged into the in-celi receptaci es while the celIs are air filled.

The output signal from each detector is transmitted to an area panel. Each panel is installed in a location (continguous to a sodium fire protection zone) which is, or can be, Isol ated f rom the smokt, and heat of a sodium fire w ith in the zone. The sodium fire protection zones are sections of the Nuclear Island where sodium-containing equipment is installed or may be transported.

These zones are I isted in Tabl e 9.13-12. Signal s received at an area panel are group retransmitted from each area panel to the sodium fire protection zone indicating panel in the control rocrn.

Two types of detectors are used: product of combustion and optical. Smoke detectors are actuated by the particul ate products of combustion or sodium or NaK. This actuation may be a resul t of: 1) reduced Ion diode current in an photoelectric type detector; or 3) reduced light received by the detecior in a photoel ectric emitter-detector type detector. Optical detectors are actuated by the infrared energy omitted by a fire.

Inocerative Detector and Area Panel Locations and Test The method of detecting a f ail ed detector is to test its performance periodically as recommended in NFPA Code 72E. Loss of power to an area panel (with subsequent loss of operation of the associated detectors) is signaled to the sodium fire protection zone Indicating panel in the control room. This is accompl ished by supervising the power circuits.

Miscellaneous Eauloment Personnel protection equipment, protective cl oth ing, set f-contained breath ing apparatus, and other equipment essential to personnel safety in event of a sodium or NaK fire, are stored et locations selected on the basis of ready avail abil Ity for rescue or repair operations.

9.13.2.3 Deslan Evaluation Those cells of the plant that normally contain radioactive sodium piping and components are equipped with steel Iiners and inert atmospheres (2% oxygen).

The I imited amount of oxygen in the cel I suppresses both spray and pool burning. The steel l iner contains the spilled I Iquid to prevent contact with the structural concrete and al so acts to l Imit the spread of combustion products. The l iner is thermal ly Insul ated f rom the cell structure to l imit the temperatures imposed on the concrete. The l Iners and the celi structure are designed to withstand the imposed thermal and pressure Ioadings associated w ith a design basis l eak, in the event of a sodium /NaK leak, the liquid pool is alicwed to cool to a temperature at which burning is unl ikely even if exposed to an ambient air atmosphere. This technique is universal ly anpl oyed in LMFBR loop-type designs and is considered the most ef fective fire suppression technique avail abl e.

O 9.13-17 a Amend. 64 Jan. 1982

Those cells containing nonradioactive sodium /NaK piping and components are

{y g normal ly air-f Il l ed. In these celIs, the basic fire suppression feature is a covered catch pan located in the floor of the cell. The use of a covered catch pan to suppress pool burning is a welI established concept and has been developed to varying degrees in several countries. Covered catch pans have been tested under simulated sodium spilI conditions at HEDL and at the Karl sruhe Nuclear Research Center and were shown to be ef fective in suppressing pool burning. In both cases, a perf orated cover pl ate was used.

In the HEDL test, there were two holes per square foot of surf ace corresponding to 1% open area and in the Karlsruhe tests there was one hole per square foot of surf ace corresponding to 0.5% open area, in the HEDL test, a nitrogen flood was added below the cover plate. In both cases, the sodium burning rate was reduced to approximately 1/10 that of an open pool. The rate at which heat was generated due to burning was lecs than the rate of heat loss f rcm the pool, and the sodium temperature which was initially approximately 1000 F decreased rapidly (<10 hours) to the freezing point, and burning was term inated. Calculations for a typical SGB cell of CRBRP indicate that with a pool burning rate of approximately 1/250 of that of an open pool, approximately 1/20 of that of the tests, the sodium pool temperature is maintained welI above the freezing temperature such that burning may continue until either the sodium is completely reacted or until it is terminated by means other than f reez ing. This prediction is characteristreally dif ferent f rom the experimental observations where the sodium cooled rapidly to the f reez ing point, and burning was terminated. The dif ference is attributed to the f act that in CRBRP, the heat losseg f rom the pool are minimal. I n CRB RP, the pool covers an area up to 5,000 f , whereas the Iargest pool area of the experiments was approximately 36 f2t . The l arge heat transfer area-to-vol ume h

d ratio of the experiments gives rise to enhanced lateral heat transfer ef fects, Also, the Iarge ratio of colI walI area to burning area increases radiation heat l osses. Thuc, in the experiments, heat losses f rom the pool were larger than the heat generated by burning, and the sodium temperature decreased to the f reez ing point. In CRBRP, the heat losses f rom the pool are so small that a much lower burning rate is suf ficient to maintain the pool temperature above freezing.

The catch pan f Ire suppression deck design for CRBRP incorporates features to reduce the burning rate to approximately 1/250 of that of an open pool and to

~

terminate burning with the sodium pool at an elevated temperature. The fire suppression deck is equipped with drainpipes (in place of holes) that extend downward to a point just above the catch pan plate. With these p! Des and the catch pan f ll l ed w Ith sodium, the of fectIve pool burning surf ace is reduced to the surf ace of the sodium exposed in the pipes ( 1/250 of the floor surf ace area). Burning is terminated whe,n the pipes become plugged with combustion products (approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). These are the essential design features of the CRBRP catch pan f Ire suppression deck design and form the basis f or testing its offectiveness.

G

)

9.13-17b Amend. 64 Jan. 1982

The compl ete spectrum of design basis sodium-NaK spil l s (Tabl e 9.13-9) for air filled cells has been analyzed to verify that: (1) the structural integrity of the buil ding is maintained; (2) the perf ormance of the pl ant saf ety-rel ated equipment is unimpaired; and (3) the site boundary combustion product aerosol conccatration is acceptably low.

In the event of a sodium spill in SGB cells, the combination of catch pans and fire suppression decks contains the spilled sodium, prevents a sel f-sustaining f ire (oxygen depl etion), and allows the sodium to cool to its f reez ing temperature where reignition wIlI not occur even when exposed to an ambient air atmosphere.

In the RSB, catch pans wilI contain spilled NaK and prevent its contact wIth concrete, in cells equipped with open catch pans, the NaK is allowed to burn until the fire is either sei f-extinguished or the entire mass of NaK is consumed. in colIs equipped wIth catch pans with fire suppression decks, the NaK w ill cool to a temperature at which reignition is unl ikely even if exposed to an ambient air atmosphere.

Analysis of the design basis sodium - NaK leaks shows that peak concrete temperatures remain below 350 F (except for two cases) and the long duration (greater than 24 hrs.) concrete temperatures are below 200 F. For two cases, the Intermodlate Heat Transfer Shield Cells Loops 1 and 3, the vall temperature peaks at approximately 600 F and is reduced below 200 F within approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />. These temperatures are acceptable for buil ding structural design. A detailed examination of sodium-NaK spills and their consequences is given in Chapter 15.

The pet aerosol release rate, to the outside atmosphere is 2 kg/sec. In the case of a design basis leak in the Intermediate heat transport loop, the total aerosol release, to the atmosphere must be maintained below 630 lbm in order to assure the perf ormance of pl ant saf ety-rel ated equipment. This is accompl ished by: 1) closing the HVAC outside openings in the SGB loop cells;

2) venting the hot gases through a controlled area vent stack; and 3) closing the vent stack af ter the initial pressure put se is over. These actions are initiated by the signal f rcrn the aerosol release I imiting Instrumentation. A detailed examination of sodium or NaK spills and d. heir consequences is given l In Sections 6.2.7 and 15.6.1.5.

NaX, used in f Iro extinguishers, has been tested by Underwriters Laboratory l and is UL l isted f or use on potassium, sodium, and sodium potassium f ires f or

! temperatures up to 1400 F. The UL listing includes NaX in 50-lb palis and NaX in hand hel d and wheel-type f Iro extinguishers of 30,150, and 350-lb ca pac.i t i es. NaX has been widely accepted in the U.S. as the pref erred f Iro extinguishing agent for sodium /NaK f f res. The use of NaX does not have any adverse ef f ects on material s util ized in the pl ant heat transport systems.

O Amend. 73 9.13-17c Nov. 1982

In assessing the suitability of NaX fire extinguishers for use in CRBRP, it (3 should be emphasized that these fire extinguishers are intended primarily for V use as an auxil iary precautionary measure during maintenance operations, during sodium /NaK loading and unloading operations, and during sodium /NaK fire recovery operations. Sodium /NaK f ire protection, in the sense of plant safety, is provided by a system of steel catch pans with fire suppression decks in air-filled cells and steel liners and low oxygen concentrations in inerted cel I s. Thus, it is expected that NaK f Ire extingulshers w Il I not pl ay a significant role in f ire extinguishing.

9.13.2.3.1 Catch Pan Structural Analysis The open catch pan system and the catch pan with fire suppression deck system wIth its support f raming are designed to contain a large sodium /NaK spilI while maintaining structural integri ty at accident temperatures and pressures resulting f rom a sodium fire. The catch pan design is based on a " floating" concept which alicws free expansion to minimize the thermal expansion ef fects due to the accident conditions, in some local areas where restraint to f ree expansion exists because of anchorsge of equipment, the thermal stresses or strains w il I be calcui ated and compared against specif led al Iowabl e val ues.

Load combination and stress-strain allowables for the dif ferent load l combinations are given in Appendix C of Section 3.8. Where required, an elastic-ps astic finite element analysis using the computer program ANSYS will be performed to verify that the specif led stress-strain I imit under the accident conditicns are not exceeded.

7 9.13.2.4 Insnecticn and Testino Reautrements (V The SFPS is designed to permit inspection of all instrumentation, and portabl e f ire extingui shers, in accordance w ith NFPA codes. Periodic inspection of the catch pans and f Ire-supprossion docks wIl I be perf ccmed, as appropriate.

O 9.13-17d Amend. 64 Jan. 1982 ,

l

s sb '

, s N' ,

i' p 9.15 EOUIPMENT AND FLOOR DRAINAGE SYSTEM -

, 1 9.15.1 Deslan Baggs '

The plant Equipment and Floor Drainage System,(EFDS) is designed to collect i the drainage f rom al l pl ant equipment such as pumps,, tanks, coolers, etc., as well as the floor drainage.  ;

Under normal operating conditions the floor drains in the plant serve for house keeping purposes. However, the EFDS is sized to accomodate the maximum postulated flooding event such as a pipe rupture, tank rupture, or sprinkler discharge and I imits water accumulation on the floor to no more than 31/2 inches. All safety related equipment is mounted on pads at least 4 Inches high which is above the maximum flood level. In locations where a postulated complete blockage of the drainage system can cause an accumulation of water above 4 inches, the safety rolated equipment is mounted on 6 or 8 inch high concrete pads as required; the height was determined in the EFDS flooding study to assure that safety rolated equipment cannot be innundated. In some cases, al so determined by the flooding study, the gaps under doors have been increased to provide an al ternative passive means f or drainage' to egress f rom a colI wIth saf ety rel ated equipment. The fIooding study has a*so determined . -

that backfIow between connected celi s w Il I not occur. The deainage system.. t siz ing has been based on the largest postulated drainage load in each cell; ,

such as fire protection system discharge flows, tank ruptures and pipe breaks. '

9.15.2 System DescrIntion O s , i O

Separate EFDS sumps are provided for radioactiva, potentially, radioactive %d non-radioactive areas of the plant. Each sump contains two vertical sump pumps with one pump serving as a full capacity spare. 3 Equipment and floor drains in the BOP buildings, i.e., where there is no ,

potential of radioactivity, are collected and discharged to the waste weteF '

disposal system. Drainage from the f Ire protection system in the Centrol Bull ding, Electrical Equipment Bull ding and Diesel' Generator Bull didg, Jwhich +

have no potential radioactivity are also directed to the waste water treatment sy stem.

'. J

\

Potentially radioactive drainage will emanate from the Reactor Containment Buil ding (RG), the Reactor Service Buil ding (RSB) and Radwaste Area inia) in -

the RSB. The drainage from the respective sump in the RG and RSB will be pumped to the main collection sump in the RWA. This sump has the storage' capacity for the single largest design drainage load. ~he potential ly radioactive waste in the RWA where radioactivit'lc ts expected area drained directly to the radwaste sump. The main collection sump is equipped with sa radiation monitor and diversion valves so that following an accident, i:

radioactive drainage is pumped to the l Iquid radwaste sump f n the RWA for O

O ,

~Anend. 73 9.15-1 Nov. 1982 L

processing. Non-radioactive drainage is pumped to the equalization ponds of the Wastewater Treatment System. A power f ail ure to the radiatic monitor or diversion valves will cause recirculation back to the sump to prevent radioactive drainage f rom entering the non-radioactive wastewater treatment sy stem.

Treated water and other process water treatment wastes which do not have the potential to be radioactively conieminated, are routed to seperate sumps foi transport to the waste water treatmwt system.

Where there is a potential for oil spills, the drainage is routed to the oil separation system prior to discharge into the waste water disposal system.

Oil spills are not allowed to drain in areas that con'ain radioactively contam inated equipment or fl uids. In this case, the oil spill Is contaminated with curbs and dikos and removed manually. Oil routed to the oil separation system is collected in a waste oil tank and renoved f rom the site for subsequent d(sposal .

l9.15.3 Safetv Evaluation The pl ant equipment and floor dra'nage system is designed so that it is not reasonably possible f or any radiorctive drainago in these systems to be discharged out of the pl ant withoto undergoing the required treatn;ent or processing.

Eval uations of radiological considerations f or normal operation and postulated spills and accidents are presented in Sections 11.2.5 and 15.0 respectively.

The plant Equipment and Floor Drainage Systems is not saf ety related except f or the piping and val ves required f or containment isol ation (Section 6.2.4).

EFDS piping within areas containing saf ety related equipment is supported with Soismic Category i supports.

There are no drains in colIs where sodium piping or equipment containing sodium is located, accordingly sodium leaks cannot enter the equipment and fI oor drainago system.

A water pipe break or f ire protection system drainage load cannot enter cells or compartments containing sodium from drain system backflow because these cel l s do not have 'any drains. The CFBRP design criteria requires that three passive barriers (or two passive and one active barrier) exist between all sodlm and water boundarios. Accordingly, leak detectors located in the drainage system are not required.

Saf oty rel ated systemscontaining water have instrumentation to detect leakage.

9.15.4 Tests and insoections EFDS pipas embedded in concrete are leak tested prior to the pouring of coneretc.' Al i EFDS p1 ping i s tested f or i eaks af ter Instal Iation. AlI leaking p1pos or joints are ropalrod before the concrote is pl acod. The piping wif I be cleaned out to insure that construction debris will not cause a blockage or reduction in the flow. All pumps are tested to ensure that their l

Amend. 73 9.15-2 Nov. 1982

performances meet the required design flows and pressures. A check source will be provided w ith the radiation monitor to ensure its operabil ity.

Periodic sampling of drainage from the WA main collection sump will be perf ormed and analyzed at the pl ant l aboratory, 9.15.5 Instrumentation Aoolication Each sump is provided with automatic control s to start and stop the operation of the sump pumps. A switch is provided to alternate the lead with the lag p ump. In case the lead pump f alls to start, a high-high level switch autcnnatically starts the standby lag pump. A high-high level switch is provided in each sump and alarms in the control room to indicate potential sump overflow. A radiation monitor is provided in the potentially radioactive WA EFDS main collection sump. The WA main collection sump pump will recirculate the drainage flow back to the sump until the radiation monitor has suf fIclent time to determine whether the drainage is radioactive.

O l

l l

O '

Amend. 73 Nov. 1982 9.15-3

(

l l Taner 9.16-2 SYSTEM PARAKTERS Subsystee Operating Operating Des!gn l Cooling Cepecify Gas Flos Presswo

! Dealgaation TIlle Gaa BTtu)5t S&M PSIG 6

PA fMTS Loop fl N 1.17 x 10 16,300 35 2

6 fB fHTS Loop #2 N 1.09 x 10 15,200 35 2

6 PC PHTS Loop #3 N 1.17 x 10 16,400 35 2

0 Ot Control Rod Drive Mechenlam N 0.28 x 10 3,200 150 2

6 M4 Sodlue Nakeup P e p and Pipeways N 1.24 x 10 18,300 15 2

6 j le Sodlue Makeup Pwp and Vessels N 0.60 x 10 9,400 15 2

6 l 7 CT Cold Trep, Mak Cell N 2 0.33 x 10 5,010 15 e 6 RC Reactor Cavity N 2 1.49 x 10 JO,950 35 6 8,730 EA EVS Loop fl N 2 0.52 x 10 15 0 12,480 j EB EVS Loop #2 N 2 0.94 x 10 15 6

EC EVS Loop #3 Ny 0.19 x 10 2,610 15 6 5,000 ET Ex-Vessel Storage Tank Cavity N 2 0.36 x 10 15 l

FH Fuel Handllag Col 1 Ar 0.55 x 106 8,000 15 5$

5e i F N

O TABLE 9.16-3 LIST OF SAFETY-RELATED VALVES REQUIRING COMPRESSED AIR TO PERFORM THEIR SAFETY-RELATED FUNCTION Valve No. Figure no. Subsystem Normal Position Fail Position 28MANV001A 9.16-4 MA Open Closed 28MANV001B 9.16-4 MA Open Closed 28MBNV001A 9.16-4 MB Open Closed 28MBNV001B 9.16-4 MB Open Closed 28EANV001A 9.16-6 EA Open Closed 28EANV001B 9.16-6 EA Open Closed 28EBNV001A 9.16-6 EB Open Closed 59 28EBNV001B 9.16-6 EB Open Closed i

Amend. 59 Dec. 1980 9.16-9a

9.17 Sewaae Disoosa System 9.17.1 Design Basis The non-safety related Sewage Disposal System is required for sewage collection and treatment during plant construction and operating periods providing a level of treatment that satisfles ef fluent guldelines and performance standards defined in the National Pollutant Discharge Elimination sy stem ( NPDES) . Permit is issued for the CRBRP by the Environmental Protection Agency (EPA).

9.17.2 Svstem DeserIotIon The construction period (temporary) system and the operating period (permanent) system provide secondary treatment wIth chlorination of the effluent. Both the construction and permanent sewage treatment plants provide biological treatment by the extended aeration modification of the activated sl udge process. Raw sewage first enters a surge tank which stores peak loads and provides downstream equipment with a constant flow. Each pl ant aerates the activated sl udge-sewage mixture in the aeration tank and settles the aerated mixture in the clarif ier. A portion of the settled activated sludge in the clarifier is continuously returned to the aeration tank by an air lif t sy stem. Excess activated sludge from the settling compartment is accumulated in the waste sl udge hol ding tank. The cuerflow from the holding tank flows to the inlet of the aeration compartment. The ef fluent from the clarifier is continuously chlorinated by a hypochlorinating system and is post-aerated to p maintain a desired dissolved oxygen level in the ef fluent to be discharged.

9.17.3 Safetv Evaluation The Sewage Disposal System is a Seismic Category lll and non-saf ety cl ass sy stem.

9.17.4 Tests and Insoections Af ter each of the sewage treatment plants is connected for operation, acceptance tests are conducted in the f leid to determine the abil ity of the equipment to meet design and guaranteed conditions.

9.17.5 Instrumentation Acolications l A cal Ibrated V notch weir is provided to measure flow through the treatment pl ant. It is provided with float and cable flow Indicator / recorder. A pressure switch Is installed in the discharge piping to start the spare blower when the discharge pressure of the operating blower f alls below the normal operating pressure of the air dif f user system.

l i

l Amend. 73 Nov. 1982

fs 12A.3.1.4 Health Physicists ALARA Reviews

. ; )

The other level of review is perf ormed by health physicists f rom TVA and Conmonweal th Edi son. There are three health physicists involved in these rev iews, two f rom the TV A ALARA committee and one f rom Commonwealth Edison.

The two TVA health physicists on the CRBRP ALARA committee satisfy the TVA commitments in PS AR Section 12A.3.2. The health physicist's ALARA review meetings are conducted tw ice a year. The health physicists review system /

component design, maintenance outl ine procedures, and the radiation exposure data and provide recommendations to f urther reduce radiation exposure based on their AL ARA experience at operating nucl ear power pl ants. The specific personnel involved in these reviews by position title, incl uding their health physics training and experience, are l Isted below:

Title Trainina/Exoerience Heal th Physics Supervisor, 23 years experience in Power Reactor Technical Services, Nucl ear, Heal th Physics Commonwealth Edison Company Chemical Engineer, Radiation 7 years experience in power reactor Section, Emergency Preparedness chemistry programs, and Protecticn Branch, Division 1 1/2 years in radiation protection of Nuclear Power, of f ice of Power, Tennessee Vai ley Authority Heal th Physicist 23 years of experience in appiled and Os Radiol ogical Heal th Staf f, technical aspects of health physics Of f ices of Management Services Tennessee Valley Authority 12A. 3. 2 CRBRP Ooerations Staco ALARA Program The purpose of TVA pol Icles and procedures is to guide the of ficial actions expected of TV A empl oyees. A pol icy or a required procedure will not serve that purpose unless it is known to all those it af fects and is understood, interpreted, and appl led consistently. Continuing guides of this nature in TVA are published and distributed in such a way as to be available to all empl oyees concerned. They are known as " administrative releases".

The TV A Administrative Release System is composed of Organization Bulletins.

TVA Codes. TVA Instructions. and TVA Announcements.

l With regard to inf ormation that occupational radiation exposures are low as is reasonably achievable, the f oll ow ing quotation is excerpted f rom TV A's Administrative Rel ease Manual :

l f- s 12A-5 Amend. 73 Nov. 1982

6 This instruction supplements the TVA Codes under VIII HAZARD CONTROL and VIII HEALTH SERVICES. It describes general respon-sibilities and administrative arrangements of ionizing radiation arising in connection with TVA's work. The detailed administra-tive arrangements in the instruction apply to all activities involving ionizing radiation.

TVA management is committed to maintaining radiation exposures to its employees and the general public, and the release of radioactive materials to unrestricted areas as low as is rea-sonably achievable (ALARA), as defined in 10 CFR Part 20. For the protection of its employees, TVA also subscribes to the ALARA philosophy set forth in the Nuclear Regulatory Commission Regulatory Guides 8.8 and 8.10 in the design and operation of all facilities utilizing radioactive materials or radiation sources.

ALARA Program - In view of the commitment in the TVA Admini-strative Release Manual, TVA has established a formal program to ensure that occupational radiation exposures to employees are kept as low as reasonably achievable (ALARA) and will apply this program to the CRBRP. The program consists of: (1) full management coninitment to the overall objectives of ALARA; (2) issuance of specific administra-tive documents and procedures to the TVA design and operating groups that emphasize the importance of ALARA throughout the ' design, testing, startup, operation, and maintenance phases of TVA nuclear plants; (3) continued appraisal of inplant rad *ation and contamination condi-tions by the onsite radiation protection staff; and (4) a 4-member cor-parate ALARA committee consisting of management representatives from the TVA design, operations and radiation protection groups, whose purpose is to review and appraise the effectiveness of the ALARA program on a plant-by-plant basis, including the CRBRP. In developing its ALARA program, TVA has closely followed the recommendations of NRC Regulatory Guides 8.8 and 8.10.

The responsibility for implementing the ALARA philosophy in the operation of TVA nuclear power plants is assigned to two divisions. The Division of Power Production has the responsibility of implementing the operational procedures described in Section C.4 of Regulatory Guide 8.8.

Fun.her in the implementation of Section C.4, the Division of Environmental Planning provides the radiation protection staff for TVA nuclear facilities and has the ultimate responsibility for determining that TVA maintains radiation exposures as low as reasonably achievable (ALARA) as 52l defined in 10CFR Part 20. The radiation protection program management and staff in the Division of Environmental Planning will, as a minimum, meet the qualification and training guidelines set forth in Regulatory 49 Guides 8.8 and 8.10.

12A-6 0 i9 l

15.6 1QDIUM SPILLS - INTRODUCTION p/

\

U Postulated sodium fires could possibly result in the dispersion of some radioactive material to the atmosphere. Fires involving primary sodium coolant are of most concern since this sodium circulates through the reactor core and accumulates radioactivity due to neutron activation and entrainment of fission products leaking from defective fuel. Postulated fires involving sodium used in the Ex-Vessel Storage Tank (EVST) cooling system could also result in radiological releases. The EVST sodium is essentially non-radioactive at the beginning of plant life. However, during refueling a small quantity of primary sodium is tranferred to the EVST along with each irradiated assembly, resulting in a slow buildup of radioactivity in the EVST sodium.

Besides the potential radiological impact of postulated sodium fires, these fires can result in pressure / temperature transients. Therefore, for each fire the consequences are evaluated in terms of: 1) ihe potential Individual whole body and organ doses at the site boundary and low population zone and 2) the pressure / temperature transient in the affected cell / building. The possibility of occurrence of any of the fires considered in this section is extremely unlikely. As such, it will be shown: 1) that 1he potential off-site doses are well within the guideline limits of 10CFR100, and 2) that the pressure / temperature transient does not exceed the design capability of the af f ected cell / building.

The computer codes utilized in the analysis of sodium spills and fires are N

/ SPRAY-3D, GESOFIRE, SOFIRE-II, SPCA, and HAA-38. These codes are described in Appendix A with identification of supporting references.

I Sodium spills at potential locations other than those discussed in this section have been examined. However the results of these spills were considered to be less severe in terms of radiological consequences and cell temperature / pressure transients and for this reason are not presented.

Since cells containing either primary or EVST sodium are normally closed and inerted, the potential for large postulated radioactive sodium fires exists only during maintenance, when these cells are opened and deinerted, and sufficient oxygen is available to sustain combustion. A spectrum of fires, both in inerted and de-Inerted atmospheres, is investigated in this section.

The consistent application of conservative assumptions throughout the analyses presented in this section provides confidence that the consequences of the fires are within the predicted results. A number of these assumptions are generic to all the fires evaluated in this section, and are summarized below:

1. The radioactive content of the sodium is based on continuous plant operation for 30 years. The design basis radioisotope concentrations were assumed present in the sodlur 'or the O

Anend. 64 15.6-1 Jan. 1982

accident analyses. Included in the basis and discussed in PSAR Section 11.1.5 is a design limit of 100 ppb (parts per billion) for plutonium content of the primary coolant.

2. Retention, fallout, plateout, and agglomeration of sodium aerosol in ceiIs or butIdIngs, whose design does not incIude spectfIc safety features to accomplish that function are not accounted for in the analysis. Neglecting these factors (an assumption that all of the aerosol is available for release to the atmosphere) leads to over-prediction of potential off-site exposure.

3. No credit for non-safety related fire protection systems is taken.

4. Dispersion of aerosol released to the atmosphere was calculated utilizing the conservative atmosphere dilution f actors (X/Q) applicable to discrete time Intervals provided in Table 2.3-38 (the 95th Percentilo Values). Guidance provided in NRC Regulatory Guide 1.145 was folIowed in calculating the X/Q values. Detailed descriptions of the atmospheric dilution factors estimates are provided in Section 2.3.4.

5. Fallout of the aerosol during transl! downwind was neglected.

6. The cells will be structurally designed to maintain their Integrity under the accident temperatures and pressures and the weight of the spilled sodium. For radiological calculations, no cr edit is taken for cell atrnosphere leak tightness.

7. The cell liners, catch pans, and catch pan fire suppression decks are designated as Engineered Safety Features and will have design temperatures equal to or greater than the sodium spilI temperature, thus confining the sodium spill.

8. The design basis liquid metal spill for either Inerted or air filled cells is defined as that spill resulting from a leak in a sodium or NaK pipe / component in the celi producing the worst case spill / temperature condition. The leak is based on a Moderate Energy Fluid System break (1/4 x pipe diameter x pipe thickness) as defined in branch technical position MEB3-1 with the sodium or NaK system operating at its maximum normal operating tanperature and pressure.

9. The only credit for operator action in mitigation of postulated sodium spills is shutdown of the Na overflow system makeup pumps 30 minutes after plant scram for a postulated leak in the Primary Heat Transport System (See Section 15.6.1.4).

O 15.6-2 Amend. 73 Nov. 1982

l l

10. The analysis of postula1 red liquid metal fires in air-filled cells does not include reaction of the liquid metal with postulated water released from concrete. The validity of this approach is presently being verified in conjunction with the large scale sodium fires test program discussed in Section 1.5.2.8 of the PSAR. If the test program does not support the present analysis approach, the appropriate effects of water release from concrete will be included in subsequent analyses.

Table 15.6-1 provides a summary of the Initial conditions for each fire considered and the maximum off-site dose as a percentage of the 10CFR100 guideline limits. As the table Indicates, a large margin exists between the i potential off-site doses and 10CFR100. A discussion of the pressure /

temperature transient for each event is provided in the following sections; In no case do the fires result in conditions beyond the design capablity of the cell / building.

The Project is assessing the impacts of a design basis NaK spill in the  ;

Reactor Service Building and will provide the results in the PSAR when the assessments are completed.

O O

15.6-2a Amend. 73 Nov. 1982

O 1 I

l 500 l

1 Steel Liner 2 1.5* Into structural Concrete l \ '.

400 4

4 b

l 3

2 e <

O,

- 300 2

1 N  :

" 2 200 .

4 4

1 1

.I 1

1

^ ^^^^^^ -

100 ^^^^^^^^^ ^^^^ ^^^ ^^ ^^^^^ ^^^^^^^^ ^^ ^^^^^^ ^^ ^ ^^^ ^^^^^

0 10 20 30 40 50 60 70 80 T'me (Hrs.)

Figure 15.6.1.4-12. Reactor Cavity - Wall Temperature Anend. 64 Jan. 1982 15.6-32

15.6.1.5 Intermediate Heat Transoort System Ploe Leak 15.6.1.5.1 Identification of Causes and Accident Descriotion it is expected that results of inservice inspection, pipe fabrication and instal l ation qual ity assurance measures, fracture mechanics analyses and tests, and leak detection provisions will lead to the conclusion that a sudden large f ail ure approaching the complete severence of an IHTS pipe is not credible. In particul ar, data f rom tests of leak detectabil Ity Indicate that the selected methods of leak detection ensure early detection of small IHTS leaks. The Design Basis IHTS leak selected on the basis of the existing inf ormation is that equivalent to the flow frcrn a sharp edged circul ar orif ice whose area is equal to one-hal f the pipe diameter times one-hal f the pipe wall thickness. (For the 24 inch IHTS piping the orif ice ar ea is 2.85 square inches.) This pipe break is consistent with the Moderate Energy Fluid System (MEFS) leak f or piping w Ith low stored energy identif led in NRC Branch Technical Position MEB 3-1, " Postulated Break and Leakage Locations in Fl uid System Piping Outside Containment."

Thermal and Aerosol Consecuence Assessment A sodium leak in the 24-in.-0D main loop hot leg piping in Cell 226 was selected as the limiting case for the design of the SGB; leaks in branch lines or thermowells would f all within the magnitude of this Iimiting analysis.

Leaks in the main loop piping in other cells have been evaluated; however, the leak in Coli 226 represents the 1 imiting case for design since the potential cell pressure and the potential combustion product aerosol release to the outside atmosphere are maximized. The leak is assumed to occur while the lHTS is operating at maximum normal operating temperature and pressure. The pipe break location was chosen to be at the low point of the main loop hot leg piping. This Iocation maximizes the spilI vol ume. The spil I parameters were generated by considering the system hydraul Ic behavior during the pipe break.

A conservative assumption is made that no operator action is taken to trip the pump in the leaking loop or to drain the loop to the dump tank. This assumption disregards the probable alarm of any leak by the extensive detection provisions of the Sodium-to-Gas Leak Detection System which are discussed in Section 7.5.5. A reactor trip is caused by a Plant Protection System signal from a Primary-Secondary flow mismatch. Loop f l ow is assumed to l continue under pump head until the pump tank is emptied through the leak. The leak continues at a decreasing rate determined by the cover gas pressure and grav ity head. The initial sodium discharge flow rate is 129 lb/sec, and the total spill quantity is approximately 300,000 lb of sodium. The spill duration is approximateIy 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The ieak rate time history Is depicted in Figure 15.6.1.5-1. The temperature of the initial sodium discharge is 936 F, and the average bulk temperature of the sodium is 800 F. The reactor decay heat is removed through the two remaining loops via the condenser by-pass or via steam venting and the protected air-cooled condensers. This accident is ci assif led extremely unl ikely.

This assessment has not included potential sodium Jet impingement on SGB concrete wal l s. The Project is investigating techniques to mitigate the ef fects of sodium Jet impingement on SGB concrete walls and will incorporate discussions of mitigation f eatures into the PSAR as they are developed.

15.6-33 Amend. 73 Nov. 1982

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Radioloalcal Consecuence Assessment I\

J An even more conservative assessment was made to demonstrate the potential

! radiological consequences of an lHTS pipe leak do not pose an undue hazard to i publ ic heal th and saf ety. The IHTS Design Basis Leak was combined with the i 1 1

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15.6-33a Anend. 73 '

Nov. 1982

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THE CLINCH RIVER BREEDER REACTOR PLANT 1 PRELIMINARY SAFETY ANALYSIS REPORT 1

O-IAFTER 17.0 - QUALITY ASSURAN&

l APPENDIX J i

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A DESCRIPTION OF THE ESG - RM i

j QUALITY ASSURAN PROGRAM r

i l ENERGY SYSTEMS GROUP f A DIV ISION OF ROO(WELL INTERNATIONAL CORNRATION l

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Amend. 73 t

! Nov. 1982 i

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CL INOi RIVER BREEDER REACTOR PLANT s )

A DESCRIPTION OF THE ENERGY SYSTEMS GROUP MANUFACTURER QUALITY ASSURANCE PROGRAM CONTENTS Page

0.0 INTRODUCTION

.......................................... 17J-I 0.1 SC0PE................................................. 17J-1 0.2 BASIS................................................. 17J-1 0.3 APPLICATION........................................... 17J-1 1.0 ORGANIZATION.......................................... 17J-2 2.0 OUALITY ASSURANCE PR0 GRAM............................. 17J-12 3.0 D E S I G N CO NTR0 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17J-18 4.0 PROCUREMENT DOCUMENT CONTR0L.......................... 17 J-21 5.0 lNSTRUCTION

S. PROCEDURE

S. AND DRAWINGS................ 17J-22 6.0 D OCU ME NT CO NT R0 L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17J-23 7.0 CONTROL OF PURCHASED MATERIAL. EOUIPMENT. AND SERVICES.............................................. 17J-25 8.0 IDENTIF.l CATION _ AND CONTROL OF MATERI ALS. PARTS.

A ND CGMP 0 NE NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 J-27 9.0 CONTROL OF SPECIAL PROCESSES.......................... 17J-28 l

j 10.0 INSPECTION............................................ 17J-29 J

1 11.0 TEST CONTR0L.......................................... 17J-31 12.0 CONTROL OF MEASURING AND TEST E0UIPMENT............... 17J-32 13.0 HANDLING. STORAGE. AND SHIPPING....................... 17J-34 -

14.0 INSPECTION. TEST. AND OPERATING STATUS................ 17 J-34.1 s

Amend. 73 17J-i Nov. 1982

-.mm _ ., _ - . . _ . , _ . , _ _ _ - . - _.

CONTENTS (continued)

Page 15.0 NONCCNFORMING MATERIALS. PARTS. OR COMPONENTS......... 17J-34.2 16.0 CORRECTIVE ACTl0N..................................... 17J-34.4 17.0 OUALITY ASSURANCE REC 0RDS............................. 17J-34.5 18.0 AUDITS................................................ 17J-34.6 18.1 EXTERNAL AUDlTS....................................... 17J-34.6 18.2 JMTERNAL AUDITS....................................... 17J-34.6 18.3 ACT I V I T I E S AUD I TED. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17J-34.7 FIGURES 17J-1 Overall Energy Systems Group Reactor Manuf acturer Qual ity Assurance Program Functional Organization of Progr am Par t i c i pa t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17J-35 17J-2 Energy Systems Group Quality Assurance h Department Orga n iz at i on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17J-36 17J-3 Major Elements of the Energy Systems Group Reactor Manuf act urer Q ual ity Assurance Program. . . . . . . . . . . . . . . . 17J-37

. 17J-4 Qual ity Assurance Procedure Index vs Req uirements of 10 CFR 50, Appendix B. . . . . .......... 17J-38 ATTACHMENTS Qual ity Assurance Program Manual Procedure I Descriptions.......................................... 17J-51 O

Amend. 73 Nov. 1982 17J-il

0.0 INTRODUCTION

k/s 0.1 SCOPE This appendix provides a description of the Quality Assurance Program conducted by Rockwell International Energy Systems Group (ESG) as a Reactor Manuf acturer (RM) for portions of the Nuclear Steam Supply System. Through the practices described herein, ESG discharges its responsibil ities to assure the qual Ity of systems, components, and structures provided by ESG and ESG's subcontractors.

ESG provides an annual review of the Quality Assurance Program description contained in this appendix, and modification as necessary to keep it current.

Changes in the Qual ity Assurance organization are provided to the Lead Reactor V..'uf acturer within 30 days of issuance of the approved organization chart.

0.2 BASIS The program defined herein is based on ESG having been assigned execution responsibil Ity for the Qual ity Assurance Program appl ied to design, procurement, and manuf acturer of systems, components, and structures as shown by Figure 17J-1. ESG is not assigned responsibil ities f or site construction and instal lation.

0.3 APPLICATION r3 The practices described herein wil l be appl led to the planning, design,

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procurement, and manuf acture of those systems, components, and structures defined in Sections 3.2, 7.1, and 9.13 of the PSAR that are assigned to the ESG scope of work.

1.0 ORGANIZATION Energy Systems Group, a division of Rockwell International, has been assigned RM responsibil Itles f or the systems, components, and structures def ined in Section 0.3 of this appendix. The organization of Individuals and groups perf orming qual ity-rel ated activities is shown and def ined in Section 1.4 of the PS AR. Figure 17J-2 depicts the organizational structure of the ESG Qual ity Assurance Department. This organization chart shows only lines of administrative control (sal ary review, hire-fire, position assignments). The

, separation of the organizational elements of Engineering, Procurement, Manuf acturing, and Qual Ity Assurance (which includes all inspection f unctions), with separate 1 Ines of administrative control from the Energy Systems Group President, provides the authority, independence, and f reedom f or each ef fectively to perf orm qual ity-rel ated activities.

Qual ity Assurance responsibil Itles f or CRBRP are assigned and executed by the f ollowing f irst l ine organizations of ESG:

Atomics International Division Engineering and Test Operations

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V Qual ihr Assurance Amend. 73 17J-1 Nov. 1982

The Atomics International Division contains the CRBRP Progran Of f ice and associated financial and program planning and control f un ct i ons. The Operations organization contains the Purchasing and Manuf acturing Departments which are responsible for CRBRP procurement and Internal fabrication activities. Engineering and Test contains the ESG design as well as development and design verif ication testing f unctions. Qual ity Assurance has the responsibil Ity for developing a qual ity assurance program meeting CRBRP project requirements and assuring its of fective execution. Qual ity Assurance al so provides resources f or inspection, examination, and test of suppl ier and ESG fabricated items.

1.1 The responsibil ity and authority of key managers involved in quality-related activities is as f ollows:

1.1.1 Atomics International Division Vice President and General Manager The Atomics International Division V ice President and General Manager has overalI responsibil ity for the management of the LMBR programs and the Nuclear Products Facil ities and Services. LMBR prograns incl ude the CRBRP RM activities as well as Large Breeder Reactor, Sodium Technology, LMFBR Ccrnponent Devel opment, and Saf ety prograns. Theref or e, the responsibil Ity for ESG's overall perf ormance on the CRBRP is vested in the General Manager.

1.1.2 jyFBR Proarams Director The LMFBR Programs Director has overalI responsibil ity for the LMFBR business segment, incl uding CRBRP Program activ ities, large pl ant design projects, and LMFBR Base Technol ogy.

1.1.3 CRBRP Procram Manaaer The CRBRP Program Manager is responsible for the management of the CRBRP Program at ESG. In this capacity, he is responsible fcr managing the CRBRP Program work in'accordance w ith the contract requirements and providing direction to the f unctional organizations within ESG for CRBRP development, design, and procurement.

1.1.4 Enaineerirta and Test Vice President The Engineering and Test V ice President is responsible f or the management of ESG's central ized engineering activities. On the CRBRP program, engi neeri ng work in support of conceptual design, preliminary design, and final design is assigned to the Engineering Department. Engineering and design work conducted by the Engineering Department includes: Mechanical Design, Drafting and Checking, El ectr ical and Control Engineering, Material s and Process, Piping and Structural Design, Thermal and Process Systems Pressure Components Stress Analysis, Structural Systems Stress Analysis, Specif ications and Manual s, Engineering Assurance and Data Management, and the verification of design through devel opmental and acceptance testing O

Amend. 73 W-2 Nov. 1982

1.1.5 Ooerations Director U The Director of Operations is responsible for the product manuf acturing, material purchasing and warehousing in support of the CRBRP in accordance with the control l ing programatic documents. The material purchasing f unction is responsibl e f or selecting sources, procurement, subcontract administration, assuring adherence to work statements, prices and del ivery schedul es, receiving, inspect i on, storage, Issuance, payment of invoices, and observing the perf ormance qual ity of the articl es purchased. The manufacturing manager is responsible for reviewing engineering and design work performed by ESG to assure manufacturabilIty.- On the CRBRP program, as wIth other programs, the Manager of Manuf acturing Engineering is responsible for conducting on-the-board reviews, participating in design reviews, reviewing vendor design Information, and assuring component designs can be f abricated and assembled expeditiously and at minimum cost.

1.1.6 Finance and Administration Vice President and Controller The CRBRP administration is under the cognizance of the Finance and Administration V ice President. The Finance Controller reports administra-tively to the Finance and Administration V ice President and organizationally to the Al Division V ice President and General Manager. Within the Finance and Administration Organization, the Program Business Management f unction is responsible to the Individual projects f or assistance in the budgeting and pl anning of manpower and dollar expenditure rate; f or maintaining and "

reporting project costs and remaining balances; for monitoring and satisfying (q

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contractual requirements; f or maintaining contract data control systems; and for providing assistance in preparation of project schedul es. On the CRBRP program, Program Administration provides the CRBRP project management with detailed weekly summaries of manpower expenditures, monthly cost information, projection of figure costs at various subaccount levels, commitment control system reports, and various other reports required by the project and the customer.

1.1.7 Oualltv Assurance Director The Qual ity Assurance Director is responsibl e f or the Qual ity Assurance activities within ESG, which include the Quality Assurance f unctions for the CFBRP project. He is responsible for establishing and maintaining a quality system that meets the requirements of all contracts received by ESG, incl uding meeting the requirements of RDT F2-2 for the CRBRP project. The authority for achieving these responsibil Itles is through the issuance of Standard Operating Pol Icles and Procedures f rom the President of ESG.

The Qual ity Assurance Director has the authority to prevent issuance of drawings and specif ications, and to terminate work where qual ity requirements are not being met. He Interraces directly with the Atomics International Division Vice President and General Manager to assure that quality program requirements are being met by ESG personnel working on the CRBRP project.

O Amend. 73 17J-3 Nov. 1982

The Qual ity Assurance Director manages a number of organizations and f unctions wIthin the Quality Assurance Department to provide assurance that the ESG and CRBRP Qual ity progrms are ef fectively Impi mented. A description of the responsibil ities of the managers of these organizations and f unctions is given in the f ol low ing sections.

The Qual ity Assurance Director reports directly to the President of ESG.

l 1.1.8 CRBRP OualItv Assurance Proaram Manager The CRBRP Qual ity Assurance Program Manager is responsible to the Quality Assurance Engineering LMFBR Progres Manager for defining and assuring that the Qual ity Assurance Program fcr CRBRP Reactor Manuf adurer activities assigned to the Energy Systems Group is ef fectively executed within ESG. This responsibil ity al so extends to assuring that subcontractors def Ine and impi ment contractual ly appl led qual ity assurance programs. He is also responsibl e f cr cost, schedul e, and technical perf ormance of the Qual ity Assurance cost accounts of the Energy Systems Group Perf ormance Measurment Sy stem.

l 1.1.9 Oualltv Assurance Audits and Controls Manager The Energy Systems Group Audit Program responsibilities of the Quality Assurance Director are impimented through the Manager, Qual ity Assurance Audits and Control s. The Manager, Qual ity Assurance Audits and Control s, is responsible for:

1) Maintaining and administering the Quality Progra, Auult System by preparing and maintaining audit schedules.

2) Arranging for check!Ists and conducting or arranging for audit teams to conduct audits.

3) Insuring preparation of audit reports.

4) Foiicwup to verif y correciive action impi mentation.

5) Maintenance of audit case history fil es.

6) Devel opment, issuance, control, and revision of Qual ity Assurance Manual s and procedures.

7) Review of operating procedures, and revisions thereto, prepared by other qual Ity-af fecting vrganizations, to assure compatibil Ity with overall ESG Qual Ity Assurance Progrm requirements.

8) Perf orming suppl ler qual ity surveys of procurment sources f cr materials and f abrication services and maintenance of the approved I ist of such suppl Iers.

9) Administering a Material Review system f cr nonconf ccming items.

O Amend. 73 Nov. 1982 17J 4

O 10) Administering a Corrective Action system to assure prompt and ef fective correction of conditions causing nonconformance to technical req uirements/ procedures.

11) Chemical, physical, and mechanical property testing services to support other Qual!ty Assurance Department units.

12) Qual if ication programs f or wel ders and welding procedures.

13) Performing surveillance of warehouse areas and manuf acturing control stations tc' assure that only accepted items, properly identifled and protected f rom damage and deterioration, remain in storage. Assure corrective action for any unsatisf actory conditions observed.

l 1.1.10 Oual Itv Assurance Engineering LWBR Proarams Manager The Qual ity Assurance Engineering LMFBR Programs Manager is responsible to the Quality Assurance Director and provides quality assurance engineers to support the CRBRP Qual ity Assurance Program Manager. Qual Ity Assurance Engineering personnel perf ccm the following activities:

1) Qual ity Assurance Program administration for specif Ic portions of the CRBRP activities, to mcnitor and assure of fective implementation of qual Ity requirements f rom design through procurement and f abrication.

2) Qual ity Assurance engineering support for change control boards, design reviews, and design document review and approval .

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3) Nonconforming item review board coordination.

4) Developing and implementing statistical test programs and analyses as req u i red.

5) Evaluating inspection and test data and report quality trends.

6) Reviewing and evaluating bid Invitations and returns for quality impact.

7) Participation on capabil ity eval uation teams f or prospective suppl iers of major items.

8) Procurement document review and suppl ier qual ity surveys f cr material s and f abrication services and maintenance of the approved Iist of such suppliers.

9) Receiving inspection pl anning.

10) ESG f abrication Inspection pl anning.

11) A qual Ity data and records collection and storage system fcr procured and ESG-f abricated items.

12) Data packages f cr ESG-f abricated items.

Amend. 73 tiov. 1982 17J-5

13) Source inspection and surveillance of suppliers.

14) Qual if ication and certif ication programs f or nondestructive examination (NDE) personnel and procedures.

15) Nondestructive examination technical support and consultation to ESG organizations and suppl Iers.

16) Qual Ity Assurance instructio" ?cr compl ex inspection, tests, and process control operations.

17) Development of nondestructive examination methods f or the Inspection and Test Unit.

1.1.11 Oualltv Assurance Engineering Util ity and Enerov Procrams Manager The Qual ity Assurance Engineering util Ity and Energy Programs Manager is responsibl e to the Qual Ity Assurance Director. This organization has no involvoment in the CRBRP program.

Tl 1.1.12 Insoection and Test Unit Manager The inspection and Test Unit Manager is responsible to the Quality Assurance l Director and, along w ith his assistant managers, is responsibl e f or:

1) Perf orming receiving inspection of procured items and services, identifying and documenting nonconforming conditions of these items and services, and assuring conformance to the establ ished nonconf crmance dispositions.

2) Perf orming inspections and tests of ESG f abrication and subassembly operations, final inspections, and performing or witnessing perf ormance of acceptance and qual If Ication tests of ESG-f abricated items.

3) Perf orming nondestructive examination and acceptance of ESG-fabricated i tem s.

4) Making inspection acceptance and release acceptable ESG-fabricated items f or del Ivery to the next operation. Reject and withhol d nonconf orming items. Document nonconf orming conditions f or Material Review eval uation and assure prompt conformance to Material Review disposition.

5) Perf ccming inspection of purchased or ESG-manuf actured tooling.

6) Perf orming inspection of packaging, preservation, and identification of items prior to shipment.

7) Maintaining a system for calibration of measurement instruments used f or product inspection and test, incl uding appl Icabl e procedures and records. Perf orming periodic cal lbration of measuring instruments, in accordance w Ith establ Ished requironents.

Amend. 73 17J-6 Nov. 1982

l 1.2 Qual ity assurance pol icy originates w Ith the President of RockwelI Q International, through the issuance of a Corporate Policy statement covering l Product integrity. The Qual ity Policy is issued to each division of Rockwell International in a Corporate Directive, prepared and authorized by the Senior Vice President, Corporate Staf f s, which directs, each division to take action to impl ement the Corporate Qual ity Pol icy. The President of the Energy Systems Group implements the Corporate Quality Policy Directive through Standard Operating Pol Icles, which provide qual ity assurance direction consistent with Corporate Policy, as well as the Quality Assurance Program requirements appl icabl e to ESG business objectives and contract requirements.

The overalI Quality Program is implemented in the operating manuals of the qual Ity-af fecting organizational units by the managers of these units. The Qual ity Assurance Director reports directly to the Energy Systems Group President and verifles compi lance of the qual ity-of fecting organizations to the Qual ity Program, under the authority granted in the Standard Operating Policles.

1.3 The Qual Ity Assurance Director, by virtue of being at the same level of management as the highest level manager of other major Energy Systems Group I

functions, has the necessary unimpeded communication path to bring qual Ity matters to the attention of the president and executive level management.

Dif ferences of opinion on qual ity matters that cannot be resolved at lower management levels are ref erred to the Energy Systems Group President by the Qual ity Assurance Director f cr f inal resol ution. Qual ity Assurance Department Managers or Qual ity Engineers attend scheduled and ad hoc status meetings to assist in resolving probl ems, report qual ity resul ts, interpret qual ity

/ requirements, and provide a basis f or providing adequate staf fing.

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1.4 Qual ity Assurance f unctions implemented w ithin ESG are def ined in Standard Operating Procedures. A!I functional organizations (Program Of f ices, Engineering, Purchasing, Quality Assurance and Manuf acturing) are assigned responsibil Ity fcr:

1) The preparation and issuance, in the operating manual s, of written instructions and procedures which establish the methods and responsibil ities f or perf orming qual Ity-rel ated activities, and for verifying satisf actory performance of such activities.

2) The Indoctrination and training of their personnel in these procedures, as appl Icable to their work assignments.

l 1.5 in addition, the Qual Ity Assurance Director is assigned the following specif ic qual ity assurance f unctions:

1) identifying those procedures which cover the perf ormance and verif ication of qual ity-rel ated activ ities.

2) Conducting audits of the implementation of such procedures.

3) Identifying quality def telencies and problens in the Program and reporting them, with any recommendations, to the responsible ESG executive, f unctional and program managers.

Amend. 73 Nov. 1982 17J-7

4) Verifying that solutions +o reported qual ity probl ans or def iciencies are achieved.

5) Stopping nonconf orming work and control l ing f urther processing, fabrication, and del ivery of nonconf orming items.

6) Submit overall status reports on the ESG Quality Assurance Prograns to the ESG President, as well as concerned program and f unctional managers.

l 1.6 Communications fl ow directly between the ESG Quality Assurance Department and the Qual ity Assurance organization of subcontractors, and are documented, as appropriate, by the Purchasing organization buyer assigned f or each subcontractor. The I ines of communication are def ined in Internal procedures, and in procurement and qual ity assurance administrative specif ications contractual ly appl led to each subcontractor. The ESG Contract Data Management organization tracks and provides management reports of all communications requiring action, on either the part of ESG or subcontractor, to provide a means of insuring timely resol ution of probl ems.

l 1.7 Verification of conf ormance to establ ished quality requiranents is the responsibil ity of the Qual Ity Assurance Department, through the actions of ceview and approval of design documents (specif ications and drawings),

V scurement documents (purchase requisitions and purchase orders, along with their referenced documents and attachments), and manuf acturing documents (travelers and processing procedures). Additionally, the Qual ity Assurance Department is responsibl e f or verif ication of conf ormance to qual Ity requirements of hardware items during source inspection / surveil lance, reco lv i ng, in-process, and final inspections and process surveil l ance. As shown by the organizational structure and the f unctional descriptions of the ESG organization in Section 1.4 of the PSAR, the Qual ity Assurance Department is divorced f rom the qual ity-af fecting organizational units perf orming the design, procurement, and manuf acturing activities, with the Qual ity Assurance Department having a hierarchal position at the same or higher level than the perf orming organizations.

The authority and responsibil Ity for stopping unsatisf actory work, or the control of f urther processing, del ivery, or installation of nonconf orming material, is an expl icit f unction of the Qual Ity Assurance Director in the Standard Operating Policy covering the ESG Qual ity Assurance Program and issued by the ESG President.

The ESG Qual Ity Assurance Department reporting l evel, and the Standard Operating Pol icy covering the ESG Qual ity Assurance Program, are structured and expl icitly provide f or the Qual ity Asst e Director to:

1) Identify qual ity probl ans

2) Initiate, recommend, or provide sol utions to qual ity probl ems

3) Verify impl ementat!cn of sol utions O

Amend. 73 Nov. 1982 17J-8

A 1.8 The qual if Ication requirements f or the Qual ity Assurance Department management positions are as f ollows:

1) Minimum qual if Ication requirements f or the Qual Ity Assurance Director are (a) a Bachelor of Science degree in Engineering, Science, or Technology from an accredited college or university, (b) 15 years experience in qual Ity assurance or engineering in an advanced technology industry, of which at least 5 years will be in quality assurance; and, of th is 5 years, at l east 2 years wil l be in the nucl ear area, (c) experienced in the direction of personnel, and the planning and management of resources needed to conduct a Quality Assurance Program, and (d) possess a knowledge of industry and government codes, standards, and regul ations def ining qual ity assurance requirements and practices; qua!ity assurance administrative methods and technology and their appl ication; and be experienced in pl anning, def ining, and perf orming qual ity assurance practices and application of procedures.

l 2) Minimum qual if Ication requirements f or the Qual ity Assurance Engineering LMFBR Progrms Manager, Quality Audits and Controls Manager, and Quality Assurance Engineering Util Ity and Energy Programs l

Manager are (a) a Bachelor of Science degree in Engineering, Science, or Technology from an accredited college or university, (b) 5 years experience in or related to the field of his educational major, of which at least 2 years will have been in quality engineering or technology; and (c) possesses a knowledge of at least two of the p f ollowing areas of specialty: statistics /rel iabil ity, nondestructive examination, physical / mechanical properties measurement, metal f abrication, measurment technol ogy, instrument and control f abrication and testing, chemical processing and analysis, f ail ure analysis, and qual Ity program development and implementation.

l 3) Minimum qual if Ication requirenents for the Inspection and Test Unit Manager are (a) 10 years experience in a manuf acturing Industry of which 5 years will have been in qual ity control / assurance; and (b) have a general knowledge of manuf acturing and inspection methods and techniques including dimension and electrical measurements, nondestructive examination, qual ity pl anning, and f abrication and assembly methods.

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4) Minimum qual if Ication requirements f or the CRBRP Qual Ity Assurance Program Manager are (a) a Bachelor of Science degree in Engineering, Science, or Technology from an accredited college or universi+y, (b) 5 years experience in qual Ity assurance or engineering in an advanced technology Industry, of which at least 3 years will be in quality assurance in the nucl ear area, (c) experienced in the direction of personnel, and the planning and management of resources needed to conduct a Qual Ity Assurance Program, and (d) possess a knowledge of Industry and government codes, standards, and regul ations def ining qual Ity assurance requirements and practices; qual Ity assurance administrative methods and technology and their appl Ication; and be experienced in pl anning, def ining, and perf orming qual ity assurances practices and appi Ication of procedures.

Amend. 73 17J-9 Nov. 1982

1.9 Adequate staf fing of the QA Department is the responsibil ity of the QA Director and managers reporting to the Director. Basically, staf f size is a f unction of business l evel . For each project or progrm, the Q A Director provides an estimate of quality engineering, inspect i on, and supervision f unding needs to the project or program manager. These estimates are prepared by members of the QA Department staf f and negotiated when necessary by the QA Director w ith the project or program manager. Issuance of the funding to the QA Department is then through normal accounting channel s to Qual Ity Assurance Department Managers who then staf f appropriately. Certain overhead f unctions, such as cal ibration, procedure development or audit are staf fed to an adequate size based on negotiations between the Q A Director and the Control ler.

Qual ity Assurance personnel are involved in day-to--day pl ant activ ities to assure adequate QA coverage. For ESG f abrication, both the assigned Qual ity Engineer and inspection Manager attend schedul ed meetings with Manuf acturing and Purchasing management on status of work in progress. These meetings are normally schedul ed weekly and may be hel d daily during periods of intense act i v i ty. Fl oor l evel inspection and manuf acturing managers al so interact daily to ensure adequate inspector avail abil Ity to meet current work schedul es. Qual ity Engineers are assigned to specific portions of the CRBRP activities at ESG, e.g., systems and/or components, and these engineers Interact daily with their counterparts in Program Off ice, Engineering and Purchasing. The qual ity engineers al so attend schedul ed and ad hoc meetings and are on distribution f or appropriate correspondence, reports, drawings, and specifications.

2.0 OUALITY ASSURANCE PROGRAM The Qual ify Assurance Program described herein compi les with the requirments of Title 10, Code of Federal Regul ations, Part 50, Appendix B, " Qual ity Assurance Critoria fcr Nucl ear Power Pl ants," f or the ESG scope of work as a O<BRP Reactor Manuf acturer. The el ments of the CRBRP Qual ity Assurance Progrm to be executed by ESG are shown in Figure 17J-3. The Qual Ity Assurance Progra is appl led to individual structures, systems, and components in a def ined, graded approach, according to their importance to safety. This program is issued and made mandatory by direction of the President of Energy Systems Group by Standard Operating Pol Icles that require the issuance of operating procedures, and provides f or verif Ication of thelr enf orcement through a system of qual ity progrm audits. ESG delegates execution responsibil Ity of appropriate Qual Ity Assurance Program elments to suppl Iers of mater ial, equi pment, and services, but retains responsibil ity for their impi mentation by these suppl Iers. Such delegation is controlled as described in paragraphs 8.1, 9.1, 10.1, 11.1, 12.1, 13.1, 14.1, 15.1, 16.1, and 18.1 of this appendix.

l 2.1 Management assessment of the scope and of fectiveness of the Quality Assurance Program is accompl ished by two independent audits. One of these is perf ccmed atyearly interval s, and specif ical ly addresses the 10 CFR 50, Appendix B, requirements as they are impimented through that portion of the Qual ity Assurance Progrm that addresses Section NCA-4000 of Section lli of the ASK Boller and Pressure Vessel Code. The second audit occurs at 18-month interval s and is conducted by sonicr of ficials frm other divisions of Rockw el l international. This latter audit is to assure compliance with contractual and statutory qual Ity assurance requirements.

Amend. 73 17J-10 Nov. 1982 l

p Continuing involvement of the ESG President in Qual Ity Assurance matters is Q achieved by three routinely scheduled interactive associations with the Qual ity Assurance Director. These are: (1) periodic staff meetings, during which each member of the President's staf f, which incl udes the Qual ity Assurance Director, must report on signif icant probl es, accompi ishments, and status of activities, (2) periodic Program Review Meetings, in which formal and in-depth reports are presented by Program Managers, and during which time the Qual ity Assurance Director addresses signif icant qual Ity probl ms, wIth recommendations f or corrective action, and (3) submission of a monthly quality status report to Executive Management that covers qual ity progress accompi Ishments, probi ms, and audit results, and to the customer as required by contract.

l 2.3 Qual ity pol icy originated at Rockwell International, with the issuance of a " Product Integrity" pol icy statement, in which the President of the Corporation states..."It is the policy of the Corporation that its product w il l meet or exceed appl icabl e standards and requirements f or qual ity, rel iabil ity, and saf ety," The Senior V ice President, Corporate Staf f s, issues a directive appl Icabl e to all Division Presidents of the Corporation, which requires actions to be taken to impiment this Corporate Pol icy, incl uding:

1) Provi@ g engineering activities f or def ining saf e and rel iabl e products.

2) Providing verification or qualification testing of new products and any subsequent significant design chsnges prior to introduction into the market.

a 3) Providing purchasing activities that are responsible for procuring material s, components, and end items that comply with specified req u irements.

4) Providing manuf acturing activities that are responsible f or the manuf acture of products that comply with specif led requirments.

5) Providing qual Ity assurance activities at each manuf acturing location to ensure compi lance w ith specif led requirments.

6) Preparing and maintaining clear and correct descriptions of products to be used in advertising and sal es l Iterature, proposal s, contracts, customer I iterature, service manual s, Iabel ing, and other necessary documents.

7) Providing prompt f eedback of f iel d data regarding f ail ures, compl aints, and accidents to the appropriate f unctional organizations.

8) Developing procedures to ensure that appropriate government agencies and customers are promptly notified of product conditions that could be hazardous and timely resolutions of such conditions.

1

9) Notify ing the Of f ice of the V ice President - Communications, Corporate J Of fices, rel ative to all product conditions which could be hazardous. j l

Amend. 73 l j7pj j Nov. 1982  !

I

10) Establ ishing measurement techniques to provide management visibil ity of the adequacy of product Integrity activities.

11) Preparing and maintaining appropriate written operating procedurse, to impiment the requirments of this Directive.

12) Maintaining a record retention progre in conpliance with the appropriate Corporate Finance Pol icy which w il l support the integrity of company products.

13) Conducting periodic audits throughout alI activities having a direct impact on product integrity to measure compl iance w ith establ ished operating procedures.

l 2.4 The Corporate Qual Ity Pol Icy is impimented at Energy Systems Group through Standard Operating Policies issued by the President, Energy Systens Group. This Group Pol icy states: "The managers of Engineering, Material (Purchasing), Manuf acturing, Qual ity Assurance, and Program Offices wIlI be responsibl e f ce:

1) The preparation and issuance, in their operating manual s, of. written instructions and procedures which establ Ish the methods and responsibil ities f or perf orming qual Ity-rel ated activItier and f or verifying satisf actory perf ccmance of such activities;

2) The indoctrination and training of their personnel in these procedures, as appl icable to their work assignments;

3) Assurance that the instructions and procedures covering quality-l related activities meet the Quality Assurance Program reqAroments of the appl .icabl e government regul ations and/or contract provisions; j 4) Requiring that each Individual be responsibl e f cr perf orming qual Ity-rel ated activities in accordance w ith the appl icabl e instructions and procodures."

Based on the previously described qual ity pol icy, the department managers provide procedural coverage in their department manual s f or qual ity-af fecting activities.

l 2.5 The Qual ity Assurance Director has overalI responsibil Ity for assuring conf ccmance to the procedures of the Qual ity Assurance Progrm Manual. He has the f urther responsibil ity, authority, and organizational freedom to stcp non-conf orming wcrk, and control further processing, f abrication, and del Ivery of nonconf orming items. If the dif ferences of opinion occur that cannot be resolved, these are ref erred to the President of Energy Systems Group for f inal resol ution. Changes to department manual s may be proposed by any individual or crganization, but f inal review and approval resis with the department manager. Changes to ESG ASW Code Section t il Manual and the basic Qual Ity Assurance Department Manual receive f inal review and approval by the Qual ity Assurance Director. Additionally, Standard Operating Pol Icles and Procedures, CFERP Program Directives, Engineering Management Procedures are rev iewed f cr concurrence by QA Department personnel . All procedures decl ared as qual ity-af fecting are submitted to the lead reactor manuf acturer and owner.

Amend. 73 17J-12 nov. 1982

6 xy ' { ,

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2.6 Provisions 'f or controll ing the distribution of Department and Qual Ity s

. Assurance Manuals are addressed in each manual . These provisions provide f or N

l'c seria(lzation of each manual ,in use and maintenance of a record of the

' x recipients of each manual .' Revisions of procedures in the manual s are distributed to each manual holder of recora, along with an updated table of contents. '

l 2.7 The CRBRP Qual Ity Assurance Project Manager Icentif les the procedures f rm Departmert and Qual ity Assurance Manual s that constitute the Qual ity Assurance Progrm for the ESG CRBRP Project Reactor Manuf acturer scope of work. These procedures are documented in a Quality Assurance Program Index

.. that is approved by the ESG Qbal My Assurance Director and CRBRP Program S Manager. The index is issued f or use by managers and key personnel in organizations perf crming activities that af fect qual Ity. Changes to this g  : Index must be approved by the EEG Quality Assurance Director and the CRBRP Program Manager. A brief synopsis of euch procedure contaired in the CRBRP Qual Ity Assurance Progrm Manual is given in Attachment 17J-1 of this

! appendix.

The saf ety-rel ated structures, systems, and components tasks controlled by the ESG Qual Ity Progrm during engineering, design, and procur ment are defined in Section 0.3 of th is appendix.

! l 2.8 Contractors of component designs and/or f abricated items are required to

(; submit their qual Ity assurance program descriptions for these items for review and approval . This revlew is made against contractual ly appi led qual Ity O assurance program requirements. Additionally, audits of these program activities are conducted by ESG. The requirments f or qual ity assurance program description submittal, and notification of the right of audit, are contained in administrative specifications, which are made part of each component contract.

2.9. Porsenel perf ccming qual Ity-rel ated activities f or CRBRP receive a 4

tralding and Indoctrination course covering the CRBRP QA program incl uding qual Ity assurance for nuclear f acil ity projects in the United States, the overall Cl inch River Breeder P.eactor Pl ant project, and the impl ementation of th is Q A program at ESG. This training includes quality concepts; CRBRP design f mll lorization, major participant responsibil ities, and organization interrelationships; and procedure requirments f or each ESG organization.

Additional ly, personnel involved in ASE Code Section lli activities receive training courses' as to the specif ic procedures appl icabl e to their f unction, and their content, scope ard purpose. Contents of the courses, attendees, and dates of attendance are documented.

2.10 Specif ic categories of personnel responsible f or verify ing activities af fecting qual ity require f ormal training in the principles, techniques, and '

requirments of the activity being perfccmed. Certif ication as written testimony of qual if ication is provided in accordance w ith the appropriate

code, standard or procedure, and course content, attendees and dates of attendance documented. Proficiency tests are given to obtain evidence of

proper tr,aining and qual if Ication. Certif Ications of qual if Ication are issued O that del ineates the specif ic f unctions personnel are qual lf led to perf orm.

The cr.iteria for qual if ication are provided in appl icabl e procedures, and resul ts f cr each indiv Idual are maintained in Training Department f Il es.

Amend. 73 17J-13 nov. 1982

Prof Iclency of personnel is maintained by retraining, rooxamination, or continued satisf actory perf ormance in accordance w ith specified procedural requirmonts, and recortif Ication is documented along with the basis for rocertification. Qual ity verif ication personnel involved in the cortification progrzrn are os f ol lows:

Personnel perf ccming nondestructivo examinations, and establ Ishing

. NDE techniques (SNT-TC-1 A)

Personne: perf ccming wel ding operations (ASkE S-IX and AWS)

Personnel leading qt ;l Ity audit Itms (ANSI N45.2.23)

Personnel perf crming visual examination (ASFE S-Ill, Subsection NF)

Personnel perf ccming dlmonsion inspection Personnel cortif y ing Design Specif Ications, Design Reports, Overpressure Protection Reports, and Load Capacity Data Sheets (ANSI /ASFE N626.3) 2.11 Proceduros that provido instructions f or qual ity-rel ated activitios such as cl oaning, wol ding, nondestructivo exam ination, inspect i on, and test, specify equipment and f acil itles to be used as well as any appropriato environmental conditions to be maintained during those activities, e.g.,

temper at uro, humidity, and cl eani inoss. The sequence of events to be followed is specif led in the work Instruction documents (Tost Proceduros and bianuf acturing Travel ers), and verif icat!on of conf ormance to th is sequence is perf ormed to assure preroquisitos havo boon mot price to successive operation.

2.12 The Qual ity Assuranco Progrm described herein is reviewod and revised annual ly as appropriate. Changes in the QA Department organization are transmitted to the load reactor manuf acturer and owner within 30 days of issuance of the crganization chart. The overall ESG organization given in Section 1.4 of the PSAR is reviewed and revised annually. Al so, the l ead reactor manuf acturer is notif led of key personnel changes bef cro the changes are announced.

1 2.13 Devel opment, control, and use of computer progres f cr design and design verification are covered by a procedure under the control of the Engineering Department and which is incl uded as part of the CRBRP QA progran. Adherence to this proceduro is audited by Quality Assuranco using knowledgeable and Indopondent auditcrs.

2.14 The dockot date of the CRBRP PSAR was April 11, 1975. Regul atory guidos to be addressed prior to that date and other f actors to be considered are as f ol i cw s:

1) Regul atory Guidos in Subsection V of Section 17.1 of WREG 0800, as

described in PS AR Sections 1.1,17.0, and 17.1.2.1 and the answers to j Quostions 411.1 and 411.2.

2) 10 CFR Part 50, 50.55a, as described in PSAR Sections 17.1.2.1, 3.1, l 3.2, and 7.1.

Amend. 73 17J-14 Nov. 1982

1

3) 10 CFR Part 50, 50.55(e) in accordance with the qual ity assurance program, as described in PSAR Section 17A.15.1.

4) 10 CFR Part 50, Appendix A, General Design Critoria 1, as described in PSAR Sections 17.0.5, 17.1.2.6, and 3.1.1.

5) ASE B&PV Code Section .f il, as described in PSAR Sections 17.1.2.6 and 3.2.2.

O O

Amend. 73 17J-15 Nov. 1982

3.0 DESIGN CONTROL l 3.1 ESG utilizes a Cognizant Engineer concept to assign engineering responsibil ity fcr the various systems and subsystems f or wnich ESG is the Ructor Manuf acturer. Each Cognizant Engineer, under the direction of his m anager, has the responsibil Ity for pl anning, directing, and control lIng alI ef fort in conf ormance w ith the contract work scope for the system, subsy stem, or component under his jurisdiction. This responsibil Ity includes the coordi-nation and integration of all activities related to systems requirements definition, system engineering, component design, interf ace control, and change control . The Cognizant Engineer is supported in this ef fort by the f unctional engineering groups, such as the structural, electrical, and design groups. Written procedures describe the methods to be used in carrying out these activities.

l 3.2 Appi Icabl e regul atory requirments and design bases are def ined in principal design documents. The top level design requirement document is the Overall P1 ant Design Description (OPDD-10). This document describes the over-alI CRBRP technical, f unctional, and qual Ity parameters. OPDD-10 is written, released, and control led by the Lead Reactor Manuf acturer.

System Design Descriptions (SDDs) provide the principal means of design def inition and control fcr each CRBRP system for which ESG has system responsibil ity. The SDDs reflect the OPDD-10 requirments and are used to def ine the various technical, operational, and saf ety considerations invol ved, identify Interf aces, and serve as the basic technical document f or the system.

Specif ications and procedures are prepared to def ine the requirments for the design, f abrication, qual ity ;ssurance, testing, handi ing, sh ipping, instal l ation (where appl icabl e), construction testing, and preoperational testing of components and structures Ir compl iance w ith the SDD and all approved basel ino documents.

Engineering drawings are developed to meet the requirements of the SDD, approved basel Ine documents, and component specif Ications, and f urther to def Ine and establ Ish engineering permeters, characteristics, and design functions.

Design drawings and specifications are reviewed prior to release by Quality Assurance engineers. Th i s rev iew is perf ormed in accordance w ith a procedure that provides approval requirements established by senior manapment of the Engineering, Operations, Qual ity Assurance and Progrm Management organ iz ati ons. The Qual ity Assurance engineering review is conducted to assure compl iance to Engineering and progrm procedures which specify that draw ings and specifications contain qual ity assurance requirements such as inspection and test requirments, acceptance requirements and inspection and test resul ts documentation. Deviations or changes from these drawing and specif ication requirements are processed as specified in Sections 15.0 and 16.0 of th ,3 appendix.

1) Design characteristics can be controlled, inspected, and tested.

2) Inspection and test (including any design verification testing) l criteria are identified, i Amend. 73 17J-16 Nov. 1982

l 3.4 Identification and control of design interf aces is accomplished by the Cognizant Engineer and documented by means of System Design Descriptions (SDDs), Component Specif ications, and interf ace Control Documents (ICDS). The f undamental control document f or f unctional interf ace data is the SDD, which identifies the system interf aces including referencing supporting control documents (e.g., ICDS) .nd together w ith the ICDS, compl etely def ines requirements f or every Interf ace w ithin a system.

ICDS are drawings or documents that identify the physical interface characteristics necessary to ensure compatibil ity between mating pieces of eq u i pment. ICDS are distributed to, and used by, project participants for assuring compatibil ity of system and/or components. Interf ace requirements are transmitted to Interf acing organizations, and concurrence obtained prior to issue. Proposed changes are coordinated with interf acing organizations prior to implementation.

3.5 The preparation of design documents (SDDs, ICDS, specifications, and draw ings ) Invol ves input f rom appropriate technical discipl ines incl uding sy stem, saf ety, stress, thermal, fl uid fl ow, mechanical, material s and process, electrical, control, manuf acturing, and qual Ity engineering.

Qual ified representatives of these discipl ines review and approve design documents before issue. Additionally, drawings are checked f or dimension accuracy by an independent checking f unction before issue.

l SDD drawing and specification changes are reviewed and approved by the same gs discipl ines as the original issue. A method is used by the releasing f unction

( to check the approvers and the f unctions they represent to assure alI the same discipt inos review and approve revisions.

3.6 Verif ication of designs is perf ccmed by formal and independent design reviews at various stages of the design to insure that all significant f actors af f ecting perf ormance, rel labil ity, saf ety, operabil Ity, and maintainabil ity of a component cr system are properly considered. The Design Review Board is establ ished on an ad hoc basis to provide an expert eval uation and is com-prised of a Chairman and special ists in Design, Material s, Saf ety, Qual ity Assurance, and other discipiines. Members of the board are selected f rom any organization on the basis of their knowledge of the subject but are not responsible for the work. Action items are assigned during the meeting, and the followup is provided by the Design Review Board Administrator to assure that the action is taken and the action Items closed out. Analyses and calcu-lations having significant ef fect on the design are subject to verification.

The compl eteness, adequacy, and appropriateness of assumptions, input data, and analytical or calculation method used are evaluated. Certain aspects of designs are verified by test to supplanent independent design reviews, in those cases where the adequacy of a design is verified by a qualification test, testing is identified and documented. Testing is conducted using a prototype unit under the most adverse design conditions for which an item is required to perf orm its saf ety function. The resul ts of design verif ication are cl early documented, with the verif ier identi f ied. Documentation of the results is auditable against the verif ication methods identif ied.

Amend. 73 Nov, 1982 17J-17

The design engineer, assisteo by the materials process, and quality engineering, is responsibl e f or determining the appl icabil ity of material s, parts, and equipment used in the design. This selection of hardware is reviewed during the design review.

One of the basic purposes of the design review system is to find and correct errors and deficiencies, prior to the release of the engineering document for procurements, manuf acture, construction, or to another erganIzation for use in other design activities, in all cases, the design verification is completed prior to relying upon the components system, or structure to perform its function. Documentation of the def iciency, and the resulting corrective action, are incl uded in the records of the design review.

l 3.7 The methods f or the collection, storage and maintenance of desis documents, review records, and related engineering data are describsd in Section 17.0 of th is appendix.

4.0 PROCUREMENT DOCUMENT CONTROL l 4.1 ESG uses a system of procedures which describe the sequence of actions to be taken in preparing, review ing, approving, and controll ing procurement documents. The basis for all procurement actions is the Purchase Requisition, which is prepared by the organization requiring the material, service, or com-ponent being purchased. Each Purchase Requisition is reviewed and approved by qual if led Qual Ity Assurance Department personnel to assure that correct and compl ete qual Ity requirements are stated or ref erenced. This review is documented. Drawings, specifications, design reports, and other documents which are ref erenced in the Purchase Requisition are reviewed and approved as described in Section 6.0 of this appendix. The requirments of the Purchase Requisition are transferred to a Purchase Order, which is of fered to the suppller. Purchase Orders are reviewed by Qual Ity Assurance Department personnel to assure no changes of requirements f rom the Purchase Requisition.

l 4.2 Purchase Orders f or structures, systems, and components Identify appropriate requirments, which must be addressed in the suppl ler's qual Ity assurance program description. The supplter's program is reviewed against contract requirements and approved by qual ifled Qual ity Assurance Department personnel prior to start of activities af fected by the Quality Assurance Program.

The Purchase Order and its ref erenced documentation contain al I necessary design basis technical inf crmat l on. They addltional ly identi fy al i documenta-tion to be prepared, maintained, and submitted to ESG fcr review and approval .

The Purchase Order al so identifies those records which must be retained, control l ed, maintained, or del ivered to ESG. Provision is mado in the Purchase Order to ensure ESG's right of access to the suppl fer's f acil Ities and records f or source surveil lance and audits.

l 4.3 The Purchase Requisition - Purchase Order cycle described here is al so used to process changes and revisions to the contract. The same review and approval is required of changes as is required of the original Purchase Requisition and Purchase Order. Procurement documents pertaining to spare or repl acement parts are treated in the same manner as that uspd f or initial production parts.

Amend. 73 17J-18

l 4.4 Applicable elements of 10 CFR 50, Appendix B, are applied to suppliers by invoking. government or industry Qual ity Assurance standards in whole or in part, or by inserting specif ic qual ity requirements in the Procurment Speci f Ications.

Procurement specif Ications contain the design basis technical requirements; identif ication requiroments of components, subcomponents, and material s; applicable codes, standards and specifications; test and inspection require-ments; and appropriate special process requirments covering critical processes such as welding, braz ing, heat treatment, eloctropi ating and thermal surf ace coating, cleaning, and nondestructive examinations. Appl icabl e regulatory technical requirements are incl uded in the procurment specif ica-tions rather than specify ing these by ref erence to regul atory documents.

5.0 lNSTRUCTION

S. PROCEDURE

S. AND DRAWINGS l 5.1 Policy, procedures, and Instruction documents are prepared to cover actIvitles af f ectIng qual Ity. These qual Ity-of fectIng activities loci ude management, design and engineering, procurement, qual ity assurance, and manuf acturing. Policies, procedures and instructions are collected and issued in oporating department and qual ity assurance manual s. The manua! s contain provisions f or preparation, review, control, and rev.Islon of procedures and instructions comprising the manual .

The manual s containing procedures and instructions f or qual ity-af fecting actlvItlos at Energy Systms Group are:

Standard Operating Pol Icles Manual 00RP Program Management Directives Manual Engineering Management Procedures Manual Corporate Material (Purchasing) Procedures Manual ASW Code Section iII Manual Qual ity Assurance Operating Procedures Manual Manuf acturIng Manual Methods f or comply ing w ith qual Ity assurance criteria appi Icabl e to the ESG scope of work aro def ined in the preceeding manual's. A correl ation of procedures, pol icles, and instructions f rm these manual s w ith the criteria of 10 CFR 50, Appendix B, is given in Figure 17J-4, and a summary description of the contents of each document referenced in this figure is given in Attachment 17J-l to th is appendix.

Acceptance criteria for important activities def ined by the af ormentioned procedures and instructions are a part of each procedure, as appi Icable. For exampl e, document formats and content are specif ied, as are release, approval, and distribution control req ui rements.

l 5.2 The requirments f or activities af f ecting qual Ity as well as appropriate quantitative and qualitative criteria for determining that important activi-ties have been satisf actorily accompi Ished, are specif led in instructions, procedures, and drawings, including the f ollowing types of documents:

(1) Design Specif ications, (2) SDDs, (3) Procurement Documents, and (4) Test Procedures.

Amend. 73 j73_jg Nov. 1982 l

l 5.3 Provisions f or preparation, content, quantitative and qual itative roquirements rev iew, revision, and control of drawings are contained in Sections 3.0 and 6.0 of this appendix. They provisions f or manuf acturing end inspection Instructions and procedures are . antained in Sections 6.0, 9.0, 10.0, and 13.0 of th is appendix.

6.0 DOCUMENT CONTROL l 6.1 Documents, such as design specif ications, design drawings, computer programs, manuf acturing drawings, equipment specif ications, construction and preoperational test specif ications, material processing specif ications, and nondestructive examination procedures, are prepared, reviewed, approved, and issued in accordance w ith written procedures. Review methods may vary from a series of f ormal ized reviews by a Design Review Board to individual reviews by per sonnel from involved organizations and the Quality Assurance Department.

Organizations responsible f or review and approval functions for a specific type of document are identif ied in a written procedure. Originals, prints, and/or reproducibles of these documents are controlled by the Engineering organization, which releases, distributes, stores, and maintains f iles and records of these documents. Document changes are prepared, reviewed, and approved in accordance w ith appl Icable procedures, only under the authority of the organization or f unction that prepared, reviewed, and approved the original. Drawings and drawing changes distributed to Manuf acturing and Qual Ity Assurance f or items being f abricated by the Energy Systems Group require return of a document recolpt to Engineering, as evidence that the documents were received by those organizations. Draw ings f or manuf acturing and inspection purposes are f urther controlled through the Manuf acturing Production Control Station. The personnel of this organization Insure that correct draw ings and revisions thereto are available f or manuf acturing and inspection planning, as well as f or the subsequent manuf acturing and inspection operations. Periodic audits are conducted to verify that active documents are in use and obsolete issues have been removed f rom use. The Engineering data base, containing the latest issues of drawings, specif ica-tions, and design basis documents, is updated daily. Terminal s are avail able to al l f unctions f or assuring that obsolete issues are not used.

l 6.2 Procurement documents are controlled as described in Section 4.0 of this appendix. Source and receiving inspection documents are controlled as described in Section 7.0 of this appendix.

ESG Qual ity Assurance Manual s and department operating procedure are distributed and controlled in accordance with a procedure contained within each manual.

l 6.3 Manuf acturing, inspection and testing instructions, and testing procedures are designated in Manuf acturing Production Orders (MPGs) by instruction or procedure number and by appi icable revision letter or revision number. The instructions and procedures either accompany the MPO or are maintained avail able at the location where the work is perf ormed. Changes to MPOs, necessitated for any reason, require the prior review and approval of Qual ity Assurance, as do changes in manuf acturing inspection and test instructions, and test procedures.

O Amend. 73 Nov. 1982 17J-20

6.4 A listing is periodically issued of design documents and their revisions V which incl udes system design descriptions, drawings, specif ications, engineer-Ing reports, engineering orders, nonconf ormance reports, manuf acturing process procedures, test procedures, and nondestructive examination procedures. The

, administrative polIcles and procedures l Isted in Figure 17J-4 are contained in the CRBRP Qual Ity Assurance Program Index. These I istings are used to assure that obsolete issues of the aforementionod documents are not used.

l 6.5 Assurance that receiving and source inspection is performed to the latest purchase order change is achieved through a system that routes purchase requisitions and orders and changes thereto to the Quality Assurance Engineer-Ing Department function. At the time that a change is received by this organiz ation, it is reviewed for quality requirements, and the source or receiving inspection instructions are revised as necessary. Copies of revised Inspection instructions and the change orders are sent to the Receiving and Source Inspection f unctions.

l 6.6 Assurance that approved changes are incl uded in specif ications, draw ings, and procedures prior to their Implementation is achieved through review and approval of the implementing documents (purchase requisitions and manuf ac-turing travelers) by Quality Assurance Engineering Department personnel and enf orcement actions of the Q A Department inspection f unctions. Qual Ity Assurance Engineering Department personnel review and approve purchase requisitions and manuf acturing travelers prior to their release to !nsure that the correct revisions of specifications, drawings, and procedures are given p therein. Af ter issuance, purchase orders are reviewed to assure that there are no unauthorized changes f rm the purchase requisition. Source, Receiving, and in-Process inspection Inspects to the requirernents document revision given in the purchase orders and travelers.

6.7 As-buil t draw ings and documentation are a requirement of contracts f or components and are required to be del Ivered w Ith the item. Qual Ity Assurance source surveillance and document review prior to authorizing shipment to the GBRP site assures that as-bull t documentation is received in a timely manner.

7.0 C_0NTROL OF PURCHASED MATERIAL. EOUlPMENT. AND SERVICES l 7.1 Each suppl ier of material s, structures, systems, and components is eval uated to assess his capabil ity to provide acceptable services and products. Eval uation of major item suppl lers for which there is no recent capabil Ity inf ormation is performed by a team, consisting of representatives of Purchasing, the Program Of f ice, Qual ity Assurance, and Manuf acturing Departments as appropriate. Representatives of Design Engineering, Material s and Processes Engineering, and other units of the Engineering Department participate in the eval uation as necessary.

The detail s of the eval uation incl ude reviews of past perf ormance, eval uation of procedures and capabil Ity descriptions provided by the suppi f er, survey i of the suppl ler's f acil Ity and Qual ity Assurance Progran in operation, and/or experience of other CRBRP participants w ith the suppl ier. The snluation considers the suppller's capability to supply a product which satisfies all Resul ts of th is eval uatIon are documented and retained on f Il e O req ui rement s.

at ESG.

Amend. 73 17J-21 Nov. 1982

l 7.2 ESG Qual ity Assurance Department personnel perf orm surveil lance of suppllors during fabrication, processing, inspection, testing, and shipment of products. These survellIance actIVItles are pl annod and perf ormed in accord-ance w ith written procedures. The plans provide Instructions which specify the characteristics or processes to be witnessed or verified, the documenta-tion required, and the acceptance criteria which must be met. SuffIclont survelliance is perf ormed to verify that qual Ity is achieved in items which cannot be inspected upon receipt. This surveillance ends with written approval to ship the item to ESG or the construction site, given by appro-priate Qual Ity Assurance Department personnel .

l 7.3 Receiving inspection is perf ormed on products del Ivered to ESG to assure their acceptabil Ity prior to use. This inspection is carried out in accor-dance w lth wrltten inspection pl ans. The product is eval uated to determine that it is properly identif led, that it meets inspection criteria, that necessary inspection and testing records are included with the product, and that the accepted product is identifled as to its acceptabil lty before being rel eased f or use or storage. Nonconf orming items are segregated, control led, and clearly identif led pending proper disposition. ESG Qual Ity Assurance Department personnel provide written Instructions f or receiving inspection of items purchased by ESG and del Ivered directly to the construction site frcm the suppller.

7.4 ESG requires that the suppl ler f urnish, as a minimum, certif ! cations that idontify (e.g., oy the purchase order number) the product and the gclfIc requirements (codes, standards, specif Ications) met by the item. l'he supplle-is f urther required to submit a report, Identify ing any requirements which have nc. been met, and indicating his disposition of such nonconf ormances.

Certif ications and test reports are reviewed and approved by appropriate Qual ity Assurance Department personnel . Acceptable certif icates of compi lance, and data reports as required, are provided to the pl ant site with equipment del ivory.

7.5 Procurement of spare or repl acement spare parts w Il l be conducted under the qual ity assurance progrun that is in of fect at the time of oraer pl acement. Technical req u i rements, if not the same as f or the initial plant item, wIlI be eval uated to insure that they are eaual to or better than those f or the initial pl ant item.

7.6 "Of f-the-shel f" items are subjected to special receiving or source inspections f or critical character i stIcs. Specif Ic Inspection instructions are prepared on a case-by-case basis to accommodate the unique characteristics and use of each item.

l 7.7 Suppl lors' certif Icates of compi lance are val idated by an establ Ished program of audits, independent inspections, and surveillance and overchecks.

This is accompl ished using Itinerant or resident Qual ity Assurance site representatives or source inspectors, hol d point release, and suppl ler audits.

Additionally, procurement specifications require supporting technical data for certi f icates of compl iance, and these data are rev iewed f or compl eteness bMore use of any item.

O Amend. 73 N v. 1982 W -22

8.0 IDENTIFICATION AND CONTROL OF MATEMALL_f.N1TS. AND COMPONEN 3 ,

l 8.1 For purchased items, ESG delegates execution responsibil Ity for activities of identif Ication and control of material s, parts, and components to suppl iers and assesses the ef fectiveness of these suppl ier activities, as described in Section 7.0 of this appendix.

8.2 For items f abricated by ESG, procedures and instructions establish identif ication and control requirements of material s (incl uding consumables),

parts, and components, f rom design through f Inal assembly, identif Ication requirements begin w Ith spect f Ications and draw Ings. Drafting procedures require that notes and location Indicators appear on drawings that specify identification Information and exact location. Specifications describe how the identif ication is to be accompl ished (e.g., name pl ates, impression stencil, el ectrochanical etch ing). Identif ication requirements f rom draw Ings and specif Ications are ref erred to on the Manuf acturing Production Order (MPO).

8.3 Traceabil ity of parts, assembi les, components, and structures to drawings l and specif ications is achieved through the practice of the drawir.g number becoming the part number. Completed component and structure name pl ates reference the design or component specification number, in manuf acturing and assembly, the MPO, which ref erences draw ings and specif Ications and directs the identif ication to be appl led to the items, provides data for traceabil ity to nonconf ormance reports, special process procedures, inspection procedures, purchase orders, and mil I test reports.

l 8.4 Any adverse ef fect of the location or method of identif ication on qual Ity or function of items identified is prevented by specifying these requirements in engineering draw ings and specif ications. These documents are reviewed and approved by special ists in stress, material s, processing, manuf acturing, and qual Ity assurance, to assure identif ication markings do not af fect qual ity and function.

8.5 Verification of the correct identification of materials, parts, and components is perf ormed by the Inspection f unction of the Qual ity Ascurance Department f or ESG f abricated items. Qual ity Assurance Engineering Department personnel are responsible for issuing instructions to inspection for this verification to appear on the MPO. Upon completion of f abrication and assembly, Qual ity Assurance Engineering Department personnel review the MPO to assure the steps specifying identif ication and its verification are initiated and stamped to show completion of these operations. For suppller fabricated items, Identif ication verif ication is accompi ished by ESG Itinerant or resident Q A representatives.

9.0 CONTROL OF SPECIAL PROCESSES l 9.1 For purchased items, ESG delegates execution responsibil ity for special process control activities to suppl lers and assesses the of fectiveness of suppl ler special process control activities, as described in Section 7.0 of this appendix.

%./

i Amend. 73 17 J-23 Nov. 1982

9.2 For items f abricated by ESG, special processes, incl uding but not l imited to wel ding, braz ing, heat-treating, cl eaning, bonding, coating, sol dering, pl ating, hard surf acing, forming, cl ean room operations, and nondestructive testing are controlled to the degree required by appl Icable codes, standards, specif ications, and regul ations. This control is accompl ished by several means:

1) Fabrication Procedures are written by Manuf acturing Engineering, and reviewed and approved by Design Engineering and Qual ity Assurance.

Nondestructive examination procedures are reviewed and approved by certif ied NDE Level l l i Exam inors.

2) Detail instructions in the Manuf acturing Production Order (MPO), which serves as ESG's shop traveler, are written by Manuf acturing Planning and reviewed and approved by Qual ity Assurance.

When Processing Procedures are used, they are made part of the MPO by reference.

9.3 Procedures, equipment, and personnel perf orming special processes are qual If ied and certif ied by Qual ity Assurance Department personnel .

Qual if ication is accompl ished in accordance w ith appi Icabl e codes, standards, specif ications, or internal requirements. Qual if ications are reviewed and approved by Qual ity Assurance.

Speci al processes are perf ormed by trained, qual if ied personnel working to .

l written qual if ied instructions using qual if ied equipment. Evidence of perf ormance or verif ication is recorded on the MPO which accompanies each structuro, system, or component during manuf acture. Evidence of perf ormance is either recorded or verif ied by qual if ied Qual ity Assurance personnel .

9.4 Qual if ication records of procedures, equipment, and personnel for l

perf orming special processes are establ Ished, filed, and maintained current in compl iance w ith written ESG procedures. Periodic audits of these records are perf ormed by Qual ity Assurance to ensure their adequacy.

10.0 INSPECTION l 10.1 For purchased items, ESG del egates execution responsibil Ity for inspec-tion activities to suppl lers and assesses the ef fectiveness of these inspec-tion activities, as described in Section 7.0 of this appendix.

10.2 For items f abricated at ESG, inspections examinations, and qual ity l verif ication testing of systems, structures, and components are perf ormed by inspection and Test Unit personnel of the Qual ity Assurance Department. The manager of this f unction reports directly to the Qual ity Assurance Director, who reports directly to the President of Energy Systems Group, tnus providing the inspection f unction f reedom ef fectively to perf orm its responsibil ities.

10.3 The shop travel er f or the control of manuf acturing and inspection l activities is the hbnuf acturing Production Order (MPO). The MPO is a single document that authorizes and directs both manuf acturing and inspection activities. For inspection, the MPO serves as the test and inspection checklist.

Amend. 73 17J-24 Nov. 1982

l l

, The MPO specif ies the characteristics to be inspected and 1he specific point in the manuf acturing process where the inspection must be accompi ished. It al so specif ies, by line entry, the specif ic department and group responsibl e f or perf orming the operations, incl uding inspections and tests. Inspect ion points are sel ected by Qual Ity Assurance Department staf f.

Acceptance and rejection criteria and the description of the method of in spect i on, incl uding any special requirements such as use of particul ar equipment, are specif ied on the MPO or are contained in documents specif ically referenced by the MPO. These are entered on the MP0 by Qual ity Assurance Department staf f.

The inspector who perf orms the inspection operation stamps the MPO entry when he completes an inspection activity. When the manuf acturing and inspection ef f ort on the MPO is completed, the MPO is reviewed by the Qual ity Assurance Engineering personnel of the Qual ity Assurance Department to verify and certif y acceptabl e compl etion of al l specif ied manuf acturing, inspection, ano test operations.

Each system, str uct ure, component, or subtler detail is f abricated against an indiv idual MPO. Establ ished procedures require that a copy of each draw ing and procedure ref erenced on the MPO be at the manuf acturing and inspection work station f or use by personnel during the work operation.

l 10.4 Inspectors are trained and indoctrinated, as required, to assure prof iciency in their assignments. in addition, nondestructive examination personnel are f ormal ly trained, qual if ied, and certif ied to SNT-TC-1 A as suppl emented by Section lli of the ASME Boiler and Pressure Vessel Code (see paragraph 2.10) .

l 10.5 Modif ications, repairs, and repl acements are f abricated under the same Manuf acturing-Inspection control sy stem as new Itans, and receive the same rev lews and approval s as original item fabrication.

l 10.6 Hol d points f or w itness by the authorized Code Inspector and/or customer representatives are provided f or and estabt Ished, as required by these agencies, on the MPO by Qual ity Assurance Engineering personnel, prior to rel ease f or f abrication.

10.7 Procedures require Qual f ty Assurance Department perso.'nel monitoring of special processes, where direct inspection is not possible. Process proce-dures are used which specify control measures and acceptabil ity requirements.

11.0 TEST CONTROL l 11.1 For purchased itemt, Energy Systems Group del egates execution responsi-bil ity f or test programs to suppl lers, and assesses the ef f ectiveness of these programs through surveil lance actions, as described in Section 7.0 of th is appendix.

k ,

)

Amend. 73 l N v. 1982 17J-25

l 11.2 For items produced by Energy Systems Group, test programs are Identified by Design Engineering, as appropriate, to demonstrate that items will perf orm satisf actorily in service. Testing is accompl ished in accordance with written and control led procedures. These procedures are prepared by Engineering or Qual ity Assurance Department personnel frm the group or unit responsible for conducting the test. They are reviewed and approved by the cognizant Qual Ity Assurance Department personnel having responsibil Ity that qual Ity and qual ity assurance requirements are met and by Progrm Of f ice cognizant engineers hav ing responsibil ity that technical requirments are aet.

l 11.3 Test procedures incl ude appropriate requirments f or test articl e Identif Ication, test purpose and objectives, test prerequisites, test condition lImits, instruments and cal ibration, equipment, environmental warnings and cautions, authority for test restart af ter interruptions, accept /

reject criteria, data type, method of documentation, and records collection, and storage requirements, Qual ity Assurance Department, authorized inspection, or custmer w itness requirements, personnel qual if Ication requirements, and step-by-step procedure requirements with provision f or perf ormer signof f and Qual ity Assurance Department w itness verIf Ication signof f or stamp.

[ 11.4 Test data are analyzed by qual if led personnel and a written report prepared in which results are documented, eval uated, and the acceptabil Ity of the item for perf orming its f unction satisf actorily in service stated.

11.5 Tested items that have subsequently been modif led, repaired, or have l been repl aced in whole or in part are retested to the original test require-monts. If the repair, modif ication, or repl acement invol ves a design change and modif ied testing requirements, all design and test documents are revised prior to this work in accordance with the procedures and control described in Sections 3.0, 5.0, and 6.0 of Appendix J.

12.0 CONTROL OF MEASURING AND TEST EOUIPMENT l

12.1 For purchased items, Energy Systems Group delegates execution responsibil ity for control of measuring and test equipment to suppiIers and assesses the ef fectiveness of these activities, as def ined in Section 7.0 of this appendix.

l 12.2 For items produced or tested by Energy Systems Group, procedures define the requirements and responsibil ities f or cal ibration, cal Ibration standards, and control of measuring and test equipment used f or f abrication, testing, and inspection. Tho Qual ity Assurance Department has the responsibil ity for implementing and maintaining the program for cal ibration and control of measuring and test equipment. Cal ibration operations are conducted by the Qual Ity Assurance Department, Engineering Department, other Rockwell International DivIslons, and qual if led suppl Iers.

[

12.3 Each liem of measuring and test equipment is given a unique serial number, and the records containing cal ibration and test data are identif ied and f lied by that serial number.

O Amend. 73 Nov. 1982 17J-26

The cal lbration system procedures require that measuring and test equipment be

(--) cal ibrated at specif ied Interval s and that these Intervals be based on usage, stabil Ity, accuracy, and h istory. Cal ibration procedures are prepared by the cal Ibrating f unction and are reviewed and approved by cognizant Qual ity Assurance management.

The complete cal ibration status of measurement and test equipment is main-tained, using a computerized cal ibration inventory and recall system, which provides the basic cal ibration system control, by forcing a l Isting of equip-ment requiring cal Ibration and the periodic recall notif ication to the instrument user and cal ibration f unction.

Measuring and test tool s and instruments are labeled to show cal Ibration status, i.e., out of use, Indication only, and next cal Ibration due date f or in-use equi pment. Out-of-use tool s and instruments are l abeled " Cal Ibrate Bef ore using. "

12.4 Cal ibration procedures specif ical ly state that the cal ibration standards l

against which the measuring and test equipment is cal ibrated have an error no more than one-fourth of tolerance of the equipment (including standards) being cal ibrated, unl ess proh ibited by the state-of-the-art. A greater error may be permitted af ter discussion between management of the using organization and

• the Manager of Inspection and Test.

Energy Systems Group maintains working standards against which measuring and test equipment are cal ibrated. Working standards are cal ibrated f or trace-(-'s g ) abil ity to the National Bureau of Standards. This is accompl ished by procuring standards or cal ibration services directly from the NBS or from suppl lers which, in turn, can demonstrate N3S traceabil Ity. Where N3S standards do not exist, cal ibration of standards is accompl ished by such .

methods as inter-laboratory comparisons or internal development of a standard.

12.5 When discrepancies f rom accepted tolerance are found f or measuring and l test instruments during cal ibration, this f inding is reported to the Manager of the using organization who ini1;ates an investigation of items inspected since the prev ious cal ibration. The val idity of previous inspection perf ormed w ith the suspect instrument is eval uated, and the resul ts, along with appropriate actions, documented f or the record and f ol low-up.

13.0 HANDLING. STORAGE. AND SHIPPING l 13.1 For purchased items, Energy Systems Group delegates execution responsi-bil Ity for cleaning, handl ing, storage, and shipping activities to the suppl iers and assesses the ef fectiveness of these activities, as def ined in Section 7.0 of this appendix.

l 13.2 For items produced by Energy Systems Group, special handl ing, preserva-tion, storage, packaging, and shipping roquirements are specified by packaging engineering special ists. Any special cleaning requirements are specified by manuf acturing pl anning. Operations invol ving thesc activities are accomp-

, lished by qual if ied individual s, in accordance with written work and inspec-tion instructions. Handl ing and cleaning instructions are detailed in proce-s dures ref erenced in the Manuf acturing Production Order (MPO).

Amend. 73 17J-27 Nov. 1982

All specif ications and instructions covering cleaning, handl ing, preservation, storage, packaging, and shipping refleci design and specification requiranents of the material, components, or system being processed. Special attention is given to prevention of loss, damage, or deterioration due to adverse environ-mental conditions, such as temperature or humidity.

By the time of shipment to the construction site, Instructions f or handi Ing and storage are transmitted to the Constructor.

14.0 _lflSPECTION. TEST. AND OPERATING STATUS l 14.1 For purchased items, Energy Systems Group delegates execution responsi-b il ity f or identifying and maintaining Inspection, test, and operating status.

Assessment of the of fectiveness of inspection, test, and operating status is obtained fran surveillanca activities described in Section 7.0 of this appendix.

l 14.2 For items produced by Energy Systans Group, the inspection and test status of structures, systems, and components, throughout manuf acturing, is identif led by the util ization of a shop traveler, known as a Manuf acturing Production Order (MPO). The MPC is a comprehensive manuf acturing, inspect i on, and testing planning document written by the Manuf acturing Planning Unit of the Manufacturing Department. It is reviewed and approved by Qual Ity Assurance Department personnel to assure that adequate inspection and test control s are incl uded, inspections and tests are performed or witnessed by qual ified Qual ity Assurance Department inspection personnel, and the status of the inspection er test is indicated on the MPO with the inspector's stamp.

Finished Items z.I so receive the Qual ify Assurance Department inspector's stamp; or, if too smalI to be stamped, are bagged and tagged wIth the status indicator appl led to the tag.

Qual Ity Assurance Department personnel perform periodic and f inal reviews of the MPO, to assure that all inspections and tests have been performed and their status properly indicated. Thus, bypassing of Inspections, tests, and other critical operations is precl uded. Appi Ication and ranovel of Inspection status indicators, such as tags, markings, label s, and stamps are perf ormed or w itnessed by Qual ;ty Assurance Department personnel . Wel ding stamp indica-tions are appl led by the wel der, as required by the MPO and are verif led by Qual Ity Assurance Department personnel .

l 14.3 The status f nonconf orming, inoperative, or mal functioning structures, systems, or components is identi f led by Qual ity Assurance Department personnel to prevent Inadvertent use. Detail s of the control system are described in Section 15.0 of G Is appendix.

15.0 h0NCONFORML 4 MATERI ALS. PARTS. OR COMPONENTS 13.1 For purchased items, Energy Systems Group delegates execution responsi-l bil Ity for nonconf orming material s, parts, or components control measures to suppllers. Assessment of the ef fectiveness of these measures is obtained f ran surveillance activities described in Section 7.0 of this appendix. Noncon-formances that af f ect saf ety-rel ated f unctions or util Ity that are proposed 9

Amend. 73 n v. m2 17J-28

I O for " accept as is" or " repair" dispositions are submitted to Energy Systems b Group f or approval; and if ESG approval is granted, then to the customer for approv al .

l 15.2 For Energy Systems Group f abricated items, procedures are implemented whereby nonconf orming item identif Ication, documentation, segregation, review, and disposition are perf ormed. The administrative system for nonconformance control routinely provides f or notif ication of appropriate af fected organiza-tions (Manuf acturing, Purchasing, Engineering, Qual ity Assurance Engineering LMFBR Programs) of the existence of nonconforming conditions.

The shop traveler, or Manuf acturing Production Order (MPO), described in Section 10.0 of thIs apper. dix initIatos idontifIcation of a nonconformance of

inprocess items, wIth the Quality Assurance Department inspector af fixing his

! discrepancy stamp to the I ine items on the MPO for the Inspection operation.

ThIs Idontif Ication of the Item, the nonconf ormance, and the acceptance i criteria involved are transf erred to a ncnconf ormance report f orm by Qual ity Assurance Department personnel, and the serial number of this report is transcribed onto the MPO. The nonconformance report form and the procedure controlling its use provide for documentation of the disposition of the nonconformance, signature approval of individual s authorized to determine dispositions, and distribution of the report. A simil ar approach is used f or suppl ler nonconf ormances detected at receiving or source inspection.

Nonconf ormance procedures def ine the Individual s and groups responsible f or the disposition of nonconforming items.

l 15.3 Nonconf orming items are physically segregated f rom acceptable items in control led access hol d roms. The hold roms are controlled by the Qual ity Assurance Department inspection Unit, items too large to be placed in the roms are prminently tagged to identify their hold status. Rel ease f rom hol d areas or removal of hol d status tags can only be perf ormed by appropriate Qual Ity Assurance Department personnel, af ter receipt of an approved noncon-formance report.

Nonconf ormances in services w il I normally be written against af fected h ardw are. Where that is not practical (e.g. , def ective computer codes), the Corrective Action Request (see Section 16.0) is used to control firther operations and/or hardware as appropriate, and to track resol ution.

l 15.4 Repair and rework operations of material s, parts, components, systems, and structures is accompl ished by a revision to the original MPO. This revision of the MPO is prepared, reviewed, and approved in the same manner as the initial issuance, which is described in Section 10.0 of this appendix.

This revision specifies the repair, rework, and inspection procedures to be used. The inspection methods used cre, as a minimum, those used f or the original inspection.

l 15.5 Nonconf ormances that af fect saf ety-rel ated f unctions or util Ity of the items that are proposed f or " accept as is" or " repair" dispositions are submitted to the customer for approval . Approved nonconf ormance reports, with O

V the dispositions, " accept as is" or " repair", are maintained by Qual Ity Assurance, and are submitted with the item at the time of shipment, in accordance w Ith contract requirements.

Amend. 73 17J-29 Nov. 1982

l 15.6 Nonconf ormance reports are summarized and analyzed for trends at least monthly by QA Audits and Controls and Quality Assurance Engineering and the summary is di stributed to managers of Qual ity Assurance, Manuf acturing, and Purch asi ng. Nonconf ormance reports are submitted to the custmer as required by contract.

16.0 CORRECTIVE ACTION l 16.1 For procured items, Energy Systems Proup delegates execution responsi-bil ity to suppl Iers f or establ ishing and maintaining corrective action measurec. Assessment of the of fectiveness of these measures is obtained f rm the supplier surveillance activities provided for in Section 7.0 of this appendix.

l 16.2 For activ ities w ith in Energy Systems Group, a documented corrective action system, under the control of the Quel ity Assurance Department, is establ ished in accordance w ith procedures f or handl ing nonconf ormance to technical rquirements and technical procedures. Technical requirments are those contained in design draw ings, specif ications, fabrication procedures, and inspection and test procedures. Technical requirment nonconf ormances, theref ore, are refl ected by hardware nonconf ormance. Tcchnical procedure requirements are those that guide the general processes of documenting and disseminating design, perf ormance, conf Iguration, procurment, manuf acturing, and inspection requirements. These technical procedures are those in the Qual ity Assurance Manual s and f unctional manual s of qual ity-af fecting organ iz at i ons.

l 16.3 CorrectIvo actions f or technical requirment v f olations are an integral part of the nonconforming item, system described in Section 15.0 of this appendix. Corrective action f cr technical procedure nonconf ormance are defined in procedures covering audits and the basic corrective action system.

Corrective action is initiated during (a) nonconformance evaluation and resol ut ion and (b) following the determination of a condition adverse to qual Ity, to precl ude reoccurrence. Appropriate completion periods are assigned as parts of the corrective action commitments. To assure timely resol ution, corrective action completion dates are monitored by the Qual ity Assurance Audits and Control s f unction; and, in the event of a del Inquency, these f acts are brought to the attention of the management of Qual Ity Assurance and the af fected organizations, impimentation of corrective action is verifled by Quality Assurance, and thIs is the basis f or close-out of corrective actions.

All corrective actions are based on conditicas that do or may adversely af fect q ual i ty . These conditions and their causes are strnmarized in monthly reports to management, along with status of the corrective action impimentation

( i. e. , compl ete, on schedul e, or del inquent).

O Amend. 73 17J-30 Nov. 1982

i 17.0 OUAllTY ASSURANCE RECORDS 17.1 Pol icies, pl ans, and procedures have been implanented by Energy Systems Group to obtain appl Icable qual ity assurance records in ANSI N45.2.9 (1974).

These pol icies, pl ans, and procedures al so provide for storage and preserva-tion of the qual ity assurance records while at ESG. Generic qual Ity record categories have been Identified and organizational retention responsibil Ity assigned f or these. At the time of contract award for equipment items, a specif Ic I ist of qual Ity records to be obtained is prepared based on the generic Iisting. Qual Ity records incl ude system design descriptions, speci-fIcations, draw ings, design reports, design verif ication test procedures and reports, purchase orders, design revlew reports, manuf acturing process procedures and instructions, material test reports, personnel and process qual if Ication resul ts, nonconf ormance reports, audits, inspection resul ts, acceptance test reports, cal Ibration procedures and records, and qual ity surveil l ance reports. The records program procedures al so provide f or responsibil ities f or its management and operation, records col lection, def inition of terms unique to the records program, verif ication of such characteristics as I eglbil Ity, compi eteness, inventory control, and transf er to the Owner.

17.2 The organizations involved in the qual Ity records program are Qual Ity Assurance, Engineering, Purchasing, and Manuf acturing. Responsibil Itles of these organizations f or specifying, generating, collection, verif ication, fil ing, storage, and preservation are given in appropriate procedures.

17.3 Inspection and test records for items examined contain the following information:

1) The inspection or test performed

2) The date and results (acceptable / unacceptable) of the inspection or test

3) A notation of the acceptabil ity of parts, assembi les, or operations

4) A signature or stamp of the Individual perf orming or verifying Inspections and tests

5) Notif ication that nonconf ormances exist, inf ormation rel ating to nonconf ormances, and disposition of the nonconf orming item, and specif ic repair or rework actions.

17.4 Record storage f acil ities and f fles minimize the possibil Ity of destruction by f ire, fl ooding, thef t, blodegradation, and deterioration by environmental conditions such as temperature, humidity, and corrosive f umes.

O Amend. 73 Nov. 1982 1

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

18.0 AUDITS 18.1 EXTERNAL AUDITS Energy Systems Group has an eudit program for auditing suppl lers of structures, systems, and components. Qual Ity Assurance Department personnel perf orm audit pl anning, schedul Ing, audit team selection, audit coordination and contact, report issuance, and f ol Iow-up to verify ImpimentatIon of offective corrective action. Audits are pl anned on an annual basis.

Unschedul ed audits may be perf ormed when deemed necessary. Audits are schedul ed, based on suppl ler activity status, to eval uate the ef fectiveness of suppl ler Qual ity Assurance Programs. Checkl ists are prepared to guide the conduct of audits. Personnel experienced in the conduct of audits are selected as audit team leaders.

The responsibility for the execution of audits within their own and subtler suppl lers' is delegated to suppl Iers in procurement documents.

18.2 INTERNAL AUDITS l 18.2.1 Internal qual ity assurance audits are conducted in accordance wIth pre-estabiIshed procedures and checktists. Personnel experienced in the conduct of audits perf orm the audits, or are team leaders when the team approach is used. Audit personnel are selected to prevent their having direct responsibilitles in the areas being audited.

Auditors document their f indings, and these f indings are reviewed with managers having responsibil ity for the area audited. At the time of this review, the af fected manager accepts a commitment to implement corrective action f or def iciencies, and a specif ied date when impimentation will be compl ete. Upon, notif ication of completion of a corrective action commitment, that area is re-audited to assure the corrections have been accompi Ished.

l 18.2.2 Audits are conducted of systems and procedures, processes, and products. The procedures audited are f Irst eval uated against code, standard, and contract requirments, and then the ef fectiveness of their impimentation to on-going work of fort is establ ished during audits. A review of documents and records is an integral part of all audits.

Qual ity audits are perf ormed by personnel f rm the Qual ity Assurance Department, or, in the Instance of team audits, personnel frm other f unctions under the direction of a Qual ity Assurance Department lead auditor certified to the requirements of ANSI N45.2.23.

Audits are scheduled yearly, in advance, to cover all elments where there is on-goi ng act iv i ty. The audit activity is initiated concurrent with initiation of conceptual design and is conducted throughout the l Ife of the program, so that discrepancies noted can be corrected early enough that end products will not be atfected.

O Amend. 73 17J-32 tiov . 1982

l 18.2.3 Audit results and status are reported monthly to program and f unctional managers. A summary report of problems af fecting timely corrective action is sent to the ESG President and executive level functional managers monthly.

Yearly summarization and analysis of CRBRP Audit Results are conducted and reported to management f or review and assessment and as required by contract req ui rements.

18.3 ACTIVITIES AUDITED Activities audited are those Quality Assurance program elements indicated in Figure 17J-3.

O l

O Amend. 73 17J-33 Nov. 1982

O OWNER I

LEAD RE ACTOR MANUFACTURER I

RE ACTOR MANUF ACTURER ENERGY SYSTEMS GROUP

' I SYSTEMS SERVICES COMPONENTS I

SUBSERVICES

= SERVICES SERVICES -

=

MATERIALS M ATERI ALS l

- COMPON ENTS I

i i MATERI ALS SERVICES l

Figure 17J-1. Overall Energy Systems Group Reactor Manufacturer Quality Assurance Program Functional Organization of Program Participation Amend. 73 Nov. 1982 0

17J-34

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OUALITY ASSUR ANCE DEPARTMFNT l

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I OUAL IT Y ASSUR ANCE OUALITY ASSURANCE b FNGINEERING ENGINEERING INSPECTION & TFST QUALIT Y ASSURANCE LMrBR PROGRAMS UTILITY & ENERGY MANAGER AUDITS & CONTROLS W PROGRAMS g MANAGER MANAGER

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PROGR AM MANAGE MENT OVALITY ASSUR ANCE PROGRAM ORG ANIZATION DOCUME NT AT ION AUDITS AND RE s tE WS CORRECTIVE ACTION ENGINE E RING HOLOS

1. Plann ng 1 Respons,tisaity and AutFonty 1. Pci c.es and Prwedures 1 Quao e v Aum es i J Quahey Anu<erwe Progam inden 2 Tresning and indu tonation 2 Qual ty Records 2 Vanagement Rev.ewws UNUSU AL OCCURRE NCE REPORTING '

3 Personnet Quaistication 3 0941 tv Status Rec,o ts DE StGN AND DE VE L OPME NT PHOCUH E UF N T MA%UF ACTURING F ARH C ATION AND ASSE MRL Y D,s.gn Pianning Piocurement Plann.ng Plann.ng Design Def.netioe and Control Protusement Requiremen's inspection and Test Plan 1 Design Criteria , yg yg g,, , , , , , , ,, g g,g

2. Cc=1es, Standards and Pras.t.ces 3 E nwncenng Stud,,s E waivation and Selection of Pewueement Control of P un esses 4 Parts. Ma rce. ass and Proresws S*'

1 F40 .ut.cin and Aswmt9v Proc esses 5 Des.gn Descoptrons 1 0,ner a, Reous,,m,n es 2 Pr oi ca Quant < 4 tion 6 Spec eNations. Draww.ngs and 2 Acccotatae Source List 3 Newstruct.ve E mamination Instruc teuns 3 Pre Anand E watuation 4 Clean.ng L Identificat-on 4 Interchanry of Suurg.e 8 Acceptance Cr.iena Capatal,ty info,mation inwtion and Tests 9 Interf ate Control 1 N Control of Configuration O",'**',N*N"'**"

p Document Revieww anrs Control 1 Contract Change Contros 2 p,0 ,qu ,,

3 Compietes tiem inspeci.on and Tesi W 1 Document Review's 2 As Built Vent. cation 4 In5Det t'on 51atus initsution G 2 Document Control Equipment Ca siteation and Standa<ds 5 Ce'tification 3 E nyneenng 04aww.ng Lisrs Source Sueveillane.e end Inspection Dwumnt Control Design Reviewvs g g , g g,,,g DeMopment Receiving inspect.on

' 1 E quipment E waluat.on Feature Reporting and Correct.we **'"8*^ ' " ' " * ' ' ' "

2 Documentation 2. Control of Inspect.on Measunnq and Tesi Action

3. Disposetioning of Received items E ausament 3

Controf of Nonconfosmeng Items 4 D.screpanc y E quipment Control of Received items Statistical Quality Contros and Analyses Control of Nonconforming items Corrective Action Handhng. Preservation, Pack ag ng. Storage and Shipping j 1. Handhng

2. Preservation. Pack ag ng and Stora

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< o Figure 17J-3e Major Elements of the Energy Systems Group Reactor Manuf acturer

a. Quality Assurance Program 00 N N La3 O O O

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d ESG Implementing Doceent cr Procedure Appendim B

& l ter lon Neber Title I. Organiration SOP M-10 Progre Management SOP Q-10 ESG Quality Assurance Progre QMP NI.21 Quality Assurance Pl ans

11. Qual Ity SOP A-01 ESG Policies and Procedures Assurance SOP M-10 Progre Managerent SOP Q-10 ESG Quality Assurance Progre SOP Q-16 Quality Assurance (QA) - Progre Support Functions SOP Q-12 Quality Assurance Progre Audits SOP Q-18 ESG Quality Records SOP Q-26 Product Integrity F90 No.16 Quality Assurance Managment Reviews FM) No. Il CRBRP Document Hold Status System F90 No. 20 DERP Training and Indoctrination F90 No. 27 OE RP Document Status System y EW 3-1 Engineering Doceentation Process p 04P 2.35 Case File Doceentation w Q AOP N1.00 Pref ace to Quality Assurance Manual N Quality Assurance Department Functions QMP Hl .01 QAOP Hl.03 VIslon Requirments fcr Quality Assurance Personnel Q AOP N1.21 Quality Assurance Plans QMP N1.23 Quality Status Reporis QMP N6.02 Qualif Ication and Certif Ication of Handestructive G3N2.4 Examinetton Personnel Q AOP N7.02 Quellf Ication and Certification of V isual and Dimensional Inspection Personnel ,

QMP N8.00 Statistical Quality Control Progre QMP N13.02 Quality Assurance Data Packages a

CS3M2.3 Training end Indoctrination Figure 17J-4 Quality Assurance Procedure index vs Requirements of 10 CFR 50, Appendix B yy (Sheet 1 of 12)

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ESG tepleronting Document cr Procedure A& pendia B Cri ter ion Amber Title it. C al I ty CS3W17 Guality Asswance Records Asse ance Progr su F90-13 CFERP L Iconsing Assinistratcr (cort'c)

Iff. Design Contrcl SCP M-13 Program Management SCP h-16 ContIguration Management F90 ho. 1 00RP Ccrrespondence Contrcl F90 No.11 QERP Document Hold Status System F90 ho.15 Sctiedule Develoreent and Control F90 ho.19 OBRP SDD Preparation and Revision F90 ho. 21 CFBP Develc*eert Activ ities F90 ho. 25 DeRP Ports Standardiz ation F90 ko. 26 Use of Controlled icf creation Date Transmittal (ClhCT)

FSO No. 21 OBRP Document Status System F90 ho. 30 QERP Specifications

- FSO ho. 32 OBRP Design Revlees and Release N F90 No. 34 Application of Additions to ASE Code Requirements I

w F90 No. 36 F90 ho. 40 Engineering Dramings Materials and Processes fcr DBRP C.O F90 No. di Basel f alog of Documents F90 No. 5 4 SHR$ RelIabII ity Program l F90 No. 56 Acceptance Test Requirements and Specifications EMP 1-0 Pref ece to Engineering Managesent Procodures Manua1 EMo 2-8 Engineering Studies E W 2-9 Design and Acceptance Criteria i EMP 3-5 EMP 3-42 Engineering Release System Engineering Management System fcr spectf: cations Figure 17J-4 Quellty Asswance Procedure Index vs Regulraments of 10 CTR 50, Appendix 8 (Sheet 2 of 12) 23 >

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j ESG lept asenting Docment w Procedure CrIter iren h aber TItie i

fil. Dusign Control EW 3-21 Englesering Change Control i icontinued) EMP 3-22 Interface Control EMP 3-24 Control of Engineering Drawings

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EMP 3-25 Engineerleg Droers - Preparation Instructlons EMP 3-21b Preparation and Control of Supporting Documents EMP 3-28 Compor.ent Traceability EMP 3-29 Engtreoring Requirements tw serialization EMP 3-51 teimont Oeckt ist E W 3-52 Engineering Release Plan of Action EMP 3-63 Doceentation Release and Control of Scientific and j Techalcal Computer Progres

EW 5-3 Design Revlows E M 5-17 Decking of Engineering Drawings

• EMP 5-21 Meterial s and Processes Control Systen E W S-24 Application of Standards g CS3M 3, 6 Design and Doceent Control u 5 3-13 habering and Control of Manuf acturlag Meterial C-, Processing Procedures (MPP) n I

$ IV. Procw eent Document 50P J-12 50P

• 10 Properaticn and Processing of the Purchase Ranulsition Progra Management Cuatrol f90 No. 22 Use of QBRP Aelnistrative $secifIcation la Procurments

. f90 No. 23 $6bcontract Proprocurment Planning

) 790 No. 24 Preparation, Revies, Approval, and Processing of Purchase Requisitf#s Ai r 1.1.4 Procurement Policy AlW 3.103.1 Procurment fram Approved Suppi ter Figure 17J-4 Quality Assurance Procedure inden vs Requirements et 10 CFR 50, Appendix B Gheet 3 of 12) i 1

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ESG toplamenting Document r Procedure Appendix B Criterion N e ar Tltie IV. Procuresnent OsP 2.14 Osanges to Purchase Orders and Other Direction to Suppl fers Documert Control 04P 2.35 Case File Docimentation (conttrued) QAOP h4.00 Procurement Documents QAl h4.00A GERP Procurenent Document Review C534 4 Subcontracter Fabricated items CS3M, Appendix A Contracting f x the Fabrication of a Code item as an h-Certif icate Holder Retalning Overall Responsttill ity fcr Certification and Stmping V. Instr uct ions, SOP A-01 ESG PolIcles and Procedures Procedures, SOP Q-10 ESG Qual Ity Assurance Program i and Drawings SOP Q-28 Unusual Occurrence Reports - RDT Progems I SOP Q-16 ESG Quality Records 50P Q-20 Reports to the Nuclear Regulatwy Ccmalssion ORC)

Concerning Def ects and Nonccepilances g FSO No. 35 Change Control N FSO No. 36 Engineering Drawings Cs FSO No. 48 Unusual Occurrence Reporting o

( E MP 2-9 EMP 3-1 Design and Acceptance Criterla Engineering Docimentation Process EMP 3-4 Nebering of Engineering Documents EMP 3-5 Engineering Release System EMP 3-42 Engineering Management Systems f w Specif ications E W 3-29 Engineering Requirements fw Serialization SOP L-12 Laboratcry and Engineering Notebooks EMP 4-4 Test Procedures EMP 4-5 Test Reports Figure 17J-4. Quality Assurance Procedure index vs Requirements of 10 CFR 50, Appendix B (Sheet 4 of 12) 2>

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ESG lept ewentIng Docusent ur Procedure Appendia B --

Cr i ter ion N eber Title l v. i nstr uct l ons, O# 2.35 Case Fit e Dccumentation Procedures, QOP N1.21 Qual i ty Assurance Pi an s and Dren ings Q O P N1.22 Qual Ity Assurance Acceptance Procedures (cont i nue d ) Q OP hl.23 Qual ity Status Reports C 3M 5.11 Cleaning Procedures CS3M 9 Control of Fabrication Procesws Q C P h6.01 Qualification of eeldipg ProceJures and eelding Personnel QCP h6.02 Qualification acd Certification of Nondestructive E mm ina tion Per sonne!

CS3M 2.4 Qualification and Certification of hondestructive EaminatIon *ersonne!

Q oP :,c .05 QualIf Ication of Special Processes CS3M S.4 holding Procedures CS3M 9.3 Control of Welding Operations CS3M 5.5 Heat-Treating Procedures CS3M 5.9 handestructive Emmination Procedures CS3M 7.10 Subcontracted Nondestructive Exami% tion Services Q3M 10,11, 5.10 In-Process and Final Examination and Tests CS3M 2.6 Authcr f red Inspectw

,_. l CS3M 17 Quality Assurance Recoras y

• M-3-15 Qualif Ication of nelders, welding Operator s, and welding Procedures C CS3M 3, 6 Design and Document Control e

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- V I. Document Control . 50P J-12 Preparation and Processing of the Purchase Requlstilon F50 ho. 1 00RP Ccrrespondence Control FND No. 36 Engineering Drawings FND No. I2 Quality Assurance Revle and Approval of EnglneerIng Requirernents Documents FND No. 35 Change Control 150 No. 56 Acceptance Test Requirments and Specifications Figure 17J-4. Quality Assurance Procedure inden vs Requireents of 10 CFR 50, Appendix B (Sheet 5 of 12)

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ESG lept ementing Docwent a Procedure Appendia B Cr i ter ion Nat,or TItie vi. Document Control EMP 3-42 Engineering Management System f ar Specifications (continued) EMP 3-21 Engt reering Change Control EMP 3-24 Control of Engineering Drawings EMP 3-25 Engineering Orders - Preparation instructions EMP 3-20, Preparation and Control of Supporting Documents EMP 3-36 Request f cr Docwent Change EMP 3-52 Engineering Release Pl an of Action EMP 3-63 Documentation, Rel ea se. ard Control of Scientific and Technical Caputer Progres 04P 2.14 Changes to Purchase Order and Other Directions to Suppliers QM)P N2.03 Doceent Control CS3M 3, 6 Design and Docment Control M-3-13 hsbering and Control of Manuf acturing Material Processing Procedures (MPP)

Vll. Control of Pur- SOP J-12 Preparation and Processing of the rurchase Requisition chawd Mater f al, SOP K-90 Receiving and Inspection of inccntng Material and Equipment Equigment and SOP K-84 Warehousing of Direct-Charged Purchased Mater f els by Serv ice Traf fic and Warehousing

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SOP P-46 50P K-78 Handl Ing and Stor age of Pra'ect Critical Harovare Procurment and Control of Supplier Data 8

F40 ho. 23 Subcontract Proprocurment Planning M Pb0 No. 43 F90 No. 55 Review of Supplier Data instructions f rr Required Documentation and Procedures f cr Shipment of Camponents to CRBRP Site cr Other Designated Areas 04P 3,121 Source SeIaction Q AOP h4.01 Suppi ter Evaluation and Approval Q AOP N4.02 Procurement Quality Verification Instructions Figure 17J-4. Quality Assurance Procedure Index vs Requirements of 10 CFR 50, Appendix 8 (Sheet 6 of 12)

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ESG lapimenting Doceent w Procedure Appendix 8 Cri ter ion het.or Title Val. Control of Pur- QADP h4.03 Procurment Qual Ity Assurance - Source inspection /Survell s ence chased Material, Equipment and Q AOP N4.04 Procurament Quality Assurance - Receiving inspection j Service (cont'd) QAOP h4.04C 06RP Receiving inspection Overcheck Requirements CS3M 7.2 Approved Procurem4ct Sources CS3M 4 Procurement Docwent Control CS3M 5.3 Procurement Quality verification instructions CS3M 7.3, 7.4 Procurment Verification (Source and Receiving Verification)

CS3M 8 Identification and Control of Materials and items V il l. identification SOP K-90 Receiving and inspection of lacom ag Material and Equissent and Control of SOP K-84 Waremousing of Direct-Osarged Purch ased Material s by I

Materials, Traf fic and Warehousing Parts and SOP P-46 Handling and Storage of Project Critical Harerare Components EMP 3-28 Ccaponent TraceabliIty EMP 3-29 Engineering Requirements fcr Serialization Q ADP N4.02 Procurment Quality Verification Instructions

, QACP h4.04 Procuranent Quality Assurance - Receiving inspection

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c QAOP N5.01 QADP N6.04 Manuf acturing Productton Order (Shop Traverters)

Weld Material Control

' s D Q AOP N9.00 Stamp Control g CS3M 14.2 issuance, Use, and Control of Stamps Q ADP N9.02 Serlai tzation of Harosare QAOP N10.0 Nanconfcraing Materials and items CS3M 4 Procurement Document Control CS3M 5.3 Procurement Qual sty Verification Instructions 4 CS3M 7.3, 7.4 Procurment Verification (Source and Receiving Verification)

CS3M 8 Identification and Control of Materials and items l

Figure 17J-4. Quality Assurance Procedure index vs Regulrunents of 10 CFR 50, Appendix 8 (Sheet 7 of 12)

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ESG leplanenting Document cr Procedure Appendia B Criterlon N et,er Title y Vill. I denti f icat ion G3M 9 Control of Construction Processes and Cwtrol of CS3M 15 peonconf aming Materials and items Mater i al s, Par t s 644 M-2-4 Control Stations and Cczeponents >#4 M-3-6 Material Control (continued)

IX. Control of EMP 5-21 Material s and Processes Control Sy st em Special Proces se s Q AOP N3.02 ESG Special Tooling Q ACP N5.01 Manuf acturing Production Order (Shop Travelers)

CS3M 9 Control of Construction Processes Q ACP N6.01 Qualification of Welding Procedures and Welding Personnel Q AOP N6.02 Qualif ication and Certif ication of Nondestructive Examination Personnei CS3M 2.4 Qual if ication and Certif ication of Nondestructive Examination Personnel CS3M 5.11 Clsaning Procedures Q A0P N6.03 Nundestructive Examination Procedures CS3M 5.9 Nondestructive Exam ination Procedures H

Q AOP N6.05 Quellfication of Special Processes

[i CS3M 5.4 Q3M 9.3 Wel ding Procedures, Specif Ications, and Personnel Control of Welding Operations A CS3M 5.5 Heat-Treating Procedures

• CS3M 7.10 Subcontracted Nondestructive Examination Services

>94 M-3-15 Qualification of Welders, Welding Operators, and Welding Procedures X. Inspect ion SOP K-90 Receiving and inspection of incoming Material and Equipment Q AOP N1.21 Quality Assurance Pi ans Q ACP N1.22 Quality Assurance Acceptance Procedures Q AOP N4.02 Procurement Quality Verification Instructions Figure 17J-4. Quality Assurance Procedure index vs Requirements of 10 CFR 50, Appendix 8 (Sheet 8 of 12) 2: >

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ESG Impleenting Doceent a Procedure Appendix 8 Criterion Neber TItie X. 6 nspect ion Q AOP h4.04 Receiving inspectlon (continued) Q AOP h4.03 Procurment Quat Ity Assurance - Source inspection / Surveillance Q AOP h4.04 Procurement Quality Assurance - Receiving Inspection

, g Q AOP N4.04C CBRP Receiving Inspection Overcheck Requirements Q ADP N5.01 Manuf acturIng Product ton Order (Shop Travelers)

' QAOP N6.03 handestructive Examination Procedures CS3M 5.9 Nondestructive Examination Procedures Q ACP N6.05 Qualification of Special Processes Q ADP N7.00 Product Acceptance Tests Q ADP N7.01 Pressure Testing CS3M 5.3 Procurement Quality Verification CS3M 7.3, 7.4 Procurement inspection (Source and Receiving InspectioM CS3M 9 Control of Construdion Processes CS3M 10 Examination, Tests, and inspections l

CS3M 2.6 Authorized inspector XI. Test Control SOP L-12 Laboretwy and Engineering Notebooks EMP 4-4 Test Procedures EMP 4-5 Test Reports

• c_. Xil. Control of SOP Q-24 Callbretion of Measuring instruents and Equipment 8 Measuring and Q A)P N3.00 Control of Measuring and Test Equipment (M&TE)

$ Test Equipment QAOP N3.02 CS3M 12 ESG Special Tooling Control of Measurement and Test Equipment Xill. Handl Ing, Storage SOP P-46 HandlInn and Storage of Project Critical Hardware and Shipping 50P K-44 Shipping SOP P-48 Material Hand!Ing Equipment peiE)

CS3M 5.11 Clocting Procedures F90 55 Instructions For Required Documentation and Procedures f or Shipment of Components to OERP Site or Other Designated Area F90 57 Storage, Maintenance, and inspect ton of Material, Parts, and Components Figure 17J-4 Quality Assurance Procedure index vs Requirements of to CFR 50, Appendix B (Sheet 9 of 12)

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ESG Implementing Document or Procedure Appendla B Cr i ter ion Numt,or Title Xill. Handling, Q ACP N12.00 Packaging and Shipping inspection Stw age and CS3M 13 handl ing, Preserv ation, Stcr age, and Shipping Shipping W M-2-4 Control Stations (cont i nued) m M-3-10 Packaging and Shipping XIV. Inspect i on, SDP K-90 Receiving and Inspection of incoming Material and Equipment Test and 50P E-84 Warehousing of Direct-Osarged Purchaud Material s by Oper ating Traf fic and Warerousing Status SOP P-46 Handling and Storage of Project Critical Harcheare SOP Q-18 ESG Qual Ity Records Q AOP N1.21 Qual Ity Assurance Plans Q AOP N3.02 ESG Special Tooling Q AOP N4.04 Procurernent Quality Assurance - Receiving inspection Q AOP N5.01 Manuf acturing Production Order (Shop Travelers)

CS3M 9 Control of Construction Processes Q ACP N6.04 Wold Material Control Q AOP N7.00 Product Acceptance Tests Q AOP N7.01 Pressure Testing y Q AOP P6.00 Stamp Control y l CS3M 14.2 lasuance and Control of Stamps Cs Q AOP N9.02 Serialization of Hard.are

[ Q AOP N10.00 Nonconf orming Materials and items y y C33M 7.3, 7.4 Source Qual Ity verif Ication and Receiving Inspection r CS3M 8.0 tdontIfIcation and Control of Materials and Items CS3M 9.3 Control of Welding Operations CS3M 5.5 Heat-Treating Procedures CS3M 10,11 Examination, Tests, and inspections, Test Control l CS3M 15 Noncontcrming Materials and items CS3M 2.6 Authwired inspecta H4 k-2-4 Control Stations H4 M-3-6 Material Control Figure 17J-4 Quality Assurance Procedure Index vs Requirements of 10 CFR 50, Appendix B (Sheet 10 of 12) if. >

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ESG lapiamenting Document cr Procedur e Appendix B Criterlon N aber Title XV. Nancontcra lng 50P K-90 Receiving and Inspection of inccaning Material and Equipment Mater ial s, SOP Q-18 ESG Quality Records Perts, or QMP N5.01 Manuf acturIng Production Order (Shop Travelers)

Ccaponents QAOP Nf 0.00 Nanconfcralng Materials and items QA1 N10.000 06RP Nardeare Nonconfcrmance Processing QEP N13.02 Quality Assurance Data Packages CS3M 9 Control of Construction Processes CS3M 15 Noncontweing Materials and items CS3M 17 Quality Assurance Records XV I. Corrective SOP K-90 Receiving and inspection of Incoming Material and Equipment Action SOP Q-14 Corrective Action System SOP Q-28 Unusual Occurrence Reports - RDT Programs l F90 No. 48 Unusual Occurrence Reporting EMP 5-19 Failure Reports EMP 5-20 locident Reports Procurement Quality Assurance - Source inspection / Surveillance

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c Q AOP N4.03 Q AOP N4.04 Procurement Quality Assurance - Receiving Inspection a Q AOP N10.00 Nonconfwming Materials and Items A QAOP N14.00 CorroctIwe Action

" CS3M 16 CorroctIwe Action 50P Q-20 Reports to the Nuclear Regulatcry Ccomission (NRCI Concerning Def ects and Noncompilances XV 3 e. Qual Ity SOP Q-18 ESG Quality Records Assurance SOP k-78 Procurement and Control of Suppller Data Recor ds CS3M Quality Assurance Records FNO 18 06RP Quality Records Management System N099QRPn00001 Quality Records Management Plan f cr 00RP N099CW P410001 Quality Records Management Procedures Figare 17J-4. Quality Assurance Procedure index vs Requirenonts of 10 CFR 50, Appendix 8 (Sheet 11 of 121 Z>

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ESG lept ementirg Document or Procedure Appendix B O f terlon N w.ber Title XV ill. Audits SOP Q-12 Quality Assurance Progra Audits Q ACP N1.04 Qual Ity Assurence Audits CS3M 18 Audits Figure 17J-4. Quality Assurance Procedure Index vs l Requirments of 10 CFR 50, Appendix 8

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O 17J-49 Amend. 73 Nov. 1982 i

OUALITY ASSURANCE MANUAL PROCEDURE DESCRIPTIONS STANDARD OPERATING POLICIES (SOP's)

SOP A ESG Pol icy and Procedures This S0P def ines the types of ESG administrative policies and procedures auth or iz ed, and establ ishes minimum format and distribution requirements f or such pol icies and procedures, it identi f ies the h ighest l evel of managenent, corporate or otherw ise, responsibl e f or establ ish ing qual ity pol icies, goal s, and objectives. A clear path of communication between Qual Ity Assurance organization and corporate management is def ined.

Positions and groups responsible f or def ining both content and changes to the Qual ity Assurance Program and manual s are identif ied, in addition to the management l evel responsible f or the approval of the Qual ity Assurance Program and manual s. Provisions are establ ished f or control l ing and distribution of Qual ity Assurance manual s and revisions.

S0P J Preparation and Processing of the Purchase Requisition This S0P establ ishes methods and pol icies appl icable to the preparation and processing of the Purchase Requisitions (Form N25-R-2). The requisition is used f or authoriz ing procurement, through Purchasing, of mater i al s, eq u i pment, and services from suppliers.

Procedures are establ ished that del ineate the sequence of actions to be accompl ished in preparation, review, approval, and control of the Purchase Req u i si ti on.

S0P K Receiv ing of incoming Material and Equipment Receiv ing inspection of suppl ier-f urnished material and equipment is perf ormed in accordance w Ith the f ol l ow Ing. The material is properly identif ied and corresponds with receiving documentation. Inspection is perf ormed and judged acceptabl e, in accordance w Ith predetermined instructions, prior to use. items accepted and released are identified as to their inspection status, prior to rel ease. Nonconf orming items are segregated, control led, and identif ied until proper disposition is made.

S0P K Warehousing of Direct-Charged Purchased Material s by Traf fic and Warehousing Methods are specif ied to identify and control material s. Verification of correct identif ication of material, prior to release, is required.

Material shall be protected against loss, damage, and deterioration from env ironmental conditions.

Amend. 73 17J-50

S0P P Handl ing and Storage of Project Critical Hardware Special handl ing, preservation, storage, packaging, and shipping requiremants are specif ied and perf ormed by qual if ied personnel under predetermined instructions.

S0P K Shipping i Special packaging and shipping requirements are specif led and accompl ished by qual if led Individual s, in accordance w Ith predetermined instructions.

Procedures are prepared in accordance w ith design and specification requirements which control the packaging and shipping of materials, components, and systems to precl ude damage, loss, and deterioration.

l S0P P Material Handl ing Equipment Special handl ing requirements are specif led and accompi Ished by qual if led indiv idual s, in accordance with predetermined instructions. Procedures are prepared in accordance with design and specification requirements which control the handl ing of material s, components, and systems, to prevent damage.

S0P Q Cal ibration of Measuring Instruments and Equipment Procedures describe the cal ibration technique and f requency, maintenance, and control for alI measuring instruments and test equipment which are y used f or obtaining data, where traceabl e cal ibrations are required, measuring and test equipment is identified, end the calibration test data is identified with the associated equipment. Measurement and test equipment are cal ibrated at specif led interval s, based on the conditions af fecting the measurement. When measuring and test equipment is found to be out of cal ibration, any items measured with this equipment are withheld until the accuracy of the results is eval uated. The complete status of al I items under the cal ibration is recorded and maintained. Ref erence and l

l transfer standards are traceable to national standards. If national standards do not exist, the basis f or cal ibration is documented.

S0P Q Product Integrity impl anents Rockwel l International qual ity pol icy and directive f or ESG operations by establ ishing the Product Integrity Program. Def ines 14 areas to be covered, makes ESG Qual Ity Assurance Director Product integrity Progran Coordinator, and establ ishes a Product integrity Ccomittee consisting of ESG executive management.

S0P L Laboratory and Engineering Notebooks it is the policy of the company to record all scientif ic and laboratory research and development activitles in Iaberatory and engineering notebooks to be used by scientif Ic and engineering personnel, primarily to l record results of scientif ic studies and lab wcrk, whether company or custcmer oriented. innovations, inventions, discoveries, and improvanents l

Amend. 73 17J-51 Nov. 1982

r wil l be recorded f or the purpose of f ul fil l ing contractual obl igations and protecting company Interests.

S0P K Procurement and Control of Suppl ier Data Procedures are establ ished f or preparation, review, and control of instructions, procedures, drawings and changes thereto. These documents and changes thereto are procedurally controlled to assure adequacy.

Provisions are establ Ished, identify Ing the personnel responsibl e f or these activities. Changes are reviewed by the same organizations that perf ormed the original review, unless delegated by the appl Icant to qual if ied responsibl e organizations. Approved changes are promptly incl uded in the appropriate documents.

S0P M Program Management This S0P sets f orth principles and guidelines for the managements of Energy Systems Group Business Programs. The Guidel ines incl ude organiz ational framework, program management processes, perf ormance monitoring, and reporting systems.

l S0P N Conf iguration Management This SOP establ ishes the pol icies, metnods, and responsibil itles f or the preparation, issuance, and use of Conf iguration Summary Reports.

The primary purpose of this report is to aid the Manuf acturing, Quality Assurance, and Engineering f unctions in determining conf iguration and of fectivity requiranents for product hardware.

S0P Q Corrective Action Systan Eval uation of nonconf ormances and determination of the need f or corrective action f ol low establ ished procedures. Prompt corrective action is initiated, following the determination of nonconf ormance to procedural or technical requiranents. Adverse conditions significant to quality, their causes, and corrective actions, are reported to the appropriate levels of managanent.

S0P Q ESG Qual ity Assurance Progran This procedure def ines the Qual Ity Assurance Program to be appl ied to all ESG products and services, in compi lance w Ith appl Icabl e contract, federal, or state requirements. Management (above or outside of Qual ity Assurance and to the highest corporate level) regularly assesses the Qual ity Assurance Program ef fectiveness. The establ ishment of Indoctrination and training progress review is specif ied.

S0P Q Unusual Occurrence Reports - RDT Programs l

This SOP establ ishes methods and responsibil Itles f or reporting to the customer of unusual occurrences af fecting ESG programs under the requirements of RDT Standard F 1-3T.

Amend. 73 17J-52 Nov. 1982

S0P Q Reports to the Nuclear Regul atory Commission (NRC) Concerning V Defects and Noncompilances The purpose of this S0P is to comply with requirements of 10CFR21 including requirements to adopt procedures to 1) provide for: a) eval uating deviations or b) Informing i Icensees or purchasers of devlations; and 2) assure that a responsibt e of fIcer is informed of:

a) f ail ures to comply with the Atomic Energy Act of 1954, as amended, or any appl icable rul e, regul ation order or l icense of NRC rel ating to Substantial Saf ety Hazard, or b) defects in the construction or operation of a f acil Ity or actlvIty Iicensed or otherwIse regulated pursuant to the Atomic Energy Act of 1954, as amended.

This S0P designates the President, Energy Systems Group, as the responsible of ficer to be informed and provides methods for informing the President, Energy Systems Group, and provides for delegating his authority for reporting to the NRC.

SOP Q Qual ity Assurance Progran Support Functions This procedure establ ishes pol icy on the util ization of ESG Qual Ity Assurance Department f unctions on ESG prograns and describes the Qual ity Assurance Department f unctions and interf aces with other ESG departments.

It summarizes the provisions f or resolving disputes arising from a

(] dif forence of opinion between Qual ity Assurance - Qual ity Control and

() other department personnel .

The procedure outi ines the saf ety-rel ated structures, systems, and components control led by the Qual Ity Assurance Program, and the respective organization executing Qual ity Assurance - related f unctions on these items during the design, engineering, procurenent, in spect i on, manuf actoring, construction, and testing phases. Qual Ity-rel ated activ ities (Inspection and test, etc. ) perf ormed with appropriate equipment and under suitable environmental conditions are described.

S0P Q Qual ity Assurance Program Audit Procedures and responsibil itles f or assuring the adequacy and of fectiveness of the ESG Qual ity Assurance Program through audits of procedures, standards, methods, and practices used in producing ESG hardware or sof tware products are establ ished by th is SOP.

Audits are perf ormed in accordance w Ith pre-establ Ished written procedures or checkl ists and are conducted by trained personnel not having direct responsibil Itles in the areas being audited. The audits include an objective eval uation of qual ity-rel ated practices, procedures, and Instructions, and the ef fectiveness of impl anentation and conf ormance w ith pol icy directives.

O Amend. 73 Nov. 1982 17J-53

Audit data are analyzed and reports indicating quality trends and the of fectiveness of the Qual ity Assurance Program are provided to management.

The audit results are documented and then reviewed with management having responsibil ity in the area audited. Subsequently, responsibl e management takes the necessary action to correct the defIclencies revealed by audit.

SOP Q ESG Qual ity Records ESG Qual ity Records are def ined, and responsibil Itles f or their retention are establ ished by th is SOP. Its purpose is to establ Ish standards for meeting ESG and customer requiranents f or fil ing, storing, and retrieving of qual ity history information on ESG products and services.

Qual Ity Assurance records incl ude: 1) operating iogs, 2) results of rev iew s, inspections, tests, audits, and material analyses, 3) monitoring of work perf ccmance, 4) qualification of personnel, procedures, and equipment, and 5) other documentation, such as drawings, speci f ication, procurement documents, calibration procedures and reports, and nonconf orming and corrective action repor ts. The records are to be readily identif iabl e and retrievabl e.

Requirements and responsibil ities f or record transmittal s, retention and maintenance subject to work compl etion must be consistent w ith appl icabl e codes, standards, and procuronent documents.

Record storage f acil itles are to be constructed, located, and secured to prevent loss or destruction of the records or their deterioration by environmental conditions.

O Amend. 73 17J-54 Noy. 1982

( CRBRP PROGRAM MANAGEMENT DIRECTIVES (PM)'s)

V PV0 GBRP Correspondence Control Tnis procedure del ineates the method f or identify ing, control l ing, and accounting f or all incoming and outgoing correspondence, and f or capturing commitments on the Commitment Status Report system.

PMD CRBRP Document Hol d Status System This procedure applies to holds and TBD's on all released (for project use) Principal Design Data f or which ESG is responsible. The current status of each Hold and TBD in these documents which impacts Level 2 cr Level 3 activities is maintained in the Document Hold system as described in this directive.

PMD Qual ity Assurance Review and Approval of Engineering Requirements Documents This directive establ ishes the requirement and procedure for formal review and approval by Qual ity Assurance personnel of ESG-generated 1) drawings,

2) speci f ications, 3) specif ication amendments, 4) Engineering's Change Proposal s, 5) System Design Descriptions (SDD), and 6) Engineering Orders.

PMD CRBRP L iconsing Administrator p],

\

This directive def ines the responsibil ities of the ESG CRBRP Licensing Administrator for impl ementing and controlling licensing criteria in accordance w Ith Section 9.0 of the Management Pol icies and Requirements (MPR).

PMD Schedul e Devel opment and Control This directive del Ineates the method f or development, processing, approval, maintenance and change control of the ESG schedule hierarchy which def ines the CRBRP ef f ort w ith in the requirements of ESG Program Management System.

This directive def ines both the vertical integration of schedul es f or CRBRP f rom the contractual interf ace to the detailed work package structure and the horizontal breakout over the time of the various schedul ar l evel s and documents.

PMD Qual ity Assurance Management Reviews This procedure impl ements a Qual ity Program requirement f or periodic qual ity assurance management review meetings to assess CRBRP Project qual ity accompl ishments, discuss program qual Ity audits, and resolve management probl ems af fecting qual ity, o

Amend. 73 17J-55 Nov. 1982

PMD CRBRP Qual ity Records Management System This procedure impiments the quality records requirments of the CRBRP Management Pol Icles and Requirments Document, Section 11.0, " Project Records Management", fcr ESGRM activities.

PMD CRBRP SDD Preparation and Revision This procedure def ines the methods f or preparation and maintenance of CRBRP Systm Design Descriptions.

PMD CRBRP Training and Indoctrination This procedure impiments CRBRP Project requirements for training and indoctrination of personnel whose activities may have an ef fect on q ual ity.

PMD CRBRP Devel opment Activities This directive defines the methods f or initiating and controlling development activities required for the CRBRP Program and incl udes directions f or 1) preparation, revlew and release of development activities, 2) revision and control of approved development activities, 3) review and control of development activities, and 4) control of devel opment hardware.

PMD Use of CRBRP Administrative Specifications in Procurments This procedure describes the use of administrative specifications for Qual ity Assurance administration of purchase orders between Energy Systems Group and the sellers of services or items.

PMD Subcontract Preprocurment Pi anning This procedure provides the guidel Ines required to accompi ish a thorough subcontract preprocurment pl anning f unction by the Purchasing Department.

It outi ines purchasing pol icles that are consistent wIth requirments establ ished in the Management Pol Icles and Requirments (MPR) f or the Cl inch River Breeder Reactor Pi ant (CRBRP), and w Ith pol Icles del ineated in the Rockwell Corporate Material Procedures (CMP's).

PMD Preparation, Rev iew, Approval and Processing of Purchase Requisitions This directive describes the procedure for preparing, reviewing, approving and processing Purchase Requisitions. These instructions augment those in S0P J-12.

The directive appiles to Purchase Requisitions for CRBRP ltems prepared by the CRBRP Program Of f ice or the Engineering Department, it does not apply to Purchase Requisitions prepared by Manuf acturing in support of hardware "make" Itons.

O Amend. 73 17J-56 Nov. 1982

[J PPO CRBRP Parts Standardization All CRBRP design activities perf ormed within ESG will util ize the ESG parts standardization system as described in " Preferred Parts and Design Standards", publ Ished by the Checking and Design Standards f unction.

Changes to that publ ication w il l be appl icabl e to the CRBRP Progran immediately upon release for general ESG use and will not require revision to th is directive.

PS0 Use of Control led inf ormation Data Transmittal (ClNDT)

This procedure establishes a method for the controlled dissemination of OBRP technical information and to assure that information used as a basis for design is obtained only from controlled sources.

PMD-Z7 - CRBRP Document Status System This. procedure def ines the operation of the Documentation Status System (DSS) modul e (WARD-D-0059) and the ESG responsibil ities and Interf ace with the Westinghouse ARD computer. The DSS assures that principal design data is identified, measured and statused to provide inf ormation required to manage said CRBRP Program data.

PMD CRBRP Specif Ications This procedure modif les the requirements of the standard ESG specif Ication g revicion system to certain specific requirements of the CRBRP Project.

PMD OBRP Design Reviews and Release This procedure impiments the CRBRP pol Icy relating to design reviews of systems and components, to supplanent the standard ESG design review practice. -

PMD Appi ication of Additions to ASE Code Requiranents This directive covers all CRBRP components including piping systems designed and constructed under ASE Section Ill, ASE Section Vill, and ANS I B31.1.

! PMD Change Control This procedure provides direction f or revIslon of ali ESG documents which have been def ined to be part of the CRBRP Basel ine.

PMD CRBRP Engineering Draw Ings This procedure def ines the methods to be used f or release and rey!sion of QBRP ongineering draw ings.

O O

Amend. 73 Nov. 1982

PMD Material s and Processes f or CRBRP This direC.ive is established to ensure that all CRBRP design work will be based upon one common set of materials data as welI as on consistent extrapolations and interpretations of these data.

PMD Basel ining of Documents This procedure gives the method f or def ining documentation as part of the CRBRP basel Ine.

PMD Rev lew of Suppl ler Data This directive establishes specific requirements for the review of suppl ler data and augments the general requirments of S0P K-78.

PMD Unusual Occurrence Reporting The purpose of this procedure is to provide for DOE Unusual Occurrence Reporting and f or identification of those occurrences which require special consideration as def iciencies reportable under 10CFR50.55(e) and 10CFR21.

PMD SHRS Rel iabil ity Program This directive def ines the requirements of the rel labil ity progrm at ESG on CEBRP.

PMD Instructions f or Required Documentation and Procedures f or Shipment of Components to CRBRP Site or Other Designated Areas This directivo describes the required documentation and the submittal sequence to be f ollowed prior to and during shipment of components and equipment to the CRBRP Constructor, Stone and Webster Engineering Company.

PMD Acceptance Test Requirments and Specifications This directive def ines the requirements f or systems acceptance testing specif Ications which are to be prepared by Al-ESG.

PMD Storage, Mai ntenance, and Inspection of Material Parts and Cm ponent s.

l This directive describes the requirments and responsibil itles f or storage, mai ntenance, and inspection of material, parts, and components f or CRBRP that are under the cognizance of ESG.

l O

Amend. 73 17J-58 Nov. 1982

ENGINEERING MANAGEENT PROCEDURES (EMPs)

(V)

EMP 1 Pref ace to the Engineering Management Procedures Manual This procedure describes the sccpe of the Engineering Management Procedures (EMP) Manual .

EMP 2 Engineering Studies This procedure establishes the requirement for conducting studies to estabi Ish that the design meets the design criteria, is based upon ' proven practices or analysis, and is adequate for the intended service. It describes the method f or preparing, releasing, and controlIing Engineering Studies.

EMP 2 Design and Acceptance Criteria This procedure delinoates the need for design and acceptance criteris to be def ined and publ Ished in the appropriate design basis documents.

EMP 3 Engineering Documentation Process This procedure describes the scope of the procedures which control the preparation, release, and control of specif ications, drawings, and reports p by Engineering.

EMP 3 Engineering Release System This procedure provides Instructions f or the preparation, numbering, release, and control of drawings f or the Engineering Release System, and prov ides guidel ines f or appl Ication of the standard release. EMPs 3-5.1, 3-5.2 and EMP 3-5.3 provide for procedural detail s f or the ASE Code, standard and experimental release systems.

EMP 3 Numbering of Engineering Documents This procedure and its sub-procedures (3-4.1, 3-4.2, 3-4.3, 3-4.4, 3-4.5, '

3-4.6 and 3-4.10) defines the requirements and means for uniquely numbering various types of ESG engineering documents including drawings, specif ications, supporting documents, O&M manual s, subcontractor memos, and sof tware control documents.

EMP 3 Engineering Change Control This procedure def ines the method f or requesting, eval uating, approving, and executing engineering changes.

EMP 3 Interf ace Control This procedure establishes the criteria for interf ace def inition and the methods f or describing and controlling the interf ace in appropriate O' documentation draw ings and specif Ications.

i Amend. 73 17J-59 Nov. 1982

i l

l EMP 3 Control of Engineering Documents This procedure describes the methods f or control of drawing original s and prints, released by both the Standard or Limited Release Systems.

EMP 3 Engineering Orders - Preparation Instructions This procedure describes the preparation and use of an Engineering Order to rel ease draw ings or specif ications, and def ines requirements. EMP's 3-25.1 through 3-25.17 provide detail s f or various types of Engineering Orders.

EMP 3 Preparation and Control of Supporting Documents This procedure establ ishes the types of supporting documents and def ines the requirments for their preparation, rel ease, and change.

EMP 3 Component Traceabil ity This procedure describes the elments and responsibil Ity for establ ishing item traceabil ity.

EMP 3 Engineering Requirements f or Serial ization This procedure sets conditions under which Engineering requires serial Ization of components or parts f or traceabil Ity purposes.

EMP 3 Request f or Document Change This procedure describes the formal means for requesting a change to a released draw ing or specif ication and the approval and processing of that req uest.

EMP 3 Engineering Management System for Specif Ications This procedure def ines the method f or the preparation and control of Engineering specif ications.

EMP 3 Wel dment Checki ist This procedure provides the checkl ist to be completed f or critical wel dments, and the system for its impimentation.

EMP 3 Engineering Rei ease Pl an of Action This procedure gives the f ormat and requirments f or a plan describing the means of preparation and release and approval of program documents.

O Arend. 73 17J-60 nov. 1982

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Amend. 73 u v. 1982 17J-61

EMP 3 Documentation, Release, and Control of Scientif ic and Technical Ccrnpu:ar Progrms This procedure describes the documentation formats for scientific and technical (S&T) computer programs used and/or produced within the Research and Engineering Department. Those S&T progrms that are developed outside of ESG shal l al so be documented to the same extent specif ied by th is EMP al l ow ing f or vendor documentation f ormats.

EMP 4 Test Procedures This procedure gives the f ormat for preparation of Test Procedures.

EMP 4 Test Reports This procedure gives the f ormat for preparation of Test Reports.

EMP 5 Design Revlows This procedure establishes the requirements for independent design reviews, and the means of their schedul ing, conduct, and reporting.

EMP 5 Checking of Engineering Drawings This procedure establ ishes the responsibil itles f or checking of al I engineering draw ings.

EMP 5 Faii ure Reports O

Fail ure Reports are to be used when a component or system under test has f ailed or deviated f rom expected conditions on all ESG programs as def ined in Paragraph 3.1.

EMP 5 Incident Reports incident Reports are to be used when an incident or f ailure occurs in a test other than on the component being tested on all ESG programs as defined in Paragraph 3.1.

EMP 5 Material s and Processes Control System l This procedure establ Ishes the pol icy and responsibil itles f or control of material s and processes.

j EMP 5 Appl Ication of Standards l This procedure provides guidance and direction for the appl Ication of l codes and standards. It categorizes various types of standards and establ ishes responsibil ities f or their col lection and appl ication.

l O

Amend. 73 17J-62

i 1

• Q)RPORATE AND AI MATERl AL PRO &DURES (CMP's/AIMP's)

! AIMP 1.1.1 - Procurement Pol icy This procedure describes the procurement policy of Rockwell International ,

and supplements it to cover procurement reflecting DOE requirements.

WP 3.121 - Source Sel ection This procedure def ines Rockwell International's practice concerning selection of procurement sources and making commitments.

WP 2.14 - Changes to Purchase Orders and Other Directions to Suppllers This procedure establ ishes standards f or accompi Ishing changes to purchase orders and ef fecting other direction to suppliers.

l WP 2.35 - Case Fil e Documentation

' This procedure establ ishes the documentation required to be accumul ated in procurement case f Il es.

AIMP 3.109.1 - Procurement f rom Approved Suppl lers This procedure requires procurements to Code requirements, to ensure that Qual ity Assurance-approved suppl lers are obtained.

4 i O Amend. 73 17J-63 Nov. 1982

OUAllTY ASSURANCE MANUALS

• PROCEDURES QAOP N1.00 - Pref ace to Qual ity Assurance Manual The pref ace to each Qual ity Assurance Manual del ineates the purpose and authority of the manual .

QAOP N1.01 - Qual ity Assurance Department Functions This document outi ines the f unctions of the Individu.at groups within the Qual Ity Assurance Department.

QA0P N1.03 - Vision Requirments f or Qual ity Assurance Personnel This procedure establ ishes vision standards f or Qual ity Assurance Department personnel and def ines responsibil ities f or administering an eye examination progrm.

QA0P N1.04, CS3M 18 - Qual ity Assurance Audits These procedures outi ine the Qual ity Assurance responsibil itles f or impimenting and maintaining an audit program to determine the overall of fectiveness of the ESG and suppller qual Ity programs and to identify areas where corrective prevention action is required.

QAOP N1.21 - Qual ity Assurance Pl ans This procedure def ines Qual ity Assurance Department responsibil itles f or participating in the preparation of Qual ity Assurance Program Plans or Qual Ity Assurance Program Indexes and for preparing Qual ity Assurance Functional P1ans.

QA0P N1.22 - Qual ity Assurance Acceptance Procedures This procedure def ines requirements and responsibil Itles of the Qual Ity Assurance Department f or the preparation, rel ease, and control of Qual '#-l Assurance Acceptance Procedures (QAP's).

Q/0P N1.23 - Qual ity Status Reports This procedure establ ishes Qual ity Assurance Department requirments and responsibil itles f or preparation of periodic Qual ity Assurance Progrm Status Reports and f or submittal of the reports to Energy Systems Group customers.

• Energy Systems Group Qual Ity Assurance Department Procedures (QAOP)

Energy Systems Group ASW Code Section ill Manual (CS3M)

O Amend. 73 17J-64 Nov. 1982

(* QAOP N2.03 - Document Control This procedure provides direction for the control of engineering and shop draw ings, incl uding customer drawings appl icabl e to products to be f abricated in the ESG Manuf acturing Shops. The purpose of such control is to assure the f abrication, processing, inspect i on, and testing of products to the proper drawings.

l QA0P N3.00, CS3M 12 - Control of Measuring and Test Equipment These procedures def ine requirements f or cal ibration control of tool s, gauges, instruments, and test equipment used by Manuf acturing and Qual ity Assurance to measure products (material s, parts, components, and appurtenances) or to control processes related to the product.

QAOP N3.02 - ESG Special Tool ing This procedure def ines the requirements and responsibil itles for control of tooling used by Manuf acturing and Qual ity Assurance Departments in product fabrication.

l QA0P N4.00, QAl N4.00A, CS3M 4 - Procurement Document Control These procedures def ine requirements and responsibil ities f or preparation, revlew, and approval of procurement documents associated with the purchase of material s, parts, and services.

L/

QAOP N4.01, CS3M 7.2 - Approved Procurement Sources These procedures def ine Qual ity Assurance Department requirements for eval uation and approval of procurement sources (suppi lers) of material, parts, and services used in ESG products.

QAOP N4.02, CS3M 5.3 - Procurement Qual ity Verif ication Instructions These procedures def ine Qual ity Assurance Department requirements and responsibil ities f or preparing inspection instructions appl icabl e to procured items and services.

QAOP N4.03 - Procurement Qual Ity Assurance - Source Inspectlon/Survell Iance This procedure def ines Qual Ity Assurance Department requirenents and responsibil ities f or qual ity verif ication of procured items and services at a suppl ter's f acil Ity.

QAOP N4.04 - Procurement Qual ity Assurance - Receiving Inspection These procedures def ine Qual Ity Assurance Department requirements and responsibil ities f or Ir.specting and testing incoming procured items and services.

(3 Amend. 73 17J-65 Nov. 1982

QAOP N5.01 - Manuf &turing Production Order (Shop Travellers)

This procedure def ines the requirments and responsibil ities f or the preparation and utti Ization of the Manuf acturing Production Order (MPO).

CS3M 9 - Control of Construction Processes These procedures def ine the guidelines used to authorize and control the process, fabrication, instal l ation, inspection, examination, and testing of components, parts, and appurtenances.

l QA0P N6.01, CS3M 5.4 - Wei ding Procedures These procedures establ Ish requirments and responsibil ities f or qualifying welding and brazing procedure specifications and welding and braz ing personnel (wel ding, wel ding operators, brazers, and braz Ing operators) employed in f abrication of Code items.

QAOP N6.02, CS3M 2.4 - Qual if ication and Certif ication of Nondestructive Examination Personnel These procedures establ Ish requirements and responsibil itles f or the training, examination, qual if ication, and certification of Energy Systems Group personnel engaged in the f ollowing nondestructive examination processes:

Radiographic Liquid Penetrant Magnetic Particl e Eddy Current Ultrasonic Leak Detection Q AOP N6.03, CS3M 5.9 - Nondestructive Examination Procedures These procedures estabi ish requirments and assign responsibil itles f or preparing and controlling nondestructive examination (NDE) procedures used f or determining compl iance of products to rnquirunents of appl Icabl e codes and standards.

QAOP N6.04 - Wel d Material Control This procedure def ines requirements and responsibil itles f or issuance and control of wel ding material s (el ectrodes, rods, spool s, and fI ux).

QA0P N6.05 - Qual if Ication of Special Processes This procedure def ines requirments and responsibil ities f or qual if ication of special processes used during f abrication or inspection of products at Energy Systems Group.

QAOP N7.00 - Product Acceptance Tests This procedure def ines requirments and responsibil itles of Qual Ity Assurance Department personnel in perf ccming acceptance tests, or Amend. 73 17J-66 Nov. 1982

(3)

( witnessing acceptance tests perf ormed by others on parts, material, subassembl ies, assembl les, subsystems, and systems (items) that require acceptance by Qual ity Assurance.

QAOP N7.01 - Pressure Testing This procedure def ines the requirements and responsibil ities f or perf orming hydrostatic or pneumatic tests of ESG-f abricated ASNE Code or other products.

QAOP N7.02 - Qual If ication and Certif ication of V isual and Dimensional Inspection Personnel This procedure def ines requirements and responsibil Ittes to provide a mandatory program of training, exanination, and certif ication f or personnel perf orming dimensional inspect i on. The program will provide periodic updating to accommodate changes in requirements and maintain the l evel of knowledge necessary to perf orm dimensional inspection assignments.

_Q ,AOP N8.00 - Stati stIcal Qual ity Contral Program This procedure establ ishes Qual ity Assurance Department requirements and responsibil ities f or impl ementing and maintaining a Statistical Qual ity Control Progran.

(%

(m-) Q AOP N9.00, CS3M 14.2, 14.3, 14.4 - Issuance, Use, and Control of Stamps These procedures def ine the requirements and responsibil ities f or the issuance, appl ication, and control of stamps used f or markings that identify personnel perf orming examination, inspection, test, welding, and braz ing operations.

Q AOP N9.02 - Serial ization of Hardware This procedure def ines Manuf acturing and Qual ity Assurance Department requirements associated with the serialization of parts and assembi les that are f abricated or procured by Manuf acturing.

QAOP N10.00, CS3M 15 - Nonconf orming Material s and items These procedures def ine requirements and responsibil ities f or control and disposition of nonconf orming materials and items in the product manuf acturing/ procurement processes.

QAl N10.00D - CR3RP Hardware Nonconf ormance Processing This Instruction suppl ements Procedure NIO.000 by providing specif ic detail s f or CRBRP nonconf ormance items in accordance w Ith LRM and Owner req u i rement s.

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Amend. 73 17J-67 Nov. 1982

QA0P N12.00 - Packaging and Shipping inspec+ Ion This procedure def ines Qual Ity Assurance Department responsibil itles f or inspecting and packaging and the preparation f or shipment of ESG products.

it appl los to products requiring Quality Assurance acceptance that are shipped f ece ESG, to an ESG construction site, to an ESG customer, or to an ESG supplfer.

QA0P N13.02 - Qual ity Assurance Data Packages This procedure provides f ormat requirments for the preparation of Qual ity Assuranco Data Packages f or transmittal to the customer. Contractual requirments take procedence over this procedure, in case of confl ict.

QAOP N14.00, CS3M 16 - Corrective Action f or Noncor.f ccmance Products Thoso procedures estabi Ish requirments for taking ::ction to correct conditions causing nonconf crming material, parts, and components, its purposo is to provide increased assurance thsr ESG products will meet design, configuration, and perf ccmance requirements.

CS3M 2.3 - Training and Indoctrination This proceduro dof inos requirments and responsibil itbss f or training and indoctrination of personnel perf orming activities af fecting quality or Codo compl iance, as necessary, to assure that suitabl e prof iciency is achieved and maintained.

l CS3M 3, 6 - Design and Document Control Those procedures establ Ish the requirements and responsibil itles as an Owner's Agent, and f or the control of design activities and documents associated wIth items being constructed in accordance with the requirments of the Code.

CS3M 7.3, 7.4 - Procurement Verif ication (Source and Recolving Verification)

These procedures def ino requirements f or source and receiving inspection, examination, and test of procured material s, parts, and serv ices.

CS3M 8 - identif Ication and Control of Materials and items These proceduros def ino requirements and responsibil itles f or impi menting and maintaining material checkl ists required by the Code.

CS3M and Appendix A - Contract Ing f cr the Fabrication of a Code item as an N-Certi f Icate Hol der Retelning Overal l Responsibility fcr Certif ication and Stamping This procedure covers the situations where ESG as an N-Certificate holder retains overal I responsibli Ity fcr a Code item, inctuding design, certif Ication, and stamping can contract f or f abricetion of the Itms.

Amend. 73 17J-68 Nov, 1982

____.c_ _ . _ _ _ _ _ _. __ _ . _____ _ _ _ _ _ __ . _ _ . _ ..

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(') CS3M 13, 5.7, 5.8 - Handl ing, Preservation, Storage, and Shipment These procedures establ !sh measures f or handl ing, preservation, packaging, storage, and shipping to prevent damage to Code items.

CS3M 8.4 - Material Checkl ists i ThIs procedure def ines requirements and responsibil itles f or implementing I and maintaining material checkl ists required by the Code.

CS3M 8.3 - Welding and Brazing Materials i These procedures def ine requirements and responsibil ities f or control of i Code wel ding and braz ing material s (electrodes, fil ler wire, f l uxes, i gases, and weld insert materials) used in f abrication and assembly of Code i tem s. ,

CS3M 9.3 - Control of Welding Operations Those procedures def ine requiroments and responsibil Itles f or controlling production welding and brazing operations on Code items.

CS3M 5.5 - Heat Tratilng Proceduros Ti ese procedures def ine requirements f or controlling heat treating prctesses perf ormed by Energy Systems Group. It is applicable to heat-i O treating processes other than wold preheat and interpass tanperature, whlCh ero control' led in accordance with methods specif ied in qual lf led wol d procedure specif ications.

CS3M 7.8 - Subcontracted Furnaco Braz ing Services This procedure def inos requirements and responsibil itles f or control of subcontracted f urnace braz ing services.

CS3M 7.9 - Subcontracted Heat Treat Services This proceduro def inos relulroments and responsibil Itles f or control of 2

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subcontracted heat treat services.

CS3M 7.10 - Subcontracted Nondestructivo Examination Services This proceduro def inos requiromonts and responsibil Itles f or control of subcontracted nondestructivo examination operations perf ormed on Code material s and Itans.

i CS3M 10, 11, 5.10 - In-Process arid Final Examination, Tests, and j , i nspect ions f

i These proceduros def ino requiromonts and responsibil Itles f or examinations 3 and tests of Codo items, during f abrication and upon completion of f abricetion to assure their compliance with Code requirements.

Amend. 73 17J-69 tio v . 1982 i

_ , . _ _ , , , , _ _ _ _ __ _ ___- _ _,- - - _ ~ .

l CS3M 2.6 - Authorized Nuct car Inspector This procedure def ines Energy Systems Group requirements and responsibil itles f or assisting the Authorized Inspector in perf crming his duties, in accordance w Ith Code requirements.

l CS3M 7.11 - Procuranent Qual Ity Yorif ication Records This procedure def ines requirements and responsibil Itles f cr accumul ating records generated during design and/or f abrication of Code items at Energy Systems Group, transmitting records to the owner er customer, and retention of records by Energy Systems Group.

CS3M 5.11 - ClcanIng Procedures This procedure def ines requirements and responsibil ities f or preparing and control l ing cl eaning procedures.

O O

Amend. 73 17J-70 Nov. 1982

l O

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_ CURVE FIT K=mRe l  ;

HOLE l SIZE m b  % STD. ERROR l H0LE DIAMETER (in.) -

I -

A 42.031 -0.1087 1.26 1.000

. B 63.487 -0.1236 3.42 l .840 _

C 55.046 -0.1029 1.54 .776 0 37.264 -0.0654 0.64 l .720 RUN l N Orifice 6.0 15.865 -0.0593 3.50 ,

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Figure A.36-2 Fuel Assembly Inlet Nozzle Flow Resistance vs Reynolds Number; l Single Plate Orifices and No Orifice Plate (from Reference 4 ).

A-113 Amend. 45 July 1978

A.37 f_QPE -2M FORE-2P, which is an improved version of FORE-II, is a coupled thermal-hydraul ics point-kinetic digital computer code designed to calcul ate

,ignif Icant reactor core paraneters under steady state conditions or as f unctlons of 1Ime durIng transients. Variabt e Inl et cool ant f Iow rate and temperat ure are considered. The code cal cul ates the reactor power, the Individual react iv Ity f eedbacks, and the temperature of cool ant, el adding, fuel, structure, and additional material for up to seven axial positions.

Various Plant Protection System trip f unctions can be simulated, and the control rod shutdown worth prescribed as a f unction of time f rom the trip signal. By specif ying appropriate hot channel / spot f actors, the transient behav ior of an average, peak and hot f col rod can be analyzed. The heat of f usion accompany ing f uel melting and the spatial / time variation of the f uel-cladding gap coefficient (e.g. , due to changes in gap size) are considered.

The f eedback reactiv ity incl udes contributions due to the Doppler ef fect, coolant density chenges and dimensional changes (inct uding bowing and radial l expansion). FORE-2M is val Id while the core retains its initial gectnetry.

The original FORE-Il computer model (Reference 1) was renamed FORE-2M folicwing the incorporation of several major changes which were made to the program (Ref erence 2) . Since then, additional modif Ications have been made to the code. These incl ude updated medel ing of the gap conductance heat transf er, changes affecting material properties, modifications in transient cool ant fI cw characteristics, simulation of inter- and intra-assembiy flow and heat redistributlon, reactiv ity feedback and decay heat modif IcatIons, model changes to al Icw for al tornate f uel rod characteristics and program improvements to provide user flexibil ity. These changes are described in Ref erence 3 which al so provides the required input variables associated with these modifications.

hYALLatLLLity The FORE-2M code described in References 2 and 3, is available on the Westinghouse Power Systems CDC-7600 and CRAY-1 computers located at the MonroevIlIe Nuclear Center.

Verification FORE-2M transient results have been compared to thermal-hydraul Ic and nuclear l cal cul at ions of other codes (e.g. , DEM0, FX-2, IANUS and TAP-B). Other checks by hand cal cul ations were made f or quasi-steady state temperature distributions. The original FORE-Il code has been used extensively over the l last 17 years in the nuclear industry.

ApplItation l The FORE-2M code is used to calculate the nuclear kinetic response of the core as well as the average, peak and hot rod behavior at steady state conditions, or as a f unction of time during transients.

O A-114 Amend. 73 Nov. 1982

References

[%.s)

1. J. H. Fox, B. E. Lew ler, H. R. Butz, " FORE-f l; A Computational Program f or the Analysis of Steady State and Transient Recctor Performance," GEAP-5273, September, 1965.

2. J. V. Hil l er, R. D. Cof f iel d, " FORE-2M: A Modif ied Version of the FORE-Il <

Computer Program for the Analysis of LMRBR Transients," W ARD-D-0142, May,  !

1976. l

\

3. J. V. Mil l er, R. D. Cof f lei d, K. D. Daschke, et. al . , " Suppl ementary Manual !

f or the FORE-2M Computer Program," CR3RP-ARD-0257, September,1982.

1

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l v A-114a Amend. 73 Nov. 1982

TABLE B-1 o

PRELIMINARY DESIGN DUTY CYCLE EVENT FREOUENCIES Event Frecuency

! 1. Normal Events tb1 Dry system heatup and cool down, sodium 5 total system + 8 fili and drain per ioop + 17 additional for entire intermediate loop exclusive of lHX N-2a Startup f rm ref uel ing 140 N-2b Startup f rom hot standby 700 tb3a Shutdown to refueling 60 N-3b Shutdown to hot standby 21 0

'l N-4a Loading and unloading 9300 (loading) t 9300 (unloading)

I

N-4b Load fIuctuations 46500 (up)

.\

46500 (down)

N-5 Step load changes of i 10% of f ull 750 (+10%)

Ioad 750 (-10%)

N-6 Steady state temperature fl uctuations 30 x 10 6 N-7 Steady state flow induced vibrations 1010 (sodium)

2. Uoset Events U-la Reactor tr Ip f rom f ul I power w Ith h normal decay heat III U-Ib Reactor trip f rom f ul I power with 180 minimum decay heat U-1c Reactor trip f rom partial power w ith kj minimum decay heat /

U-2a Uncontrol led rod insertion 10 0-2b UncontrolIed rod w Ithdrawal from 10 100% power O II'

- The total frequency for U-1 is associated w lth normal decay heat f rom V f ulI power so as to balance the trips associated wIth partial decay heat f or events U-2 through U-23.

B-25 Amend. 73 Nov. 1982

TABLE 8-1 (Continued)

Event Frequency U-2c Uncontrolled rod withdrawal from startup with automatic trip 17 47 U-2d Uncontrolled rod withdrawal from startup to trip point with delayed 3 manual trip U-2e Plant loading at max. rod withdrawal rate 10 47l U-2f Reactor startup with excessive step 48 power change 50(2)

U-3a Partial loss of primary pump 2 per loop U-3b Loss of power to one primary pump 5 per loop U-4a Partial loss of one intermediate pump 2 per loop U-4b Loss of power to one intermediate pump 5 per loop U-Sa Loss of AC power to one feedwater pump motor 10 U-5b Loss of feedwater flow to all steam generators 5 4d U-7a Primary pump speed increase 5 U-7b Intermediate pump speed increase 5 U-8 Primary pump pony motor failure 5 per purr 1 U-9 Intermediate pump pony motor failure 5 per pump U-10a Evaporator module inlet isolation valve closure 4 per loop U-10b Superheater module inlet isolation valve closure 2 per loop U-10d Superheater module outlet isolation valve closure 2 per loop

() - These events are part of the startups specified for event N-2b and should 47 not be added as separate startups.

Amend. 48 B-26 Feb. 1979

-w _

--- _a O

CLINCH RIVER BREEDER REACTOR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT O

APPENDIX J PRA PROGRAM PLAN l

PROJECT MANAGEMENT CORPORATION iO l

TE LE OF CONTENTS EAGE

1.0 INTRODUCTION

1 2.0 OVERVIEW OF THE RISK ASSESSENT PROGRAM 1 2.1 Accident initiator Development 1 2.2 Plant Logic Model Development and Quantification 3 2.2.1 System Functional Event Tree Development 3 2.2.2 System Functional Fault Tree Development 4 2.2.3 Analyses of Plant Response 5 2.2.4 Accident Sequence QuantifIcation 6 2.2.5 Uncertainty Analyses 7 2.2.6 Common Cause Fail ure Analyses (CCFA) 8 2.2.6.1 Expl icit Model ing 8 of Dependencies 2.2.6.2 Qualitative CCFA 8 2.2.6.3 Detailed CCFA 9 2.2.6.4 Special CCFA 10 investigations 2.3 Core and Containment Accident Model Ing 11 2.3.1 Phenomenological Event Trees 11 2.3.2 Source Term Evaluation 12 2.4 Heal th Consequence Analysis 13 O

Amend. 73 Nov. 1982 J-I

}

TMLE OF CONTENTS (Continued) 2.5 Risk Analysis 13 2.6 PRA APPLICATIONS TASKS 13 2.6.1 Operator Action Event Trees 14 2.6.2 Assessment Of The Ef fectiveness 14 Of Postulated Design Variations incl uding Consequence Mitigation Features 2.6.3 improve Underst anding of the Pl ent 14 2.6.4 Characterization Of Risk From Early 15 Life Fail ures 2.6.5 Impianentation Of A Continuing Risk 15 Management Program 2.6.6 Input To The Site Emergency Procedures 16 2.7 Interaction W Ith the NRC 16 2.8 Accidont Delineation 17 2.9 Study Limitations 18 3.0 PRA PERFORMANCE AND REV IEW 19 4.0 SWEDULE, MILESTONES, AND RESOURCE ALLOCATION 20 TE LE I CRBRP PRA PRODUCTS 21 TM LE 2 RESOURCE ALLOCATION 22 TMLE 3 CURRENT SWEDULE OF UNCOMPLETED TASKS 23 FIGURE 1 PRA FLOfWART 24 FIGURE 2 GBRP PRA ORGANIZ ATION 25 FIGURE 3 ANTICIPATED PRA INTERNAL REV IEW PLAN 26 FIGURE 4 EXPECTED SWEDULE OF PRA PRODUCTS 27 Amend. 73 y

Nov. 1982

h w/

PROGRAM R.AN FOR THE CLINCH RIVER BREEDER REACTOR PLANT PRGM31LISTIC RISK ASSESSENT (PRA)

1.0 INTRODUCTION

This Program Pl an describes the pl an for the CRBRP PRA and appi Ications of the study. The PRA was initiated principally related to the desire of the project to perform an Integrated saf ety assessment as one ingredient in the decision process leading to safe design and operation. The PRA w il l al so satisfy the requirements of NUREG-0718, Section ll.B.8 and it is consistent with the current direction of the NRC in development of safety goals and PRA applications. As detail s are developed, they wilI be incorporated into this

'rogram Pl an. This review process is discussed in Sections 2.6 and 3.0. The PRA will be used as an aid in evaluation of the current design and alternative designs and as a tool to provide f urther assurance of safe pl ant operation.

The PRA w il l be, in the terminology of the PRA Procedures Guide (W REG /CR-2300, Rev. 1), a Level lll PRA. In addition to the tasks which comprise a Level ill PRA, several tasks will support appl Ication of the study.

These tasks incl ude use of the PRA model s to support operational programs such as emergency procedure preparation and operator training. The purpose of using the PRA in this way is that it can serve as a basis for defining potential operational incidents, thereby hel ping to reduce uncertainties caused by the I imited availabil Ity of LMFBR operating experience.

2.0 OVERVIEW OF THE RISK ASSESSMENT PROGRAM The PRA can be divided into the following major elements: accident initiator development, pl ant logic model s f or system f unctional event trees, and system functional f ault trees, phenomenological event trees, and release and consequences analysis. The phenomenological event trees are divided into two groups; those that describe the phenomena frcrn core melt to breach of the reactor vessel (i.e., core damage phenomenological event trees) and those that describe the phenomena from breach of the reactor vessel to containment integrity f ail ure (i.e., containment phenomenological event trees). Refer to Figure 1 for a flowchart depicting how major elements will be tied together.

These elements trill be discussed in the following subsections.

2.1 ACCIDENT INITIATOR DEVELOPMENT The approach being taken to logic model construction emphasizes the investigative nature of the task and results in an iterative model building process which ensures the accuracy of the f Inal Iogic model s. The folIowing describes the iterative investigative method to be used.

A prei iminary Iist of Initiating events wilI be developed by extracting inf ormation f rom a variety of rel evant sources. These sources include:

o Comollations of Generic Exoerlence: Examples are NUREG-0460 and EPRI p NP-2230. In addition, NSAC has produced a screened Iist of LERs which d identifles a number of risk significant FWR Initiating events.

Amend. 73 J-1 Nov. 1982

o Prevlous PRAs: A numtur of other PRAs have been completed or are on-going. Each of these PRAs has compiled a I ist of inl+1ating events ,

(of ten f rom the generic sources l Isted above). I o CRBRP Pro fect Documentation: A number of design-specific documents are being scrooned to identify potential initiating events. These incl ude the PSAR, pl ant design descriptions, Key Systems Review, Avail abil ity  ;

Analyses, and the existing Reactor Shutdown System (RSS) and the Shutdown Heat Removal System (SHRS) rol labil Ity assessments, (see PS AR Appendix C).

o Breeder Reactor Exoerlence: Incl uding f oreign and domestic sodium and/or breeder experience.

The resultant list of initiating events will allow the event tree and f ault tree analyses to commence, but is not considered the final list. It is important that information gained during the event tree / fault tree analyses be continuously fed back into the task of Identifying initiating events. By definition, an important initiating event is one that can evolve into an important sequence, it is impossible, theref ore, to conf idently list al l the important initators before the event tree and f ault tree analyses have been performed. This process systematically util Izes knowledge gained in the event tree /f ault tree analyses to ensure that all important initiating events are identifled. This approach is based upon the recognition that: (1) Important initiators are either relatively high frequency events or are events which adversely impact the abil ity of the safety systems to respond, and (2) an initiating event must, by definition, require an active plant response to avold core damage. The cut-sets of the f ault tree model s w11I be systematically examined for their relationship to event tree headings (developed f or the prel iminary l ist of initiators). It is then possible to identify any failure events which both call for an active pl ant response and adversely impact the perf ormance of the saf ety systems. These types of events will be considered as potential initiating events.

The initiating event development requires performance of a f ault tree analyses of the Initiating event where appropriate. By comparing the cut-sets of the initiator wIth those of safety systems required to respond to the initiator it is possible to ascertain whether the specif Ic cause of the initiator could al so impact the abil ity of saf ety systems to respond. This additional step in initiating event identification ensures the accurate quantification of conditional probabilities and allows an Initiating event to be broken down into subevents to highl ight potential dependencies between the initiator and subsequent events. As an example of this step, the " loss of of fsite power" event would be identif ied in the cut-sets of a " loss of feedwater" initiator and al so show up as an element in the cut-sets of the Shutdown Heat Rmoval Systems; accordingly, " loss of of fsite power" is always identif ied as a separate Initiating event.

Thus, the proposed approach is en iterative process of Initiating event identif ication which starts w ith the cppl Ication of avail able compil ations of operating experience and design information and f eeds back crucial inf ormation f ran the ensuing event tree /f aul t tree analysis. In this way, completeness is assured not only by searching available compilations of data but al so by expl icitly and systematically investigating the CRBRP pl ant design.

Amend. 73 J-2 Nov. 1982

e' 2.2 PLANT LOGIC MODEL DEVELOPMENT AND OUANTIFICATION Considerable attention will be devoted to the task of constructing accurate logic model s. The design of the CRBRP which is being analyzed was that in pl ace as of February 1,1982. The analysis will be updated to reflect the current design at several stages during the PRA; specif ically, any changes derived from the Construction Permit licensing review will be included in the f Inal model s.

The specific activitles included under the general heading of Plant Model Development and Quantif ication are the development of system functional event tree and f ault tree logic models, the analyses of plant response, accident sequence quantif Ication, uncertainty analyses, and common cause f ail ure anal y ses. The result of the plant model development and quantification tasks is a compilation of those probabilistically quantified accident sequences which lead to core damage and their particular plant state. These plant states are then the entry point to the analysis of the severity of damage resul ting f rcrn an uncool able core w ith in the primary cool ant boundary (i.e.,

core damage phenomenological event trees) and to the severity of challenge to containment integrity associated with the accident sequences which breach the primary coolant boundary (i.e., containment phenomenological event trees).

Damage analyses will be discussed in subsequent subsections.

2.2.1 Svstem Functional Event Tree Deveicoment System functional event trees (ET) will be constructed for the initiating O events discussed in Section 2.1. Bef ore actual event tree construction begins, ti,e numerous Individual initiating events will be grouped into prel iminary cateaorles based upon their impact on the plant and the subsequent demands upon the pl ant saf ety systems. The approach to be used to perform the event tree analyses f ol low ing prel iminary categorization of Initiating events is as f ol lows:

1. Determine the f unctional requirements which must be met in response to the initiating event. Examples of such functions are reactor shutdown and decay heat removal .

2. Def ine the pl ant systems avail abl e to perf orm each of the necessary functions.

3. List alI of the supporting systems which are common to the pl ant systems identif led in Step 3. Examples of supporting systems are service water, instrument air, and electrical power. This information will assist in the identification of potential dependencies between systems and wIlI af fect the IdentIf Ication and ordering of event tree headings.

4. Identify operator actions associated with the systems identified in Steps 2 and 3.

5. Define potential functional dependencies between the pl ant systems identifled in Step 2. This is the f irst step in an iterative process of IdontIfyIng important system interrelationships which are not apparent by merely listing common hardware.

Amend. 73 J-3 Nov. 1982

6. Perf orm the necessary analyses to determine timing of events, systems requirements, and the corresponding f ail ure state definitions.

7. Construct the system level event trees. The two el ements of constructing system event trees are determining: (1) the definition of the individual event headings and (2) the ordering of the events to produce the logic model for the event tree.

8. Document the event trees. The assunptions and reasoning which produced the event tree f orm are caref ully documented.

Al though the above outl ine describes a step-by-step process, the event free construction process will be an iterative one in which the f ailure state def initions, timing, and system headings are continuously influenced by inf ormation f ed back f rom the f ault tree development and the pl ant response anal y ses.

2.2.2 System Functional Fault Tree Develooment Fault trees will be drawn f or most of the event tree headings. Decisions concerning the necessity to develop individual f ault trees w li l be based upon the recognition that the purpose of a f ault tree is to: (1) quantify the probabil Ity of an event f or which no statistically acceptable data exist by logically breaking down the event into its constituent parts for which acceptable data do exist, and/or (2) Identify potential dependencies among mul tipl e sy stems. Faul t trees w il l not be drawn f or systems f or which: (1) acceptable data exist f or the event heading and no significant dependencies coul d exist between this event and subsequent headings, or (2) the event heading could not be involved in any risk-important accident sequences even if its conditional f ail ure probabil ity were extremely high.

The f ault tree (FT) analyses will be perf ormed using procedures and symbols presented in NRC's Fault Tree Handbook (NUREG-0492) or in the PRA Procedures Gu ide ( NUREG/CR-2300, Rev. 1 ) .

A package of FTs w il l be provided f or:

o Shutdown Heat Removal System Top Logic o Primary and Intermediate Heat Transport Systems o Steam Generator System o Main Feedwater and Condensate o Turbine Bypass Valves and Cordenser o Steam Generator Auxil iary Heat Removal System o Auxil iary Feedwater o Direct Heat Removal Service o Normal & Emergency Chilled Water 9 Amend. 73 J-4 Nov. 1982

p o Pl ant Service Water

'G o Class IE Electrical o Containment Cleanup o Annulus Filtration o Annul us Air Cool Ing o Compressed Gas o Containment isol ation Each FT w il l include all known support systems such as electric power, instrument and control, instrument air, and service water.

In addition, a f ault tree data base will be produced which will allow quantification of all f ault trees. The data base will be derived f rom the same sources as those Identified for accident initiator development.

2.2.3 Analvses of Plant Resoonse Analyses of realistic plant responses to postulated accidents will be perf ormed throughout the course of the PRA to ensure that the pl ant logic model s f or ETs and FTs represent an accurate picture of the pl ant response and that all dominant risk contributors have been identified. Analyses wil l al so d be performed to assure that the system success criteria (i.e., successf ul system f unction) is real istically based on the physical capabil ities of the plant.

A key element of the approach to this task is the ef ficient, systematic identification of specific analytical needs. Due to the costs and time delays in obtaining best-estimate pl ant response data, it is crucial that the analyst be able to determine: (1) what analyses are necessary for the PRA, and (2) what are the specific inputs and desired outputs of the analyses. Event trees will be systematically used to make these determinations. At each branch point in the event tree, the analyst will ask:

o What is required of the plant systems to maintain the necessary functions?

o What are the existing plant conditions important to maintenance of these f unctions?

o What are the real Istic capabil ities of the systems under these conditions?

These questions will determine the basic analytical requirements. When answers to these questions are not readily available to the PRA analyst, one of three avenues will be pursued to supply the needed information:

(1) Locate applicable analyses in available documentation.

Amend. 73 J-5 Nov. 1982

( 2) Perf orm hand calculations or extrapolation of existing analyses.

Analytical needs not satisfied by available docurentation can be satisf lod by hand calculations where appropriate.

(3) Perf orm additional computer analysos. This avenus will be used only when documented analyses are not available and hand calculations would be insuf ficient. The role of the risk analyst at this point is to ensure that the results of the PRA are truly sensitive to the results of the desired analysos and to define the analytical requirements as caref ully as possibl e.

The pl ant logic model s w ill be modif ied where incorporation of best-estimate cal cul ations w ill more accurately model actual pl ant behavior under accident conditions.

The heat transport and heat sink systems wIll be characterized in terms of their heat removal capabil itles. Best-estimate cal cul attens w il l provide the heat removal rate f or one , two , and three-loop operation of the heat transport system under f orced and natural circulation conditions. The characterization w lli al so assess the anount of heat transported to the steam drum in excess of the capabil Ity of the closed-cycle sinks (i.e., Protected Air Cooling Condensers). In regard to the heat sinks, the capabil ities of the PACCs w ill al so be characterized f or f orced and natural draf t and capabil itles. The Steam Generator Auxil iary Heat Removal System vents will be characterized f or operation with dif ferent Auxil lary Foedwater pumps and vent s. Characteristic values of the heat transport and heat sink systems will allow determination whether various Loss of Heat Sink sequences will satisfy the success criterion. .

The success criterion (prevention of core damage) for the l oss-of-heat-sink event will be f ormulated and the analysis supporting the rationale leading to the criterion w lil be provided. The analyses w il l be best-estimate and the criterion w lil be simply-stated but w Ill bo broadly appl Icable to accident sequences l eading to a l oss-of-heat-sink.

The core damage event tree will trace through the phenomenological steps f rom the initial conditions in the core to the release of core material from the reactor vessel, i f any. Event frees w Ill be developed f or all accidents identif ied by the event trees as leading to core damage. The pl ant states identified in the event trees wIll be carried through the coro damago event treo to the containment event trees. Quantif ication at each node will be derived f rom an understanding of the basic phenomenological processes invol ved in the event.

2.2.4 Accident Seouence Oaantification Using the initiating events, ETs, FTs, and associated data bases, the dominant accident sequences w Il l be quantif ied. The products of th is task w il l incl ude:

o A description of the process used to "I Ink" the f aul t trees together and to ensure that all dependenclos among system f ault trees are idontif ied and incorporated into sequence quantif ication.

Amend. 73 J-6 Nov. 1982

q o A l Isting of the dominant accident sequences and a description of each incl uding the Individual events comprising each sequence.

o A systematic justification for anitting any ET sequence fra the dominant l ist (e.g., sequence Z, while producing the same ef fects, has an occurrence frequency four orders of magnitude less than sequence A).

The computer code which will be used in the generation of cut-sets and accident sequence quantification is COMCAN lli (0014non Cause Analysis) developed by ldaho National Engineering Laboratory (INEL). Quantification wIlI be an Iterative prtacess in which early analysos are used to help focus the more detailed common cause f ail ure analyses, (see Section 2.2.6).

2.2.5 Uncertaintv Analvses Early in the PRA sensitivity studies will be utilized to provide information on the relativo importance of equipment and human f ail ures. Detail ed uncertainty analyses w il l be del ayed until later in the PRA progran.

Af ter the best-estimate quantification, a detailed uncertainty analysis will be perf ormed in order to establ Ish uncertainty bounds in the overalI results of the PRA study. Estimates wilI be made of the probability distributions or conf idence l imits for those component f ail ure rates and event f requencies which are potentially important to risk. The l Imitations of the uncertainty analyses w il I be addressed. The assessment methodology wIlI include eval uation of the uncertainties on the importance of accident sequences which are known to have Iarge uncertainties. The work w Il I al so establ Ish V conf idence I imits f or total core damage f requency and a cumul ative probabil Ity distribution curve. Probabil Ity distributions or conf idence I imits w11I be estimated only for those component f ailure rates and event f requencies which are potentially important to risk. For sequences, such as seismic, which are known to have large uncertainties, the assessment methodology will include eval uation of the of fec;., of these uncertainties on the importance of the accident sequences. Any uncertainties which cannot be quantified will be qual Itatively discussed.

Sensitivity analyses will be performed using the plant logic models to identify those input data or assumptions which significantly af fect the dominant accident sequence l ist. These sensitivity analyses will focus on (1) uncertain f ail ure rates, or (2) uncertain assumptions concerning success criteria or system dependencies. The final product of this task will incl ude:

o A description of each sensitivity analysis, incl uding motivation, o The results of each analysis, o interpretation of each analysis including the impact on detailed common cause f ail ure analyses.

l l

Amend. 73 J-7 Nov. 1982

2.2.6 Ccann Cause Failure Analvses (CCFA)_

The common cause f ail ure analyses are broken down into four subtasks. These include: expl icit model ing of dependencies, qual Itative CCFA, detailed rCFA, and special CCFA Investigations. These subtasks are delineatea in the f of Iow ing f our sections.

The CCFA is an assessment of common f alI uro susceptibII Ity and opportunity, incl uding quantIf IcatIon of common cause events f or:

a. Internal events (such as temperature and pressure extrmes and transients, common locations, proximity to degrading influences, fires); and

b. external events (such as seismic events, tornados, floods, 1Ightning, chemical, radiation, explosions and aircraf t or misslie impacts).

2.2.6.1 fyolIcit ModelIna of Deoendencies This portion of the CCFA entalls those of forts required to ensure that alI common support systems and f unctional dependencies between and among plant systems are expl icitly and accurately incl uded in the plant logic model s.

This task will be carried out in the process of constructing the event and fault trees.

All known Inter- and intra-system dependencies w il l be model led (such as common support systems identif ied in the f ault trees), and performance of prel iminary analyses of f unctional deperdencies in which the physical response of the plant to the f ailure of one system has an adverse af fect on the abil ity of another system to operate, and potential operator errors which can disable multiple systems or provide the l Ink between f ailure of one system and another. The specif Ic products of this task wlll Inctude:

o A description of all known inter-system dependencies resul ting f ran common support or interf act ng systems.

o A " system level" f ail ure modes and ef fee.ts analysis (FMEA) in which the physical plant response to ali event tree f alI ure modes Is described in terms of the behavior of major parameters (e.g., f ail ure of system A to start w ill result in a rapidly rising pressure in syctem B).

o A l ! sting for each event tree of the potential impacts of major parametric changes (e.g., a rapidly rising pressure will trip of f pump C and cause f ailure of system B).

o A description of the more detailed dependency analyses to be performed and a prel iminary approach f or perf orming these analyses. An approach to treating external events such as seismic events will be included.

2.2.6.2 OualItatIve CCFA There may be f ailure causes common to multiple components which f all below the Amend. 73 J-8

(" practical level of resolution in the event trees and f ault trees. Examples of

( ,)) such potential common f ailure causes are:

o manufacturing, installation, or maintenance errors o adverse environmental infl uences such as high temperature, humidity, or radiation o corrosion, carburization, rust, or other chemical degradation processes.

In this subtask, these types of common cause f ail ures which coul d potentially have a significant Impact on plant risk will be identified. This will be achieved by a conservative f iltering process. This filtering process will allow the subsequent, more detailed analyses decribed below to f ocus on those dependencies which could actually be important to risk. This filtering process is based upon the recognition that for a common cause f ailure to occur, two criteria must be met:

(1) Both components must be susceptible to f ailure by the common cause (e.g., two dif ferent val ve operators might both be susceptible to f ail ure due to flooding, but a pipe and valve do not share a common susceptibil Ity to f ailure by flooding).

( 2) The common cause must have the opportunity to af fect both components.

1 Based on th is recognition, a two-step qual Itative CCFA w il l be perf ormed. The s_s/ first step involves determination of which redundant components are susceptible to f ail ure by the common cause. The second step Involves determination of which susceptible redundant components (identified in the f irst step) are located such that an event could subject them to the common cause. Any susceptible redundant components so located are considered common cause candidates.

2.2.6.3 Detailed CCFA The input to this subtask will be the relatively small number of common cause candidates which survive the screening process discussed above. In this subtask a more detail ed assessment of common f ail ure susceptibil ity and opportunity will be perf ormed and probabil Itles estimated for these common cause events.

The detailed CCFA will address an extensive I ist of potential common f ailure causes (e.g. , vibration, high temperature, etc. ) for each component and will determine the potential for these mechanisms coincidentally af fecting the com ponent s. For redundant components in dif ferent locations, this will entall eval uating the i Ikel Ihood that causes can be coincidentally present in both l ocat i ons. For components in the same location, this will entall a determi-nation of whether the components are both (or all) susceptible to the same causes and the I ikel thood of those causes existing in that particular l ocat ion.

O Amend. 73 J-9 Nov. 1982

2.2.6.4 Special CCF]Lifutestigat torts The above three subtasks will allow a practical, of fectivo CCFA to be perf ormod f or most potential falIuro mochan!sms. Howover, there are additional potential causes f or multiple f ailuros which will be addressed separately. These are f Iros, solsmic ovents, and other signif icant external events.

fires The cmmon location analysis perf ormed (soo Section 2.2.6.2) will form the basis f or the fire analysis. The location analysis will provido:

o List of key locations with potential for exposure to combustibl es, an oxidizer, and an Ignition sourco.

o Key components in these locations o Fire related f ail ure modes of these key compononts.

Based on this Inf ormation, a preliminary scoping f iro analysis will be perf ormed to determine if f iro-rel ated accident sequences coul d contribute signif icantly to risk at the CEBRP.

Should the above scoping analysis identify any single or double location cut-sots which coul d real istIcally support a f Ire of suf fIclent size and duration to f all the components associated with these cut-sets, a more detailed f iro analysis will be perf ormed f or these specif ic locations. Thus, the scoping f ire analysis will be used to focus any detain ed fire analyses which are required on those particular fire-related sequences which could potentially contribute to risk.

Seismic and Other External Events External events such as seismic events, tornados, floods, explosions, aircraf t or missile impacts, etc. will be ovaluated to ascertain their signif icance to risk. Al so, a detalI ed seismic methodology wIl I be devel opod.

The pre!Iminary analysis will be comprised of six basic steps:

(1) Estimate the f requency of occurrence of each external event.

( 2) Identify the specif ic components or systems which could be adversely impacted by the event.

(3) Cal cul ate the f alI ure probabII Ity of such equipment and recalcui ate the probabil Ity of core damage wIth these components or systems unavailabie.

(4) Multiply the f requency of occurrence f rcm (1) by the conditional probabil Ity of core damage f rcm (3),

(5) Compare the resul ts of (4) to the basel Ine core damage f requency.

(6) If the f requency comparison in (4) indicates the sequence is risk significant, considor phenomenological aspects of the sequences.

Amend. 73 J-10 Nov. 1982

If the results of the preliminary Investigation Indicate that there are I potentially risk significant sequences initiated by one of the external events b) a more detailed analysis will be performed. The approach to this more detailed analysis will be very similar to that outl ined above for the prel iminary analysis. However, conservatisms in the prel iminary investigation w111 be repl aced by real istic eval uations of the impacts of the initiating event on pl ant systems. The overall risk shall include +he contribution f rom external events.

2.3 CORE AM) CONTAINMENT ACCl DENT MODELlNG The result of the tasks on Plant Logic Model Development and Quantification will be a set of probabilistically quantified dominant accident sequences each of which is expected to produce damage to the core. Sequences which do not iead to core damage wilI be identif led. AU.aciated w Ith each accident sequence ieading to core damage w!!I be a plant state which w11I include:

1. An Indication of the successf ul operation of the Plant Protection Sy stem.

2. An Indication of the availability of mitigating systems, (i.e.,

Containment isol ation, Annul us Air Cool ing, etc. ).

3. An Indication of the capabil Ity of structures and surf aces to act as static and convective heat sinks during the accident sequences.

The def inition of pl ant state f or accident sequences terminated by core damage O w ilI aliow two eval uations to be perf ormed. First, the potential for varlous degrees of mechanical damage to the primary system resulting f rom energetic disassembly of the core can be eval uated. Second, the potential for f ailure of the containment system to malntain its integrity following a variety of severe accident sequences can be eval uated.

2.3.1 Phenomenoloalcal Event Trees Phenomenological event trees w ill be prepared f or both core damage and containment behavior resulting f rom accident sequences that lead to core damage (i.e. core damage phenomenological event trees and containment phenomenological es a trees). The accident sequences wilI include those which could potenti . ly lead to core energetics as wel! as those which have no significant energetics associated with the core disruption. The combined core damage and containment event trees w IlI sequential ly start w Ith a def inition of the plant state and sequentially terminate with a description of either a j stable coolable state for the core debris or the time and size of the containment f ail ure. As part of this eval uation, the radioactive source term above the operating floor at the time of a stable end point or containment f all ure w IlI be def ined. The event trees w 11 i describe, in detaII, the major physical processes occurring within the primary system and containment which precede, cause, and follow, hydrodynamic core disassembly and/or loss of core cool abil ity. Th is w il l incl ude consideration of the thermal margins provided by the CRBRP design to mitigate the consequences of core damage as welI as the structural margins to mitigate energetic ef fects and minimize a direct release (s of sodium and radionuct Ides f rom the primary system through the reactor vessel

( head. Both the core damage and containment event trees will be quantified.

l Amend. 73 J-11 Nov. 1982 i

The bases f or selecting probabil ities f or each node will be documented.

Development of the phenomenological event trees w il l incl ude analyses, as f ol l ow s:

o Thermal-hydraul ics eval uation of the loss of decay heat removal following reactor shutdown including thermodynamic and heat transfer eval uation of the primary heat transport system.

o Extrapolation of currently available CACECO analyses to apply to the loss of decay heat removal fol low ing reactor shutdown.

o Structural calculations to assess the structural integrity of systems and components wnere necessary to support the phenomenological event trees.

The products of these analyses w ill incl ude def inition and probabil istic quantifIcatIon of the range of potential sequences by which Iarge quantities of radionuci Ides might be released f rom containment following a variety of accident sequences which produce core damage. These containment phenomenol ogical sequences will define the conditions under which detailed analyses of the radionuci Ide source term frm containment will be completed.

The core damage event tree will trace through the phenmenological steps f rm the initial conditions in the core to the release of core material from the reactor vessel, i f any. Event trees w il l be developed f or al l accidents identif ied by the event trees as leading to core damage. The pl ant states identified in the event trees will be carried through the core damage event tree to the containment event trees. .Quantif ication at each node w il l be derived f rm an understanding of the basic phenmenological processes involved in the event.

These phenomenological processes and the considerations leading to the sol octIon of the quantIf Ication w ll l be documented. The resulting core damage states will be grouped on the basis of energetics (abil ity of structurally loading the primary system) and on the basis of the degree of core melting (abil Ity of thermally loading the primary system).

The containment event tree wIlI combine the phonmonology of the moiten core and soitum reacting with the environment outside the reactor vessel and the response of the containment / conf inement system, beginning with the pl ant states previously identif ied and a small number of core damage states. The quantifIcation of the phenmonologIcal branch w11I be based on the understanding of the processes, the results f rom existing analysis and new supporting calcul ations where necessary. Quantif ication of the avail abil ity of containment system responses w ill be provided in the pl ant states.

2.3.2 Source Term Evaluation An analysis will be performed to def ine the environmental source term for each of the unique paths through the containment phenomenological event trees for which signif icant releases of radionuci Ides are expected. Existing computer codes which w ll l be util Ized in th Is analyses incl ude CACECO, HAA-3, and CO W ADEX.

O Amend. 73 J-12 Nov. 1982

o The potential for release and related health ef fects f rom ex-core sources of radionuct Ides wIlI be def ined. These sources w ll l be eval usted using f ault tree analysis techniques and the appropriate source terms given various plant system responses. Ex-core sources incl ude radioactive cover gas, ex-vessel spent f uel storage, and other auxil lary systems.

2.4 HEALTH CONSFOUENCE ANALYSIS The ex-plant consequence analysis wif I characterize the disfribution of publ Ic health ef fects which can result f rom accidents involving core damage and signif icant radionucl ide releases to the environment. Results f rom this analysis will assess the uncertainties in publ Ic health ef fect distributions which result f rom uncertainties in predicted accident sequence probabil ities and radionuct Ide releases f rom containment.

The characterization of health consequences will be accompl ished using the CRAC 11 computer code together with the meteorological and demographic data f or the CRBRP site.

The health consequences associated with the release will be defined based on the source terms derived for each release category. The study wil l use state-of-the-art model Ing codes which accurately project doses for the LMFBR source i terms. The health consequences w ill examine both acute f atal ities and l atent cancer fatalities.

2.5 RISK ANALYSIS O Based on health consequences and sequence probabilities derived in the above V tasks, an overal l assessment of the risk shal l be provided, including a breakdown of the major contributors to the risk. This assessment of the dominant contributors to risk shall incl ude design and operational aspects, and sensitivity to key 6ssumptions, and shall be kept current with knowledge of the pl ant and the PRA model .

2.6 PRA APPLICATIONS TASKS The purpose of this section of the plan is to summarize the tasks which will be used for appl Ication of the PRA.

A number of PRA appl ications w il l be implemented. These appl ications rely on two characteristics of the PRA results:

1. The PRA is a complete description of the accident sequences which have the potential to cause damage to the core;

2. The PRA incorporates suf ficient Information to provide a quantitative ranking of the importancc of equipment f ail ures and human errors to both the frequency of core damage and the publIc health risk.

The use of these characteristics in a variety of appl Ication tasks is discussed below.

O Amend. 73 J-13 Nov. 1982

4 l

2.6.1 Ooerator Action Event Trees f)perator Action Event Trees (OAETs) wil l be developed. 0AETs are a method to investigate the role of the plant operation staf f in important accident sequences (Ref. NUREG/CR-1440). The analysis addresses three f undamental questions:

1. What actions can (or must) the operator take in respor.se to a specif Ic accident condition?

2. What information is required by the operator to take this action?

3. What instrumentation is necessary and suf ficient to provide this inf ormation?

By developing logic model s and supporting information which allow these questions to be addressed systematically, a very detailed description of the operator's role in managing an accident sequence can be developed. This description will al so provide inf ormation about the specif ic role of pl ant instrumentation in inf orming the operator of the status of the pl ant. The complete set of OAETs WILL consist of one tree appl icable f or each dominant accident sequence. Common characteristics of a number of dominant sequences will allow the total number of OAETs required to be reduced to f ewer than the number of dominant sequences.

2.6.2 Assessment of the Effectiveness of Postulated Design Variations including Consecuence Mitigation Features Models developed during the PRA will be util ized to assess the potential benef its or Iack thereof associated wIth postui ated changes in pl ant design.

These changes may be oriented toward reducing the f requency of events which produce core damage or mitigating the consequences of these events.

The present design of the CRBRP containment includes a number of systems designed to mitigate accident consequences. A quantitative display of the ef f ects of these f eatures on the risk f rom the CRBRP w il l be developed. Such a comparative eval uation is cal led f or in NUREG-0718, item II.B.8. This eval uation w il l include sensitivity studies in which the ef fectiveness both of currently designed and of postuiated consequence mitigation systems can be assessed.

In addition, a search of dominant accident sequences will be conducted to assess whether cost ef fective modifications to the existing design can be post ul ated. Where such potential ly usef ul modif ications are identif ied, a more detailed evaluation of alternative approaches to reducing the risk contribution f ran one or more dominant accident sequences can be performed.

This eval uation can incl ude assessment of feasibil ity, ef fectiveness, and cost of a variety of postui ated changes.

2.6.3 Imorove Understanding of the Plant Ar:ditional PRA appl ications will be undertaken to f actor insights gained in ihe conduct of the PRA Into the design and operation of the plant. These appl Ications w il I :

Amend. 73 J-14 Nov. 1982

l h

U

1. Supplement the existing programs designed to address operator aids including Reg. Guides 1.47, 1.97, and NUREG-0497. The PRA w il l be used to define and rank the risk significance of alarms and instrumentation which are designed to improve the operator's abil liy to prevent and mitigate the consequences of severe accident sequences.

2. Assist in the development and val Idation of emergency procedure guidelines.

3. Provide inf ormation on the Integrated perf ormance of pl ant systems and instrumentation for use in evaluating the design and util ization of the pi ant simul ator, as welI as to train operators and other piant personnel.

4. Assess of the sensitivity of the CRBRP risk to uncertainties in the rol labil Ity of equipment required to perform its f unction in a degraded environment, if appropriate, alternative design features intended to reduce the sensitivity of the overall plant risk to these uncertainties w ill be def ined and eval uated.

5. Evaluate the risk contribution and sensitivities to the testing interval of equipment and to the allowable on-line maintenance i nterval . This eval uation will allow Technical Specifications to be implemented in a manner which assures the minimum plant risk without unnecessarily restricting plant operation during.the maintenance of saf ety-rel ated equipment.

2.6.4 Characterization of Risk From Early !_Ife Failures The approach used to assess with the f ailures anticipated to occur during early years of operation will incl ude two important elements. Both of these elements involve the careful screening of avalIable operational data and resul ts f rom the PRA. The f irst element is to analyze the data to focus on the potential for systematic recurring f ail ure causes and to identify measures which have been successf ully used in the past to contend with these causes.

The second element is to focus on the equipment which has or is expected to produce the most signif icant operational problems and to define operational, maintenance, or training programs which might reduce the severity of these specif ic equipment f ail ures.

2.6.5 lmolementation of a Continuina Risk Manaaement Proaram The PRA will have appl ication as a tool to evaluate operational experience and to address l Icensing issues which will arise during the operation of the pl ant. Implementation of such a continuing risk management program will incl ude:

1. Formal Ization of the model s and documentation developed during the PRA to f acil itate ease of long-term util Ization;

2. Transfer of the PRA technology and associated tools to the TVA p operatons staff; Amend. 73 J-15 Nov. 1982

3. Definition of a TVA program by which the PRA and its associated documentation cna be updated to reflect the current state of the pl ant design and operation as well as current operational experience.

The risk management program will allow applications to be carried out by the TVA plant staf f throughout the i If e of the pl ant. The program wilI include eval uating operational experience and addressing i icensing issues which might arise during operation of th9 plant.

The program will al so provide assurance that operational and back-fit decisions will be based on a realistic and complete understanding of the important saf ety characteristics of the pl ant. ThIs understanding wIlI be infl uenced by experience gained in the operation of the pl ant.

2.6.6 input to the Site Emercency Procedures The PRA will be used in the development and implementation of. the site emergency procedures. The use of the PRA in this role is supported by the f act that it embodies a description of important accident sequences which includes estimates of the timing of significant radionucl Ide releases relative to the occurrence of the initiating event and the subsequent system f ail ures which lead to significant core damage.

By using the PRA estimates for the timing of the accident together with a description in the GAC-Il code of the ef fect of meteorology and demography on population exposure, various strategies will be developed and assessed to g determine the combination of evacuation and shleiding (i.e., non-evacuation) W which minimize population exposure given a set of meteorological conditions.

2.7 INTERACTION WITH THE NRC The Project plans to support an interactive, phased review process on a schedule acceptable to both NRC and the Project. The review will prcrnote en improved understanding of the PRA complexities, uncertainties, and val Idity.

NRC is expected to provide comments on schedule, scope, and detailed Implementation for consideration as the work progresses.

This review process will be carried out at appropriate Intervals during the PRA program in a two-stege format. The f irst stage will be an overview to provide information to NRC management on the overall status of the ef fort and the signif Icant resul ts. The second stage w111 Invol ve inf ormal detailed discussions of methodol ogy and interim resul ts. This latter stage is aimed at provIding technical detail to the NRC staf f and its consultants.

As r.oted earl ter, as the PRA progresses a more detailed def inition of methodology to be used in such analyses as seismic risk characterization will be devel oped, it is expected that this more detailed methodology will be presented at appropriate NRC review meetings and that comments will be considered on the selected methodology as it relates to issues which the NRC considers to be candidates for resolution or prioritization using the PRA.

O Amend. 73 J-16 Nov. 1982

2.8 ACCIDENT DELINEATION The GBRP Project is currently pursuing a program to assure that alI appropriate sequences are incl uded within ti;e pl ant design bases. A complete set of initiators together with a well formulated and quantified set of event trees and f ault trees will provide a set of accideni sequences in a probabil istic context. An assessment will be performed to determine if all appropriate accident initiators and accident sequences are incl uded in CRBRP's Design Basis Accidents. This assessment will provide the decision criteria f or the concl usions presented. Th is rev iew w il l incl ude the following considerations:

1. Development of a set of criteria on which definition of sequences comprising the design envelope should be based will be presented in the ref erenced project documentation.

2. Characterization of accident sequences identif led in the PRA (including sequences which do not result in core damage) by:

a. Occurrence f requency;

b. Number of active f ailures f ollowing the initiating event (minimum);

c. Number of passive f ailures following the initiating event (minimum);

J d. Severity of sequence impact on the environment surrounding saf ety-rel ated equipment;

e. Severity of sequence challenge to systems designed to remove d.

heat (e.g., how many dif ferent systems are capable or avalIabl remove decay heat at the end of the sequence);

f. Severity of sequence chalIenge to reactor structures, including the containment bull ding;

g. The availabil ity of support systems in important sequences at -

point in the sequence at which a particular system is required to perform its f unction.

3. Consideration of both plant induced and external Initiating event sequences in this analysis;

4. Select sequences based on the criteria in (1) and the characteristics in (2);

5. Compare the selected sequences with those which currently comprise the design envelope and group the sequences by the various measures of severity def ined in (2). The result of this comparison and grouping should be a reduced set of sequences. Any signif icant new events O which are identified will be added to the design basis.

b Amend. 73 J-17 Nov. 1982

The product of this ef fort will be presented and discussed ai one of the NRC program rev lew meetings.

2.9 STUDY LIMITATIONS A brief listing of the study limitations is presented below:

1. COMPLETENESS AND LEVEL OF DETAll 0F THE lODELS- The coapleteness and level of detail of the model s will be limited as a result of the state of the design and the unavailabil ity of detail s of construction.

These l imitations can, however, al so be viewed as strengths since there is the opportunity to util ize the PRA In reviewing these design detail s for their risk Impi Ications as they are estabi Ished.

2. HUPN' FACTORS ANALYS IS- The role of the pl ant operations staf f in the inI11ation, aggravation, and mitigation of an accident will be modeled in the f ault troos and event trees developed to describe the sequence of events. The abil Ity of the model s and quantif ication methods is l im ited. However, a suppl ementary approach w il l be util ized. In this approach, operator action event trees will be used to investigate the role of the operations staf f in severe accident sequences.

3. EXTERNAL EVENT QUANTIFICATION- Experience with the analysis of risk f rcm external events (e. g., seismic events) has shown that the associated uncertainties are signif Icantly larger than f or accident sequences initiated by in-plant causes (which typically I. ave a less pervasive of fect on equipment rel labil Ity). Nevertheless, external events have in some cases been assessed to be significant contributors to pl ant risk. This analysis with its inherent uncertainties has, theref ore, been incl uded with in the scope of the PRA.

4 FAILURE DATA- A signif icant quantity of f ail ure data is available for eq u i pment i n many sy stem s ( i . e. , steam, f ire protection, el ectrical, control, and communication) of the CRBRP. These data have uncertainties no greater than those associated wIth L 's. Components in the i Iquid metals systems and at the interf ace between the sodium and other systems are Iess welI charactorized. Although uncertalntles in the rol labil Ity of these components exist, the impl ications of these uncertainties to the risk prof lie can be characterized using sensitivity studies carried out w ithin the PRA. Other areas in which .

significant uncertainties exist A ch may be important to the overall description of pl ant risk Incl de -

a. Initiator f requency;

b. Equipment rel labi! Ity an a degraded environment;

c. Equipment repair time distributions and alicwable on-line maintenance interval s.

5. ACCIDENT OiARACTERISTICS- The response of the cot'o to conditions which will produce core degradation and the response af the Amend. 73 J-18 Nov. 1982

a containment to severe accident sequences are somewhat uncertain.

(

These uncertainties are being handled by the use of phenomenological event treen developed to describe physical processes which can lead to accident onergetics and to containment f ail ure following core damage.

. 6. SITE SPECIFIC OiARACTERISTICS- The offects of uncertainties in site f

meteorology and demography as welI as in emergency response procedures can produce significant uncertainties in overalI risk. Again, the of fect of these uncer+aintles wIlI be investigated using sensitivity an,al y sI s.

3.0 PRA PERFORfCfE AND REVIEW A program of plans and actions will be implemented to control and verify qual ity of the resul ts f rom the PRA.

This program will include measures and documentation to assure that:

1 data is reviewed and evaluated systematically to verify completeness and ccrrectness with respect to the PRA requirements. This includes assurance that analyses are verif led by the PRA analysts and appropriate design organizations that the work has been performed satisf actorily; and

2. ine methods used f or verif ication are identif ied and the verif ication resul ts are documented.

f V Organizations which are presently involved in implementation of the study are shown in Figure 2. Al so shown on the figure in a box separated f rom the PRA performers are the design organizations. These organizations will serve to provide information about the plant features to the PRA performers and to revlow the technical resul ts of the study for accuracy and compieteness.

The overall review program is pictured in Figure 3. As shown, four level s of review by the CRBRP Project Of f ice. The f irst level is a working review by the performing organizations designed to assure the technical accuracy, clarity, and consistency of elements of the product. The second level is a review by CRBRP Project Interf acing organizations to assure consistency of the elements of the analysis with plant design and operational characteristics.

The third level of review will be conducted by a project management review committee to assess the val Idity of the approach taken to project Integration and to assure proper implementation of the approach. Finally, the fourth step is an overall review by a peer review-group made up of participants external to the Project. The purpose of this f Inal review is to assess the adequacy of the program integration and to eval uate the consistency of the methods used w ith the state-of-the-art.

The CRBRP Project Of fice will ul timately be responsible for util izing the results and insights f rom the risk assessment to help ensure that the systems designed to shutdown the pl ant, to cool the core, and to mitigate the ef fects of severe accident sequences are designed and operated to be consistent with p the resul ts of the PRA. Decisions on possible changes to the plant design

( wIlI be made through estabt Ished Project procedures. These procedures ensure Amend. 73 J-19 Nov. 1982 l

both consideration and review of proposal s by all af fected personnel throughout the project.

4.0 SCHEDULE. MILESTONES. AND RESOURCE ALLOCATION The PRA products are l Isted in Table 1 and the schedule for. key milestones for the entire PRA is depicted on Figure 4. As shown, the program is expected to produce a f Inal report in I ate 1984.

Table 2 provides estimates of the resource allocation for each task. The val ues in the table will fluctuate over the course of the PRA, but the table communicates relative l evel s of ef fort f or each task. Table 3 provides the planned compietIon dates f or the tasks. The results of these tasks wilI be avalIable for NRC revlow af ter the compietIon dates.

O Amend. 73 J-20 Nov. 1982

TE LE 1 CRBRP PRA PRODUCTS y

PRODUCTS o INITI ATING EVENT TOP LOGIC AND INITI ATOR COMPLETENESS ANALYSIS o PRCBMILISTICALLY QUANTIFIED ACCIDENT SEQUENCES AND THEIR BASIS A. SYSTEM FUNCTIONAL EVENT TREES B. FAULT TREES C. DETAILED COMMON CAUSE FAILURE ANALYSIS (SYSTEMS INTERACTION EV ALUATION)

D. EXTERNAL EVENT EV ALUATION (SEISMIC, ETC. )

o CORE DAMAGE FHEN0EN0 LOGICAL EVENT TREES AND QUANTIFICATION o (X)NTAINENT FHENOENOLOGlCAL EVENT TREES AND QUANTIFlCAT10N o UNCERTAINTY ANALYSlS o RADIONUCLIDE RELEASE ANALYSIS o HEALTH CONSEQUENCE ANALYSIS o ANALYSIS OF EX-(X)RE SOUR %S OF RADIONUCLIDES o DEFINITION OF PROGRAM TO SUPPORT CONTINUING OPERATIONAL APPLICATIONS o OPERATOR ACTION EVENT TREES AND APPLICATIONS TO OPERATIONS SUPPPORT AND

TRAINING PROGRAMS o DEFINITION OF AN ON-GOING RISK MANMEENT PROGRAM o EV ALUATION OF POTENTI AL RISK REDUCTION ASSOCI ATED WITH SUGGESTED DESIGN W ANGES o EVALUATION OF RISK CONTRIBUTION AND SENSITIV ITIES TO EQUIPENT TESTING INTERNALS (TEW SPEC. IMPACT) l o DETAILED DOCUENTATION OF STUDY AND FINAL REPORT O

Amend. 73 J-21 Nov. 1982 l

TABLE 2 RESOURCE ALLOCATION Listed bel ow is each major task associated wIth the risk assessment and the manpower estimated f or that task.

IAik man-Months Study Direction 28.0 Accident initiator identif ication 4.6 Event Tree Development 25.7 Initiating Event Data 3.3 System Faul t Tree Development 65.8 Component Fail ure Data 1.0 Faul t Tree Quantif ication 25.6 Event Tree Quantif ication 5.3 Common Cause Fail ure Analysis 22 Sensitivity Analysis 10.3 Dependency Analysis 12.2 Systems Fail ure Criteria 6.3 Core Response Event Trees 11.2 Containment Response Event Trees 11.2 Source Term 27 Conseq uence 12 Risk Analysis 9 Ex-core Sources 7 Uncertainty 7 OAET 13 Site Emergency 8 Input LOOS 7 Accident Delineation 6 On Going Risk Management 9 1

O l

Amend. 73 J-22 Nov. 1982 l

i T/BLE 3 CURRENT SCHEDULE OF UNCOMPLETED TASKS DESCRIPTION DUE DATE Provide final written Accident Sequence Definition Review 3/31/83 Provide final written RadionuctIde Reiease Analysis 1 2/31/83 Provide final written Uncertainty Analysis 10/31/84 Provide final written Detailed Common Cause Fail ure Analysis 10/31/ 84 Provide final written Accident Delineation Report 10/31/ 84 Provide final written Health Consequence Analysis 10/31/84 Provide written Risk Management Program Report 1 2/ 31/ 84 Provide written Operator Action Event Trees Report 1 2/31/84 Provide written input to Operational Procedures and Testing Interval 12/31/ 84 Provide written input to Site Emergency P1 an 1 2/31/84 Provide final written Report 12/31/84 O

Amend. 73 J-23 Nov. 1982

FICURE 1 PRA FLOWCHART ASSIGNMENT ASSIGNMENT ASSIGNMENT OF PLANT OF RELEASE OF HEALTH STATES CATAGORIES EFFECTS I I I I I I I I I EXTERNAL I I EVENTS l l

l l 1 1 I I U l l l l l l

' I CORE CONTAIN- ^'

h SYSTEM '

DAMAGE MENT WNC l MENT- CONSE-ACCIDENT PHENOME- PHENOME- RISK INITIATIONS + N NOLOGICAL

• NOLOGICAL ^

NUCLI E ANALY S TREES RELEASE TREES TREES ANALYSIS n

?N fa 2

3~

FAULT TREES is n arti.,

9 O O

Figure 2 CRBRP PRA ORGANIZATION l CRBRP PROJECT '

l OFFICE I

1 ACCIDENT SEQUENCE DEFINITION AND ACCIDENT PROCESS QUANTIFICATION:

ANALYSIS: FAUSKE &

EG&G-IDAHO; WOOD-LEAVER ASSOCIATES, INC.

& ASSOCIATES, INC.

l PROJECT SOURCE TERM ll DESIGN AND EVALUATION AND

?i. OPERATIONAL HEALTH CONSEQUENCE j; ORGANIZATIONS ANALYSIS:

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FIGURE 3 ANTICIPATED PRA INTERNAL REVIEW PLAN ELEMENTS PURPOSE

• ASSURE QUALITY, INTERNAL REVIEW BY CLARITY, AND ORIGINATING ORGANIZATION CONSISTENCY OF ELEMENTS OF PRODUCT l

REVIEW BY PROJECT

• ASSURE CONSISTENCY OF INTERFACING ORGANIZATIONS ELEMENTS OF ANALYSIS b ERATIONAL l L CEkSING CHARACTERISTICS

• ASSURE PROPER OVERALL REVIEW BY INTEGRATION OF PRODUCT DESIGNATED PROJECT ELEMENTS AND ACCURACY COMMITTEE OF APPROACH TO INTEGRATION

• PEER REVIEW OF

[E OVERALL REVIEW BY PROJECT INTEGRATION APPROACH

'a AND EXTERNAL PEER FOR ACCURACY AND 3;

REVIEW BOARD CONSISTENCY WITH STATE-OF-THE-ART IN PRA O O O

Figure 4 EXPECTED SCHEDULE OF l PRA PRODUCTS ACTIVITY 6/81 12/81 6/82 12/82 6/83 12/83 6/84 12/84 IIiIi IiIiI iiiii iiiii IIIII Iii11IIil 1I CRBRP '

• CRBRP CONSTRUCTION i LICENSING SAFETY PERMIT ISSUED l MILESTONES EVALUATION REPORT U V g SUPPLEMENTAL

• INITIATING DRAFT REVISED SAFETY EVALUATION J EVENT INITIATING INITIATING REPORT ISSUED EVENTV V EVENT DEVELOPMENT ()

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AMENDMENT 73 LIST OF RESPONSES TO NRC QUESTIONS There are no new NRC Questions in Amendment 73.

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Ouestion 001.245 (15.7.1.2.1)

! Identify all safety related valves or instruments which require a compressed air supply.

Response

See updated Table 15.7.1.2-1 Q

G Amend. 73 Q001.245-1 Nov. 1982

Ouestion CS760.60 m

The f uel !Ife and coolant bollIng constralnts are quantifled by defining equivalent limiting temperatures which shall not be exceeded. The constraints are defIred for Plant Expected Operaiing Conditions (PEOC). The uncertainty factors are at the 2-sigma level of confidence. Assembly Ilfetime/burnup goals are achieved whea both the cladding inelastic strain and cladding CDF are within established limits of 0.2% for the ductility strain limit and 0.7 for the CDF during steady state operation. Strain Equivalent Limiting Temperature (SELT) is defined as the end-of-life temperature which, if maintained throughout life, would cause for the particular assembly an EOL cumulative strain of 0.2%. A Damage Equivalent Limiting Tenperature (DELT) is defined similarly as the equivalent EOL temperature corresponding to a CDF of 0.7 for f uel assembi les and 0.5 for bl anket assembi les.

The calculation of the Transient Equivlant Limiting Tenperature (TELT) is perf crmed in three steps:

A. to provide an adequate margin to bolling, a temperature of 1550 F is defined as the maximum coolant temperature allowable during the natural circulation transient at a 3-sigma level of confidence assuming Thermal-Hydraul Ic Design Val ue (THDV) conditions.

B. this limiting temperature is translated into a temperature T whichM is defined as:

the maximum steady state coolant temperature corresponding to a 1550 F transient maximum coolant temperature for

(

(')T -

PEOC at the 2-sigma conf idence l evel -

C. finally, T is g translated into the Transient Equivalent Limiting Temperature TELT by multiplying the dif ference betweenuT and the inlet temperature T by the ratio of the coolant temperature Fises at EOL and the time in l Ne when the maximum transient temperature occurs, consider-Ing also the axial position where this temperature is reached, and adding to this the inlet temperature and correcting for the ID temperatures needed. The assemption for this correction is that the temperature difference Tg-T would increase / decrease with time in the same manner as the iomperature;01 f ference T ~

Cool In*

lt is understood that the orificing is an iterative process whereby a flow distribution is assumed which yields SELTs, DELTs and TELTs. These numbers in turn provide for a new flow distrubution which yields new values for these temperatures, etc.

The design basis requires no f uel centerline melting at 115% overpower conditions.

Why has this criterion not been Jsed for the flow orificing?

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QCS760.60-1 Amend. 71 Sept. 1982

Resoonse:

Fuel centerline temperatures are only a weak f unction of the cladding temperature; therefere the no f uel centerline melting would have been a

" flow-insensitive" criterion if used for or f f icing. Rather, detailed ad hoc analyses were performed to guarantee that the no melting criterion is satisfied, as discussed in Section 4.4.3.3.6.

It shoul d al so be t

'arified that the orificing is only partially an iterative process, and the astion statement "It is understood..." is only partially true. A flow d' 'bution which y iel ds SELTs, DELTs and TELTs is actually assumed. Ho ever, once the l Imiting temperatures are determined, the corresponding flow is calculated by the OCTOPUS code and this represents the I imiting flow adopted in the orificing process. Thus, there is no iteration process on the temperatures constraints; once they are determined the orificing conf iguration f ollows through. However, final verification that all design constraints are indeed satisfied is performed following calculation of the detailed performance prediction parameters reported in Section 4.4.3.3.

Thus, orifIcIng constraints are quidelines not Iimits. Guidel Ine vai ues have margins to any limits and provide guidance in core orificing (i.e.,

establishing the optimum flow allocations in core). Subsequent detailed structural analyses (PSAR Section 4.2) determine the design adequacy of core components, using as input the design data f rom Section 4.4.3.3.

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Ouestion CS760.106

!n the event of pipe breaks, what would cause the various isolation valves to close? It is stated in Section 5.6.1.2.1 that in at least one such break It is necessary for the operator to close an Isolation valve to save the Inven-tory of the PWST. How many such postulated breaks required operator interven-tion? How does the operator determine the break location?

Resoonse in the event of a large pipe break, in a steam generator loop or in an AFW loop downstream of the AFW check valves, the AFW isolation valves to the affected SGS loop will automatically closo following the steam drum depressurization to <200 psig. All postulated breaks that do not allow steam drum depressurization will require operator action to isolate the AFWS If tho breaks are large enough to initiate SGAHRS.

The information available to the operator to isolate these pipe breaks is described in PSAR Section 5.6.1.2.1.1 " Identification of Active and Passive

• Components which Inhibit Leal.s".

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i QCS760.106-1 Amend. 71 Sept. 1982

l CLINCH llVER O

BREEDER REACTOR PROJECT PRELIMINARY SAFETY ANALYSIS REPORT O

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l O CUNCH RIVER BREEDER REACTOR PROJECT l

PRELIMINARY SAFETY ANALYSIS ,

. REPORT  !

O VOLUME 27 PROJECT MANAGEMENT CORPORATION O

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