BAW-1280 Ca-7

THE CNSG _JI_ A CONCEPTUAL^MERCHANT SHIP

NUCLEAR REACTOR DESIGN

September 1963

Facsimile Price $ '/1 f^y £3

Microf i lm Price $ 2 . * S^J

Avai lable from the Of f ice of Technical Services Department of Commerce Washington 25, D. C. '

AEC Contract No. AT(30- l)-3206 B&W Contract No. 59-3085

Submitted to THE UNITED STATES ATOMIC ENERGY COMMISSION

By THE BABCOCK & WILCOX COMPANY

Atomic Energy Division Lynchburg, Virginia

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER

Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

The Babcock & Wilcox Company Atomic Energy Division Report BAW-1280

ABSTRACT

The Babcock & Wilcox Company has designed the Consolidated Nuclear Steam Generator II (CNSG II), an improved p re s su r i zed water r eac to r that is more economical than conventional power plants for high shaft horsepower and high load factor appl icat ions. The ent i re p r i m a r y sys tem is contained in one p r e s s u r e vesse l , resul t ing in an ex t remely compact and light sys tem. The refueling cycle is completed in only 10 days .

The CNSG II design presented here is for 66, 000 shaft horsepower , but basic plant information is given for designs of 15, 000, 22, 000, and 30, 000 shp.

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CONTENTS Page

1. INTRODUCTION 1-1 1.1. Design Fac tors . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 = 1 1.2. General Reference Design . . . . . . . . . . . . . . . . . . . . 1-3

2. ECONOMICS. 2-1 •2. 1, Capital Cost Es t ima te . . . . . . . . . . . . . . . . . . . . . . . 2-2 2 . 2 . Fuel Cost . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2. 3. Weight and Space of Reactor Plant . . . . . . . . . . . . . . 2-4

3. SAFETY CHARACTERISTICS. ..................... 3-1 4. OPERATIONAL PROCEDURES. .................... 4.-1

4. 1. General Charac t e r i s t i c s . . . . . . . . . . . . . . . . . . . . . 4-1 4 . 2 . Normal Operation . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4. 3. Reactor Plant Warmup and Cooldown. . . . . . . . . . . . . 4-2 4 . 4 . Decay Heat Removal. . . . . . . . . . . . . . . . . . . . . . . . 4-2 4. 5. Effect of Ship's Attitude and Motion. . . . . . . . . . . . . . 4-3 4. 6. Refueling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4. 7. Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4. 8. Response to Accidents . . . . . . . . . . . . . . . . . . . . . . 4-6

5.- DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5. 1. General Descr ip t ion . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 . Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 = 2 5. 3. ■ Reactor Vessel Arrangement . . . . . . . . . . . . . . . . . . 5 = 2 5.4 . Auxiliary Systems . . . . . . . . . . . . . . . . . . . . . . . . . 5-9 5. 5. Shielding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-13 5. 6. Instrumentat ion and Control . . . . . . . . . . . . . . . . . . . . 5-15

• 5 . 7 . Erect ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15

6. DESIGN HISTORY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6. 1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6 .2 . P r e s s u r i z e d Water Reactor Design . . . . . . . . . . . . . . 6-1 6. 3. Integral Boiling Reactor . . . . . . . . . . . . . . . . . . . . . 6-3 6 .4 . Wet Containment . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 6. 5. CNSG I, Phase I . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5 6 .6 . CNSG I, Phase I I . . . . . . . . . . . . . . . . . . . . . . . . . . 6-6 6 .7 . Burnable Poison Core Design. . . . . . . . . . . . . . . . . . 6-7 6. 8. CNSG ILDesign . . . . . . . . . . . . . . . . . . . . . . . . . . . 6.-8

APPENDIX Plant Information Sheets . . . . . . . . . . . . . . . . . . . . . A-1

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List of F igures F igure

The Following Figure Follows Page 1-4 1-1. CNSG II General Ar rangement

The Following F igures Follow Page 2-4 2 -1 . CNSG Cost Es t ima te 2-2. Annual Fuel Cost (66,000 shaft horsepower) 2-3. Unit Fuel Cost (66,000 shaft horsepower) 2-4. Comparative Size of Containments 2-5. Weight Vs Horsepower Comparison for Nuclear Plant

The Following Figure Follows Page 4-6 4 - 1 . Heat Balance and Flow Diagram

The Following Figures Follow Page 5-6 5 -1 . CNSG II Reactor and Containment Ar rangement 5-2. Core Plan and Control Rod Posi t ions 5-3. CNSG II Hot Channel Design (66,000 shaft horsepower) 5-4. P r e l im ina ry Systems Design

The Following Figures-Fol low Page 6-9 6 - 1 . PWR Plant Ar rangement (30,000 shaft horsepower) 6-2. IBR Plant Ar rangement 6-3. IBR Containing Bayonet Tube Boiler 6-4; IBR General Ar rangement 6-5. Dual Purpose Plant (82.5/400 MW-t) 6-6. CNSG I, Reactor and Containment Ar rangement 6-7. Advanced CNSG Scheme I

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1. INTRODUCTION

In 1962 The Babcock & Wilcox Company (B&W), working under an AEC contract , developed the Consolidated Nuclear Steam Generator (CNSG) design. This was an advanced p re s su r i zed water r eac to r design that included a large number of new ideas , such as wet containment, consolidation of the p r i m a r y sys tem into a single p r e s s u r e vesse l , and once-through s team g e n e r a t o r s . The CNSG achieved economic par i ty with conventional power plants for high-speed, long-distance cargo ship se rv ice .

B&W's f i rs t CNSG design was for a 20, 000 shaft horsepower (shp) plant. However, after ship application studies by George G. Sharp, Inc. showed that p roper application for nuclear power plants was in the 60,000-shp range, the Atomic Energy Commission requested B&W to modify the original design to produce a 66, 000-shp CNSG. The Com­miss ion also requested B&W to study means of simplifying the refueling operation and making the power plant more compact, l ighter , and l e s s cost ly . In addition to studying the reference design at 66, 000 shp, B&W was asked to extrapolate the bas ic information to power plants producing

. from 15, 000 to 66, 000 shp. The resu l t s of these efforts is the CNSG II design presented in this r epor t .

1 .1. Design Fac tors

The CNSG II was designed specifically as a mar ine reac to r and includes many features that a re of major importance to mar ine opera­t o r s :

1. Reliability - The plant has backups.for all rotat ing, e lec t r i ca l , and electronic equipment and can continue to operate with a section of the boi ler isolated. The design is based to a major degree on components or concepts that have been proved in operating power p lants . The power plant is bas ica l ly self- l imit ing and cannot be dam­aged by control sys tem fa i lu res .

1-1

2. Simplicity - El iminat ion of all f r i l ls from the design provides a plant that is eas i ly understood by the opera tor and is simple to operate and maintain.

3. Maintenance - The containment and all equipment exposed to radioactivity a re a r ranged for easy, quick a c c e s s . Wher­ever poss ible , sensit ive equipment is located outside of the containment for ready a c c e s s .

4 . Radiation safety - The radioactivity level in the machinery space during operation is 1.0 m r e m / h r . All sources of radioactivity a re closely contained.

5. Ease of refueling - Arrangement of the r eac to r plant has reduced refueling t ime to 10 days or l e s s .

6. Weight and space - Both of these factors a re signifi­cantly improved over those of previous mar ine power p lants .

7. Cost - The CNSG II is designed for minimum cost con­sis tent with the foregoing fac tors . The low-enriched UOz core reduces fuel costs significantly. Through simplici ty and attention to such p a r a m ­e te r s as t empera tu re and p r e s s u r e , capital costs have been minimized.

8. Ease of instal lat ion - To reduce the t ime required for plant assembly and shipyard instal lat ion, the sys tem can be instal led . in a smal l number of individual uni t s . For ins tance , del ivery of the ent i re p r i m a r y sys tem as a single, sealed, clean unit, would minimize cleanl iness requi rements in the shipyard during erec t ion .

B&W achieved the des i red c ha ra c t e r i s t i c s by basing i ts design on a f i rm foundation of past exper ience . Elements contributing to the excel ­lence of the design a r e as follows:

1. Consolidated design - Containment of the ent i re p r i m a r y sys tem in one vesse l provides grea t s implici ty and control over rad io­activity, reduces weight and space r equ i r emen t s , and provides an easi ly instal led sys tem.

2. P r e s s u r i z e d water - Being a p r e s s u r i z e d water r eac to r the CNSG II benefits from U. S. Navy and NS Savannah exper iences afloat, and from the experience of many other r eac to r s a sho re .

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3. Once-through steam genera tor - The once-through steam genera tor facil i tates consolidation and great ly simplifies steam gener ­ator construct ion.

4. Se l f -pressur iza t ion - The se l f -p ressur iza t ion e l iminates the need for a separa te p r e s s u r i z e r and its con t ro l s .

5. Wet containment - Wet containment is a great advance­ment in r eac to r safety and design s implici ty .

6. Low-enriched UOz core - This co re , of the type used in the Savannah, provides very low fuel costs and inherent se l f - regu la ­tion.

7. Simplified refueling - Ar rangement of the r eac to r plant minimizes the number of operat ions requi red for refueling. The control rod drives a r e not removed during refueling, nor is it neces sa ry to make and break the usual e lec t r ica l and piping connect ions. These , along with other r e q u i r e m e n t s , great ly minimize the t ime spent before and after actual fuel handling.

1.2. General Reference Design

A c r o s s section of the CNSG II is shown in Figure 1-1. The p r e s ­sure v e s s e l , which contains the r eac to r c o r e , steam genera to r , and circulat ing pumps , is largely filled with water for t ranspor t ing heat and moderat ing neu t rons .

The c o r e , cent ra l ly located in the r eac to r ve s se l (see F igure 1-1), consis ts of 16 fuel e lements assembled from Zircaloy tubes containing low-enr iched U0 2 . Water flows upward around these fuel tubes , absorbs heat , and then leaves the core at a t empera tu re of 544 F . Above the core the water turns radial ly outward, pas ses through the pumps , and then flows down through the steam genera to r , which consis ts of he l ica l -coi led Inconel tubes . The heat absorbed by the water is t r ans f e r r ed to the steam genera tor to make superheated steam (400 psig and 523 F)\ which in turn dr ives the tu rb ines . Leaving the steam genera to r , the water flows downward for a short d is tance , turns radial ly inward through the flow baffle, and then turns upward and r e - e n t e r s the c o r e .

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To provide for maximum access ibi l i ty , the reac tor vesse l is a r ranged so that maintenance and repa i r functions, except for control rod d r ives , a re per formed at the top of the sys tem. . The control rod dr ives a r e maintained through access tubes from the outside of the vesse l so that the containment will not have to be drained,

The refueling t ime has placed severe economic penal t ies on nuclear r eac to r s previously proposed for mar ine application. The CNSG II incorpora tes the following fea tures , which el iminate these objectionable cha rac t e r i s t i c s of nuclear r e a c t o r s .

The core is a r ranged so that a minimum of equipment must be dismantled for the refueling p r o c e s s . Fuel e lements a re retained by bottom la tches , thus eliminating the conventional upper grid plate; con­t ro l rod dr ives a r e placed below the co re . The reac to r must be refueled once every 4 y e a r s ; refueling is es t imated to requi re 10 days or l e s s .

The s team genera tor , placed concentr ical ly around the core , has inlets and. outlets at the top of the vesse l to pe rmi t easy plugging of the tubes if n e c e s s a r y . Ver t ica l positioning of the s team genera tor r e l a ­tive to the core was selected so that N16 activity will be low in the s team to the propulsion turb ine .

The reac tor vesse l is surrounded by a water- f i l led containment (Figure 1-1). The containment is required for safety in case of a fai lure in the p r i m a r y sys t em. The water in the containment will absorb the energy re leased from a sys tem rupture , as well as shield personnel from radiation during operat ion.

1-4

Figure 1-1. CNSG II General Arrangement

FEEDWATER

STEAM

STEAM GENERATOR PRESSURE

VESSEL

2. ECONOMICS

The CNSG II, in selected applicat ions, will, propel a ship more economically than a conventional power plant.

The major factors involved in an economic analysis a r e the cap­ital cost and the fuel cost . The weight of the power plant and its fuel a re a lso involved, since it may affect the size of the ship and the h o r s e ­power required to propel i t .

At the p resen t t ime, construct ion of a nuclear ship will probably include government par t ic ipat ion. The nonrecur r ing costs associa ted with the f irs t nuclear power plant of a new kind a re too grea t to be recovered in building the f i rs t s eve ra l p lants . The federal government has general ly assumed the burden of these nonrecur r ing costs to encourage the development of nuclear power. For example, the govern­ment const ructed the NS Savannah as a n e c e s s a r y step in the develop­ment of mar ine nuclear power that could'not possibly be commerc ia l ly justified. The government has a lso financed severa l pro jec ts to develop technology for more advanced nuclear power p lants .

Government policy for construct ion of a second nuclear ship has not been completely formulated. An indication of i t s possible form can be. obtained from the Garmatz Bill (HR 1071), which is p resen t ly before Congres s . This bi l l , an amendment of the 1936 Merchant Marine Act, provides that the Secre ta ry of Commerce can, under cer ta in c i r c u m ­s tances , grant the following financial aid to the cons t ruc tor of a nuclear ship:

1. Provide payment of development cos t s . 2. Provide payment for all or pa r t of the excess

cost of the nuclear ship over the conventional ship.

3. Waive or reduce charges for nuclear fuel. 4 . • Ass is t in planning or designing shore fac i l i t ies .

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These aids a r e given at the d iscre t ion of the Secre ta ry of C o m m e r c e . Before an aid is granted, the Secre ta ry must de termine whether it will be beneficial to the government according to c r i t e r i a specified in the bi l l . The financial aid of the Garmatz Bill or s imi la r bi l ls is not p r o ­vided after nuclear power is fully developed. The mer i t s of nuclear power must then f i rs t be judged by the no rma l economic ru les . One major factor the federal government cons iders in deciding whether to grant aid to a nuclear ship is the economic m e r i t of the r eac to r concept when fully developed. This is judged by cost es t imates for the second ship of a given kind.

The second-of-a-kind costs a re used in the economic analys is given in the following sections since these a r e the figures of r ea l in te res t to prospect ive ship owners and the AEC.

2 . 1 . Capital Cost Es t imate

The capital cost es t imate is the es t imated cost of the r eac to r plant as instal led in the ship. This includes the equipment and se rv ices supplied by the r eac to r manufacturer and the shipyard. Costs a r e given for a range in plant s izes from 15, 000 to 66, 000 shp. Since the r e fe r ­ence design for the study was the 66, 000-shp s ize , cost e s t ima tes were made for this size and extrapolated for the other s i ze s .

In an es t imate of second-of-a-kind cos t s , it is assumed that an identical r eac to r plant has been constructed previously and the f i rs t plant has absorbed al l of the nonrecur r ing c o s t s , such as design, devel­opment, and safeguards rev iews . The only engineering needed for the second-of-a-kind construct ion is that required to p rocu re , expedite, and test equipment and to modify the safeguards repor t s for application to the second ship. The l a rges t single entry in the cost es t imate is the r eac to r ve s se l and i ts i n t e rna l s , including the s team genera to r . (The cost of this i t em was es t imated by B&W's Barber ton Works. )

The cost of p r i m a r y c i rculat ing pumps for the CNSG II was based on es t imates from pump manufac tu re r s . Cost of control rods and con­t ro l rod dr ives were based on recent cost e s t ima tes for s imi la r com­ponents . Cost of auxil iary sys tems was based on the flow d iagram (Figure 5-4) . The cha rac t e r i s t i c s for the requi red components were developed from this d iagram. Their cos ts were then es t imated by using s imi l a r , recent cos ts as a b a s i s . The ins t rumentat ion i tem in

2-2

the cost es t imate includes the automatic valves shown on the flow dia­g ram, the nuclear ins t rumentat ion sys tem, the safety sys tem, and the power plant control sys tem. Cost of these i tems was based on recent quotes for s imi la r equipment. The sum of these en t r ies is the cost of nuclear cont rac tor - furn ished equipment. Nuclear contrac tor engi­neering cost was based on Reference 1. Shipyard cos ts were ex t rapo­lated from s imi l a r information developed by a naval a rch i tec t . The shipyard cost of a second ship is the sum of the expense of the following i t e m s : erec t ion of the nuclear equipment; p rocuremen t of piping, fit­t ings , and valves for auxi l iary sys t ems ; p rocuremen t and erec t ion of the containment and shielding; shipyard engineering; and overhead.

Indirect cos ts (s tar tup tes t ing, owner ' s engineering, and changes under the contract) a re es t imated at 7-1/2% of d i rec t cos t s .

Figure 2-1 shows the es t imated capital cost of the nuclear power plant for a range of ho r sepower s .

2 . 2 . Fuel Cost

The fuel cost for the CNSG II is much lower than that of conven­tional ships and also lower than other nuclear power plants of the same s ize . These low fuel costs a r e achieved through the efficient use of low-enriched uranium. Fuel cost for the 66, 000-shp CNSG II a re e s t i ­mated at 1.84 m i l l s / s h p - h r .

Nuclear fuel i s bought for a ship at i n t e rva l s . In the case of the CNSG II, it i s n e c e s s a r y to buy a new core or batch of fuel once every 4 y e a r s . Actually, since only the federal government may own nuclear fuel by law, the ship owner l ea ses the fuel itself. He only buys the assembly containing a form of the fuel that can be used safely in the r eac to r . Only a port ion of the fuel in each core is consumed. After this port ion of the fuel i s consumed, nuclear poisons build up to the point where the nuclear cha in- reac t ion s tops . At this t ime, the core must be removed from the ship, and the remaining fuel must be recov­ered and purified. The portion of the nuclear fuel p resen t , but not con­sumed, has a high value. Hence, the lease charges on fuel that is not consumed a re an appreciable port ion of the total fuel cost . The fuel costs can be divided into the following ca t egor i e s :

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1. Fabr icat ion ■» Charge paid to the core manu­fac turer when the core is del ivered.

2. Fissionable mate r i a l ­ Charge for the fuel actually consumed during operat ion.

3. Inventory = Continuing charge on the amount of fuel presen t in the c o r e .

4. • Transpor ta t ion, insurance , r ep rocess ing , and conversion = Charges concerning the p r o c e s s e s required to re tu rn the fuel to the government in such a form that it can be used in another r eac to r .

Figure 2­2 shows the es t imated annual fuel cost for 66, 000 shp. Because of the inventory cost , which does not vary whether the core is used or not, it has been n e c e s s a r y to show the util ization factor on the graph (Figure 2­2) . This is the average percent of full power at which the reac tor opera t e s . For approximate s tudies , it may be assumed that the annual fuel bil l is proport ional to the horsepower . Figure 2­3 shows the same information in the form of m i l l s / s h p ­ h r . These fuel cos ts are based on the presen t p r i c e s . The future is expected to bring further reduction in fuel cos t s , due to continuing r e s e a r c h p r o g r a m s being ca r r i ed out by the government and pr ivate indust ry . For example, B&W is constructing a l abora tory to c a r r y out r e s e a r c h on recycled fuel. Because of these p r o g r a m s , it i s es t imated that fuel costs 10 y e a r s from now will be about 1.3 m i l l s / s h p ­ h r , or about 70% of the present leve ls .

2. 3. Weight and Space of Reactor Plant

Figure 2­4 presen t s the size of r eac to r plants for various h o r s e ­

powers . Figure 2­5 presen t s the reac tor plant weight, excluding col ­

lision b a r r i e r , as a function of horsepower .

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F i g u r e 2 - 1 . CNSG Cos t E s t i m a t e

0 10 20 30 40 50

Shaft Horsepower (in thousands)

60 70

Figure 2-2 . Annual Fuel Cost (66, 000 shaft horsepower)

w u ex)

CO

T3

CO

O

■(->

CO

o o

1200

1100

3 1000 o

T3 4-1 O

900

800

^. 700

600

500

0 ^^075 0.6 0.7 0.8 0.9 1.0

Utilization Fac to r

Figure 2 - 3 . Unit Fuel Cost (66,000 shaft horsepower)

o a CO

O

CO

CO

o O 0)

3.0

2.0

1.0

0 — / / — 0.5 0.6 0.7 0.8

Utilization Fac tor

0.9 1.0

Figure 2­4. Comparative Size of Containments

• 1 8 ' ­ 0 " OD »■

s 3 0 ' - 0 '

f

2 2 ' ­ 0 " O D '

\

3 7 ' - 0 '

15,000 SHP 30,000 SHP 66,000 SHP

Figure 2­5 . Weight Vs Horsepower Comparison for Nuclear Plant

a o M « O

60

800

700

600

500

400 20 30 40 50

Shaft Horsepower (in thousands)

60 70

u CO

o On <u CO

o 43

43 CO

43 ao CD

80

60

40

20

10 20 30 ■■},- 40 50

Shaft Horsepower (in thousands)

60 70

3. SAFETY CHARACTERISTICS

The CNSG II pos se s se s severa l improvements that enhance the plant ' s safety. Basical ly, however, this concept r ep re sen t s wa te r -cooled reac to r technology incorporat ing safety fea tures that have been extensively reviewed and approved by var ious regulating bodies. In addi­tion, many yea r s of operating experience in numerous water -cooled r eac to r plants have demonst ra ted the safety and re l iabi l i ty of this sys tem.

The CNSG II co re design is s imi l a r in most r e spec t s to the low-enriched UOz rod-type core for the NS Savannah and other water reac to r power plants . This concept pos se s se s the following s imi la r fea tures : (1) a la rge prompt negative power react ivi ty coefficient due to the Doppler effect, (2) a la rge negative coolant t empera tu re coefficient of react ivi ty, (3) a la rge heat capacity in the fuel oxide and metal , (4) a la rge heat capacity in the r eac to r coolant, and (5) a single coolant -modera tor ma te r i a l .

The concept of se l f -p ressur iza t ion with a gas o v e r - p r e s s u r e p r o ­vides a r eac to r having no significant voids in the co re . Slow excursions owing to smal l react ivi ty inser t ions (<0. 3% of react ivi ty) a r e readi ly l i m ­ited by the negative coolant density coefficient without control action. Power excurs ions with r eac to r per iods as low as 3 mil l iseconds and react ivi ty additions of approximately 1. 5% Ak can be exper ienced without core damage with a s c r a m action of many seconds. Kinetic exper iments in the SPERT I facility, using NS Savannah c r i t i ca l experiment fuel pins, have verified these conclusions. This s e r i e s of exper iments demonst ra ted that a r eac to r period of 3. 2 mil l iseconds caused by a step input of r e a c ­tivity of 1. 6% Ak produced a peak power of 7, 480 MW, an integrated energy of 114 MW-seconds, and a peak p r e s s u r e of 15 ps i . However, the peak fuel t empe ra tu r e was more than 100 F below melt ing; the core received absolutely no damage; and the reac to r remained opera t ive .

Reactivity induced excursions as ser ious potential accidents a r e ruled out in the CNSG II due to the absence of a source of cold, dense

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water , the self-regulat ing p roper t i e s consis tent with slow rod speeds and simple control , and the inherent capability of the core to withstand large nuclear excurs ions without damage. In addition to the favorable effects of the water coolant in limiting nuclear excurs ions , the p r i m a r y system water -ac ts as a large t he rma l flywheel, providing an attenuation of the rmal per turba t ions between the reac to r and the s team side of the boi ler . This provides a ready source of s tored energy for inc reased load demands or a heat s torage medium when needed, thus allowing a la rge misma tch of r eac to r power and boi le r output during t r a n s i e n t s . This simplifies control r equ i rements (minimizes sources of react ivi ty excursions) and reduces the probabil i ty and extent of t empe ra tu r e and p r e s s u r e excurs ions .

These many strong, self-regulat ing safety fea tures have a l ready reduced the str ingent r equ i rement s placed on external safety sys tems by ear ly r e a c t o r s . For example , ea r ly r e a c t o r s required safety s y s ­t ems capable of sc ramming in only a few thousandths of a second, whereas p resen t day water r e a c t o r s containing oxide fuel rods a r e ade­quately protected by s c r a m s that will act in s eve ra l seconds. F u r t h e r ­m o r e , p resen t day technology gives every indication that as more exper ­ience is gained this dependence on ex terna l safety sys tems may be reduced even fur ther , thus providing a safer , as well as a more rel iable power plant.

The concept of se l f -p ressur iza t ion with a gas o v e r p r e s s u r e p r o ­vides a r eac to r having no voids in the c o r e . Therefore , the motion of the ship in heavy seas cannot affect the r eac to r stabili ty through react ivi ty changes brought about by the effect of gravity var ia t ions on the voids .

The maximum credible accident for the CNSG II is assumed to be a rupture in the p r i m a r y sys t em. In compar i son with other water r e a c ­t o r s , the probabil i ty of this accident is great ly reduced by providing a p r i m a r y sys tem that is essent ia l ly completely within the reac to r vesse l , the most rel iable p r e s s u r e containing component in a r eac to r sys tem. Should a p r i m a r y sys tem rupture occur , such as by failure of one of the vesse l pene t ra t ions , the energy re leased will be absorbed by the water -surrounding the r eac to r ves se l . In this manner the p r e s s u r e buildup in the containment vesse l is kept to a minimum. This p r e s s u r e suppres ­sion a r rangement is designed to function r e g a r d l e s s of the ship 's at t i tude.

3-2

4. OPERATIONAL PROCEDURES

4 . 1. General Cha rac t e r i s t i c s

When comparing the CNSG II to other nuclear reac tor plants , i ts outstanding cha rac t e r i s t i c is s implici ty. Exper ience with the NS Savannah has not only shown the p rob lems inherent in operating a complex power plant, but it has also shown that great simplification can be effected. At the time the Savannah was built , many of the cha r ­ac t e r i s t i c s of the nuclear r eac to r were not known; therefore , provis ions were made for a grea t many problems that have been proved nonexistent .

A second cha rac t e r i s t i c of the CNSG II that will simplify i t s ope ra ­tion is the inherent stabil i ty of the sys t em. As was explained in the s e c ­tion on safety ana lys is , inherent c h a r a c t e r i s t i c s tend to oppose any tend­ency of the sys tem to get out of control . This tendency great ly simplifies the functions of the opera to r s and automatic con t ro l s .

4 . 2 . Normal Operation

To simplify operation of the ent i re plant and to make the adapta­tion of automation as simple as poss ib le , the s team plant heat balance, as shown in Figure 4 - 1 , was adapted. The flow cycle has been s impl i ­fied even to the extent of eliminating h i g h - p r e s s u r e feed h e a t e r s .

The power plant can easi ly adapt i tself to changes in power l eve l s . The once-through s team genera tor , which is unique to shipboard power plants , is fitted with an automatic control that is bas ical ly s imi la r to the t h ree -e l emen t feedwater controls now employed in shipboard power p lants . The only difference is that one of the inputs is s team p r e s s u r e r a the r than s team genera tor water level . Control studies have e s t ab ­lished that this control will accommodate the fast power changes requi red in merchant mar ine se rv i ce . The ability to manually control the s team genera tor has been retained.

The nuclear r eac to r plant is inherently self-regulat ing and tends to follow changes imposed by the s team plant. Owing to the t he rma l

4 - 1

capacity of the sys tem, the p r i m a r y sys tem can meet any power change demanded by the sys tem. In p r ac t i c e , the plant will be control led by the turbine throt t le . The r e s t of the plant will adjust itself to meet the demands imposed by the thro t t le . This power plant will require close control over p r i m a r y water and feedwater quality. Exper ience with o ther r eac to r s and power plants on board has shown that such water quality control is feasible and p rac t i ca l .

4 . 3. Reactor Plant Warmup and Cooldown

The reac tor vesse l is filled with high puri ty water from the s e c ­ondary water s torage; the vesse l is vented at the pumps and nozzles to bleed out a i r . Hydrogen is added (Section 5 .4 .2) as the water level is lowered in the reac tor ve s se l to the hot water level . When the p r e s s u r e in the vesse l reaches the net positive suction head of the p r i m a r y pumps, the pumps a re turned on. The hydrogen sys tem i s shut off and discon­nected after the p rede te rmined amount has been added.

The sys tem is heated to a t empera tu re corresponding to zero power by nuclear heat . As the water expands it i s bled off to s to rage . To cool the sys tem, the decay heat removal sys tem functions as descr ibed in Section 5. 3. Water must be added to the reac tor vesse l during cooldown to maintain pump suction and to a s s u r e coverage of the co re .

4 . 4 . Decay Heat Removal

The power plant will be provided with a decay heat removal sys tem to remove heat generated after the reac tor is shut down. The sys tem shown in Figure 5-4 will a lso be used for emergency cooling when al l e l ec t r i ca l power supplies fail . Decay heat is normal ly removed by con­tinued operation of the s team genera tor and p r i m a r y pump(s) , the heat being dissipated in the condenser . For emergency decay heat removal , the sys tem takes advantage of the s team genera tor being a cooling coil . The secondary sys tem becomes an in termedia te sys tem under emergency conditions. Steam or hot water flows through the emergency cooler when the turbine throt t le i s closed and is condensed and cooled by sea wate r before enter ing the deaerat ing feedwater hea te r . The deaerat ing feed-wate r cooler is used because it normal ly contains a large volume of

4 - 2

wate r . The cooled water is pumped back to the s team genera tor by the emergency cooling pump' to complete the cycle .

For normal decay heat removal when the plant is shut down and e lec t r i ca l power is avai lable, one or more pumps will c i rcula te the p r i m a r y fluid and the secondary sys tem will remove the heat using the main condenser as the heat sink. Power supplied to the secondary equipment will pe rmi t the completion of the cycle . Having no power available to the p r i m a r y pumps and emergency power available to the emergency cooler pumps, heat will be removed by the s team genera tor by na tura l convection in the uppermost water and condensation plus some na tura l circulat ion in the s team space . To aid na tura l c i rculat ion, por t s a re provided between the upper water and the shell side of the s team genera tor .

Emergency cooling engine-dr iven pumps to be connected in p a r ­allel with the motor driven pumps a r e provided to ensure coolant c i r ­culation if no power is avai lable . The salt water pump takes its suction from an overboard l ine.

4. 5. Effect of Ship's Attitude and Motion

The power plant has been designed to meet the conditions usually imposed by the Mar i t ime Administrat ion with r ega rd to pitch, rol l , and heaving. None of these motions will have any harmful effects on the reac to r core or on hea t . removal from the p r i m a r y sys t em. Having a gas o v e r p r e s s u r e , the water in the core is subcooled, which gives l ess voids in the core than if the water was at sa turated p r e s s u r e . The ship 's ex t reme motion in rough seas will have l i t t le or no effect on the reac to r stability or change i ts react ivi ty by gravity var ia t ions on the voids.

4 . 6 . Refueling

The refueling t ime of a nuclear mar ine plant extends from the t ime the power is shut off to the t ime that power can be generated with a new co re . This refueling cycle includes cooling the r eac to r plant down, opening the containment ves se l , disconnecting any ins t rumenta ­tion or equipment that must be removed before acces s to the r eac to r vesse l can be accomplished, unbolting of the r eac to r vesse l head, removing any in ternals that a r e in the way of the removal of the fuel

4 - 3

e lements , erect ing any t empora ry shielding, locating any positioning divices used in the fuel handling p rocedure , and the actual fuel hand­ling (unloading and loading of the co re ) . After the new core is instal led, the r e v e r s e procedure must be followed (disassembly of the fuel hand­ling mechanism, t e m p o r a r y shielding, instal lat ion of any in t e rna l s , assembly of the p r e s s u r e vesse l head, a ssembly and connection of all equipment that had been previously d i sassembled , and instal lat ion of the containment cover . At this point, all equipment that was discon­nected or reconnected during the refueling cycle must undergo a thor ­ough test ing to ensure that each reconnected piece of equipment is func­tioning as it should. Subsequent to the equipment check, the containment t ightness must be es tabl ished. After the p r e c r i t i c a l tes t ing, the nuclear c r i t i ca l test ing can begin; after i ts completion, the r eac to r vesse l can then be warmed up to i ts operating t e m p e r a t u r e s . At this point, the refueling cycle is complete and the r eac to r is ready to go on the l ine.

For most r e a c t o r s to date, the long sequence of events occurr ing during a refueling cycle has extended from weeks to months . A major purpose of the CNSG II design has been to el iminate events or reduce the t ime spent on them by changes in the major components . A reduc­tion in the refueling cycle has been accomplished in the following ways :

1. The large bolted c losures usually assoc ia ted with p r e s s u r i z e d water r eac to r s have been el iminated, resul t ing in the reduction of the r eac to r t empera tu re cooldown t ime .

2. Except for a few ins t rumen t s , d i sassembly of equipment p r io r to opening the r eac to r vesse l has been el iminated. This has been accom­plished mainly by locating the control rod dr ives in the lower half of the r eac to r v e s s e l and not in the top head.

3. The refueling cover of the CNSG II is re lat ively smal l and has relat ively smal l bo l t s , which make the removal operat ion s imple r and fas te r .

4 . There a r e no in ternal s t ruc tu res above the fuel e lements to be removed p r i o r to refueling. This has been accomplished p r i m a r i l y by utilizing a bottom holddown latch for each of the fuel e lements in place of the top holddown grid usually assoc ia ted with p r e s s u r i z e d wate r r e a c t o r s .

4 - 4

5. The p r e c r i t i c a l test ing of equipment that has been d i sassembled and reassembled during the refueling p r o c e s s has been great ly reduced. Other than the main p r e s s u r e sys t ems , p r e c r i t i c a l testing involves only a few ins t ruments that have to be checked out p r io r to cr i t ica l i ty .

An analysis of the refueling cycle of the CNSG II d isc loses that as a con­sequence of the design improvements made in this reac tor , the refueling cycle has been reduced to an es t imated 10 working days or l e s s .

4 . 7 . Maintenance

One of the major objectives of this study was to design a r eac to r having the major components a r ranged to allow easy access to any com­ponent requir ing maintenance. The p r i m a r y pumps , the control rod d r ives , and the s team generator a re readily access ib le for maintenance.

The p r i m a r y pumps located in the top head of the p r e s s u r e vesse l may easi ly be removed for maintenance or replacement after access is gained to the containment. After disconnecting the e lec t r i ca l leads the pump may be unbolted, lifted direct ly off of the reac tor vesse l head, and removed to the maintenance a r e a .

Provis ion has been made to maintain or remove any pa r t of the rotating e lements of the control rod dr ive sys tem without having to dra in the wate r out of the r eac to r ve s se l or the containment shel l . This is accomplished by canning the control rod dr ives in the containment and providing a t empora ry sealing device for removing the pinion gear and shaft. The operat ion will be accomplished by personnel standing out­side of the containment shell .

The s team genera tor , consisting of approximately 844 separa te hel ical coi l s , is anticipated to be essentially, t rouble- f ree during i ts ent ire l i fet ime. Unlike fossil-fuel-fired gene ra to r s , the s team genera tor in a p r e s s u r i z e d water r eac to r environment can never experience over ­heating of any tube, and since Inconel tubing is used in this s team gen­e r a t o r , cor ros ion p rob lems a re prac t ica l ly nonexistent . The possibi l i ty that one or more of the 844 tubes may develop a leak due to faulty e r e c ­tion or some ma te r i a l problem cannot be overlooked, and for this reason the CNSG II design affords ready access to the tube sheets for tube plug­ging if a leak does develop in a tube. As can be seen in Figure 1-1, both the inlet and out le t ' secondary s team genera tor headers and tube sheets

4-5

a re located in the top head of the reac tor where the tube sheet is a c c e s ­sible after removing a smal l port ion of the flange-connected piping. The inlet and outlet secondary s team genera tor piping is a r ranged to allow for t he rma l and p r e s s u r e expansion without the use of expansion jo in ts .

4 . 8. Responses to Accidents

Owing to the self- l imit ing cha rac t e r i s t i c s of the power plant, all anticipated accidents due to failure of equipment can occur without damage to other p a r t s of the plant. Complete fai lure of all normal e l ec ­t r i c a l power will cause the reac to r to s c r a m and the p r i m a r y circulat ing pumps to shut down. In this ca se , the decay heat removal sys tem will automatical ly come into operat ion preventing any damage to the reac tor plant. Control rod withdrawal accidents , whether caused by improper operation or by a failure of the control sys tem, will be l imited by inherent cha rac t e r i s t i c s of the r eac to r and by a safety sys tem that will prevent damage to the r eac to r plant .

Simultaneous fai lure of a l l p r i m a r y circulat ing pumps will resul t in a reac to r s c r a m , and no plant damage will occur during the pump sys tem coastdown.

4-6

Figure 4 ­ 1 . Heat Balance and Flow Diagram

Lost 0= 3360

2­2000 kw 1 Running

385 PC 518 Fr

Q = 43750'

628, 664, 000 Btu/hr

h= 1260 397 PG 523 F ,

Q=605300

| h = 22l.4 261 Q=605300

/ HTurbo­Gen 0 = 43500

29.3P

/ 4

h= 1257.5 385PG 518 F 0=537690

Q= 538290

h= 11065 Separator

Q = 0

85 1; L 0=5670 ~~7

0 = 0

12 PG

P

ST o

4r ..J!

Q=46700—

J&

Q = 49152

•Distilling Plant

4"720 1b/hr F. W

Deaerating Feed

Heater 15.3 PG 2 50 F

218.8 = h Leak off CDR

Q=652000 „385PG 518 F

Q= 18600

Q= 19900 Q=1300 P H r * *

| 200 F J

Q=650

Legend

Superheated Steam tm^m^mm^^m Low Pres su re and Bleed Steam Drain

Desuperheated Steam Feed and Condensate ■ Gland Leakoff and Vent

Stop V EO ; Check V M ; Orifice H !H; Control V C&3 ; Back Pressure V E&3 "Port

PG = psig; PA = psia; Q = lb/hr flow (flows are total per ship); H = Btu/lb; F = temp in Fahrenheit

Data •

Steam and Feed Conditions: Superheater Outlet, 397 psig and 523 F Main Turbine Throttle, 385 psig and 518F Main Condenser Vacuum, 28.5" Hg at 75 F inj. Feedwater Temp, to Boiler, 251 F

Calculated Heat Rate: 9522 Btu/shp hr, based on main turbine non­extr steam rate of 7. 98 lb/shp hr

Distilling Plant Load: 13, 600 gpd Turbo­Generator Load: 1455 kw

Injection: Pump

5. DESIGN

5. 1. General Descript ion

The CNSG II shown in F igures 1-1 and 5-1 consis ts of a p r e s s u ­rized water r eac to r , a s team genera tor , and a p r e s s u r i z e r combined in a single p r e s s u r e vesse l . These components a r e a r r anged within the p r e s s u r e vesse l in such a way that the nuclear core in the center of the vesse l is completely surrounded by the s team genera tor tubes, which a re in turn a r ranged in a s e r i e s of hel ical coi ls . The section of the p r e s s u r e vesse l above the water line is the p r e s s u r i z e r portion of the reac tor sys tem.

The p r i m a r y coolant circui t and the secondary boiler a re also contained in the same p r e s s u r e vesse l . The p r i m a r y coolant flow path is indicated by a r rows in Figure 5 -1 . The p r i m a r y coolant is heated in the core a r e a in the center of the vesse l ; it then r i s e s through baffles into the outer annulus of the pump stalk and up to the impel le r of the pump, which forces the water down through the cent ra l section of the pump stalk and into the to rus - shaped s team genera tor space immediate ly above the he l ica l -coi led s team genera tor . The flow proceeds down­ward through the s team genera tor and back into the core a rea through openings in the core support cylinder to complete i ts cycle .

The s team genera tor is par t of the secondary sys tem, which begins outside of the r eac to r when feedwater en te rs from a header . The feedwater en ters the s team genera tor , through two separa te tube sheets and then flows through the ver t ical tubes , which convey the water down the side of the vesse l wall to the bottom of the s t eam genera to r . F r o m h e r e , the tubes sp i ra l upward. The economizer section of the s t eam genera tor extends from the point where the downcomer tubes begin t r ans fe r r ing heat to the point where nucleate boiling begins . The secondary fluid r is ing in the spiral ing coil pa s se s through the s t eam generat ion section to the superhea te r . Four s team-out le t headers

5-1

collect the superheated s team (400 psig and 523 F) . This s team, piped out of the containment vesse l in four s team l ines , dr ives the turb ines . The p r i m a r y sys tem is designed for a ship 's roll of 30 deg rees . At 30-degree l is t , no joints between the pump suction plenum and suction pipe a re exposed. Maximum assu rance applies s ince the pumps a re located l ess than 45 degrees from the fore and aft center l ine of the ship. The inlet holes in the pump inlet plenum are designed so that a roll of 45 degrees will not uncover them. There a re no holes located 90 degrees from the fore and aft center l ine .

5 .2 . Containment

The reac tor plant containment consis ts of a water- f i l led vesse l approximately 22 feet in d iameter and 37 feet high. The containment is divided into two a r e a s ; the centra l a r ea approximately 12 feet 7 inches in d iameter extends the full length of the containment and houses the reac tor p r e s s u r e vesse l . This is the dry section of the containment. The annular section between the cent ra l d iameter port ion and the outer shell of the containment is par t ia l ly filled with wa te r . The water in the outer section of the containment se rves as a vapor suppresso r and as a shield water tank for biological shielding. The proport ion of water to void volume in the outer port ion of the containment was es tabl ished on the basis that all of the s t eam re leased from a maximum credible acc i ­dent in the reac tor vesse l would be condensed in this water without ra is ing the p r e s s u r e above the design p r e s s u r e of the containment, approximately 100 ps i . P ipes or ducts convey the r e l eased vapor from the dry portion of the containment to the wa te r . No pipes a r e located on the sides so that they do not become uncovered during a ship 's ro l l . If the ship is laying on i ts side at 90 degrees , the suppress ion ducts a re st i l l submerged. Check valves or covers prevent a back flow into the dry containment.

5 . 3 . Reactor Vessel Ar rangement

5 .3 . 1. Genera l Descr ipt ion

The design and fabrication techniques for the CNSG II reac to r p r e s s u r e vesse l a r e well known and have been proved in s e rv i ce . The reac tor core , s team genera tor , and circulat ing pumps a re con­tained within the p r e s s u r e vesse l , which is la rge ly filled with water to

5-2

t r anspor t heat and modera te neu t rons . The space at the top of the vesse l provides a surge volume to maintain a constant p r e s s u r e when changes in water volume occur .

The core , centra l ly located in the r eac to r vesse l , con­s i s t s of 16 fuel elements assembled from Zircaloy tubes containing low-enr iched UOz. The s t eam genera tor , which is located concent r i ­cally around the core , has inlets and outlets at the top of the vesse l so that tubes can be easily plugged if nece s sa ry . Vert ical positioning of the s team generator re la t ive to the core was selected so that N16

activity will be low in the s team to the propulsion turbine . Specifically, the r eac to r p r e s s u r e vesse l is designed to

be fabricated from carbon steel that has been clad on the inside s u r ­faces with s ta in less s tee l . The top head of the reac tor vesse l is pene­t ra ted by four p r i m a r y pump nozzles , two secondary sys tem feed nozzles , four secondary s team outlet nozzles , and a relat ively smal l opening for refueling (approximately 50-inch ID). The refueling por t c losure is a s imple flanged and bolted design that is sufficiently sma l l ' to a s s u r e a rel iable seal with conventional flexitallic gasket or meta l l ic O-r ings ' . 'The reac tor vessel ' is suppor tedfrom the bottom by a cylin­dr ica l ski r t , which extends downward to a ring flange that is bolted to the inner containment bottom head.

The most significant feature of this r eac to r vesse l is the elimination of the la rge bolted c losure flanges usually found in p r e s ­sur ized water reac to r vesse l s of this d iamete r . The use of the smal l d iameter refueling port having smal l c losure flanges r a the r than the usual ful l -diameter opening head having la rge c losure flanges resu l t s in a savings in weight and space as well as a reduction in the cooling and heating t ime , which is a significant factor in judging the effective­ness of the design.

5. 3. 2. Reactor Core

5 .3 .2 . 1. Core Phys ica l Design

Basical ly , the core consis ts of 16 fuel e l e ­ments ; 12 have a square c ross section and 4 have a t r iangular c ross section. Each of the square c ros s - sec t i ona l fuel e lements has incorpo­rated in it four T-shaped channels through which control rods p a s s . The

5-3

fuel elements a r e designed to provide a combination of simple and economic fabrication along with s t ruc tu ra l ability to withstand handling and operat ional accelera t ion loads . Essent ia l ly , the fuel e lements con­sis t of Zi rca loy-c lad fuel rods held together in a uniform pitch lat t ice by grid spacers located at in tervals along the length of the element . The grid s p a c e r s , which a re brazed assembl ies of s ta inless s tee l , a re connected s t ruc tura l ly by means of axially disposed t i e - r o d s . These t i e - rods also connect to the upper and lower adap te r s , thus providing the neces sa ry s t ruc tu re for fuel handling location and operat ional rods . Each of the square fuel elements contain approximately 470 to 490 fuel rods .

The T-shaped control rod configuration was selected for severa l r ea sons . F r o m a s t ruc tu ra l standpoint this a r r ange ­ment pe rmi t s a much s turd ier fuel element configuration, since in this design all port ions of the element grid space r s can be connected through the center of the element. In addition, this configuration offers a s ig ­nificantly improved nuclear effectiveness as compared with the more conventional central ly located cruci form rod. Each group of four rods is connected together at the lower end by means of a coupling that also connects to the control rod drive l ine.

The fuel elements a re located within an oc tag­onal-shaped core container . This container , which is fabricated from s t a in l e s s - s t ee l cas t ings , is sized on the in te r ior contour to provide the neces sa ry control of coolant flow along the outer per iphery of the core . The container not only se rves as a flow control baffle but also as a main support m e m b e r for the core being supported by means of a flange from the centra l permanent s t ruc ture within the vesse l . At the lower end of the core container is a grid plate that includes accura te ly located r e c e p ­tacle holes which receive the lower ends of the fuel e lements . In addi­tion, T-shaped slots through which the control rods penet ra te a r e machined into this grid plate . Each fuel element is latched to the grid plate through its lower adapter by means of a holddown latch. This holddown latch having the ability to r e s t r a i n the fuel e l emen t from upward movement due to any anticipated condition of hydraul ic forces , ship 's atti tude, and acce lera t ive forces is designed as a highly re l iable and simply actuated mechanism. The latch is re leased by a s imple

5-4

ver t ica l motion of an actuating shaft passing upward through the center of the fuel e lement . At the upper end of this shaft is a handling knob that can be engaged for fueling operat ions by a s imple , long-handled tool. This latch is of a type developed and tes ted for other reac to r applications by B&W. It has a self-energizing cha rac t e r i s t i c that p r e ­vents inadvertent r e l ease of the element unless the actuating rod is pulled upward.

5. 3. 2. 2. Core Nuclear Design

The two means of react iv i ty control in the CNSG are control rods and burnable poison. There a re 12 control rods , as shown in F igure 5-2. The burnable poison, which consis ts of ZrB 2

dispersed in Z r 0 2 , is contained in rods identical external ly to fuel rods . These poison rods (6 16) a re d i spersed throughout the core . Burning out the poison in these rods par t ia l ly offsets the reduction in react ivi ty due to burnout of uranium and to the accumulation of fission products . Burnout of the poison maintains the react ivi ty more near ly constant, thus, reducing the number of control rods requi red and permit t ing longer core life.

To provide an even distr ibution of power, four enr ichment zones a r e provided. The average enr ichment at the begin­ning of core life is about 5. 3%. For initial analys is , it has been assumed that the use of these four enr ichments gives a peak- to -average power of 3 .0 . This peak- to -ave rage power is a conservat ive a s sump­tion based on analysis of a s imi la r co re .

5. 3. 2. 3. Core Thermal Design

P r e l i m i n a r y the rma l design of the CNSG II was based on the requ i rement that the core not be damaged by ope ra ­tion of 130%. full power. Damage was defined as centra l melting of the fuel or clad burnout (depar ture from nucleate boiling or DNB). Because the overpower condition of 130% might be reached by a rapid t rans i t from low power, it was further requi red that the overpower condition be satisfied using the higher core inlet t empera tu re conditions c o r r e ­sponding to normal operat ion at 25% power. These requ i rements may be modified as detailed design p roceeds . Additional requ i rements will be made as additional operat ional conditions (flow coastdown, ramp

5-5

additions of react ivi ty, and load changes) a r e invest igated. At the p resen t t ime, the rmal design must be sufficiently descr ip t ive to show that the general cha rac t e r i s t i c s of the core fit the plant design.

At 100% power the cent ra l fuel t empera tu re is 4250 F (manufacturing to lerance factors included) at the core hot spot. The central t empera tu re at the same location for 130% power is 4850 F , which is still below the melting t empe ra tu r e of 5000 F . In real i ty overpower conditions only resu l t when a t rans ient and a safety ac t ionoccurs before the t empera tu re is s tabil ized to the above max i ­m u m s . Hence, the t e m p e r a t u r e s given, which were based on steady s tate operation, a re somewhat pes s imi s t i c .

The heat t ransfe r design l imits for the core were based on boiling burnout curves presented in APED-3892. 2 The thermal design analysis was based on the heat r e l eased in the hottest channel in the core . This heat r e l ease is approximately 1.9 t imes that of the average channel and is due to the radial and local peak- to -average power fac tors , plus a hot channel factor based on manufacturing to l e r ­ances .

The hot channel will be located near a control rod, since the local peaking factor is based on improved neutron mod­erat ion in the water gap between the control rod and the nea re s t fuel pin. The presence of this water gap also induces additional water flow through the hot channel, which improves heat t ransfe r in the hot channel. This effect, while substant ial , could not be evaluated in the p re l imina ry analys is . Therefore , it was ignored, thus providing a degree of con­se rva t i sm in the core design.

Based on the analysis indicated above, the peak heat flux is 459, 000 Btu/f t 2 -hr at no rmal power and 598, 000 Btu/f t 2 -hr at 130% power. F igure 5-3 shows the design l imit heat flux curves (cj>DL) and the surface heat flux curves (<f>g) for 150% power plotted against exit quality for the hot channel. It is apparent that even at 150% power there is ample marg in to burnout for the condition corresponding to a step inc rease in react ivi ty from a steady state power level of 25%. At this stage of design some conserva t i sm has been purposely included to cover coastdown accidents , various rod posi t ions, and different modes of operat ion. More detailed design will probably show that these m a r ­gins are unnecessar i ly high.

5-6

5 . 3 . 3 . Steam Generator

The CNSG II s team genera tor is of the fo rced -c i r cu la ­tion, once-through type that is widely used in fossil-fueled s ta t ionary power p lants . It offers the advantage of not requir ing l a rge , bulky s team separa t ion equipment that would make the reac to r vesse l for this pa r t i cu la r design concept considerably l a r g e r . Since the s team leaving the coils is superheated, the genera tor ensures that the s team is dry . The s team genera tor is a wound coil of continuous tubes . This design allows more square feet of heating surface per cubic feet of s team generator space than a more conventional design. The tubes a r e made of Inconel, which have thinner wal l s , be t te r heat t r ans fe r , and l ighter weight than s ta inless s tee l . This genera tor pe rmi t s the s team turbine to opera te at constant s team p r e s s u r e , permit t ing the s team sys tem to have a lower design p r e s s u r e . Studies have been made to de te rmine the lowest s team p r e s s u r e for the mos t des i rab le s team t e m p e r a t u r e . Superheated s team may be produced with the loss of one p r i m a r y pump with only a loss of about 30 degrees of superheat .

The once-through s team genera tor is located within the reac to r vesse l in the space between the core support cylinder and the reac to r vesse l wall . It is divided into two sect ions; each has a feed-water inlet and two s team out le ts . Each section may opera te independ­ently. It is possible to opera te with one section if leaks or other diffi­culties make it des i rab le .

Each of the feedwater lines connect to a tube sheet as they enter the top portion of the r eac to r . The tubes pass down through the s team void and the p r i m a r y water in the outer annulus to the bottom of the tube bank. At this point the tubes s t a r t to sp i ra l upward around the core support cylinder to the top of the core , emerge from the p r i ­m a r y water , and again pass through the s team zone. At the top of the ' r eac to r the tubes for each s team genera tor section connect the two tube sheets as they pass through the vesse l head. Four s t eam outlets a re provided to produce a more compact a r r angement . Steam leaving the tube sheets is channeled into four s t eam lines routed outside the contain­ment . The tube sheets a re access ib le from outside of the r eac to r vesse l for plugging opera t ions . Detailed data on the s team genera to rs a re l is ted in the Appendix to this r epor t .

5-7

5 .3 .4 . P r i m a r y Pumps

The p r i m a r y sys tem contains four p r i m a r y pumps located on the reac to r vesse l top head and spaced near ly 90 degrees apar t . The number of pumps selected for this study was based on the space available in this pa r t i cu la r a r r angement . Pumps located in the top head can easi ly be maintained. The suction of all four pumps a re taken from a to rus - shaped plenum through the outer pipe of two concen­t r i c p ipes . When the p r i m a r y water reaches the pump, it makes a 180-degree turn inside the pump casing and then is forced down by the p rope l le r - type canned motor pump. The water is d ischarged down the centra l duct to another to rus - shaped plenum before being discharged to the s team genera tor .

Each pump is driven by a canned e lec t r ic motor cooled by in termedia te cooling wate r . A total developed head of 28 feet is produced at a capacity of 18 X 106 l b / h r . During cold s ta r tup , the required net positive suction head for the pump is supplied by p r e s ­surizing the reac tor vesse l with hydrogen. The p ressu r i za t ion fills the pump and pipes with water , and by venting the pump, the p r ime is com­ple te .

5. 3. 5. Control Rod Drives

Twelve control rod drive mechan isms a re located below the reac to r core and enclosed within the containment vesse l . The con­figuration of the dr ive mechan ism is s imi l a r to the rack and pinion dr ive mechanism being developed for the Advanced Tes t Reactor (ATR), located at Idaho Fa l l s , Idaho. The upper end of the rack component of the dr ive mechanism is connected to the lower end of the ver t ica l con­t ro l rod element , providing for withdrawal of the rods in an upward (out of the top of the core) direct ion and inser t ion ( sc ram) in a down­ward direct ion.

The canned drive motor mechan ism and pinion shafting is mounted in a horizontal position and made access ib le for maintenance purposes through water t ight nozzle c losures located in the side of the containment vesse l . Removal and rep lacement of these components without draining the r eac to r vesse l is done by utilizing a se rv ice tool and procedure s imi la r to that designed for the ATR.

5-8

If power failure or any other malfunction shoud occur , the control rod drive sys tem will fail safely by gravity sc ramming the r eac to r . Fas t inser t ion of the control rods at the same speed as the normal rod withdrawal ra te may be accomplished with the ship in a 90-degree posit ion. Posi t ive inter locks hold the rods in the fully inser ted position after a s c r a m , r ega rd le s s of the ship 's posit ion.

The s c r a m delay t ime from the rece ip t of the s c r a m signal to two-thirds completion of rod t ravel does not exceed 2 seconds. The design of the mechan ism and s c r a m circui t enables the drive mech­an ism to withdraw, in se r t , or s c r a m the rods individually, s imul tane­ously, or in groups . A remote position indication sys tem provides for posit ive indication of individual rod posi t ions .

5. 4. Auxiliary Systems

5.4. 1. Genera l

To consolidate and simplify the design of the CNSG II, some auxil iary sys tems superceded other sys tems and some were el iminated.

5. 4. 2. P r i m a r y Fi l l and Makeup System

Initial fill of the p r i m a r y sys tem is accomplished by using the makeup pump (see F igure 5-4) and a t empora ry line through the open containment and reac tor vesse l ha tches . The water is demin-era l ized secondary water and except for the oxygen content, the quality of the fill water and the makeup water a r e the s a m e . An oxygen scavenger , such as hydrazone, is added to the init ial fill to help remove any dissolved oxygen. To vent the p r i m a r y sys tem of a i r , the initial fill of cold water should completely fill the reac tor ves se l . P r i m a r y water is pumped or bled from the sys tem to the hot water level . As the reac to r is brought up to zero power and water heated, expanded water is bled out of the sys tem through the makeup l ine.

Makeup required after the r eac to r is placed into ope ra ­tion is supplied from the secondary sys tem makeup tanks . Rec ip ro­cating pump(s) located outside the containment vesse l a r e the makeup pumps . P r i m a r y water level is indicated on the console so the opera tor may cut the makeup pumps on and off. The level of the p r i m a r y water will be allowed to drop 1 foot before beginning the makeup. Makeup

5-9

will continue until the level is 1 foot above no rma l water level . Water conditions for init ial fill and makeup a r e given in Section 5 of the Plant Information Sheets in the Appendix to this r epor t .

5. 4. 3. Hydrogen Addition

Hydrogen is added at initial plant s ta r tup . The p r i m a r y reason for using hydrogen is to provide an excess of hydrogen to sup­p r e s s the radiolytic decomposit ion of the r eac to r wa te r . An additional amount of hydrogen is added to provide an o v e r p r e s s u r e of 200 ps ia in the s team space so that local boiling at the surface of the fuel pins will be avoided. Hydrogen is also maintained in the r eac to r s team space to provide net positive suction, head requ i rements for the p r i m a r y pumps .

At operat ing conditions, about 460 cubic feet of hydrogen at s tandard t empera tu re and p r e s s u r e (STP) is located in the s team space while approximately 1620 cubic feet at STP is dissolved in the wa te r . Hydrogen is added from cylinders located outside the contain­ment vesse l during heatup of the p r i m a r y wa te r . It is anticipated that supply cylinders will not be c a r r i e d on board ship during normal c ru ises since the large quantity of hydrogen in the p r i m a r y water will supply the s team space should hydrogen be lost through leakage.

5. 4. 4. Containment Cooling

The containment water and the stagnant a i r space between the top port ion of the reac tor vesse l and the shielding a re cooled by sea water flowing through coils located in these respect ive a r e a s . Sea water is also used to cool in te rmedia te water , which in turn, cools the four p r i m a r y pumps and the twelve control rod drive mo to r s . Two pumps that take suction from a sea chest pump sea water through the in te rmedia te cooler or the containment cooling coi l s . Three containment water cooling coi ls , equally spaced around the reac to r vesse l , cool the water by na tura l c i rculat ion. A coil also cools the air space by natura l c i rcula t ion. Two pumps provide flow for the closed in termedia te loop to cool the four p r i m a r y pumps and the twelve con­t ro l rod dr ive m o t o r s . Valves a re located in the c i rcui ts for isolation and control pu rposes .

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5. 4. 5. P r i m a r y System and Secondary System Relief Valve Venting

The p r i m a r y sys tem ut i l izes two methods to prevent ove r -p r e s su r i za t i on of the r eac to r ves se l : (1) a d iaphragm-opera ted safety valve operat ing on a signal from p r e s s u r e ins t rumentat ion and (2) two code self-actuated safety valves . The diaphragm valve has a remote operated valve located ups t r eam for backup or isolation pu r ­poses should it leak. The code safety valves have a remote opera ted th ree-way valve located up s t r e a m for switching purposes in case a valve should leak. For the secondary sys tem, a code self-actuated safety valve is located on each of the feedwater inlet l ines to protec t the s team genera tor if the sys tem becomes full of wa te r . All of the safety valves for both the p r i m a r y sys tem and the secondary system discharge d i rec t ly to the containment water through sparging nozzles where the relief s team is quenched and collected.

5 . 4 . 6 . Purif icat ion

The design of the CNSG II does not include a pur i f ica­tion sys tem. However, it is assumed that the customer will have a purification sys tem at his dockside facility for cleaning up the p r i m a r y water after power operat ion p r io r to entering the containment vesse l for refueling or maintenance. For this reason blind flange connections a re provided outside the containment for connecting to dockside and r e c i r c u ­lating the p r i m a r y wate r .

The philosophy in operat ing a CNSG is that no pur i f ica­tion equipment is normal ly needed for operat ing the ship at sea , but is only needed in por t to clean up the p r i m a r y sys tem pr io r to opening the reac tor vesse l for maintenance or refueling. Connections to and from the r eac to r a r e such that breaking off or rupture of a connection will not dra in the r eac to r vesse l below the top of the core and at the same t ime will r e l ease p r i m a r y p r e s s u r e below the containment water level .

5 . 4 . 7 . Sampling

Since the quality of the p r i m a r y water is not being con­trol led after the r eac to r vesse l is init ially filled, a no rmal sampling sys tem is not needed. It is anticipated that a sampling sys tem is

5-11

provided in associat ion with the dockside purification facility so that cleanup of the p r i m a r y water can be monitored p r io r to refueling or maintenance work.

5. 4. 8. Decay Heat Removal

No special equipment is requi red for the no rma l removal of decay heat following a shutdown. It is accomplished by the continued generation of s t eam in the once-through s team gene ra to r s ; the s t eam flows to the condensers where the decay heat is d iss ipated. If the p r i m a r y water must be cooled down beyond the point where s team can no longer be generated, then feedwater flowing through the s team generator coils is used to remove the heat .

An emergency cooling sys tem is provided when s team can no longer be dumped into the main condenser or when the main feedwater pump is inoperat ive . Steam o r hot water is routed to the emergency cooler when the turbine throt t le and bypass valve a r e closed. ; The secondary fluid is cooled and dumped into the deaerat ing feedwater hea t e r . The emergency cooling pump takes suction on the l a rge supply of cooled water s tored in the deaerat ing feedwater hea te r and re tu rns the feedwater to the s team genera to r . The emergency cooler d iss ipates heat by using sea wa te r . Having no emergency power o r no sea water : suction available, in ternal combustion engine driven pumps a r e provided •: on ei ther sides of the ship as backup emergency pumps .

5 . 4 . 9 . Waste Handling and Venting

Because of the design philosophy of the CNSG II, a design for s toring radioact ive gaseous and liquid-wastes is not requi red . When at sea , gases collected in the containment may be diluted with outside air and disposed of through the ship 's s tack. When in por t and entry to the containment is requi red , the containment vesse l can be purged and ventilated by the ventilating fan. Air from the containment is f i l tered for suspended ma t t e r and monitored before d ischarged up the s tack. Gases formed in the r eac to r vesse l may be disposed of through a line connected to the dockside disposal facility.

Any was tes resul t ing from reac tor vesse l leakage col­lects in the containment. Should this leakage r a i se the containment water level , a line will be provided above normal water level to bleed

5-12

excess water for d isposal . Final removal of r eac to r and containment water will be done by dropping suction lines into the vesse l and pumping out.

Vents a re provided on each nozzle and pump on the reac to r top head and on the top hatch. Water may be vented direct ly to containment during init ial fill or gases may be vented to gaseous waste disposal after shutdown.

5. 5. Shielding

Of p r i m a r y in te res t in the CNSG II shield design a re the at ta in­ment of low environmental dose ra tes in personnel access spaces , min­imum volume shielding to conserve valuable shipboard space, and min­imum shield weight. Shielding and radiation considerat ions a re closely re la ted to placement of the p r i m a r y heat exchangers with respec t to the core in the p r e s s u r e vesse l , placement of secondary heat exchangers , and use of a vapor suppress ion- type wet containment. The shield has been designed to reduce full power dose r a t e s from neutrons and gamma photons outside the shield to 1.0 m r e m / h r or less when operating at a reac to r power level of 184 MW-t.

This dose ra te was reduced to pe rmi t essent ia l ly unres t r i c t ed access to the outside of the containment when the reac tor is operat ing at full power. (This was the same approach as used in the design of the NS Savannah. ) Although conservat ive , this design is considered the most ra t ional . The access requi rement may be re laxed with an attendant reduction in shielding weight when m o r e knowledge becomes available on the ship 's a r r angemen t . The containment may be placed in a reac to r compar tment , in which case , l imited access can be guar­anteed and the dose ra te at the edge of the containment ra i sed accord­ingly. Also, it may be possible to group equipment around the con­tainment to afford some additional shielding. Finally, it may be p r a c ­t ical to lay out the ship so that l imited access is n e c e s s a r y only in the immedia te vicinity of the containment.

The compact CNSG II shield (Figures 1-1 and 5-1) combines the functions of a p r i m a r y and a biological shield. This shield consis ts of p r i m a r y coolant water and the rmal shielding inside the reac to r vesse l and lead and containment water outside the r eac to r vesse l except on the top and bottom. The bottom of the core is shielded with 10 feet of

5-13

p r i m a r y water and the 7. 5 inches of s teel that forms the p r e s s u r e and containment, vesse ls . In addition, double-bottom water tanks may be filled to afford more water shielding, in this a r ea . The top of the core is shielded by 4 feet of p r i m a r y water in the r eac to r vesse l and a 14-inch-thick removable lead shield above the top of the p r e s s u r e ves se l . - 'Mos t of the shielding in this region is due to the p r i m a r y gammas of the core r a the r than the secondary gammas or the N1 in the coolant pool. This shielding is adequate to reduce the dose ra te to 1 m r e m / h r or l ess at the top of the containment. A hollow cylindrical lead shield having a 4-inch wall thickness has been provided immediate ly below this top shield to maintain adequate protect ion in the region of the piping penet ra t ions . The lead shielding exter ior to the p r e s s u r e vesse l at the radial center l ine of the core is sized at 8 inches and is backed by a 1-inch-thick steel support . At elevations above the core top and bottom, the shielding is reduced owing to inc reased water shielding and the slant th icknesses afforded by the lead. In the region near the bottom of the p r e s s u r e vesse l , the lead shield is 4 inches thick. Near the top of the vesse l , however, 6 inches of lead thickness is required because of the terminat ion of the containment wa te r . Using this shield design, the gamma dose r a t e at the containment vesse l exter ior is 1 m r e m / h r or l e s s . The neutron dose ra te contribution at the exter ior of the containment is negligible in comparison with the gamma dose r a t e s .

The the rmal shield, p r i m a r y water , secondary water , and s team generator metal a r e sufficient to reduce the fast neutron flux on the inside surface of the r eac to r vesse l to a 20-year nvt of approximately 1017. The a r r angemen t and design of the s team genera tor with its high secondary water and s t eam velocit ies hold the N16 in the secondary sys tem to such a level that a dose ra te of 3 m r e m / h r o r less can be expected at the surface of a 6-inch Schedule 40 s team line upon its exit from the top of the r eac to r vesse l . Depending on the s team t rans i t t imes and the line access ib i l i ty in the final layout some additional local shielding may be requ i red on these l ines .

4

5-14

5.6 . Instrumentat ion and Control

The major i ty of the ins t rumenta t ion and control of the reac tor and its auxil iary sys tems is ent i re ly conventional with special emphasis placed on design in accordance with U. S. Coast Guard Codes and ABS Rules . The simplici ty of the CNSG II and the l imited number of auxi l iary sys tems el iminate mos t of the ins t rumentat ion and control normal ly associa ted with a nuclear propulsion plant.

Special attention must be paid to the selection and placement of neutron de tec to rs , because the s team genera tors a re located between the reac tor core and the more t radi t ional locations for nuclear detectors outside the r eac to r vesse l . The a l te rna te locations considered for nuclear detec tors a re as follows:

1. Inside the core - Space he re is very l imited, but de tec tors for this se rv ice a re commerc ia l ly avai lable. Reactor s tar tup with these de tec to rs , however, has not been demonst ra ted at the present t ime .

2. Between the core and the s team genera tor - Space l imitat ions he re a re l ess s t r ingent , but commerc ia l ly acceptable detectors for this location must be developed.

3. Outside the reac to r vesse l - The distance from the core to the vesse l wall in all direct ions will r equ i re the development of a suitable neutron window.

5. 7. Erect ion

The simplici ty and consolidation of the CNSG II nuclear plant design is di rect ly reflected in the smal l portion of the plant e rec ted in the shipyard. The following discussion re la tes to the a reas where shipyard erect ion is affected; a brief explanation of how the plant is e rec ted is also given. The number of auxi l iary sys tems used in the CNSG II plant has been kept to a minimum. A compar ison of the flow d iagram shown in Figure 5-4 with a flow diagram of a nuclear plant, such as the NS Savannah, gives an indication of the re la t ively smal l number of auxil iary sys tems that have to be e rec ted aboard ship.

One very significant i tem regarding erect ion of the CNSG II is that the reac tor vesse l containing the in ternals and the p r i m a r y sys tem

5-15

and the s team generator will be completely e rec ted in the shop p r io r to shipment to the shipyard. Thus, excepting for the four p r i m a r y pumps, s t eam piping and the control rod d r ives , the ent i re p r i m a r y and secondary sys tems may be instal led in the ship in one piece at one t ime , resul t ing in savings of cost and t ime over that experienced in the erection of the nuclear instal lat ion aboard the NS Savannah. Another benefit of a shop fabrication of the consolidated reac to r and heat exchanger is that the original c leanl iness es tabl ished in the shop can be easily maintained since no further work on the reac tor is done after leaving the shop.

A tentative erect ion sequence that might be followed is descr ibed below:

1. The bot tom-head portion of the containment shell is received and instal led in its cradle aboard ship and bolted secure ly in place.

2. The reac tor vesse l is instal led in one piece on the support ring provided in the bottom-half of the containment shel l .

3. The inner shell of the containment vesse l to which the biological lead shielding has been at tached is positioned around the reac tor vesse l and built up in sections to its full height.

4. The upper ou te r - she l l section of the containment is lowered over the reac tor vesse l and inner containment shell to a pos i ­tion immediate ly above the previously instal led lower head of the con­tainment vesse l .

5. The lower head of the containment vesse l and the upper ou te r - she l l section of the containment vesse l a re welded together . This welded joint will be in a section of the containment vesse l having a thickness not g rea t e r than that for which s t r e s s relieving is exempt.

The preceding erect ion sequence outline neglects i t ems , such as smal l p ipes , feedwater inlet and s team outlet connections, and cooling coil connections, which in the erect ion would be ent i re ly dependent upon specific shipyard p rac t i ces and, there fore , a re not covered in this repor t .

5-16

Figure 5 -1 . CNSG II Reactor and Containment Arrangement

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6. DESIGN HISTORY

6. 1. Introduction

The CNSG II design is the resu l t of a s e r i e s of design efforts s tar t ing with that of the NS Savannah. At each stage of development, new information from r e s e a r c h p r o g r a m s , from experience in the design, fabrication, and operation of the NS Savannah, and from s imi la r design studies has been incorpora ted . Therefore , based on experience and exper imenta l knowledge, g rea t advancements in r eac to r design have been made since the NS Savannah was built.

This section will desc r ibe the immedia te design h i s to ry that has been incorporated into the CNSG design. The f i r s t th ree pa r t s of Section 6 desc r ibe projects which were not a par t of the CNSG, but which contributed important concepts to the CNSG design. The las t four pa r t s desc r ibe the h is tory of the CNSG design.

6 .2 . P r e s s u r i z e d Water Reactor Design

The overal l plant a r r angemen t of this p r e s su r i zed water r eac to r 3 is shown in F igure 6 - 1 . This design has the heat exchange equipment and the reac to r vesse l a r ranged in separa te p r e s s u r e v e s s e l s , as in the NS Savannah. The design differs from the Savannah since new designs for components were considered and severa l new concepts , as d iscussed below, were included in the design. Using this approach, it was possible to make a r eac to r plant that was much smal le r , l ighter , and cheaper than the NS Savannah.

A considerable portion of the effort expended in this design was used in determining what the t e m p e r a t u r e and p r e s s u r e level of the p r i m a r y and secondary sys tems should be. It was real ized that the t rend at that t ime for shore based nuclear power plants was toward ever increas ing p r e s s u r e s and t e m p e r a t u r e s . At the same t ime, the t rend in conventional shore based power plants lay in the same di rec t ion. However, for shipboard power plants the t empe ra tu r e and p r e s s u r e levels have been stabilized for seve ra l yea r s at levels considerably below those encountered on shore . B&W, therefore ,

6-1

considered that it would be important to discover whether the same kind of t r ends could be found in shipboard r eac to r p lants . Two sets of design conditions.for the p r i m a r y sys tem were invest igated. To be represen ta t ive of shore based p rac t i ce s , one condition was set at a p r e s s u r e of 2000 psi and a t empera tu re of 636 F . To provide an adequate spread and to discover t r ends , a second condition was chosen at 812 psi and 520 F . This second condition was chosen at the lower end of the p r e s s u r e - t e m p e r a t u r e range to provide as wide a spread as possible within the l imited scope of the design study. Component s izes and costs and fuel costs were est imated for each condition. An optimum set of secondary s team conditions were a lso determined for each set of p r i m a r y s y s t e m s . As a resu l t of these studies the following t rends were discovered. These t rends were con­sidered to hold valid if the chosen p a r a m e t e r s were in the range where the reac tor would be a p r e s su r i zed water reac to r (the p r e s s u r e s were below the c r i t i ca l p r e s s u r e of water ) .

1. Fue l costs remained constant . This was. t rue within.the accuracy l imitat ions of the calculat ions per formed. The higher p r e s s u r e and t empera tu re plant had bet ter t he rma l efficiencies but this t rend was offset by a dec rease in the nuclear efficiency of the co re , As the t e m p e r ­a tu re of the core increased , it was n e c e s s a r y to i nc r ea se the amount of. U235 in the core , causing an inc rease in fuel cos t s . In addition, the high-t empe ra tu r e core must be made l a rge r because heat t r ans fe r is poore r .. at the higher t e m p e r a t u r e s . The increased core size increased the cost of fabricat ing the core , causing another addition to the fuel cos t s . Final ly , owing to the higher p r e s s u r e s , the cladding on the fuel elements had to be made th icker . This means that the re was more paras i t i c m a t e r i a l in the core , which again tended to inc rease the fuel cos t s . The combination of these th ree factors and the opposite t rend in t he rma l efficiency counte r ­balanced the fuel cos t s .

2. Capital costs of power plant was higher for the h i g h - p r e s -sure condition. This was p r ima r i l y due to the thicker construction required to withstand the higher p r e s s u r e s . The difference in costs between the two conditions studied was approximately $800, 000.

3. It was considered that lower p r e s s u r e s would dec rea se the problems of designing, fabricating, and operating a nuclear power plant.

6-2

Based on these r easons , it was concluded that the reference design for the power plant should be the 812-psi condition. It was concluded that there 'was a range of conditions in which the costs would be relat ively constant, and that so far as this optimization study had de termined it would be sat isfactory to design a future power plant anywhere within the range of from approximately 700 to 1000 ps i .

This was the f i rs t B&W design to incorpora te se l f -p ressur i za t ion . It was found that se l f -p ressur iza t ion would make the power plant inher ­ently more s imple by eliminating the p r e s s u r i z e r and its associa ted con­t ro l s in piping. This had a d i rec t effect upon capital costs and operat ing cos t s . In addition, a s e l f -p re s su r i zed plant had the lowest operat ing p r e s ­sure for a given t empera tu r e . This meant that the l ightest and cheapest components could be used for a given t empera tu re condition. The use of se l f -p ressur iza t ion entailed a core penalty that seemed worthwhile at the t ime. This penalty was due to the poorer flow distr ibution that resul ted from the voids in the core and to the effect of the voids in the core upon react ivi ty .

This power plant produced superheated s team through the use of a small superhea te r located para l le l to the boiler in the p r i m a r y sys tem circui t . By the use of this para l le l configuration the size of the supe r ­hea te r was considerably reduced. The superheated s team had a smal l thermodynamic advantage obtained at a smal l p r ice for the cost of the superhea te r . It also will improve the quality of s team going to the t u r ­bines and reduce the maintenance and capital costs of the tu rb ines .

6. 3. Integral Boiling Reactor

The B&W developed Integral Boiling Reactor (IBR) design4 , shown in F igure 6-2, was the f i rs t B&W design to have the ent i re p r i m a r y sys tem, excepting for the p r e s s u r i z e r , consolidated within one p r e s s u r e vesse l . The p r e s s u r i z e r was located in a separa te vesse l , s imi la r ly to the Savan­nah. Since this design used na tura l -c i rcu la t ion s team genera to r s , , s team drums had to be located on the outside near the p r i m a r y sys tem. This design also used natura l circulat ion in the p r i m a r y sys tem.

6. 4. Wet Containment

In November I960 the AEC suggested the idea of a wet, that is flooded, containment vesse l to B&W and requested the Company to study this concept.

6-3

The initial step in the investigations was a paper study5 of the problems inherent in wet containment. It was concluded that the concept of a wet containment would permi t consolidation of seve ra l sys tems , because the containment could be used as a heat sink for components that required cooling arid a lso as a tank.for s torage of waste water . It was a l so found that the water in the containment would se rve very effectively as shielding. A potential advantage was found in that the water in the containment would suppress : the p r e s s u r e imposed on the containment by p r i m a r y sys tem rup tu re . The basic function of the containment is to protect personnel in case of such a rup ture . This potential advantage would have to be explored by means of r e s e a r c h . The major potential problem discovered by this study was the effect of the water in the containment upon the insulation around the p r i m a r y sys tem and upon the hot components of the p r i m a r y sys tem if the insulation failed. One possible way of protecting, this insu­lation would be to instal l a water t ight meta l can around the ent i re p r i m a r y sys t em. This solution looked unat t ract ive because it would requ i re a com­plicated sys tem having many joints, that would have to be made watertight.; It a lso apparently required a. sys tem for determining when and where leaks occur red . This sys tem would in itself be complicated. It was,, therefore , concluded that the major efforts should be devoted to discovering what would happen to the insulation.if water came in contact with.it , and. to find. an insulation that could function effectively r e g a r d l e s s of whether it was • wet or dry.

A r e s e a r c h p rog ram was initiated at the Alliance Resea rch Center to explore the p r e s s u r e suppress ion aspect of wet containment. The init ial phase of this r e s e a r c h was repor ted in B&W Report BAW-1258.6 This r e s e a r c h . i s still- continuing. The f i rs t phase of the r e s e a r c h established that rup ture of a simulated p r i m a r y sys tem would be absorbed by water in the containment with a modera te p r e s s u r e pulse . These conclusions were based on simulated rup tures up to 6 inches in d iameter under r a the r r e s t r i c t ed geometry . The cha ra c t e r i s t i c s of the rup tures were not d e t e r ­mined completely, because only a l imited amount of ins t rumenta t ion was available on these t e s t s . The second phase of the tes t now being c a r r i e d out by Alliance Resea rch Center will include more extensive in s t rumen­tation and, therefore , will show the effects of p r i m a r y sys tem ruptures in g rea t e r detai l .

6-4

R e s e a r c h was also initiated on the possibi l i ty of using a wet t he rma l insulation. Two types of insulation were tested in these initial t e s t s . One group consisted of conventional insulation such as Kaowool or magnesia surrounded by a jacket of epoxy res in or a s imi la r ma te r i a l . The other group of insulations included s t a in l e s s - s t ee l wools packed to various den­s i t i e s . In some c a s e s , epoxy res in covering was used; in some c a s e s , it was not. During these t e s t s no insulation was found that would be s a t i s ­factory for l ong- t e rm operat ion. Therefore , a new set of exper iments a r e being run at the Alliance Resea rch Center to consider insulations subjected to l e s s seve re contact with wate r . This approach is made feasible by the simplified p r i m a r y sys tem due to the CNSG concept.

6. 5. CNSG I, Phase I

In November 1961 B&W presented the IBR design to the AEC. The AEC decided that they would like to have design work performed on a r eac to r concept that would include the best fea tures of the IBR and be adapted to merchant ship application. The new concept was to be called the CNSG concept. The CNSG was to incorpora te the best , not only of the IBR, but a lso of the previous design studies and r e s e a r c h and development that had been conducted by B&W.

The f i r s t phase of this new study was to de te rmine what features would be incorporated into the CNSG. The previous design studies and the background developed in the s team generat ion field by B&W provided a broad spec t rum of concepts that could be included in this concept. The major i tems included in this phase of work a r e l isted as follows:

1. F o r c e d - or na tura l -c i rcu la t ion p r i m a r y sys tem. 2. F o r c e d - or na tura l -c i rcu la t ion secondary sys tem. 3. Wet containment. 4. Se l f -pressur iza t ion .

In o rde r to make the choices indicated in 1, 2, and 3 above, severa l con­ceptual designs incorporat ing var ious of these fea tures were drawn up for the CNSG. Three of the conceptual designs drawn up in this phase a re shown in F igu re s 6-3, 6-4, and 6-5. The r eac to r shown in F igure 6-3 uti l izes na tura l c i rculat ion in both the p r i m a r y and secondary s y s t e m s . It was rejected because it was l a rge r , heavier , and more expensive than the forced-c i rcula t ion concepts . In addition, the design shown has only one section in i ts s t eam genera tor ; if a tube fai lure occurred the ent i re unit

6-5

must be shut down until that tube is plugged. The reac to r shown in F igure 6-4 uti l izes a modified once-through s team genera tor that incor ­pora tes d rums . The once-through s team genera tor is a platen configu­rat ion. It was found from control studies that the s team drums were unnecessa ry with this type r eac to r . It was also learned that the platen geomet ry for the once-through s team genera tor resul ted in poor util ization within the reac to r ves se l . In addition, this design is s t i l l l a rge r and more bulky than one using forced circulat ion in both the p r i m a r y and secondary s y s t e m s .

The configuration adapted as the re fe rence design in this phase of the study is shown in F igure 6-5. It ut i l izes forced circulat ion in both the p r i m a r y and secondary s y s t e m s . It a lso uses a hel ical coil configu­rat ion for the once-through s team genera tor and it e l iminates s team d r u m s .

F r o m studies of var ious kinds of containment made at the same t ime as the above studies were made, it was found that the wet containment when combined with the consolidated sys tem promised new economies in the efficient use of ma te r i a l , weight, and space. This feature was , t h e r e ­fore, incorporated into the CNSG.

Also, the choice of se l f -p ressur iza t ion was reviewed. This study influenced the choice of na tura l ve r sus forced circulat ion for the p r i m a r y sys tem. It was concluded that the combination of na tura l circulat ion and se l f -p ressur iza t ion introduced too many unknowns into the design. Under these c i rcumstances the flow through the sys tems could not be predicted when the ship was subjected to pitch, rol l , and heave. Since forced c i r ­culation made the unit sma l l e r , it was concluded that the best combination would be made by eliminating the separa te p r e s s u r i z e r vesse l and i ts a s s o ­ciated hea te r s and cont ro ls .

6. 6. CNSG I, Phase II

The reference design at the initiation of Phase I is shown in F igure 6-5. Although this design presented a great advantage over previous r e a c t o r s , it was considered that a detailed investigation of the design in the var ious pa r t s of this sys tem would make the ent i re r eac to r plant much m o r e com­pact. This sor t of work constituted Phase II of the CNSG I p rog ram. The re ference design as it existed at the end of this phase is shown in F igure 6-6. By comparing F igures 6-5 and 6-6, it can be seen that the size of the r eac to r and its containment was reduced markedly .

6-6

In refining the design, attention was also given to accessabi l i ty and maintainabil i ty. It was found that the pumps and the lower tube sheets of the s team genera tor were located too close to the core for easy ma in te ­nance. Therefore , the pumps were shifted to stalks projecting from the bottom of the r eac to r ves se l . These stalks extend above the level of the core so that pump maintenance can be ca r r i ed out while the core is covered with water , which is n e c e s s a r y for removing decay heat . The s team gene r ­ator inlet tube sheets were a lso shifted to this ves se l .

To make the r eac to r vesse l more compact, al l of the s team gener ­ator connections were moved to the r eac to r vesse l head. This permit ted the use of a cen t ra l chimney whose size was dictated by the space required by the control rod dr ives , and therefore , permit ted making the r eac to r ve s se l about 30% smal l e r than it would have otherwise been. This a lso meant that the s team genera tor would have to be removed from the reac to r vesse l during refueling.

6. 7. Burnable Poison Core Design

The CNSG I design utilized a soluble poison core (react ivi ty was con­trol led by a neutron poison, boric acid, dissolved in the p r i m a r y water ) . This concept had many important advantages: it allowed a great deal of flexibility in the amount of react iv i ty which could be controlled; it p e r ­mitted the use of a longer life core than would be possible based on control rods ; and it improved the power distr ibution in the core , therefore , p e r ­mitting the use of a sma l l e r core and decreased fuel cos t s . However, as the work p rog res sed it became obvious that the use of a soluble poison would requ i re a la rge development p r o g r a m to de te rmine its c h a r a c t e r ­i s t i c s . Also, the use of soluble poison introduced important questions regard ing the safety of the ship, pa r t i cu la r ly after sinking when it could be postulated that the boric acid solution could be replaced by sea water . The effects of the sea water would be an inc rease in react iv i ty and possibly the r eac to r going c r i t i ca l after it was on the bottom of the ocean. T h e r e ­fore, B&W initiated a sea rch for an a l te rna te means of control that would maximize the advantages of soluble poison and not incorpora te its objection­able f ea tu res .

The concept selected was a lumped burnable poison concept. F o r this concept a neutron poison is placed in the core in the shape of pins external ly identical to the fuel pins themse lves . The poison pins contain

6-7

Z r B 2 d i spersed in ZrO z . By ar ranging the proper spacing and concen­t ra t ion of poison in the core , it is possible to keep the react ivi ty of the core essent ia l ly constant within their l im i t s . This concept was explored under an AEC contract and repor ted in B&W Report SUP-3 7-1 . 7 This study established that an a r rangement of poison pins that met the requ i rements and did not take up excessive space in the core could be developed. It appeared that the react ivi ty control would be accomplished by a concen­t ra t ion of pins that was equal to about 10% of the total number of pins in the co re .

6 .8 . CNSG II Design

As was explained in the introduction to this repor t , the AEC requested B&W to r e -examine the CNSG I design having a number of factors in mind. The two major factors involved were compactness and ease of refueling. Again a number of conceptual designs were drawn and a choice made . The re ference design is shown in the ear ly port ion of this r epor t . The design that came c loses t to being accepted is shown in F igure 6-7. It differed from the re ference design in that it has a fu l l -d iameter c losure flange, so that the s team genera tor could be replaced if considered n e c e s s a r y . The p resence of this flange made the r eac to r ve s se l heavier , more expensive, and l a r g e r . Since the complete s t eam genera tor is considered a ve ry r e l i ­able component and since means a r e provided for plugging any tubes that become faulty, it was concluded that this flange should be eliminated, resul t ing in the reference design.

The CNSG II design also incorporated the burnable poison core that was developed and descr ibed in S U P - 3 7 - 1 . 7

To reduce s team generat ion in the core , which affects t he rma l p e r ­formance, the amount of hydrogen gas in the p r i m a r y sys tem was inc reased . This provided an o v e r p r e s s u r e on the p r i m a r y sys tem, thus suppress ing boiling.

The ease of refueling was increased by geometr ica l r e a r r angemen t . All obstruct ions over the core were eliminated by enlarging the chimney over the core to the full core d iameter , and by placing the control rod dr ives below the co re . The s team genera tor was lowered to take advantage of the space around the core and the space requi red below the core for the control rod d r ives . This position also minimized N16 activity in the secondary system, since the portion of the s team genera tor nea re s t the co re contained low density s team.

6-8

Due to the gas o v e r p r e s s u r e , it was possible to move the pumps to the top head, where they could be more easi ly maintained. The control rod dr ives were r ea r r anged so that they could be maintained by personnel outside the containment. Special maintenance tools will be required to keep water from draining out of the vesse l (s imi lar tools have been devel­oped for the Advanced Test Reac tor ) .

6 = 9

Figure 6 - 1 . PWR Plant Ar rangement (30,000 shaft horsepower)

REACTOR

SHIELD WATER

CONTROL ROD SUPPORT STRUCTURE

Figure 6-2. IBR Plant Ar rangement

SECTION A-A

E Q U I P DRAIN TANK

LETDOWM CQOL-ER

SECTION D O

4'FEEDWATER f CHEMICAL FEtD

l"B LOW DOWN I'GAGE GLASS

CONNECTION MANWAY 4 REO'D

Figure 6-3 . IBR Containing Bayonet Tube Boiler

DRAIN 24REQD- \

PARTIAL PARTITION 20REQ'D

(FOR LEAK DETECTION) SYSTEM

2I*MIN -t-&CLALi

f GAGE GLASS CONNECTION 8 REQ'D

GENERAL ELEVATION SECTION

5Jc MIN + i"CLAD

Figure 6-4. IBR General Ar rangement

• •

/3fc

• •

Figure 6-5 . Dual Purpose Plant (82. 5/400 MW-t)

^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^ ^

J X X X W * : n

- 9 3 i - -TUBE BAWK SOJ-

XT

CORE HEIGHT

t^^^^^^S^fo^^

SSS3 NSWV

o

N

L E G E N D

| \ \ \ > \ j CARBOhJ STEEL

STAIWLE5S STEEL

INSULATION

Figure 6-6. CNSG I Reactor and Containment Arrangement

36 -O " TO MS IDE OF TOP READ

LFGEND (T) CONTAINMENT FILL 6 DRAIN NOZZLE

[£) PUMPS

M J HEAT EXCHANGER NOZZLES

\4) ION i-F/SSION CHAMBER UOZZLES

( T ) CHEMICAL 6-LEVEL CONTROL NOZZLES

(jT) CONTAINMENT H£~/!TEXCHANGER

( 7 ) VENT PIPE

\&) VENTILATION NOZZLES

( 9 J STEAM LINE MOZZLES

(lO) ACCESS HATCH

(77) RELIEF VALISE NOZZLE

UZ) FEED iYATER. LINE NOZZLES

U3J CONDENSING TANKS

U4) NOZZLE PENETRATIONS FOR ELECTRICAL ^INSTRUMENT WRING

(/£) REACTOR VESSEL

(/2) CONTROL ROD DRIVE SUPPORT STRUCTURE

GT) CONTROL ROP PR/IY£

(7a) CHI NINE Y

Q9) EXTENSION ROD

(20) CONTROL ROD

(p) FUEL ELEMENT

(Z2) PLUG ROD

( S ) STEAM GENERATOR

MOTES / COMPONENTS iNOZZLES SHOIYN IN SECTIONAL ELEVATION ARE

ROTATED FOR CLARITY.

Z. FOR CNSG. CORE PLAN ^CONTROL ROD ROSIT/ONS SECTION Y-y SEE £>IV<i NZ 5/197E .

- 2 t& hf/SV. THK. HEMISPHERICAL HO. SECT/OMAL ELEVATION

Figure 6-7. Advanced CNSG Scheme I

CONTAINMENT

- STEAM GENERATOR 9' -2"

- 27'-8" -

APPENDIX Plant Information Sheets

A - 1

Plant Information Sheets

Heat Balance

P r i m a r y System Heat Generation Reactor input, 183.7 MW-t, Btu /hr P r i m a r y pumps input, Btu /hr

627,766,000 1,003,000

Total 628,769,000

P r i m a r y System Heat Losses To containment a i r space , Btu/hr To containment water space , Btu /hr

Total

35,000 70,000

105,000

Heat Trans fe r red

At s t eam genera tor , Btu /hr

P r i m a r y System Design Information

Design Conditions Number of pumps P r e s s u r e , psia Tempera tu r e , F

Operating Conditions at 183.6 MW-t P r e s s u r e , s team dome, psia Core outlet t e m p e r a t u r e , F Core inlet t e m p e r a t u r e , F Reactor flow X 106, l b /h r Flow per pump X 106, l b / h r

628,664,000

4 1,400

600

1,200 544.6

517 18

4.5

Volumes at 183.6 MW-t P r i m a r y water Core , ft3

Upper section, ft3

Steam genera tor and annulus, ft3

Lower section, ft3

Steam dome, ft3

Sub-Total

59.3 351.0 477.0

1,200.5

2,087.8

320.3 Total 2,408.1

A-3

Transi t Times Reactor core to s team gene ra to r , sec Steam genera to r , sec Steam genera tor to turn , sec Turn to reac to r co re , sec Reactor co re , sec

Total

1.13 0.80 2.76 3.22 0.55 8.46

P r e s s u r e Drops Core , psi Steam genera to r , psi Other, psi

Total

1.75 3.0

1.95 6.70

Secondary System Design Information

Design Conditions Number of s team genera to r s 1 Number of s team genera tor sections 2

. Externa l p r e s s u r e (steam genera to r ) , psia 1,400 Internal p r e s s u r e (outside reac to r vesse l ) , psia 1,400 Tempera tu r e , F 600

Operating Conditions at 183.7 MW-t P r e s s u r e , (superheater outlet) , psia 412 T e m p e r a t u r e , (superheater outlet), F 523 Degrees superheat , F 75.5 Feedwater t e m p e r a t u r e , F 251 Feedwater p r e s s u r e , economizer inlet , psia 453 Steam flow, l b /h r 605,300

Components

Steam Generator

General Type Number requi red

Once-through 1

Functional Data Heat load, Btu /hr P r i m a r y flow r a t e , l b /h r Steam flow r a t e , l b /h r Steam p r e s s u r e , psia

628,664,000 18,000,000

605,300 412

A-4

S t e a m t e m p e r a t u r e , F A m o u n t of s t e a m s u p e r h e a t , F F e e d w a t e r t e m p e r a t u r e , F P r i m a r y s ide p r e s s u r e d r o p , p s i Tube s ide p r e s s u r e d r o p , p s i a

P h y s i c a l D a t a

Tube m a t e r i a l N u m b e r of t u b e s Tube OD, in. M i n i m u m w a l l t h i c k n e s s , in. A v e r a g e effect ive l e n g t h , ft H e a t t r a n s f e r s u r f a c e , ft2

523 75 .5 251 3.0

41.0

Incone l 844 3 / 4

0.060 111.1

18,400

P r i m a r y P u m p s

Type N u m b e r of p u m p s Capac i t y at 544.6 F , g p m To ta l d e v e l o p e d h e a d , ft

Canned m o t o r , p r o p e l l e r 4

12,000 22

Rel ie f Va lve s

P r i m a r y Va lves

P i l o t - A c t u a t e d , N o n - C o d e Valve N u m b e r C a p a c i t y , l b / h r Se tpo in t , p s i g S i z e , in . O p e r a t o r

S e l f - A c t u a t e d , Code Valve N u m b e r C a p a c i t y , l b / h r Se tpo in t , p s i g O r i f i c e , in. In le t s i z e , in. Out le t s i z e , in.

1 33,200

1,300 2

P n e u m a t i c

2 33,200

1,400 0.503

1.5 2.5

S t e a m G e n e r a t o r S e l f - A c t u a t e d W a t e r Rel ie f N u m b e r Se tpo in t , p s i g

2 1,400

T h e s e v a l v e s a r e to p r o v i d e r e l i e f of w a t e r p r e s s u r e when the i s o l a t i o n

v a l v e s a r e c l o s e d and the s t e a m g e n e r a t o r is so l id w a t e r .

A - 5

C o n t a i n m e n t

D e s i g n Da ta P r e s s u r e , p s i g T e m p e r a t u r e , F

H e a t L o s s to C o n t a i n m e n t D u r i n g O p e r a t i o n

To a i r s p a c e , B t u / h r To w a t e r s p a c e , B t u / h r

To ta l

100 28 0

35,000 70,000

105,000

P h y s i c a l Da ta

D i a m e t e r of c y l i n d r i c a l s e c t i o n , ft He igh t of o v e r a l l s y s t e m , ft

V o l u m e s a t 110 F

W a t e r v o l u m e , ft3

A i r v o l u m e , ft3

T o t a l

22 37

4,730 2,404

7 ,134

R e a c t o r C o r e

G e n e r a l D a t a P o w e r , MW Coolant flow r a t e X 106 , l b / h r O p e r a t i n g p r e s s u r e , p s i g A v e r a g e outlet, t e m p e r a t u r e , F

184 N o r m a l 18

1,300 544.6

C o r e N u m b e r of fuel e l e m e n t s 16 N u m b e r of s t a n d a r d e l e m e n t s 12 N u m b e r of c o r n e r - p a r t i a l e l e m e n t s 4 N u m b e r of fuel p i n s 7 ,008 N u m b e r of p o i s o n p ins 616 A c t i v e fuel l e n g t h , in. 62 F u e l c l add ing m a t e r i a l Z i r c a l o y - 2 F u e l UO z F u e l r o d d i a m e t e r , . in. 0.43 F u e l c lad t h i c k n e s s , in . 0.017 M e t a l - t o - w a t e r r a t i o (unit ce l l ) 0.778 F u e l r o d p i t c h , in . 0.57 6 F u e l d i a m e t e r , in . 0 .394 Type of fuel f a b r i c a t i o n V i b r a t o r y c o m p a c t e d To ta l u r a n i u m (U0 2) ,

t o n n e s 8.466 To ta l u r a n i u m m e t a l U,

t o n n e s 7 .462 E n r i c h m e n t ( a v e r a g e ) , % 5.3

A - 6

Theoret ica l density of fuel, % Core life (184 MW-t),

full power days Average fuel burnup,

MWD/T Maximum fuel burnup,

MWD/T Number of control rods Absorber m a t e r i a l Fabr ica t ion

Stainless s teel clad, in. Boron-s ta in less steel, cen ter , in.

Stroke length, in. Absorber length / rod , in. Absorber length, in. Follower length, in. Zirconium

Fuel pin average 'hea t r a t e , kw/foot

Fuel pin maximum heat r a t e , kw/foot

Average surface flux, Btu/hr- f t 2

Maximum surface flux, Btu/hr- f t Hot channel surface flux (FOP 1.3)

Btu /hr - f t 2

88

1,036

25,000

48,000 (approx.) 12

Boron-s ta in less s teel 1/16 1/8 50 40 50 62

4.85 15.2

147,000 459,000 598,000

Water Conditions

P r i m a r y Fluid (initial Fill) Total sol ids , ppm pH Oxygen (hydrazine control led) , ppm Chlorides (as chloride ion), maximum, ppm

0.05 Neutra l

0.01 0.01

Makeup (Pr imary and Secondary) Total sol ids , ppm PH Oxygen, ppm Chlorides (as chloride ion), maximum, ppm

0.01 Neutral

0.01 0.01

Feedwater Total dissolved so l ids , ppm

exclusive of NH3 Suspended so l ids , ppm H a r d n e s s , ppm Organic , ppm F r e e , caus t ic , ppm Dissolved oxygen, max imum, C0 2 , ppm Chlorides (as chloride ions),

maximum, ppm Total s i l ica (as S i0 2 ) , max imum, ppm Total i ron, maximum, ppm Total copper, maximum, ppm

0.05 - 0.50 0.01 - 0.04

0 0 0

ppm (Preferably 0) 0.007 (Preferably 0) Minimum

0.01 0.02 0.01

0.002

A - 7

The pH value should be adjusted to obtain 0.01 ppm of iron maximum. This will normally requ i re a pH value within the range of from 8.8 to 9.2 between the t empera tu re range of 75-125 F .

6. Mater ia ls of Construction

Mater ia ls Contacting P r i m a r y Fluid Reactor shell Carbon steel clad with Type 304

s ta in less s teel Reactor heads Carbon steel clad with Type 304

s ta in less s teel Internals Type 304 s ta in less s teel Nozzles Carbon steel clad with Type 304

s ta in less steel Tubing Inconel Tube sheets Carbon s teel clad with Inconel Piping Type 304 s ta in less steel

Mater ia ls Contacting Secondary Fluid Tubing Inconel Tube sheets Carbon steel clad with Inconel Piping Carbon steel

A - 8

REFERENCES

Consolidated Nuclear Steam Generator - Final Report , The Babcock & Wilcox Company, BAW- 1243, Lynchburg, Virginia, August 1962.

Janssen , E. and Levy, S., Burnout Limi t Curves for Boiling Water R e a c t o r s , APED-3892, Apri l 14, 1962.

NS Savannah Upgrading P r o g r a m - Advanced P r e s s u r i z e d Water Reactor Concepts, The Babcock & Wilcox Company, BAW-1219, Lynchburg, Virginia, March 1961.

MacMillan, J. H., Graham, D., and Kulynych, G., Natural Circula­tion Reactors for Marine Propuls ion , The Babcock fk Wilcox Company, BAW- 192, Lynchburg, Virginia.

A Study of the Flooded Containment Concept and Its Application to the NS Savannah (Official Use Only), The Babcock & Wilcox Company, SUP- 26, Lynchburg, Virginia, July 1961.

Luken, R. C. and Leeman, C. A., Savannah Upgrading P r o g r a m -Vapor Suppression Test P r o g r a m Report , The Babcock &: Wilcox Company, BAW- 1258, Lynchburg, Virginia, August 1962.

Consolidated Nuclear Steam Generator - Design Review, The Babcock & Wilcox Company, SUP-37- 1, Lynchburg, Virginia, October 1962.

DISTRIBUTION

1. United States Atomic Energy Commission

Botsford, CW, Mari t ime Reactors Branch, Germantown, Md. (50) Robb, J E , Washington 25, D. C.

The Babcock & Wilcox Company

Central F i les (10) Craven, J P / B a r b e r t o n L ib ra ry Deus te r , RW (2) Dobel, HF Gumprich, WC L i b r a r y , AED (2) L i t t r e l l , LW (2) Mumm, J F P le tke , LR Purdy , DC Rock, HR

Sankovich, MF Sawyer, GA Schoessow, EE Schomer , RT Taber , AP/Al l iance L ibra ry Technical Report Group T r a v i s , CC Wascher , RE Webb, RA Wilson, CW, New York Sales (20)