ML20042A066

Forwards Assessment of SEP Topics II-1.A,III-2,III-3.A, III-4.A,III-4.C & III-5.A.Submittal Represents Last Six Topics for Which Util Has Responsibility

ML20042A066
Person / Time
Site:	Yankee Rowe
Issue date:	03/18/1982
From:	Kay J YANKEE ATOMIC ELECTRIC CO.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-02-01.A, TASK-03-02, TASK-03-03.A, TASK-03-04.A, TASK-03-04.C, TASK-03-05.A, TASK-2-1.A, TASK-3-2, TASK-3-3.A, TASK-3-4.A, TASK-3-4.C, TASK-3-5.A, TASK-RR FYR-82-34, NUDOCS 8203230051
Download: ML20042A066 (71)

Text

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t-I YANKEE AT0DIC ELECTRIC COMPANY fh 1671 Worcester Road, Framingham, Massachusetts 01701 A

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March 18, 1982 S

EMACF.WED f,

United States Nuclear Regulatory Commission MAR 2 21982> 3 Washington, D. C.

20555 Ff g o:au tav:tt umans imm WMlet IA T3 Attentlon:

Mr. Dennis M. Crutchfield, Chief g

7, Operating Reactors Branch #5 N

Division of Licensing Re f e re nc e :

(a) License No. DPR-3 (Docket No. 50-29)

Su bj ec t :

SEP Topic Assessment Completion

Dear Sir:

Enclosed please find our assessment of the following topics:

II-1.A Exclusion Area Authority and Control III-2 Wind and Tornado Loadings III-3.A Effects of High Water Level on Structures III-4.A Tornado Mi ssiles III-4.C Internally Generated Missiles III-5.A Ef fects of Pipe Break on Structures, Systems, and Components Inside Containment These assessments represent the last six topics we have responsibility for writi ng.

We t ru s t this information is satisfactory; however, if you have any questions, please contact us.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY J. A. Kay Senior Engineer - Licensing JAK: dad Enclosure 0

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YANKEE ATOMIC ELECTRIC COMPANY Topic II-1.A Exclusion Area Authority and Control I.

Introduction The safety objective of this topic is to assure that appropriate exclusion area authority and control are maintained by the licensee as required by 10 CFR Part 100.

II.

Review Criteria Section 100.3(a) of 10 CFR Part 100 requires that a reactor licensee have the authority to determine all activities within the designated area, including the exclusion or removal of personnel and property.

III. Evaluat_fon The Yankee Nuclear Power Station is located in a valley in the town of Rowe, Massachusetts on the east bank of the Deerfield River, three-quarters of a mile south of the Vermont-Massachusetts border. The site consists of approximately 2,000 acres straddling the Deerfield River in the towns of Rowe and Monroe, Massachusetts and is owned, in fee, by Yankee Atomic Electric Company (YAEC), or an affiliate, the New England Power Company. YAEC is a subsidiary of New England Power Company. The site is shown in Figure 1.

The exclusion area boundary is defined by a 3,100 f t. radius centered on the reactor vapor container, with the exception of a small segment in the southern sector where the minimum distance is approximtely 2,700 f t.

This indentation into the 3,100 f t. exclusion radius is formed by an undeveloped corner of the Monroe State Forest and is situated behind an 1,800 ft. (MSL) ridge line with respect to the plant.

Plant grade is 1,127 ft. (MSL). Figure 1 chows the exclusion area boundary.

All the land in the exclusion area is owned by Yankee Atomic Electric Company or New England Power Company, with the exception of a small parcel situated across the river and southwest of the plant which is owned by the Deerfield Specialty Paper Company. All the area within the exclusion boundary is under the control of YAEC. Written permission has been obtained from the paper mill in Monroe Bridge to have that portion of their land, which is within the 3,100 f t, radius, under Yankee control in the event of an incident. This piece of land is used as a disposal area by the paper mill and contains no permanent buildings or residents.

Two public secondary roads traverse the exclusion area. The closest is across the river frem the plant and is approximately 1,500 f t. away at its closest point. This road runs north-south along the river between Monroe Bridge and Readsboro, Vermont. The second road, which connects to the main access to the plant, is approximately 2,500 ft. away at its nearest point and runs between Monroe Bridge and Eowe, south of the plant. Provisions have been made in the station's Emergency Plan to protect the public along these roadways should it become necessary in an emergency.

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n One house is located within the exclusion area.

It is across the pond NNW f rom the plant, approximately 1,500 f t. away, on the river road. - It is owned by the New England Power Company and occupied by one of its employees.

Provisions have. been made in the station's Emergency Plan to evacuate til1s residence if necessary in the event of an emergency.

IV.

Conclusion Based on the above evaluation, we conclude that YAEC has the proper authority to determine all activities within the exclusion area, as required by 10. CFR Part 100.

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1 Yankee Nuclear Power Station Site Area Figure 1

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YANKEE ATOMIC ELECTRIC COMPANY SEP TOPIC II-3.A Effects of High Water Level on Structures I.

Introduction The design basis flood high water level was evaluated in response to SEP Topic II-3.A.

In order to ensure protection of safety-related equipment, it was necessary to install flood barriers.

The purpose of this topic is to evaluate the capability of structures used as flood barriers to perform their intended function.

t curre nt Crit eria The current criteria pertaining to this topic are

- NRC Standard Review Plan 3.4

- Uniform Building Code II.

Evaluation, The maximum water level at the site was determined to be above grade.

The only structures subjected to the hydrostatic loading are portions of the turbine building / service building complex, and the screenwell and pumphouse.

Turbine / Service Building The turbine building will be exposed to external flood levels for a small portion of its northeasterly side. The north and east sides of the service building will also be exposed to these external flood levels.

Flood levels will result from the Design Basis Flood stillwater elevation and the coincident wave activity. Wind setup is negligible (less than 0.1 foot).

Ground elevations fronting this section of the plant range f rom 1,126 f eet to 1,128 feet, which will allow the design significant and maximum waves-to reach and runup against these sections of wall.

During the maximum stillwater elevation of 1,131.7 feet, the significant wave will runup to elevation 1,134.1 feet, while the maximum wave will runup to elevation 1,136.6 feet. This maximum condition will last for a

approximately 15 minutes as the water level rises to its peak and then drops rapidly.

Those portions (masonry walls) subjected to the hydrostatic head were evaluated and determined to be not capable of withstanding the required loading.

Screenwell and Pumphouse The flood level of 1,131.7 MSL will cause water to rise to a level approximately 7 feet above the pumphouse concrete foundation which serves as the access level elevation.

Presently, both access doors have 4 foot

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high reinforced plywood barriers that are to be installed for the present flood protection plan. These barriers employ grooved fittings on each side of the door frames to hold them in place. The windows of the building are 41/2 f eet above the floor elevation.

The walls of the pumphouse are constructed of concrete blocks which rest upon the reinf orced concrete foundation.

The walls are not reinforced and are of standard masonry construction. The concrete blocks rest against-the outside of the steel framework and are unbraced between vertical support beams.

The longest wall section is 23 feet long by 12 feet high.

The force of 7 feet hydrostatic pressure will fail the wall inwards and allow the pumphouse to flood.

III. Conclusion A small portion of the northeasterly side of the turbine building, the north and east sides of the service building, and the screenwell and pumphouse will be exposed to flood levels above grade level.

These structures, as currently designed, are not capable of withstanding the postulated flood condition without failure. of a barrier structure.

The need to provide additional flood protection capability will be evaluated during the integrated assessment. Resolution of this topic is also dependent on the outcome of SEP Topics II-3. A and II-3.B i

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t YANKEE ATOMIC ELECTRIC COMPANY SEP Topic III-2 Wind and Tornado Loadings I.

In trod uc tion The purpose of this topic is to determine if safety-related structures are adequate to resist wind and tornado loadings, including tornado pressure drop loading.

II.

Current Criteria The current criteria pertaining to this topic are:

- NRC Standard Review Plan 3.3

- NRC Standard Review Plan 3.8

- Uniform Building Code III. Evaluation Safety-related structures were evaluated for the capability to resist wind and tornado induced loads. Since an established criterion for loading was not available (it is the subject of another SEP Topic), the capacity of each structure was determined.

Table I lists the structures evaluated, and the critical wind speeds for different portions of the structures.

IV.

Conclusions The conservatism of the design code is evident by the fact that the plant is 21 years old, and masonry wall structures have experienced without failure, loadings supposedly in excess of their computed capacity.

Gross structural collapse of the reactor containment and reactor support structure is not computed to occur for wind speeds in excess of a value determined by the NRC to have a 10-7 probability of occurrence.

There are no unique features of the Yankee Nuclear Power Station that make it abnormally susceptible to wind or tornado damage; the most critical structures (the vapor container and reactor support structure) are extremely robust and, thus, tolerant of severe loading without failure.

The need for modifications will be evaluated during the integrated assessment. Resolution of this topic is dependent on the outcome of SEP Topics on saf e shutdown systems (VII-3), seismic design (III-6), and establishment of a design basis wind and tornado loading.

IABLE 1 WIND CAPACITY OF STRUCTURES STRUCTURE SYSTEM WIND VELOCITY MPH 252

. STRUCTURAL STEEL VAPOR CONTAINER

> 252 SKIN STRUCTURAL 158 64 TURBINE BUI LDING SKIN 161 ROOF 192 STRUCTURAL 40 PRIMARY AUXILIARY BUILDING SKIN 165 ROOF 190 ROOF DIESEL GENERATOR BUILDING 46 SKIN NOTES:

1.

THE EFFECT OF ATMOSPHERE PRESSURE CHANGES AND TORNADO BORNE MISSILES ARE NOT CONSIDERED.

2.

STEADY AIR FLOW 3.

SPECIFIC DETAILS ON THE METAL SIDING AND DECKING ARE NOT AVAILABLEj CAPACITIES ARE BASED ON STRENGTH OF SUPPORTING MEMBERS.

i YANKEE ATOMIC ELECTRIC COMPANY TOPIC III-4.A TORNADO MISSILES I.

INTRODUCTION Topic III-4.A is intended to review the plant design relative to its ability to withstand tornado generated missiles to assure that those structures, systems and components necessary to ensure:

1.

The integrity of the reactor coolant pressure boundary, 2.

The capability to shut down the reactor and maintain it in a safe shutdown condition, and 3.

The capability to prevent accidents which could result in unacceptable of f-site exposures can withstand the impact of an appropriste postulated spectrum of tornado generated missiles.

An assessment of the adequacy of a plant to withstand the impact of tornado missiles includes:

1.

Determination of the capability of the exposed systems, components and structures to withstand key missiles (including small missiles with penetrating characteristics and larger missiles which result in an overall structural impact),

2.

Determination of whether any areas of the plant require additional l

protection.

II.

REVIEW CRITERIA l

The plant design was reviewed with regard to General Design Criterion 2,

" Design Bases for Protection Against Natural Phenomena", of Appendix A,

" General Design Criteria for Nuclear Power Plants", to 10 CFR Part 50, I

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" Licensing of Production and Utilization Facilities", which requires, in l

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i part, that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as tornadoes without loss of capability to perform their safety functions.

III. RELATED SAFETY TOPICS Topic II-2.A, " Severe Weather Phenomena" describes the tornado characteristics for the plant. Topic III-2, " Wind and Tornado Loadings" reviews the capability of the plant structures, systems and components to withstand wind loadings. Topic VII-3, " Systems Required for Safe Shutdown" reviews those systems needed to achieve and maintain the plant in a saf e shutdown condition.

IV.

REVIEW GUIDELINES The review was performed in accordance with Standard Review Plan (SRP) 3.3.2, " Tornado Loadings", 3.5.3, " Barrier Design Procedure", and 3.5.1.4, " Missiles Generated by Natural Phenomena", Revision 1.

SRP 3.5.1.4 identifies two missile sets known as Spectrum I and Spectrum II missiles, each of which contains a variety of missiles and their corresponding velocity. A plant applying for a construction permit would be required to design for one of these missile sets. This SRP states that plants which were not required at the construction permit stage to design to the missile spectrum provided in Revision 0 to the SRP should show the capability to withstand two of the postulated missiles in the Revision 0 spectrum.

The following missiles are described in SRP 3.5.1.4 as being appropriate for evaluating OL Applications for plants which were not required to be protected against the full tornado missile spectrum during the CP stage:

1.

Steel Rod,1" diameter, 3' long, 8 lbs, horizontal velocity - 0.6 x total tornado velocity.

2.

Utility Pole, 13-1/2" diameter, 35' long,1490 lbs, horizontal velocity = 0.4 x total tornado velocity.

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The systems, structures and components required to be protected because of their importance to safety are identified in Regulatory Guide 1.117.

V.

EVALUATION Critical structures, systems and components of the Yankee Nuclear Power Station can be characterized in one of the following categories:

1.

Exposed equipment 2.

Steel sided structure 3.

Masonry wall structure 1

4.

Reinforced concrete structure Exposeu equipment includes tanks, piping, valves, instrumentation, cabling, and containment penetrations exposed to the environment.

Examples include:

Refueling Water Storage Tank Demineralized Water Storage Tank Primary Water Storage Tank Fire Water Storage Tank Liquid Waste Holdup Tank Waste Cas Decay Tank Diesel Generator Fuel Oil Tank Main Steam Piping (portions of) i Feedwater Piping (portions of)

ECCS Recirculation Piping (portions of)

Vapor Container Electrical Penetrations The likelihood of tornado missile entry into the area under and immediately around the vapor container is believed to be less than what would be expected for a similar structural enclosure which was " exposed" from all sides. Tanks, piping, valves and structural walls within this area are shadowed by adjacent structures and therefore confine eligible tornado missiles to specific approach paths.. - _

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Steel sided structures include the upper section of the turbine building and the vapor container.

Masonry wall structures include the following:

Screenwell and Pump House Cable Tray House Diesel Generator Building Primary Auxiliary Building Waste Processing Building Upper Pipe Chase (non-radioactive)

Lower Section of Turbine Building Spent Fuel Building (portions of)

Reinforced concrete structures include the following:

Control Room Lower Pipe Chase (radioactive)

Fuel Chute Switchgear Room (including battery room)

Based on a review of pertinent industry and regulatory data we conclude the following:

1.

Compocents inside the vapor container are adequately protected.

The steel shell of the vapor container is a minimum of 0.875 inches thick.

2.

Components inside the control room are adequately protected. The exterior walls of the control room are reinforced concrete at least 3 feet thick. The interior wall is reinforced concrete 1 foot thick.

It is positioned such that a direct strike is unlikely.

3.

Exposed components are unprotected. Some equipment, located under and around the vapor container, is, however, shadowed by adjacent structures.

i 4.

Equipment in the diesel generator building is unprotected.

5.

Most areas of the primary auxiliary building are protected.

6.

The turbine building outside of the control room is unprotected.

VI.

CONCLUSION Based on our evaluation, we conclude that the following areas are protected against tornado missiles:

o components inside the vapor container, o

components inside the control room, and o

most areas of the PAB.

Areas not protected and vulnerable to tornado missile effects are as follows:

o exposed components, o

components inside the diesel generator building, and o

components inside the turbine building.

The need for providing additional tornado missile protection will be evaluated during the integrated assessment. Resolution of this topic is dependent on the outcome of SEP Topics on safe shutdown systems (VII-3),

seismic design (III-6) and establishmer.t of a design basis wind and tornado loading (III-2).

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i SAFETY EVALUATION 1

YANKEE ATOMIC ELECTRIC COMPANY

-YANKEE NUCLEAR POWER STATION SYSTEMATIC EVALUATION PROGRAM TOPIC:

III-4.C INTERNALLY CENERATED MISSILES

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I.

INTRODUCTION Missiles that are generated internally to the reactor facility (inside or outside containment (called the vapor container)) may lead to damage of structures, systems and components that are necessary for the safe shutdown of the reactor facility or accident mitigation and to the structures, systems and components whose f ailure could result in a significant release of radioactivity. The sources of such missiles are valve bonnets and hardware retaining bolts, relief valve parts and instrument wells, pressure containing equipment such as accumulators and high pressure bottles, high speed rotating machinery, and rotating segments (e.g., impellers and fan blades).

Scope of Review The scope of the review is as outlined in the Standard Review Plan (SRP), Section 3.5.1.1, " Internally Generated Missiles (Outside Containment)", and SRP Section 3.5.1.2, " Internally Generated Missiles (Insid e Containment)". The review specifically excludes SRP Sections 3.6.1, " Plant Design for Protection Against Postulated Piping Failures in Fluid Systems Outside Containment", 3.6.2, " Determination of Break Locations and Dynamic Ef fects Associated with the Postulated Rupture of Piping", as well as those SRP sections dealing with natural phenomena (including missiles generated by natural phenomena, missiles generated outside the facility, and turbine missiles).

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II.

REVIEW CRITERIA The acceptability of the design of protection of facility structures, systems, and components from internally generated missiles is based on meeting the following criteria:

1.

General Design Criterion 4, with respect to protecting structures, systems and components against the effects of internally generated missiles to maintain their essential safety functions.

2.

Regulatory Guide 1.13, as related to the spent fuel pool systems and structures being capable of withstanding the effects of internally generated missiles and preventing missiles from impacting stored fuel assemblies.

3.

Regulatory Guide 1.27, as related to the ultimate heat sink and connecting conduits capable of withstanding the effects of internally generated missiles.

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4 III. RELATED SAFETY TOPICS AND INTERFACES Review Areas Outside the Scope of this Topic This review specifically excluded the following:

1.

SRP Section 3.6.1, " Plant Design for Protection Against Postulated Piping Failure in Fluid Systems Outside Containment" - This matter will be covered under safety topic III-5.B. " Piping Break Outside Containment".

2.

SRP Section 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping" - This matter will be covered under safety topic III-5.A, " Effects of Pipe Break on Structure, Systems and Components Inside Containment".

3.

Natural Phenomena - This matter will be covered under safety topic III-6, " Seismic Design Considerations", and III-4. A, " Tornado Missiles".

4.

Turbine Missiles - This matter will be covered under a generic NRC study.

Interfaces with Other SEP Safety Topics Satisf actory resolution of the following safety topics will depend, at least in part, on satisfactory resolution of this topic:

1.

Topic VII-3, " Systems Required for Safe Shutdown".

2.

Topic VII-4, " Effects of Failure on Non-Safety-Related Systems on Selected Engineered Safety Features".

3.

Topic IX-1, " Fuel Storage".

4.

Topic IX-3, " Station Service and Cooling Water System".

5.

Topic II-3.C, " Safety-Related Water Supply" (Ultimate Heat Sink). -

IV.

REVIEW CUIDELINES 1.

Systems and components needed to perform safety functions were identified as those listed in SRP Section 3.2.2, " Systems Quality Group Classification".

Systems or portions of systems needed to perform safety functions (safe plant shutdown or accident mitigations) include:

a.

Main Coolant System b.

Emergency Core Cooling System c.

Chemical and Volume Control ' ystem d.

Shutdown Cooling System e.

Component Cooling Water System f.

Service Water System g.

Compressed Air System h.

Diesel Generator Auxiliary Systems 1.

Main Steam System (portions of)

j. Feedwater and Condensate Systems (portions of) k.

Fmergency Feedwater System 1.

Ventilation System for Areas Such as Control Room and Diesel Generator Building m.

Condensate Storage System -_

2.

Systems whose failure may result in release of unacceptable amounts of radioactivity:

a.

Spent Fuel Pool Cooling System b.

Sampling System c.

Liquid Waste Processing System d.

Caseous Waste Processing System e.

Containment Purge System 3.

Additionally, electrical systems which are necessary to support those fluid systems needed to perform safety functions are:

a.

Diesel Generators b.

Station Batteries and Distribution Cabinets c.

Emergency 480 V Switchgear 4.

Additional areas needed to suppoort safety system functions are:

a.

Control Room b.

Switchgear Room c.

Cable Tray Room

V.

REVIEW AND EVALUATION 1.

Systems needed to perform safety functions:

a.

Main Coolant System The main coolant system serves as the pressure retaining boundary for the primary coolant and is comprised of a reactor pressure vessel and four parallel heat transfer loops. Each loop contains one steam generator and one pump, connecting piping and instrumentation. A pressurizer and safety valves are connected to one of the reactor outlet pipes. The purpose of the pressurizer is to maintain primary coolant pressure and compensate for coolant volume changes as the heat load changes.

All components of the main coolant system are located within the vapor container. Overpressure protection is provided to assure the coolant system pressure does not exceed design limits.

The reactor vessel head is secured to the reactor vessel by special studs and nuts.

It is unlikely that any of the studs would become a missile since they are not subjected to reactor pressure and, therefore, are not exposed to sufficient pressure to create an accelerating force sufficient to cause them to become missiles.

The pressurizer safety va.ws have the potential for becoming missiles. Both the power perated relief valve and the code safety valves are mounted on top of the pressurizer. The position of the pressurizer above the loops and in a concrete compartment is such that any parts blown off the valves would strike above or to the side walls, and are not likely to damage other components or piping of the main coolant system.

Control rod drive modules are mounted on top of the reactor vessel and are considered as an extension of the reactor vessel head.

Each module is attached to a threaded connection on the reactor vessel head and seal welded. A metal encased concrete cover is placed over the control rod housing during operation as protection against potential missile damage to safety systems.

Instrumentation generally requires some penetration into the main coolant system. These penetrations are usually small and take the form of welded wells. Should one fail, it will not cause serious destructive dam _3e to the main coolant system or compromise its safety.

Also considered was the possibility that missiles may result from destructive overspeeding of one of the main coolant pumps in the event of a pipe break in the pump discharge.

It was concluded that potentially damaging impeller missile ejection is minimized by a massive steel pump casing and thus, the probability of missiles from overspeed of both the motor and impeller of a main coolant pump that could result in damage to safety-related equipment is acceptably low.

The four steam generators have manways held in position by studs on the primary and secondary sides of the shell. These stud s are not subjected to sufficient pressure to result in a significant missile source. Therefore the steam generators are not a likely missile generating source.

In summary, in considering the main coolant system, because of its equipment design features and component arrangement, it is our judgment that this system's function will not be detrimentally affected considering internally generated missile sources from the main coolant system as identified above.

Further, should a missile create a break in the main coolant system, the emergency core cooling system will keep the core cooled.

b.

Emergency Core Cooling System The Yankee emergency core cooling system (ECCS) serves as the.

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i means of injecting water for core protection in the event of primary coolant system water loss and consists of three sets of high and low pressure injection pumps in series to boost final discharge pressure and a safety injection accumulator. ECCS flow is provided from these subsystems into the main coolant system through the four cold leg reactor inlet pipes.

The initial source of water for the high pressure and low pressure pumps is the safety injection tank. This tank is located out-of mdoors behind the safety injection building.

The high and low pressure piping follows a common path from the pumps, through the radioactive pipe chase, passes through containment penetrations, well below the operating floor and branches out to the four cold leg reactor inlet pipes.

The safety injection accumulator is located in a separate cubicle of f the safety injection building, and it discharges into the low pressure injection header. This branches out into four separate lines which connect to the four cold leg reactor inlet pipes as described above. This physical separation provides inherent protection against damage from internally generated missiles.

The main components of the ECCS redundant trains are located in the diesel generator (DG) building. The most likely sources of missiles in the DG building are the pumps. They are all horizontal, multistage centrifugal pumps operating at 3,600 rpm.

Yankee believes that these pumps will not become sources of missiles because the impeller, if broken, is not likely to penetrate the thick steel casing.

The positioning of the safety injection tank behind the safety injection building would make the tank safe from internally generated missiles.

_g_

Once the safety injection piping enters containment, it is kept below the operating floor before separating and connecting to the four reactor inlet (cold) legs. While this piping may be subject to damage from internally generated missiles, a single missile should not damage more than a limited area of a single train of piping. Further, depending on where the pipe damage is encountered, upstream or downstream of the check valve nearest the primary system, determines the extent of the effect on the primary system.

If the damage (break) occurs downstream of the valve, a demand would be placed on the ECCS (LOCA) and the system will perform its design function.

If the break occurs upstream of the check valve, no demand will be placed on the ECCS as the check valves would close and a LOCA would not occur.

In summary, in considering the ECCS, because of its functional design, redundant features, and equipment design features, it is our judgment that this system will be capable of performing its design function considering internally generated missile sources as discussed above.

c.

Chemical and Volume Control System The chemical and volume control system (CVCS) controls and maintains main coolant system inventory through the process of makeup and letdown. The system consists of a regenerative letdown heat exchanger to cool the excess coolant taken from the primary system during plant heat up and plant operation. The coolant is reduced in temperature and pressure and passed to the purification system where corrosion and fission products are removed. The coolant is then returned to the low pressure surge tank (LPST). The charging pumps return the coolant to the primary system from the LPST. Additional water is supplied to the charging pumps as necessary.

_9_

The equipment for this system other than the regenerative letdown heat exchanger is located in the auxiliary building in individual rooms which contains no equipment from other systems which might produce missiles. The heat exchangers are located inside containment.

The system is not fully redundant and a missile could disable the system. Should this happen, the plant could still be shut down safely in a normal manner using the emergency feedwater system and the shutdown cooling system.

We, therefore, conclude that the possibility of internal missile damage to the CVCS is very low.

Such an event will not result in an unacceptable release of radioactivity or endanger the safe shutdown cf the plant. We conclude that additional protection is not required.

d.

Shutdown Cooling System The shutdown cooling system is brought into use during plant shutdown when the primary coolant temperature and pressure fall below 330 F and 300 psig. At that time, certain valves are remote manually aligned to allow primary coolant flow through the shutdown cooling heat exchanger. The shutdown cooling system consists of the shutdown cooling heat exchanger and the associated valves and piping. The reactor coolant is circulated through the shutdown cooling heat exchanger using the shutdown cooling pump and is then pumped back to the cold leg inlet of the reactor. The shutdown cooling heat exchanger transfers decay heat to the component cooling water system which in turn transfers its heat to the service water system which discharges into Sherman Pond, the ultimate heat sink.

l While the shutdown cooling system is not completely redundant, the LPST cooling system pump and heat exchanger can be valved in to act as redundant components. These components are all located in well shielded separate cubicles in the PAB. The only sources of missiles in the area are the pumps themselves.

The piping is also run in shielded enclosures since it contains primary coolant. Because of the separation of equipment in separate shielded cubicles, and the lack of missile sources, we conclude that the shutdown cooling system is adequately protected from possible missiles.

e.

Component Cooling Water System The component cooling water system is a closed system with two motor-driven pumps rated at 125 hp and 2,000 gpm and two horizontal heat exchangers. This equipment is located on two levels of the primary auxiliary building. Heat transferred to the component cooli', w ner system is transferred to the service water system through tue two horizontal heat exchangers and is released into Sherman Pond (see also topic IX-3, " Station Service and Cooling Water System").

The component cooling water system removes heat from the shutdown cooling heat exchanger, and reactor coolant pssps. The component cooling water system also provides cooling for the spent fuel heat exchanger and for the neutron shield tank.

From our ry.iew of the component cooling water system, we conclude that it is unlikely that this system would be a source of missiles since it is operated at less than 100 psig and the pump speeds are only 1,750 rpm.

Further, we did not identify any potential missile sources which might endanger the component cooling water system. Should a missile strike the component cooling water system, it would be necessary to shut the reactor down. Should a missile strike the branch loop which cools the neutron shield tank cooling system, its loss is not of immediate concern as the temperature rise in the tank is slow. A missile strike into the spent fuel heat exchanger branch loop can be tolerated as the fuel pool heatup is a slow process and there is time to execute repairs or provide alternate cooling before the pool temperature becomes excessive.

l

We conclude that the component cooling water system is adequately protected from internally generated missiles.

f.

Service Water System The service water system consists of three 125 hp, 2,500 gpm, 50% capacity vertical motor-driven pumps located in the screenwell house. (See also SEP Topic IX-3, " Station Service and Cooling Water System".) These pumps take their suction from the intake structure (Sherman Pond) and discharge into two redundant headers. Isolation valves are provided at each pump.

We conclude that the three vertical motor-driven service water pumps are unlikely missile generators because of their enclosure (casing) and submergence in the intake structure, their low operating speed (1,750 rpm) and low operating pressure ( 100 psig) and would not endanger the service water system piping and valves in the intake structure. At the intake structure, there are two electric, driven fire pumps and two circulating water pumps along with the controls for the traveling screens. We do not consider this equipment as likely sources of missile generation. The pumps are unlikely missile generators for the same reasons discussed above for the service water pumps.

The only service water loads of any safety significance are the component cooling heat exchangers. If service water is not available, connections are available to supply the component t

l cooling heat exchangers from the fire system or temporary sources.

Therefore, in the event of loss of this system, decay heat removal could be accomplished by use of the emergency feedwater system. The main coolant system could be placed in a natural I

circulation mode utilizing the steam generators for decay heat renoval.

I l

l l

We conclude that the service water system is not a potential source of damaging missiles nor will its loss prevent a safe shutdown of the plant.

g.

Compressed Air System The compressed air system is designed to supply oil free air to both the control air and service air systems. Three full capacity non-lubricated compressors are provided with separate after coolers and air receivers. The air receivers are int erconnect ed. Two air headers leave the air receivers and go through filters and dryers before going to instrument and control air supply manifolds. Both headers supply each manifold.

One compressor with intake filter af ter cooler and receiver supplies service air loads.

It can be cross connected as a backup to the control air systems.

All of the major components of the compressed air system are located on the ground floor of the turbine building. Because of the low pressure and low speed of the compressor units, the system is not considered a missile source. However, the main feed system is in the same area and is a missile source (rotating pump). With a loss of compressed air, safety-related equipment can be operated manually or by backup nitrogen supplies. Therefore, a loss of the compressed air system due to internally generated missiles will not prevent a safe shutdown.

h.

Diesel Generators Auxiliary Systems Three diesel generators are located in the NW wing of the safety injection building. The three units and their auxiliaries are located in separate rooms. A common fuel oil header runs along the front of the rooms in the safety injection building. The only missile sources are the safety injection pumps. As discussed in Section Ib, these are not believed to be a l

potential missile source.

l t

i j

Therefore, due to the independence and separation of the diesels and their remaining associated auxiliary systems, it is our jud gment that the diesel generator auxiliary systems will be capable of performing their function considering internal missiles.

1.

Main Steam System (portions of)

The main steam system consists of four steam generators, four 14-inch steam lines or headers, and automated non-return valves. The steam generators are located insid e the vapor container. The four main steam lines penetrate containment at Elevation 1,046 feet. The main steam non-return valves, safety valves, and atmospheric dump valves are located outside the containment at Elevation 1,039 feet. The four steam lines join to one 24-inch line in the turbine building and are split into two lines prior to the turbine throttle valves.

We conclude from our review of the main steam system that this system will not produce missiles due to its heavy walled design and construction. Should a missile from other sources cause damage to the main steam system downstream of the main steam non-return valves, the valves would be closed and the plant would be shutdown. Should the missile damage occur upstream of the valve or at the valve itself, the plant can be safely shut down.

In summary, in considering the main steam system, it is our j

judgment that this system will be capable of performing its design function considering internally generated missiles as described above.

J.

Feedwater and Condensate Systems (portions of) l The main feedwater system consists of three motor driven pumps which pump water to the four steam generators. Condensate f rom l

the hot well is pumped by three motor-driven condensate pumps ~

through the air ejectors and gland seal condensers and then through several stages of preheating before going to the motor-driven boiler feedwater pumps. The area of concern for this system is the piping from the feed pumps on the ground floor level of the turbine building, outside along the turbine building wall, and into the vapor container.

Each steam generator is prevented from blowing down by a check valve in the feed piping inside the vapor container.

If any of the remaining portions of these systems are damaged by missiles, the plant can be safely shutdown using any one steam generator, and the emergency feedwater system through the alternate path in the primary auxiliary building.

k.

Emergency Feedwater System The emergency feedwater system consists of three 100% capacity pumps, two motor-driven and one turbine-driven, all of which are capable of feeding all four steam generators through two different flow paths. The normal feed path utilizes the main feed piping in the turbine building. The alternate feed path utilizes the steam generator blowdown piping in the primary auxiliary building. The two paths join together in the feed piping inside the loop compartments in the vapor container. The pumps are similarly separated.

The two motor-driven pumps are located on the ground floor of the primary auxiliary building. The turbine driven pump is located in the auxiliary boiler room on the ground floor of the turbine building. These pumps and flow paths are separated by close to 200 feet. Based on the system redundancy and physical separation, it is inconceivable that internally generated missiles from a single source could prevent the system from performing its intended function.

1.

Ventilation System for Areas Such as the Control Room and Diesel Generator Building

1) Diesel Generator Building The diesel generator building contains the safety injection equipment and switchgear, and also separate diesel rooms. The ventilation equipment is thermostatically controlled and starts at 95 F.

It blows in outside air and air in the building exhausts through wall dampers. The diesel cubicles are ventilated by the diesel cooling fans which are driven directly off the diesels.

2) The Control Room The control room is independently air conditioned by two completely separate units. Air is recirculated and fresh air is added to create a slight positive pressure in the control room. Both fresh air and recirculated air is filtered prior to admittance to the control room.

Our review did not reveal any potential missiles being generated by the ventilation system for either the diesel generator building or the control room. While duct work can be penetrated by miselles, the total cooling capability is not lost for either area, and time is available f or action to restore adequate ventilation.

In summary, in considering the ventilation systems for the two areas, it is our judgment that these systems will be capable of performing their design function considering internally generated missiles.

m.

Condensate Storage System The condensate storage system consists of a 30,000 gallon tank -. -

E f or storage of makeup water for the main feedwater system and as a source of auxiliary feedwater.

By virtue of the tank's location in the station yard (not in a building), we conclude that the tank is not subject to internally generated missiles nor is it a source of missiles.

2.

Syst, ems whose failure may result in release of unacceptable amounts of radioactivity:

a.

Spent Fuel Cooling System The spent fuel cooling system is a closed loop system consisting of two full capacity pumps, one full capacity heat exchanger, piping valves and instrumentation. Heat from the spent fuel pool is transferred by means of the above heat exchangers to the component cooling water system (see Section e, " Component Cooling Water System").

The equipment is in the spent fuel pool building.

The spent fuel cooling system is a low energy system unlikely to generate missiles or be impacted by them because it is located in a separate building. Should the equipment become inoperable due to that missile damage, there is sufficient time to effect repair or arrange for alternate cooling. In our judgment, this system will be capable of performing its function considering internally generated missiles as desired above.

t b.

Sampling System Reactor coolant system fluid samples are passed through a cooler and pressure reducing coil before entering the sample sink.

Grab samples are taken for later analysis in the chemistry lab.

(

All of the reactor coolant sampling operations are performed in a room with concrete walls. The sampling equipment is in a separate room. The likelihood of missiles from any source caubing damage to the sampling system is considered highly

unlikely. We conclude the above system meets the requirements for design against missiles.

c.

Liquid Waste Processing System The liquid waste system is located in shielded areas of the waste disposal building which provides protection against internally generated missiles. The largest pumps are low flow and low head and, therefore, are not likely to generate missiles. Further, should a missile damage this system and if contents are drained, the resulting liquid release would be retained in the building long enough to allow a clean-up.

d.

Gaseous Waste Processing System The waste gas system is operated at pressures less than 100 psig. The system is not likely to be a source of internally generated missiles. All of the compressors and storage tanks are behind shield walls and include shield walls between clusters of equipment.

Our review of this system verified that this system was protected from internally generated missiles from outside sources and the potential for internally generated missiles from the system itself was small. Further, missile damage to the system will not affect the safe shutdown of the facility.

e.

Containment Purge System The containment purge system is provided to periodically purge the containment prior to entry. The system consists of a fan, filters, duct work and isolation valves. All of these components except portions of the duct work and the isolation valves are located in a common room in the auxiliary building.

The system is only operated with the reactor shut down.

It follows that this system could only be considered a potential missile producer at that time. The probability of this system producing missiles or being damaged by a missile is very low.

Should a missile damage this system, there would be ample time to perform repairs. The missile protection provided for the system is acceptable.

3.

Electrical systems which are necessary to support those fluid systems needed to perform safety functions:

a.

Diesel Generator See Section V.1.h.

b.

Station Batteries Battery I and 2 auxiliary systems, such as the chargers and buses, are located outside each respective battery room.

Battery 3 and its auxiliary systems are fenced in a common area within the diesel generator building. Due to the physical separation of the batteries in two different buildings, there is no single missile source that would damage all three battery sets.

The physical separation of the station batteries provides protection from internally generated missiles.

c.

480 V Switchgear The 480 V emergency switchgear ic located in the safety injection building, with motor control centers in the switchgear room and a building addition on the diesel generator building.

The major piping in the safety injection building is the ECCS system (see Section b)..-

We conclude that the location of the 480 V switchgear provides adequate protection from internally generated missile sources.

4.

Additional areas needed to support safety system functions are:

a.

Control Room The control room is located on the operating floor level of the turbine building. The major source of missiles is from the turbine generator.

b.

Switchgear Room The switchgear room is located directly below the control room.

This room contains the 2400 V switchgear, the 480 V switchgear, two of the three battery rooms (see Section b, " Station Batteries"), battery chargers, de switchgear, station batteries 1 and 2, vital bus power supply, vital bus switchgear, 480 V switchgear, emergency MCC 1, 2400 V switchgear and control rod equipment panel. The cables are routed through this room on their way to the control room. The switchgear room does not contain any piping or other pressurized sources or rotating equipment which might produce missiles.

l We conclude that there are no potential missile sources in this area which could affect safety-related equipment.

c.

Cable Tray Room The cable tray room is located directly above the control room and is constructed of concrete block. The cables are ruated through this room on their way to the control room. The cable tray room does not contain any piping or other pressurized i

l sources or rotating equipment which might produce missiles. The l

fire protection system in this room consists of a water sprinkler system and two extinguishers at each entrance. There are no potential missile sources in this room.,

l

4 85 VI.

CONCLUSIONS From our view of the systems and components needed to perform safety functions, we conclude that the design of protection from Internally

~

generated missiles meets the intent of the criteria listed in Section II, " Review Criteria", and are, therefore, considered to be acceptable.

l i _

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.w.

_m. _. _ _ _ _.

C Docket No. 50-29 SEP TOPIC III-5.A REPORT ON EFF2 CTS OF HIGH ENERCY PIPING SYSTEM BREAKS INSIDE THE VAPOR CONTAINER AT YANKEE ATOMIC POWER STATION j

ROWE, MASSACHUSETTS l

t i

)

i

- - - - - ~ _ -

0 TABLE OF CONTENTS M

I I.

I N TRO D UCT I O N.....................................................

II.

REVIEW CRITERIA..................................................

I III. RELATED SAFETY TOPICS AND INTERFACES.............................

I IV.

REVIEW GUIDELINES................................................

2 3

V.

D I S CU S S IO N.......................................................

3 A.

B a c k g r o u nd...................................................

3 B.

Analysis Assumptions.........................................

4 C.

Sa f e t y-Re l a t ed Eq u i pme nt.....................................

5 VI.

E VA LU AT IO N.......................................................

5 A.

Ap p ro a ch a nd C r i t e r i a........................................

B.

Interaction Studies..........................................

6 31 VII.

CONCLUSIONS.......................................................-.

I.

INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.A " Effects of Pipe Break on Structures, Systems and Components Inside Containment", is to assure that pipe breaks would not cause the loss of needed function of " safety-related" systems, structures and components, and to assure that the plant can be safely shut down in the event of a break. The needed functions of " safety-related" systems are those functions required to mitigate the effects of the pipe break and safely shut down the reactor plant.

II.

REVIEW CRITERIA The current criteria for review of pipe breaks inside containment are contained in Standard Review Plan 3.6.2, " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping", including its attached Branch Technical Position, Mechanical Engineering Branch 3-1 (BTP MEB 3-1).

III.

RELATED SAFETY TOPICS AND INTERFACES 1.

This review complements that of SEP Topic VII-3, " Systems Required for Safe Shutdown".

2.

The environmental effects of pressure, temperature, humidity and flooding due to postulated pipe breaks are evaluated under SEP Topic 111-12. " Environmental Qualification of Safety-Related Electrical Equipment".

3.

.The effects of potential missiles generated by fluid system ruptures and rotating machinery are evaluated under SEP Topic III-4.C, " Internally Generated Missiles".

4.

The effecta of containment pressurization are addressed under SEP Topic VI-2.D. " Mass and Energy Release for Possible Pipe Break Inside Containment".

5.

The original plant design criteria in the areas of seismic input and analysis design criteria are evaluated under SEP Topic III-6,

" Seismic Design Consideration".

6.

The effects of steam line breaks on core reactivity, primary cooldown and containment pressure response are addressed under SEP Topic XV-2, " Spectrum of Steam System Piping Failures Inside and Outside Containment (PWR)".

7.

The effects of feedwater line breaks on the main coolant system and secondary system pressurization, and core integrity are addressed under SEP Topic XV-6, "Feedwater System Pipe Breaks Inside and Outside Containment".

IV.

REVIEW CUIDELINES The SEP Branch sent a letter to Yankee Atomic Electric Company requesting an analysis of the effects of postulated pipe breaks on structures, systems and components inside containment. In that letter, the staff included a position that stated three approaches were appropriate for postulating breaks in high energy piping systems (defined as P;t275 psig or Tp_200F). The approaches are:

1.

Mechanistic 2.

Simplified Mechanistic 3.

Effects-Oriented The staff further stated that combinations of the three approaches could be utilized if justified. The details of those three approaches are described in the letter from the NRC to KMC Incorporated dated July 20, 1978.

V.

DISCUSSION A.

Background

In a letter dated May 31, 1979 (WYR 79-64), YAEC submitted a list of high energy lines inside containment. Representatives of the NRC and YAEC staff met at the Rowe site on September 17 and 18, 1979, to discuss the wo-k done by the licensee on this topic. As a result of this meeting, the licensee informally submitted, in January 1980, a report on the effects of breaks in these lines on safety-related equipment. This review utilized the effects-oriented approach for the high energy line breaks analyzed. In this approach, breaks were postulated at any location along the line, and were chosen to produce the greatest jet impingement or pipe whip loadings on essential equipment.

The purpose of this report is to summarize and document Yankee Atomic Electric Company's evaluation of the effects of high energy pipe breaks inside the plant's vapor container (V.C.).

B.

Analysis Assumptions The following assumptions were made by Yankee Atomic:

1.

High energy fluid systems are systems with operating temperature

>200 F or operating pressure )275 psig. In accordance with Branch Technical Position (BTP) MEB 3-1, breaks are not postulated in piping of systems that qualify as high energy systems for only l

short operational periods (i.e., less than 2% of the time the l

system operates as a moderate energy system, or less than 1% of the i

time that the plant operates).

2.

The worst, unrelated, single active failure occurs simultaneously with the pipe break. Unrelated passive failures are not considered in the short-term.

3.

A simultaneous, unrelated, pipe failure is not postulated with the high energy pipe break.

4.

The effects of pipe whip cg jet impingement will not damage equal diameter or larger piping which has equal or greater wall thickness.

5.

The piping boundary is taken as the first normally closed valve, check valve, relief / safety valve or first valve capable of remote or automatic closure.

6.

Effects of pipe whip and jet impingement from rupture of piping 1" nominal pipe size and smaller are not required to be analyzed.

7.

The ef fects of pipe whip or impingement from low energy pipe systems are not required.

C.

Safety-Related Equipment Safety-related equipment includes systems needed to mitigate the effects of the line breaks and to bring the reactor to safe shutdown.

Of first concern when evaluating the effects of high energy pipe breaks is to identify the lines, equipment, and instrumentation in the V.C. required to safely shut down the plant. These systems and equipment perform the following functions:

1.

Insert negative reactivity into the reactor core, 2.

Maintain reactor coolant system (RCS) and/or secondary side water inventory, 3.

, Control RCS overpressure, and 4.

Remove decay heat and control cooldown of the RCS.

Where possible, Yankee has made use of previously docketed information, as well as information developed in the NRC's "SEP Review of Safe Shutdown Systems for the Yankee Rowe Nuclear Power Plant" document. Page B-12 of that document lists the minimum required components and systems required for safe shutdown coincident with a loss of total off-site power and the most limiting single failure without a design basis event (DBE). This list is as follows:

1.

Main Steam Safety Valves (MSSV) and main steam piping inside V.C.

2.

Atmospheric Dump Valves (ADV's).

3.

Emergency Feed Pumps (EFP's) and feedwater piping inside V.C.

4_

Demineralized Water Storage Tank (DWST) and Primary Water Storage Tank (PWST).

5.

Shutdown Cooling System (SCS) and piping inside

..C.

6.

Component Cooling System (CCS).

7.

Service Water System (SWS).

8.

Emergency Power System.

9.

125 V DC Power System.

10.

Chemical and Volume Control System (CVCS) and piping inside V.C.

11.

Pressure Control and Relief System.

l 12.

Instrumentation for Shutdown and Cooldown.

VI.

EVALUATION l

l l

A.

Approach and Criteria i

Using an effects-oriented approach, the piping and equipment required for safe shutdown were surveyed in accordance with the Assumptions of Item V-B to determine if the breaking of any primary or secondary high energy lines would prevent safe shutdown of the plant. Special consideration was given to.

all piping which is not part of the RCS, which through either pipe whip or jet impingement could cause small RCS breaks, to insure a safe shutdown would not be prevent ed.

For each of the postulated break locations, Yankee evaluated the effects on the needed equipment. In addition, the effects on other impacted equipment were considered to ensure that failure of such equipment would not affect the plant's ability to safely shut down. A high energy line is assumed to break impacted lines which are smaller or equal size lines with thinner walls.

If this impacted line is also a high energy line, the potential dynamic effects of that break were concurrently considered.

B.

Interaction Studies 1.

Identification of High Energy Systems The following systems have been classified as high energy because they operate at a temperature of p_200 F ard/or pressure of gb275 psig for more than:

a.

Two percent of the time that the system operates as a mod erat e-energy fluid.

b.

One percent of the time that the plant operates.

High Energy Systems a.

Main Steam b.

Fe ed wat er c.

Steam Generator Blowdown (SCBD) d.

Charging and Volume Control System (CVCS) e.

Steam Generator Instrumentation O

f.

RCS - including interconnected systems up to first normally closed valve, check valve, safety valve, relief valve, and remote or automatic isolation valve. These are:

1) RCS Vents and Drains

2) Shutdown Cooling System (SCS)

3) Safety Injection System (SIS)

4) Normal Charging and Letdown of the CVCS

5) Pressure Control and Relief 2.

Identification of High Energy Lines a.

The high energy lines that will be considered in this evaluation are listed below.

Primary Systems Size /No.

Loading Designation 20"/4 RCS Loops 1,2,3,4 RCS Hot Legs 24"/4 RCS Loops 1,2,3,4 RCS Pump Suetions l

l 20"/4 RCS Loops 1,2,3,4 RCS Cold Legs 6"/2 RCS Loop 4 SCS Loop Connections 5"/4 RCS Loops 1,2,3,4 RCS Crossovers 4"/1 RCS Loop I to Pressurizer Surge Line Pressurizer Cubicle l

l 2"/2 Pressurizer Cubicle Safety Valve Inlet 2"/1 Pressurizer Cubicle Solenoid Relief Valve Inlet 4"/4 RCS Loops 1,2,3,4 Safety Injection 2"/1 RCS Loop 4 Normal Charging l

2"/1 RCS Loop 1 Letdown 1-1/2"/4 RCS Loops 1,2,3,4 RCS Hot Leg Drains 1-1/2"/4 RCS Loops 1,2,3,4 RCS Cold Leg Drains 1-1/4"/1 RCS Loop 2 to Pressurizer Spray Pressuri::: Cubicle Secondary Systems 14"/4 RCS Loops 1,2,3,4 Main Steam 8"/4 RCS Loops 1,2,3,4 Feedwa t er 2"/4 RCS Loops 1,2,3,4 SCBD 4"/4 V.C. Broadway V.C. Heater Steam Supply 3"/4 V.C. Broadway V.C. Heater Condensate Return b.

Righ energy lines not considered in this evaluation because their pipe size is 1" or less:

Size /No.

Location Designation 1"/4 RCS Loops 1,2,3,4 Loop Safety Valves 1"/1 Pressurizer Cubicle Pressurizer Drain 3/4"/4 RCS Loops 1,2,3,4 RCS Vents 3/4"/1 Pressurizer Cubicle Pressurizer Vents 3/4"/2 Charging / Letdown Heat Charging / Letdown Heat Exchanger Cubicle Exchanger Vents 1/2"/24 RCS Loops 1,2,3,4 Steam Generator Level Connections 1/2"/9 RCS Loops / Pressurizer RCS Instrument Lines Cubicle 1"/1 Shield Tank Cavity Reactor Head Vent 1"/1 Below RCS Loops Alternate Charging /

Pressurizer Spray - - _ _

c.

High energy lines not considered in this evaluation because they do not function above the threshold temperature and pressure values specified for high energy fluid systems for more than:

1) Two percent of the time that the system operates as a moderate-energy fluid.

2) One percent of the time that the plant operates.

1. ize/No.

Location Designation 6"/l Presurizer Cubicle Discharge Header for the Pressurizer Relief and Safety Valves 6"/l RCS Loop 4 Shutdown Cooling Suction 6"/1 RCS Loop 4 Shutdown Cooling Return 4"/2 Pressurizer Cubicle Discharge Line for Pressurizer Safety Valve 3"/l Pressurizer Cubicle Discharge Line for Pressurizer Solenoid Relief Valve 3.

Primary System Breaks a.

Large RCS Piping (>20")

Using the ef fects-oriented approach, large RCS piping breaks that are contained within a single loop compartment do not prevent the safe shutdown of the plant. However, preliminary analysis has shown that the potential exists for generating secondary missiles in the adjacent loop due to the impact forces from pipe whip into the loop compartment divider walls. These interactions are unacceptable.

-9_

b.

Small RCS Piping (<20" but >1")

A break in the 5" RCS crossover lines in any loop in the piping to the cold leg could possibly whip into and break the safety injection connection to the same loop. Safe shutdown is accomplished by safety injection into the three intact loops, as described in the existing safety analysis.

Each safety injection line has a check valve located in close proximity to the RCS loop connection. A break in the safety injection piping on the safety injection side of the check will not result in a primary leak. A break in the RCS side of the check will result in a primary leak through the 2.25" thermal sleeve.

This break has been analyzed and is acceptable.

Breaks in the remaining small RCS lines have also been evaluated.

Safe shutdown is accomplished in these cases by safety injection to the other three intact loops, as described in the existing safety analysis. The worst single failure for these breaks would be the loss of one emergency diesel generator resulting in the loss of one safety injection train. The present analysis reflects this worst case single failure.

4.

Secondary Systems a.

Main Steam Piping Any additional pipe restraints or supports installed as a result of SEP Topic III-6, Seismic Design Considerations, may significantly l

change any whipping or jet impingement effects.

The following general concerns apply to all four loops:

l

1) The effects of thrust forces acting on a steam generator resulting from a break in a main steam line elbow at elevation 1124' where the line leaves the top of the steam generator requires a more detailed evaluation.

2) Damage te the V.C. due to pipe whip or jet impingement may result from a break in either of the main steam line elbows at elevatten 1124' where each line leaves the top of its steam generator. Pipe whip resulting from a break in the elbow outside the loop compartment has been evaluated to cause the most severe damage to the V.C.

However, a safe shutdown would not be prevented as long as the consequences associated with this break only result in damage to: the V.C., cable in the vicinity of the whipping main steam line, and the main feed line of the affected steam generator. In the event containment integrity is violated, the activity released to the environment would be functions of airborne activity levels inside the V.C.,

primary coolant activity, and primary to secondary leakage.

These releases are judged to be less than 10CFR Part 100 limits even if the power-operated relief valve lif ts.

Based on the above discussion, no further analysis is believed necessary.

3) At the elevation where there is a loop in the main steam line for thermal expansion, a break at the elbow at either end of the upper leg could, as a result of whipping action, fail the feedwater supply line to the same steam generator. The primary system cooldown and containment pressure responses would be the same as a main steam line break discussed in SEP Topic XV-2 submitted by FYR 81-95 on June 30, 1981. Based on the above discussion, no further analysis is believed necessary.

4) Jet impingement on the V.C. resulting from a break at the weld where each main steam line penetrates containment could violate containment integrity. Based on the consequences discussed in (2) above, no further analysis is believed necessary.

YAEC herein describes what it believes to be the worst case accident and then shows why the other cases are less severe. The consequences of secondary system breaks are similar, whether they be in a main steam line, feedwater line, or SCBD line. Where dif ferences occur, they are limited to the rate of cooldown experienced by the main coolant, how quickly the transient affects other plant systems, and the speed of operator actions..

l i

required. A rupture in the main steam line results in the est rapid cooldown transient, and therefore, would be the worst case accident. YAEC finds no basis for the assumption that a plant transient such as a main steam line break would cause a loss of off-site power. The loss of electrical generation from a small plant the size of Yankee has minimal effect on the New England electrical grid. Yankee trips have never caused a loss of off-site power.

Nonetheless, a concurrent loss of off-site power is assumed.

The worst active single failure is considered to be a loss of the EFP or steam supply to this pump. Two new motor-driven emergency feedwater pumps are available to provide makeup to the steam generators. The power feed to these pumps would have to be manually realigned to the emergency bus after stripping of f Si loads.

Main Steam Line From #3 Steam Generator A break in the main steam line associated with #3 steam generator is considered to be the worst case break. A break can be postulated that would cause jet impingement on instrument cable routed to or near V.C. penetration blister SE.

These instrument cables provide information concerning all steam

nerator level indications, all pressurizer level indications, six RCS pressure indications, and temperature indication for RCS Loops 3 and 4.

All of these instrument cables are assumed lost. In addition, power supply cables to all four SCS valves are assumed lost.

The following describes the events which occur:

1.

The main steam line from #3 steam generator ruptures causing the following:

Blowdown of #3 steam generator causes the RCS to cool down and a.

contract.

b.

Cable routed to or near V.C. blister SE is lost due to jet impingement.

c.

As a result of RCS depressurization or the loss of pressure signals, the reactor control rods insert themselves automatically (SCRAM), turbine trips, and safety injection is ac t ua t ed. Low steam generator pressure also could scram the reactor.

d.

The non-return valve in the broken line will close as a result of reverse steam flow. Closure of this valve will isolate the 3 intact steam generators f rom the break, and prevent their further blowdown. The non-return valves on the other steam generators will automatically close on low steam generator pressure.

2.

A loss of off-site power occurs.

3.

The emergency diesel generators start as a consequence of losing off-site power.

4.

ECCS injects borated water into the RCS when safety injection is ini tia t ed.

5.

V.C. isolation occurs when containment pressure reaches 5 psig.

Initiating Signals (See section on instrument cable routing)

A reactor scram is initiated as a result of any of the following conditions.

1.

Low main coolant pressure (2 of 3 channels).

2.

Failure of the main coolant low or high pressure signals (2 of 3 channels).

3.

Low main coolant flow above a power level of 15 MWe (2 of 4 ct..nnels).

4.

Turbine trip above a power level of 15 MWe..

Q O

5.

Steam generator low water level above a power level of 15 MWe (2 of 4 channels).

6.

High rate of change in reactor power level during start-up (below 15 MWe).

7.

High reactor power level.

8.

Low steam line pressure (2 of 3 per loop).

Safety Injection Actuation Signal (SIAS) is generated as a result of:

1.

Low pressure by the main coolant pressure detector in RCS Loop 1 or a pressure switch connected to RCS Loops 2 or 3.

2.

Loss of M.C. pressure signal (Loop 1 only).

3.

V.C. pressure 25 psig. (Redundant detectors are located outside V.C.)

Containment Isolation Signal (CIS) is generated as a result of:

1.

V.C. pressure 15 psig. (Redundant detectors are located outside V.C.)

A safe shutdown of the plant is accomplished by:

1.

Inserting negative reactivity into the reactor core.

2.

Removing decay heat.

3.

Maintaining RCS and steam generator water inventory.

4.

Controlling RCS overpressure.

4 - _.

In order to accomplish the above, the f ollowing actions would be required:

1.

Negative reactivity is initially inserted into the reactor core by the control rod s.

These control rods f ail-safe. They fall into the core when power is removed from their holding coils. This may be accomplished by pushing any of the three trip buttons on the main control board, by opening the scram breakers in the switchgear room, or by any of the initiating scram signals described above.

The reactor would remain subcritical af ter the scram even if the highest worth rod f ailed to drop into the core. Addi tional negative reactivity is added to the reactor core ef ter a scram as a result of the production of Xenon, which is a neutron absorber, and ECCS injection of borated water.

Further injection of boron could be accomplished using the charging pumps when the transient has stabilized and only one safety injection train was required. Since of f-site power is assumed unavailable, this would require that the charging pumps be manually connected to an emergency bus. The boron injection flow path would be through either the normal charging line or the separate, r ed und ant, charging line to the loop fill header.

2.

Initially, the cooldown of the primary system is a function of the

~

steam line break area as well as other plant conditions. This mechanism, along with the 3 unaffected steam generators until closure of the non-return valve, provides the initial decay heat removal. Depending on the rapidity of the blowdown and the resultant cooldown of the primary system, the 3 unaffected steam generators will either be a means of removing decay heat or will, for a short time, be a heat source until primary system temperature is greater than steam generator temperature.

In either case, the intact steam generators will absorb and remove decay heat after the main coolant system and steam generators heat up following the initial cooldown..

3.

Water injected by the 3 trains of ECCS pumps refills the main coolant system. When the ECCS restores pressure to approximately 1500 psig (T

= 596 F), pump shutoff head will be exceeded SAT and injection will cease. At this time, the main coolant system will have the required overpressure to insure natural circulation.

The loads on an emergency diesel generator can be shed so that one of the two full capacity, motor-driven emergency feedwater pumps can be used to supply makeup, assuming the loss of the steam-driven EFP rather than an emergency diesel generator. The large water reserve in the steam generators insures sufficient time is available to operate the required valves. Feed flow would be regulated by using throttle valves and associated flow instrumentation. The maximum flow is more than sufficient to meet the feedwater flow requirements.

If secondary cooling is available, the plant would be shut down using methods described in the NRC's Safe Shutdown Report. The atmospheric dump valves will be operated to cool down the plant.

Once of f-site power is available, a more normal cooldown using the condensate system and undamaged steam dump system would be used.

In either case, cooldown of the main coolant to 330 F and 300 psig could be accomplished.

4.

At this point, the shutdown cooling system would be placed in service using either off-site power or emergency diesels supplying the required service water, CCW, and SCS pumps. The emergency diesels may be used since decreased decay heat removal requirements allow most diesel powered equipment to be secured (i.e., two trains of ECCS pumps). Waiting until off-site power can be restored is preferred; however, this is not required.

If any of the motor-operated valves in the shutdown cooling system failed to open, decay heat removal would continue via the steam generators until the V.C. environment allowed operator entry.

l The operator will recognize very soon af ter the transient that he has had either a major LOCA, major steam line break or feedwater line break inside the V.C.

Key information provided the operator comes from safety injection initiation, V.C. pressure, V.C. isolation, steam generator pressure, non-return valve position and loss of instrumentation.

Certain parameters will be different when comparing large break LOCA's and large steam or feed line breaks. All the steam generator pressure transmitters are located outside the V.C.,

indication is local and instrument readouts are available in the control room. A large break LOCA will not disable the same instrumentation postulated for tain steam or feed line breaks. Also, a large break LOCA will result in significant ECCS flow which will remain relatively constant until operator action is required, by procedure.

In contrast, although much more plant instrumentation could be lost as a result of a large main steam or feed line break the required ECCS flow is significantly less and the pressure in the affected steam generator is lost while the pressure in the other three is not. After only a few minutes, ECCS flow will be reduced to zero as RCS pressure approaches the shutoff head of the pumps. A low ECCS flow will occur later if either the pressurizer solenoid relief valve operates or letdown flow is re-established.

Therefore, by carefully observing ECCS flow rates and steam generator pressure, the operator can distinguish between a large LOCA, and a main steam or feed line break, as well as the af fected steam generator. In a large break LOCA there would be no attempt to feed the steam generators.

There are other means available to the operator that would allow him to determine if there had been a main steam or feed line break associated with #3 steam generator. Each of the main steam lines has three steam generator pressure switches which provides low pressure trip indication and alarms in the control room, enabling the operator to determine the affected generator.

Also, flow indication is provided in the emergency feedwater lines to each generator. The above instrumentation is located outside the vapor container and is not af fected by any inside break. Temperature indication would be available from RCS Loops 1 and 2; the operator would makeup to the unaffected steam generators. ;

)

Based on the above, safe shut down of the plant can be accomplished in the event of a break in the main steam line from #3 steam generator, coupled j

with a loss of off-site power, and a loss of the EFP along with various primary system instrumentation. Nonetheless, Yankee realizes the importance of providing the operators with as much information as possible. As part of the environmental qualification program, YAEC is presently planning to replace existing electrical penetrations in the blisters with qualified ones. Also, a qualified set of steam generator wide-range level indication channels will be add ed. As part of these modificatons, the routing of existing instrumentation cable and the new instrument cable to be added has been reviewed and plans formulated to split these instruments between blisters SE and 7E.

This will greatly reduce the instrumentation which must be assumed lost for a steam line break on #3 steam generator. As discussed above, the plant can be safely shut down with the present arrangement. However, the changes being implemented will enhance this capability.

Other main steam line breaks have been evaluated and judged to be less severe than the worst case just described. However, these breaks and the resultant damage from the break will now be discussed.

Main Steam Line From #1 Steam Generator A break at the 180 pipe bend in the thermal expansion loop or in the vertical pipe runs above and below " broadway" is assumed to damage cable and c ond ui t located just above and below " broadway". Broadway is a walkway at l

approximately elevation 1104' around the outside of the biological shield wall. This would disable RCS Loop 1 pressure and temperature indication, LTOP pressure and pressurizer level indication, il steam generator level indication j

and the power supply cable to the following valves, none of which need to be used for saf e shutdown.

CH-MOV-525 - Letdown CS-MOV-535 - Chemical Shutdown PR-MOV-512 - Pressurizer Power Operated Relief Valve (PORV) Block Valve PR-SOV Pressurizer PORV l,

This break leaves intact the unaffected steam generator level indicators and several pressure indicators. Therefore, the operator has more than sufficient information to determine the affected steam generator and accomplish a safe shutdown as previously described.

Main Steam Line From #2 Steam Generator A break at the 180 pipe bend in the thermal expansion loop or in the vertical pipe runs above and below " broadway" is assumed to damage cable and conduit located just above and below " broadway". This would disable RCS Loop 2 pressure and temperature indication, RCS Loop I temperature indication, all pressurizer level and pressure indication, level indication for both #1 and #2 steam generators and the power supply cables to the following valves, none of which need be used for safe shutdown:

CH-MOV-525 - Letdown CS-MOV-535 - Chemical Shutdown PR-MOV-512 - Pressurizer PORV Block Valve PR-SOV Pressurizer PORV This break leaves intact steam generator level indication for #3 and #4 steam generators. The operator can determine which steam generator is affected and supply the others with water. Knowing which of the four (4) steam generators is affected, the operator can accomplish a safe shutdown as previously desc ri bed.

l l

In the event there is a main steam line break in the vicinity of V.C.

l l

blister 12E, there is the potential for jet impingement to disable the power

(

supply cables to all four RCS pumps simultaneously. The simultaneous loss of all 4 pumps in conjunction with a main steam line break has not been analyzed. Further evaluation is required.

Main Steam Line From #4 Steam Generator A break at the 180 pipe bend in the thermal expansion loop or in the vertical pipe runs above and below " broadway" is assumed to damage cable and conduit located just above and below " broadway". This is assumed to disable l

[

l

Loop 1 and LTOP pressure indication, RCS Loop 4 temperature indication, level indication for #4 steam generator and the power cable to the SCS valves SC-MOV-551, 552, 553 and 554. Since these valves are located inside the loop compartment, they would not be directly affected by a main steam line break and could be manually opened when conditions inside the V.C. permitted entry.

This break leaves intact steam generator level indication for #1, #2 and #3 steam generators, pressurizer pressure and level indications, several pressure indicators and RCS Loop 1, 2 and 3 temperature indications.

Therefore, the operator has more than sufficient information to determine the affected steam generator and accomplish a safe shutdown as previously d e sc ri bed.

b.

Main Feedwater Piping Described below are several break locations for each main feed line where the consequences from either pipe whip or jet impingement associated with the postulated break at those locations would be the same for each of the four main feed lines. Any additional pipe restraints or supports installed as a result of SEP Topic III-6, Seismic Design Considerations, may significantly change any whipping or jet impingement effects.

1.

Pipe whip and jet impingement resulting fro.o a break in any of the main feed piping either inside the loop compartments or at elevation 1121' where the lines enter the loop compartments could result in damage to the associated steam generator. The effects, if any, are believed to be less severe than those associated with a break at either elbow in the main steam line at elevation 1124'.

2.

Pipe whip and jet impingement resulting from a break in either of the main feed line elbows at elevation 1121' where the line enters the loop compartment could result in damage to the V.C.

The effects, if any, are believed to be less severe than those associated with a break at either elbow in the main steam line at elevation 1124'.

3.

Jet impingement on the V.C. resulting from pipe breaks on either side of the main feed check valves or where each main feed line penetrates containment could violate containment integrity. The effects, if any, are believed to be less severe than those associated with a similar break in the main steam line.

4.

Postulated breaks in any of the 8" sain feed lines would whip into or jet impinge upon the 14" sain steam line associated with the same steam generator. Yankee does not believe this will result in damage to any of the main steam lines because of the line sizes involved.

The damage resulting from other main feed line breaks associated with a specific steam generator is now discussed.

Main Feed Line to #1 Steam Generator i

A break in the main feed line vertical pipe runs above or below broadway is assumed to damage cable and conduit located just above and below broadway. This would disable RCS Loop 1 pressure and temperature indication, l

LTOP pressure indication, pressurizer level indication, #1 steam generator level indication, and two valves neither of which would prevent the plant from completing a safe shutdown:

Cil-MOV-525 - Letdown CS-MOV-535 - Chemical Shutdown This break is less severe than a main steam line break from the same steam generator. There is more than sufficient instrumentation available from which the operator can monitor plant conditions while completing shutdown.

Main Feed Line to #2 Steam Generator A break in the main feed line vertical pipe runs above or below broadway is assumed to damage cable and conduit located just beneath broadway. This would disable RCS Loop 2 temperature and pressure indication, -

RCS Loop I temperature indication, pressurizer pressure and level indication, il and #2 steam generator level indication, and the following valves none of which would prevent the plant from completing a safe shutdown:

CH-MOV-325 - Letdown CS-MOV-535 - Chemical Shutdown PR-MOV-512 - Pressurizer PORV Block Valve PR-50V Pressurizer PORV i

This break is less severe than a main steam line break from the same steam generator. There is more than sufficient information available from the instrumentation associated with the intact Loops 3 and 4 to enable the operator to monitor plant conditions while completing shutdown.

Main Feed Line to #3 Steam Generator A break in the feed line vertical pipe runs above and below broadway is assumed to damage cable and conduit running to blister SE as well as the blister itself. This is assumed to disable RCS Loop 4 tecperature indication, all RCS pressure indications, all pressurizer level indications, all four steam generadbr level indications, intermediate and power range indications and the power supply cables to all four SCS valves, which are:

SC-MOV-551 - Shutdown Cooling Return SC-MOV-553 - Shutdown Cooling Return SC-MOV-552 - Shutdown Cooling Suction SC-MOV-554 - Shutdown Cooling Suction This break is less severe than a main steam line break associated with the same steam generator, which in this case was evaluated to be the worst case bre ak.

Although Yankee has assumed the loss of all instrumentation as well as power supply cable to the shutdown cooling motor-operated valves, the actual equipment lost is highly dependent on break location. However, in the event all equipment is lost, plant would be shutdown as described for the worst case bre ak.

Main Feed Line to f 4 Steam Generator A break in the feed line vertical pipe runs above and below broadway is assumed to damage cable and conduit located just beneath broadway. This would disable RCS Loop 4 temperature indication, RCS Loop 1 and LTOP pressure indication, level indication for #4 steam generator and power supply cabits to the motor operators for the following valves:

SC-MOV-551 - Shutdown Cooling Return SCHMOV-553 - Shutdown Cooling Return SC-MOV-552 - Shutdown Cooling Suction SC-MOV-554 - Shutdown Cooling Suction This break is less severe than a main steam Ifne break associated with the same steam generator. There is more than sufficient instrumentation available from which the operator can monitor plant conditions while completing shutd own.

Since the shutdown cooling valves are located in the loop compa rtment, they would not be affected by any main feed line break outside the loop. When required, they could be manually opened, provided conditions inside the V.C. permitted entry.

C.

Steam Generator Blowdown Lines Various pipe breaks in the SGBD system have been evaluated and are discussed below by RCS loop compartme. t areas.

l RCS Loop 1 Compartment j

The SCBD lines from #1, #2 and #3 steam generators run horizontally one 1

above the other in close proximity to one another. A break in any one line wil? not damage the others since they are all the same size and pipe schedule.

Depending upon the location chosen, a break in the SGBD line from #2 steam generator near the steam generator could, as a result of pipe whip or jet impingement, damage the 1/2" steam generator level instrumentation lines, 1/2" sain ecolant piping used to mersure RCS flow or the 1/2" instrument line to the LTOP pressure transmitter. A break in any of the straight horizontal SCBD piping associated with #2 or #3 steam generators could, as a result of jet impingement, damage the 1/2" sain coolant piping used to measure RCS flow. Given the sizes of these instrument lines, any damage sustained to either secondary or primary piping would not prevent a safe shutdown, as previously described.

A break in the SGBD line from #1 steam generator in the vicinity of RCS cold leg could jet impinge on the motor operator for letdown valve CHHMOV-525. Direct impingement is not believed possible because of reinforced concrete obstacles between the line and valve. In addition, the valve is not required to complete a safe shutdown.

Jet impingement resulting from a break in any of the 3 SGBD lines in the vicinity of the RCS Loop I hot leg could damage piping used to drain, vent or protect from overpressurization the loop RCS piping. Any damage sustained would not create serious problems because the loop drain, vent and safety valves are normally closed and located very close to the RCS loop piping. In any event, damage and/or a break between the valve and RCS loop piping in any one of these lines would constitute a small RCS break, which as previously discussed, would not prevent a safe shutdown.

The f ailure of a capped section of the SGBD line from #1 steam generator could, as a result of jet impingement, damage the motor operator for the loop bypass valve. Damage to the motor operator is of no consequence since operation of this valve is not required to complete a safe shutdown.

RCS Loop 2 Compartment The SGBD lines from #2 and #3 steam generators run horizontally one above the other in close proxinity to each other. YABC does not believe a break in either line will damage the other since they are both the same size and pipe schedule.

Depending upon the location chosen, a break in the SGBD line from #2 steam generator near the steam generator could, as a result of pipe whip or jet impingement, damage the 1/2" steam generator level instrumentation lines and the 1/2" sain coolant piping used to measure RCS flow. A break in either of the straight horizontal SGBD piping could, as a result of jet impingement, i

I dasage the 1/2" sain coolant piping used to measure RCS flow. Given the sizes of these instrument lines, any damage sustained to either secondary or primary piping would not prevent a safe shutdown as previously described.

Pipe whip or jet impingement resulting from a break in any of the #2 SGBD piping in the vicinity of the RCS loop hot leg could damage piping used to drain, vent or prc.tect from overpressurization the RCS loop oiping. Any damage sustained would not create serious problems because the loop drain, vent and safety valves are normally closed and located very close to the RCS loop piping. In any event, damage and/or a break between the valve and the RCS loop piping in any of these lines would constitute a small RCS break which, as previously discussed, would not prevent a safe shutdown.

A break in either of the SGBD lines in the pressurizer cubicle could, as a result of jet impingement, disable the pressurizer 1/2" level instrument lines. Given the size of these instrument lines, any damage sustained would not prevent a safe shutdown as previously described.

RCS Loop 3 Compartment Depending upon the location chosen, a break in the EGBD line from #3 steam generator near the steam generator could, as a result of pipe whip or jet impingement, damage the 1/2" steam generator level instrumentation lines or the 1/2" sain coolant piping used to measure RCS flow. Given the sizes of these instrument lines, any damage sustained to either secondary or primary piping would not prevent a safe shutdown as previously described.

Pipe whip or jet impingement resulting from a break in the SCBD line from f 3 steam generator in the vicinity of the RCS loop hot leg could damage piping used to drain, vent or protect from overpressurization the RCS loop piping. Any damage sustained would not create serious problems because the t

(

loop drain, vent and safety valves are normally closed and located very close l

to the RCS loop piping. In any event, damage and/or a break between the valve and the RCS loop piping would constitute a c=all RCS break which, _es previously discussed, would not prevent a safe shutdown.

j l

A break in the SGBD line from f 3 steam generator would, as a result of pipe whip, disable power cable to the motor operator for the RCS loop hot leg isolation valve. Operation of this valve is not required to complete a safe shutdown.

RCS Loop 4 Compartment A line connected to #4 steam generator that was previously used for SGBD and is now capped could, as a result of pipe whip or jet impingement, damage the 1/2" steam generator level instrumentation lines or the 1/2" main coolant piping used to measure RCS flow. Given the sizes of these instrument lines, any damage sustained to either secondary or primary piping would not prevent a safe shutdown as previously described.

The SGBD line from #4 steam generator could, as a result of pipe whip or jet impingement, damage a 3/4" drain line or 1" safety valve connected to the 2" normal chstging line connection to the RCS Loop 4 hot leg.

Since there is a check and motor-operated valve downstream of these connections, no serious consequences are anticipated. In any event, the plant would not be prevented from completing a. safe shutdown.

5.

Routing of Instrument Control Cable Inside the Vapor Container Of particular importance to the routing of instrument control cable are i

the cylindrical reinforced concrete biological shield wall which completely l

surrounds all RCS compartment areas and the reinforced concrete division walls f

between each of the RCS compartments.

l The V.C. penetrations are all located along the north side of the V.C.

circumferential1y along areas outside RCS Loops 2 and 3 either just above or below broadway. Broadway is a walkway at approximately elevation 1104' around the outside of the biological shield wall.

l The electrical and instrumentation penetrations inside the V.C. are enclosed by a steel box structure called a blister. Inside each blister the control ceble f or the various instrument channels are terminated at terminal 1

l strips which are physically separated from each other by varying distances. l

However, instrument channels measuring the same parameter are connected to different terminal strips.

Instrument signals penetrate the V.C. via multi-conductor cable connected to sealed electrical penetrations called cartridges. The penetration box on the outside of the V.C.,

also called a blister, is similar to the one on the inside.

RCS Pressure Indication There are six (6) instrument channels for measuring pressure in the RCS.

Four originate from transmitters measuring pressure Jn three of the RCS loops and the others originate from transmitters measuring pressure in the pressurizer. The process connection to each transmitter is routed from within the individual compartments through conduit sleeves in the biological shield wall (just above broadway elevation) to the respective transmitters.

Three of the channels measure pressure in the cold leg of main coolant Loops 1, 2 and 3.

All of these transmitters are powered from the vital bus.

MC-PT-100 ',for Loop 1) is located on the outside wall of the biological shield between Loop 1 and the pressurizer. Cable then runs in conduit up the biological shield wall and counterclockwise around broadway to blister SE.

MC-PT-200 (for Loop 2) is located on the outside wall of the biological shield between the pressurizer and Loop 2.

Cable then runs in conduit up the biological shield wall and clockwise around broadway to blister SE.

MC-PT-300 (for Loop 3) is located on the outside wall of ti.e biological shield between Loop 2 and Loop 3.

Cable then runs in conduit up the biological shield wall and clockwise around broadway to blister SE.

One channel measures pressure in the hot leg of main coolant Loop 1.

This transmitter is powered from emergency MCC 1 (which is normally supplied by emergency bus 1).

MC-PT-712 (LTOP) is located on the outside wall of the biological shield between Loop 1 and the pressurizer. Cable then runs in conduit up the biological shield wall and counterclockwise around broadway to blister 5E.

Two channels measure pressure in the pressurizer. Both transmitters are powered from the vital bus.

PR-PT-700 (Saturation Monitor) and PR-PT-6 (NR Press) are both located on the outside wall of the biological shield i

l between the pressurizer and Loop 2.

Cable then runs in conduit up the biological shield wall and clockwise around broadway to blister SE.

RCS Temperature Indication There are seventeen RCS loop temperature indications. Each of the four RCS loops has three resistance temperature detectors (RTD's), one in each hot leg and two in each cold leg. Remote transmitters provide wide and narrow-range temperature indication for both the hot and cold legs.

Difference amplifiers provide wide-range T indication for each loop and an averaging circuit provides an arithmetic narrow-range T average indication.

Each temperature signal is routed out of each loop compartment in MI cable through the bialogical shield wall to a terminal box outside the loop compartment. From the terminal box, instrument control cable is run in conduit just below broadway around the biological shield wall to blister 7E.

The conduits for Loops 1 and 2 temperature channels are routed clockwise, while those for Loops 3 and 4 are routed counterclockwise.

There is an additional cold lag temperature detector in RCS Loop 3 powered from a transformer supplied from emergency bus 1.

In addition, several thermocouples installed in the reactor vessel at the core outlet provide additional temperature channels. These additional temperature channels are routed in essentially the same way to blister 7E as are the RCS Loop 3 hot and cold leg temperature channels.

RCS Flow Indication Differential pressure drop across the primary side of each steam generator is used to determine RCS flow in each loop. The flow signal is converted to an electrical signal by a transmitter powered from the vital bus.

From the transmitter located in the loop compartment, each loop flow channel is routed directly out of the loop compartment in MI cable through the biological shield wall to a terminal box and then to blister SE.

The conduits for RCS Loops 1, 2 and 3 flow channels are routed clockwise, while the channel for RCS Loop 4 is routed counterclockwise.

Pressurizer Level Indication Detectors PR-LD-6 and PR-LD-8 are used to measure pressurizer level over a narrow and wide range, respectively. Each level signal supplies a separate level indication channel which is converted into an electrical signal by a transmitter powered from the vital bus.

From the transmitter located in the pressurizer cubicle, each level channel is then routed directly out of the pressurizer cubicle through the biological shield wall to a terminal-box.

From the terminal box, instrument control cable for each level channel signal is run in the same conduit and routed clockwise around the biological shield wall just below broadway to blister SE.

A third channel has also been provided for remote indication of wide-range pressurizer level.

The transmitter (PR-LT-705) is located in the pressurizer cubicle and is powered from emergency MCC 1.

From the transmitter, cable runs in conduit into the Loop 1 compartment and then through and up the biological shield wall and clockwise around broadway to blister SE.

Reactor Power Level Cables for the nuclear instrumentation, i.e.,

reactor power level, are run in two (2) conduits routed in two separate directions from the neutron shield tank. One conduit runs in RCS Loop 2 compartment, while the other runs in RCS Loop 3 compartment. Each conduit runs to a junction box on the biological shield wall outside the respective loop compartment. From each junction box cable is run in conduit just below broadway around the biological shield wall to blister 6E.

The cable exiting RCS Loop 2 compartment runs l

clockwise, while that from RCS Loop 3 runs counterclockwise.

Steam Generator Water Level Two (2) detectors measure feedwater level in each steam generator. One detector gives a wide-range measurement and is used for indication. The other detector gives a narrow-range measurement and is used as an input to the feedwater regulating system as well as indication. Each level signal is converted into an electrical signal by a transmitter powered from a transformer supplied from emergency bus 1.

From the transmitters located in the loop compartments, the level channels are routed directly out of each loop c ompartment in hi cable through the biological shield wall to a terminal box outside the respective loop compartment. From the terminal box, instrument control cable is run in conduit just below broadway around the biological shield wall to blister SE.

The conduits carrying level channels from steam generators 1, 2 and 3 are routed clockwise, while the channels for #4 steam generator are routed counterclockwise.

An additional detector for each steam generator gives a narrow-range measurement and is used as an input to the reactor protection system as well as for indication. Each narrow-range level signal is converted into an electrical signal by a transmitter powered from the vital bus.

From the transmitters located in the loop compartments, each narrow-range level channel is routed directly out of each loop compartment in conduit through the biological shield wall to V.C. blister SE.

The conduits carrying the narrow-range level channels from steam generators #1, #2 and #3 are routed clockwise, while the channel for #4 steam generator is routed counterclockwise.

Modifications Planned As discussed above, modifications planned as part of the environmental qualification program will reroute instrumentation cable for existing instrumentation so that all instrumentation does not run together in blister SE.

Pipe whip considerations will be taken into account to prevent one break f rom fr.iling all channels of various indications. Also, new wide-range steam generator level channels will be added, and the cable routings given the same considerations.

l VII.

CONCLUSIONS Yankee has evaluated the effects of high energy breaks inside containment utilizing the most conservative approach, the effects-oriented approach. Based on the results of this evaluation outlined above, the vast majority of breaks do tot prevent the plant from achieving and maintaining a safe shutdown condi tion.

The following two interactions require further analysis:

1.

The resulting thrust forces on the steam generator due to main steam line and feedwater line breaks must be analyzed.

2.

The effects of main steam jet impingement on the vapor container electrical penetrations in blister 12E.

The resulting interactions from breaks in the large RCS piping are unacceptable based on the effects-oriented approach. Yankee is presently i

evaluating possible alternatives to develop an acceptable solution to these break locations. One possible solution is contained in the guidance supplied in the Palisades SER on this topic, dated December 4,1981.

This guidance and other potential solutions are presently being evaluated. -

.