Text

Forwards Addl Info & Responses to NRC Questions for Ser. Responses Will Be Incorporated Into Future FSAR Amend
	ML17266A488
Person / Time
Site:	Saint Lucie
Issue date:	08/11/1981
From:	Robert E. Uhrig; FLORIDA POWER & LIGHT CO.
To:	Eisenhut D; Office of Nuclear Reactor Regulation
References
	L-81-348, NUDOCS 8108240256
	Download: ML17266A488 (495)
	v • d • e

REGULATORY FORMATION DISTRIBUTION S EM (RIDS)

ACCESSION'BR;8108240256'OC DATEi 81/08/11 NOTARIZED NO '!

DOCKEiTI FACILC50 389 S). Lucie'lant~ Unit 2<, Florida- Power tt L'ight g'5000389 NAME-'UTHOR AFFILIATION Co;,'AUTH",

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RKC IP s NAMEI RECIPIENT AFF ILIAT!ION<

EISENHUT'rDOG, Division of Licensing,

SUBJECT:

" Forwards'ddi info.a response's to NRC'uestions for i SER'o.

Responses will be incorporated into future. FSAR aNe hd'o .

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P.O. BOX 529100 MIAMI,F L 33152 yk)lliy ly FLORIDA POWER & LIGHT COMPANY August 11, 1981 L-81-348 Office of Nuclear Reactor Regulation Attention: Mr. Darrell G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 I,

Dear Mr. Eisenhut:

Re: St. Lucie Unit 2 Q/

Docket No. 50-389 Final Safety Analysis Report Re uests For Additional Information Attached are Florida Power 8 Light Company (FPL) responses to NRC staff requests for additional information which have not been formally submitted on the St. Lucie Unit 2 docket. These responses will be incorporated into the St. Lucie Unit 2 FSAR in a future amendment.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 5 Technology REU/TCG/ah Attachments go~>

cc: J.P. O'Reilly, Director, Region II (w/o attachments)

Harold F. Reis, Esquire (w/o attachments)

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t~p 8108iil 8108240256 05000389 PDR ADOCK P..DR PEOPLE... SERVING PEOPLE

Attachment to L-81-348 A. Responses to Auxiliary Systems Branch draft SER open items.

B. Revised response to question 451.08 C. Draft FSAR writeup and supporting documentation for underground cable qualification.

D. Control Wiring Diagram supplied in support of the response to Chapter 8.3 open SER item on power lock out to MOV's.

E. Response to Chapter 8.3,.open,;SER:. item on isolation devices.

Response to Chapter 8.3 open SER item on GDC 18 G. Response to Chapter 8.3 open SER item on MOV Thermal Overload Bypass.

H. Responses to open items from 8/11/81 meeting on Post Accident Sampling System.

I. Revised response to. question 410.19 J. Draft Environmental Report sections on the use of TBTO.

K. Tables 1.9B-3 and 1.9B-4, Evaluation of ICC Detection Instrumenta-tion.

L. Responses to Containment Systems Branch questions.

M. Response to question 492.10 N. Revised responses to question 440.25, 440.28. 440.38, 440.39, 440.41, 440.44, 440.51, 440.54$ 440.58, 440.59, 440.61, 440.62

0. Confirmation on MSIV bonnet and seat thickness conservations from

,Rockwell International.

P. Response to open item No. 1 from the Structural Engineering Branch design audit.

gl08240256 ',

RESPONSES TO AUXILIARY SYSTEMS BRANCH REQUESTS FOR INFORMATION TO COMPLETE THE ASB DRAFT SER..

Section 3.5.2 Structures, S stems, and Components to be Protected from Externall Generated tlissi1es The applicant has verbally committed to providing missile protection lfor the auxiliary feedwater cross-over piping between the steam trestles

'and outside of the missile barriers; however, documentation has not been provided. Me will report resolution of this item in a supplement to this SER. This item also impacts Sections 10.3.1 and 10.4.9 of this, SER.

~Res ense See attached amended response:-to Question 410.25.

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SL2 PSAR estion No.

410.25 . Qith regard to the main steam trestle, provide the following:

(10.3, 10 ' ') a) Verify that the Category I and main steam maximum trestle is designed to seismic tornado load requirements.

Provide a complete description, including arrangement drawings, of the main steam trestle area which. illustrates how following items are protected from turbine and tornado 'he missile s.

(1) Main Steam Isqlation Valve (MSIVs)

(2) Hain Steam Safety Valves (3) Atmospheric Dump Valves (4) Main Steam Piping up to the HSIVs (5) Safety-related portions of the main feedwater piping.

c) Provide detailed layouts of the auxiliary feedwater pump and piping areas to demonstrate how the main steam trestle provides support "for missile protection enclosing the Auxiliary Feedwater Pump rooms" (FSAR Subsection 3.8.4.1.9) and its protection from high energy line breaks (eg. mein steam or main feedwater pipe. breaks) and moderate energy pipe cracks

~Res oese The main steam trestle is provided to house the safety-related components of the Main Steam, Feedwater, and Auxiliary Feedwater System. The trestle is designed to seismic Category I requirements and is capable of withstanding the maximum tornado loadings outlined in FSAR Section 3.5. The loading combinations for the main steam trestle are provided in FSAR Subsection 3.8.4.3.

b) The main steam trestl is comprised of two compartment p FIV Syg te alii'I ii located at the west end of the Reactor Bui ing The two s CC~i trestle compartments are two total y enc ose structures which are physically separated from each other. Each trestle compartment houses the following equipment:

(1) One main steam Line (2) One main steam isolation valve (MSIV)

(3) Eight main steam safety valves (4) One main feedwater line (5) Two main feedwater isolation valves (HFIV's)

(6) Two atmospheric dump valves (ADV's)

(7) Two motor driven aux. feedwater pumps or one steam driven aux. feedwater pump (with associated piping and valves) ~

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SL2 PSAR Each of the two compartments of the testle is approximately 31 Eeet wide, 45 feet long and extends vertically from grade level to Elevation 62'W". Three sides of the main steam trestle are completely enclosed with a one inch steel plate along the entire vertical run with a nine inch opening left. on the base perimeter to provide for natural ventilation- The fourth side utilizes the containment structure as a missile barrier and is recessed several feet from the containment in order to provide adequate ventilation. The roof of the trestle structure utilizes a steel grating (several inches

. thick) for missile protection purposes. The openings in this grating have been designed to inhibit the smallest missile provided in FSAR Section 3.5 and to provide sufficient main steam Mass and'Energy blowdown area to accommodate a main steam line break outside the containment.

c) Detailed layouts of the Auxiliary Feedwater Pump and piping arrangements are provided in FSAR Figures 10.4-14, 15 and 16.

The motor driven auxiliary feedwater pumps are physically separated from the turbine driven pump by two one '(1) inch steel plates. These plates provide adequate protection against the dynamic effects .of a high energy line break. The dynamic effects associated with pipe rupture and jet impingement is provided in FSAR Section 3.6i 410.25-2 Amendment No. 4, (6/81)

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2. Section 3.6.1 Plan~t Oesf n for Protection A ainst Postu1ated Pi in ailures in F uid S stems Outside Containment s

The applicant has not provided sufficient information necessary to demonstrate that a- postulated high energy pipe break or moderate energy pipe crack will not cause a loss of function of any safety related system. The applicant has not provided sufficient information to adequately demonstrate that f1ooding due to failure of non-seismic Category I tanks will not adversely affect safety related equipment.

Me will report resolution of this item in a supplement to this SER.

This item also impacts Sections 9.3.3, 10.3.1, 10.4.5, 10.4.7, and 10.4.9 of this SER.

~Res onse FP&L has formally submitted the above input via letter dated L-81-334 dated August 4, 1981.

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3. Section 9.1.3 S ent Fuel Pool Coolin and Cleanu S stem'he applicant has verbally committed to the installation of a second spent fuel pool cooling system heat exchanger by the first refueling of Unit 2. Documentation to confirm the verbal commitment is required.

)le will report resolution of this item in a supplement to this SER.

Response

See attached amended FSAR page 9.1-10 adding the commitment for the applicant to add a, second fuel pool heat exchanger.

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SL2"FSAR lated by the fuel pool pumps through the fuel pool heat exchanger where heat is rejected to the Component Cooling Water System. From the outlet of the fuel pool heat exchanger, the cooled fuel pool water is returned to the bottom of the fuel pool via a distribution header. The cooling system is controlled manually from a local control panel. Control room alarms for high fuel pool temperature, high and low water level in the fuel pool, low fuel pool pump discharge pressure and, as discussed in Subsection 9.1.2, a high radiation in the fuel pool area, are provided to alert the operator to abnormal circumstances. Radiation monitoring for spent fuel pool area and Fuel Handling Building stack is discussed in Section 11.5. The components and piping are Quality Group C, seismic Category I.

9.1.3.2.2 Fuel Pool Purification The clarity and purity of the water in the fuel pool, refueling cavity and refueling water tank are maintained by the purification portion of the fuel pool system. The purification loop consists of a fuel pool purification pump, fuel pool filter, fuel pool purification pump suction strainer, fuel pool ion exchanger, fuel pool skimmer, fuel pool ion exchanger strainer, associated valves, and piping. Most of the purification flow is drawn directly from the fuel pool. A small fraction of the purification flow is drawn through the fuel pool skimmer. A strainer is provided in the purifi-cation line to the fuel pool purification pump suction to remove particu-late matter before, the fuel pool water is pumped through the fuel pool filter and the fuel pool ion exchanger. The fuel pool water is circulated by the fuel pool purification pump through the fuel pool filter, which re-moves particulates larger than five micron size, then through the fuel pool ion exchanger to remove ionic material, and finally through a "Y" type fuel pool strainer.

Connections to the refueling water tank provide makeup to the fuel pool through the purification loop. In addition to purifying the fuel pool water, the refueling water tank and the refueling transfer canal are cleaned through connections to the purification loop. Fuel pool water chemistry is given in Table 9.1-4. The purification loop components and main process piping are Quality Group C, non-seismic.

9.1.3.2.3 Component Description The major compnents of the Fuel Pool System are described in this section.

The principal component data summary is given in Table 9.1-6.

a) Fuel Pool Heat Exchanger The fuel pool heat exchanger is a horizontal shell and tube design with a twq-pass tube side. A slight pitch, three degrees above the horizontal, is provided for complete draining of the fuel pool heat exchanger. The component cooling water circulates through the shell side, and fuel pool water circulates through the tube side. The in-ternal wetted surface (tube side) is stainless steel. f4 ~p(t>'c~~

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4. Section 9.1e4 Fuel Handlin S stem Me require that the applicant implement the inLerim actions identified in Enclosure 2 of the generic NRC letter dated December 22, 1980, con-concerning NUREG-0612 '!Control of Heavy Loads at Nuclear Power Plants" prior to receipt of an operating license and prior to full implementa-tion of NUREG-0612. The applicant indicated that the bulkhead gates (one between the cask pool and the spent fuel pool and the other between the spent fuel pool and the fuel transfer canal) are seismic Category I. Documentation to confirm the verbal comitment is required.

ate will report resolution of these items in a supplement to this SER.

~Res ense The six month response to the NRC December 22, 1980 letter was issued to the NRC via FPGL letter L-81-338 dated August 6, 1981 (Uhrig to Eisenhut).

See attached amended FSAR page 9.1-6 adding the fact that the. removable bulkheads in the spent fuel storage pool are designed to seismic Category I requirements.

SL2-FSAR kinetic energy associated with the dropped fuel assembly is 29,000 in-lb.

This energy is conservatively assumed, to be totally absorbed by one rack module. Structural deformations of the racks are limited to preclude any I 0 possibility of criticality.

The structural design also precludes the possibility of a fuel assembly being placed in the spaces between the fuel cavities.

Adequate clearance is provided between the top of the stored fuel assembly 0,

and the top of the rack to preclude criticality in the event a fuel as-sembly is dropped and lands in the horizontal position on the top. Rack design also ensures adequate convection cooling of a fuel assembly lying horizontally across the top of the racks.

The spent fuel storage racks are designed in accordance with the AISC Specifications and the load combinations and allowable stresses specified I 0 in Subsection 3.8.4.3 for seismic Category I steel structures.

The direct dose rate at the pool surface when not refueling is less than I 0 2-5 mrem/hr. This dose rate is based on the most active fuel assembly two days after shutdown. During refueling the limit switches prevent the spent fuel handling machine frorrr raising the spent fuel assembly above a height where less than nine ft. of water provides minimum radiation shielding- If the interlock should fail and if there were no operator action, the fuel handling machine cannot raise the assembly above a nine ft- water-to-active-fuel-length height 'because of the design geometry. Under the condi-tions described above, the dose rate at the surface of the water above the assembly would be still less than 2.5 mrem/hr. The grappling tool on the spent fuel handling rnachine is designed so that a fuel assembly cannot be released accidentally. The shielding provided in the Fuel Handling Build-ing is discussed in Subsection 12-1.2.4 ~

A concrete wall to,elevation 62 ft. separates the cask storage area from the spent fuel storage area. The wall prevents the water level from uncovering if the spent fuel assemblies even w.%r'r ~"Irs CD~cm ~

rstartrLI M) 'are. cled)i c~e~sE a dropped fuel cask causes damage to the pool or pool liner in the cask storage ar6'eh. @~I/:

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percent- This is substantially higher than the enrichment for the initial and future cores. In the analysis to determine allowable edge-to-edge spacing, infinite arrays of fuel assemblies are ay~Itmed. analysis of the spent fuel storage rack design uses the CHEETAH-P~ /PDQ-7~+~ rradel as the basic engineering tool. CHEETAH-P is the PMR lattice version of 0 Nuclear Associates International (NA/) CHEETAH code which is a modified vers)g~ of the original LEOPARD code and uses a modified ENDP/.

B-II cross section library. The PDQ-7 program is the well-known few-group spatial diffusion 'theory code widely used by the industry. The CHEETAH"P/PDQ-7 model has been extensively tested by NAI by means of bench-marking calculations for several existing operating power reactors.

CHEETAH-P determines a multigroup neutron spectrum for a given homogeneous mixture of materials and uses this spectrum to weigh the cross sections and provide average few group cross sections. PDQ-7 uses as input the cross 9.1-6 Amendment No. 0, (12/80)

5. Section 10.4.7 Condensate and Feedwater S stems The applicant has not committed to performing a water hammer test in accordance with Branch Technical Position ASB 10-2. We require, the water hammer test. We will report resolution of this item in a supple-ment to this SER. This item also applies to Section 10.4.9 of this SER.

~Res onse A number of paragraphs were inadvertently omitted from the response to PSAR question 410.27. See attached page for amended response. In addition, PPGL letter L-81-318 dated July 27, 1981 (Uhrig to Eisenhut) pro-vided justification for the applicant's position that steam generator water hammer testing need not be per-formed on St Lucie Unit 2 (letter attached).

SL2-FSAR '

estion No.

410.27 State how Branch Technical position ASB 10-2, "Design Guidelines (10.4.7) for Water Hammers in Steam Generators with Top Feeding Designs" is met'iscuss the design features to minimize water hammer and the confirmatory tests to be performed.

Response

The feedwater piping and feedring have been designed to eliminate or minimize the cause and effects of possible water hammer in the feedwater system.

Feedwater enters the steam generator through the feedwater nozzle where it is distributed via a feedwater distribution ring. The feedwater ring has been constructed to include discharge nozzles called "J" tubes which are welded to the top of the ring (see Figures 5.4-6, 16:; and 17 of the FSAR). 'his construction reduces the rate at which the feedwater ring drains, helping to provide "assurance that the ring remains full of water. Thus, the

'robability of significant amounts of steam entering the feedring is greatly reduced, thereby minimizing the condition which can lead to water hammer.

In addition, the length of horizontal feedwater piping immediately to the steam generator which could pocket steam is 'xternal minimized (2 1/2 feet). This short length of horizontal piping has a downward sloping 90 elbow followed by approximately 32 feet of vertical feedwater piping. This piping arrangement minimizes the drainable volume of feedpipe. Hence, when the feedrCng and piping are drained and steam enters this region, the exposed surface of subcooled water to saturated steam is minimized.

The minimization of the exposed surface of subcooled water to the saturated steam reduces the depressurization of the steam space by slowing the rate of steam condensation on the subcooled water.

The pressure pulses generated by a water slug in the piping are initiated by steam-water interaction which causes ripple formation at the steam-water interface. This results in the formation of a water slug which isolates the steam in the feedpipe As the isolated steam condenses, pressures in the region falls and the water. slug accelerates towards it. The kinetic energy in the slug keeps increasing until the steam bubble is collapsed. At this moment, the water slug impacts with the water filling the upstream side of the pipe and pressure pulses are generated.

410. 27-1 Amendment No. 4, (6/81)

I /os'w~T Also note, that. since only a small amount of steam can be trapped in a 90 degree elbow, a steam bubble will collapse before the water slug gains significant kinetic energy duxing a steam-water interaction.

Consequently, by introducing the combination of a short length of horizontal piping and the "J" tube design on the top of the feed-ring, the intensity of the pressure pulses generated (water hammer) is reduced to negligible levels.

St. Lucie Unit 1 has conducted extensive feedwatei. hammer testing.

A review of the Feedwater Piping drawing and Steam Generator internal indicates that St Lucie Unit 1 and St Lucie Unit 2 are virtually assembly. Based on this review and the testing performed on St Lucie Unit 1, the applicant concludes that additional feed-water hammer testing is not required for St Lucie Unit 2.

Section 10.4.9 of the FSAR has been revised to include the above response along with revisions for automatic initiation of the Auxiliary Feedwater System.

P.o. BOX 529100 MIAh11, F L 33152

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FLORIDA POY/ER 6 LIGHT COMPANY r.

July 27, 1981

7. P..".;.':::::: L-81-318 Office of Nuclear Reactor Hr. D. Eisenhut, Director RegulatVbn"'ttention:

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Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Hr. Eisenhut:

Re: St. Lucie Unit 2 Docket No. 50-389 Steam Generator Mater Hammer Testin At a June 17, 1981 meeting with Olan Parr et al, Florida Power 8 Light Company

{FPL) agreed to provide justification for our position that steam generator water hammer testing need not be performed on St. Lucie Unit 2.

A steam generator water hammer test program was conducted on St. Lucie Unit 1 .

with no water hammer observed. The NRC, in a Safety Evaluatio'n.Report issued February 7, 1980, concluded that steam generator water hammer was not likely to occur at that facility.

The St. Lucie Unit 1 and 2 piping arrangements are essentially identical.

Isometric drawings of both units were compared, and dimensional differences were measurable in fractions (e.g., the horizontal sections of piping entering the steam generator, which are the sections of piping most likely to experience water hammer, are all equal in length (two feet long), with one section on Unit 1 3/8 inch shorter than on Unit 2).

The preoperational test program will verify the adequacy of the design. Pre-operational test procedures 2-0700091, "Auxiliary Feedwater Pumps 2A, 2B, and 2C Initial Run", and 2-0700081, "Auxiliary Feedwater System Functional and Endurance Test", will verify that the pumps meet or exceed the manufacturers head/flow curves and associated manual controls and alarms function as required, and also verify automatic operation of the system following an actuation signal. The functional test will be performed prior to hot functional testing of the unit. FPL intends to station an operator inside containmeB; Vi7rVng'he > < ~

initiaI injection phase to monitor for water hammer. Also, FPMMQfiihe" hs.""--:n'rc""

vibration monitoring program during the St. Lucie Unit 2 startup, and piping sI.-

vibration wi 1 1 be measured.

FPL is reviewing the San Onofre steam generator feed ring collapse;-indd5t and will inform you if any change in our position on steam generh'ter-mat~

LT nt-r hammer testing for St. Lucie Unit 2 is required.

L'ZZ Very--truly yours, I PHD 1

oQ t E. Uhrig Vice President Advanced Systems & Technology rj

~ .rgr 1.4 REU/TCG/ah cc: J.P. O'Reilly, Director, Region U~er~1 II e ~as PEO LE.~ . ~k VIMQ IIFOP[ ~

6. Section 10.4.9. Auxiliar Feedwater S stem & Post THI Task II.E.1.1 The following items pertain to Section 10.4.9 of the SER, "Auxiliary Feedwater System," for which documentation is required as indicated for items a, b, and c.

a. The Unit 2 condensate storage tank includes a dedicated water volume for Unit 2 auxiliary feedwater system missile damage to the Unit 1 condensate storagein tank.

the event of tornado There are locked closed valves in the parallel connecting has not provided the procedures delineating lines. The applicant when these locked valves will be opened.

b. The applicant has not provided the results of an analysis of the effects of a potential failure of the Unit 1 condensate storage tank and the most severe failure or operator error on Unit 1 or 2 resulting in draining the Unit 2 condensate storage tank below the Unit 2 dedicated volume.

c. Additional Short Term Recommendation 2 - The applicant has not committed to providing a copy of the pump endurance test results specified in this recommendation. Me require that these results be provided.

d. Our review is not complete with respect to the minimum dedicated, water supply for the auxiliary feedwater system, the minimum flow

. requirements, and the reliability analysis.

e. Additional Short Term Recommendation 3 - The design for emergency feedwater flow indication is un'der review by the Instrumentation and Control Systems Branch as part of item II.E.1.2 of NUREG-0737 and will be reported in a separate evaluation.

f. Lon Term Recommendation GL The design for emergency feedwater automatic .-initiation is under review by'the Instrumentation and Control Systems Branch as part of Item II.E.1.2 of NUREG-0737 and

'will be reported in a separate evaluagion.

Response

a, b and c: See attached revised FSAR pages.

d, e and f: NRC Action.

0 A) FPL intends to perform the Auxiliary Feedwater Endurance Test as part of Preoperational Test No. 2-0700081, "Auxiliary Feedwater System Functional and Endurance Test."

B) FPL will modify the St. Lucie Unit 2 FSAR, Section 10.4.9.4 to reflect Endurance Testing and Section 14.2.12.1.4E to address the specifics of Ref. (a). Attached please find copies of the proposed FSAR modifica-tions.

C) FPL will provide the NRC (after completion .of Preoperational Test No. 2-0700081 results review), a summary of the Endurance Test .consist-ing of the following:

1) Description of the test.

2) Plots of bearing temperature -vs- time.

3) Plots of Pump Room Temperature (Environment) -vs- Time.

4) A statement confirming that pump vibration did not exceed allowable limits.

5) Plot of observed pump performance (pump flow, head, speed, and stem temperature) on the vendor supplied specific equipment performance curves.

Equipment is "Qualified" for 100% humidity therefore humidity will not be rmnitored.

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SL 2-FSAR TABLE 10.4.9A-4 (Cont'd)

ACCEPTANCE CRITERIA COMPLIANCE Recommendation - The licensee should perform a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months endur- A 48 hour endurance test will be performed ance test on all AFW system pumps, if such a test or continu- on thc Auxiliary Feedwater pumps .

I ous period of operation has not been accomplished to date.

Following the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pump run, the pumps should be shut down q~g~5 ~ill hvz S~$ ~;P~ Ho and cooled down and then restarted and run for one hour. pPC.-

Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing/

bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed en-vironmental qualification limits for safety related equipment in the room.

11) .5.3.3 Indication of AFW Flow to the Steam Generators Concern - Indication o! AFW flow to the steam generators is considered important to the manual regulation of AFW flow to maintain the required steam generator water level.

This concern is identical to Item 2.1.7.b of NUREG-0578.

Recocmdendation - The licensee should implement the follow- Safety grade Auxiliary Feedwater ing requirements as specified by Item 2.1.7.b on page A-32 flow indication and safety grade, of NUREG&578: redundant steam generator level indication is available to the (1) Safety-grade indication of AFW flow to each steam operator in the control room.

cp generator should be provided in the control room. These instrument loops are powered by the 120V ac Class IE (2) The AFW flow instrument channels should be powered power source.

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~vp from the emergency buses consistent with satisfy-ing the emergency power diversity requirements for the AFW system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

12) 5.3.4 AFW System Availability During Periodic Surveillance Testing to eer - do e plants ret ire losel s st real(a ent od valves to eo d at p rd di spop os t edttanoe teste on ons AFW system train. When such plants are in this test mode and there is only one remaining AFW system train available to respond to a demand for initiation of AFW system opera-o tion, the AFW system redundancy and ability to withstand a single failure are lost.

Recomm'endation - Licensees with plants which require local Not applicable. Local manual re-manual realignment of valves to conduct periodic tests on alignment of valves to conduct op, one AFW system train and which have only one remaining AFW periodic pump surveillance tests train available for operation should propose Technical on AFS trains is not required,

0 SL2- FSAR

) The AFkS 3.6).

is designed to withstand pipe rupture effects (see Section 10.4.9 2 S stem Descri tion During normal operation, feedwater is supplied to the steam generators by the Feedwater System. The Auxiliary Feedwater System (AP4S) is utilized during normal plant startup, hot standby, and cooldown. During plant startup and hot standby, the system provides a source of water inventory for the steam generators. During cooldown, the AFMS provides a means of heat removal to bring the Reactor Coolant System to the shutdown cooling system activation temperature. brith offsite power and the main condenser available, the condenser will be used as a heat sink. The AFRS system is not utilized during full power operation.

The major active components of the system consist of one steam driven pump with greater than iull flow capacity and two full flow capacity motor driven auxiliary feedwater pumps. Both electrical and steam driven AFbiS pumps are centrifugal units with horizontal split casings and are designed in accord-ance with AShE Code, Section Ill and Quality Group C requirements.

larger pump is driven by a noncondensing steam turbine. 'Ihe turbine receives The steam from the main steam isolation valves, and exhausts to the atmosphere.

e pumps take suction from the condensate storage tank and discharge to the earn generators. The turbine-driven pump is capable of supplying auxiliary eedwater ilow to the steam generators for the total expected range of steam generator pressure by means of a turbine driver controlled by a variable speea mechanical governor.

Each motor-driven pump supplies ieedwater to one steam generator. A cross connection is provided to enable the routing of the flow of the two motor-driven pumps to one. steam generator. The turbine-driven pump supplies feedwater to both steam generators by means of two with its own control va ve ana each sized to pass the full'low. The control of auxiliary feedwater flow and steam ge'nerator level is accomplished by means of control room operated control valves. Local control stations are also provided. Each of the motor driven auxiliary feedwater pumps utilize a Class XE ac power supply (4.16 kV safety related bus). The turbine driven pump train relies strictly on a'c power supply.

10.4 ' ' Safet Evaluation The ABLS removes sensible and decay heat from the Reactor Coolant System during hot standby and cooldown for initiation of shutdown cooling. For events in which main feedwater flow is unavailable, (e.g., loss of main feedwater pump, loss of offsite power, and main steam line break), the APTS is automatically initiated to provide hot standby and/or cooldown heat removals

'%he condensate storage tank (CST) discussed'n Subsection 9.2.6, provides

~ater supply for the Auxiliary Feedwater System. The CST is sized to 1de -I50pHS gallons of demineralized water for St Lucie Un t 2 hot standby and cooldown operations; an additional 550,AGO a ons is reserved in the St Lucie Unit 2 CST only for the unlikely event that a I Q'fq 40O )25,ooo 10,4-20 Amendment No. 4, (6(81)

SL2-FSAR tornado missile somehow ruptures the St Lucie Unit 1 CST and the water contained therein (116,000 gallons per St Lucie Unit 1 Technical Specifi-cations) is unavailable to St Lucie Unit 1. 4hen no tornado ~arnings are xn effect, the St Lucie bnit 2 total capacity of 300,800 gallons is avail-able if neeaed. Ch.ib e.xe-i<<% C opo4ihg <~>+<~S 4o< the. 4o'Aozaq) Uolai>>C'.

'The quantity of water required for St Lucie Unit 2 cooldown has been deter-mined assuming a worst;case condition'herein the unit is brought to hot standby conditions ana held there for approximately two hours then cooled aown at the maximum rate until the shutdown cooling window is reached.

Vnaer this scenario, each Auxiliary Feedwater Pump has the capability of achieving an orderly shutdown consisting of two hours of hot standby followed by a regulated cooldown to the shutdown cooling entry point within next five hours. The quantity of condensate required for this scenario 'he is approximately 129,000 gallons as shown on Table 10.4-2 (Case 2) 4 The. conaensate storage requirements for the Auxiliary Feedwater System were compared with the requirements of Regulatory Guide 1.139 "Guidance for Resiau 1 heat Removal System". Vnder this scenario, the unit is brought to hot standby conditions and held there for four hours then cooled down at the maximum rate of 75F/hour until the shutdown cooling window of 350F is reached. The condensate storage requirement for this scenario is 149,600 gallons as shown on Table IG.4-2 (Case 1)and Figure 10 '-9 ~

D<<ing emergency blackout conditions (except the hypothetical tornado missile which drains the St Lucie Unit 1 CST) there is sufficient water in the CST to allow hot standby operation for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months and a subsequent cooldown to 350 F over four hours (see Figure 10.4-1G). 'Ihe condensate requirements and the auxiliary feedwater flow rate basis is discussed in FSAR Appendix 10.4.9A.

The steam generated during decay heat removal'nd cooldown after a loss of offsite po~er will be discharged through the atmospheric dump valves, ex-cept for the steam used by the turbine driven auxiliary feed pump. There are two ac/ac motor operated atmospheric dump valves (ADVs) located on each main steam line. The ADV's are capable of automatic modulating service using ac power and are capable of open/close service from the control room using dc power only. Each ADV is sized to pass 50 percent of the flow re-quired to bring the Reactor Coolant System to. the shutdown cooling system entry temperature, assuming that onl 4&5~0 gallons of condensate is avail-able from the condensate storage tank.

12 t>0 The auxiliary fei duster pumps are located underneath ...; steam trestle.

The AFl'S is designed to withstand natural phenomena as described in Sec-tions 3.3 ana 3,5. The condensate storage tank is a structure. It s.~ Category 1 is surrounded by a structural barrier whii: ovides missile and tornado protection for the tank. Components in the ASS are protected from flooding as components are located above the probable'maximum flood level (refer to Section 3.4) ~ The design provisions utilized to protect the AFLS against the dynamic effects of pipe rupture and jet impingement effects are providea in Section 3.6. The Auxiliary Feedwater System piping layout and the steam trestle configuration is provided in Figures 10.4-14, 10.4-15 and 10.4-16 ~

10.4-21 Amendment No, 4, (6/81)

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~~The Consdensate Storage Tanks (CST) are intertied between Units 1 and 2.~~

Each CST is a seismic Category 1, Safety Class 3 structure designed to store suf7icient water to bring each plant from power operation to the inititation of shutdown cooling conditions. The Unit 2 tank is located within a concrete structure designed to withstand the DBE and the im-pact of tornado missiles. The Unit 1 tank, which has been designed to withstand the DBE and horizontal missiles, is not provided with protec-tion from the vertical. missiles. In the unlikely event that a tornado missile ruptures the Unit 1 CST, an intertie with the Unit 2 tank is provided. The Unit 2 CST, which stores sufficient condensate to cool down both plants, can be connected directly with the suctions of the Unit 1 Auxiliary Feedwater Pumps. Should Unit 1 require condensate before Unit 2, valves A, D, E & F could be opened. The location of the nozzle on the Unit 2 tank insures that the Unit 2 supply of condensate is not compromised while at the same time providing sufficient coolant for Unit 1. Alternately, if Unit 2's condensate had been previously consumed, valves B, C, D, E & F are opened to supply Unit 1. Check valves are placed in the Unit 1 suction lines between the tank and the interties to prevent the backflow of condensate to a ruptured tank.

The provision" of redundant locked closed manual valves precludes the accidental loss of condensate. The entire intertie line that runs between the Units is buried, thereby providing protection from the effects of tornado missiles. 'I

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~~Additional request for information regarding the trestle grating missile proteation.~~

~~~Res ense See amended FSAR page 3.5-40~~

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~~'e e~~

~~l SL2 PSAR TABLE 3.5-3 (Conttd)~~

~~Location FSAR Systt'ia~~

~~~tui ent Sld /Elevation (ft) ~teecri tien ~PRRR Pi ure Enclosure Atmospheric Dunp Valves Trestle Area/~~

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~~0 Revision II]~~

~~8/3/81 NRC uestions on St. Lucie FSAR g i 45l.OB The terrain correction factors presented in Table 2.3-102 indicate that the straight-line~~

'nnual average atmospheric dispersion model may not adequately represent the regular spat-ial and temporal vari ati ons in a irf1 ow in the vi cini ty of the St. Lucie site. However, the puff-advection model on which these correction factors are based is most useful when meteorological data from multiple sources can be used to describe spatial and temporal variations in airflow. Identify the meteoroQical data used as input to the puff-advection model, and discuss the appropriateness and reasonableness of correction factors at distances of 7.5 miles and beyond.

~~The puff-advection model {MESODIF) was used on the FSAR analyses to r~~

~~develop site-specific terrain/recirculation correction factors. Tnese~~

~~~ ~ ~~~

~

adjustments were developed for application to the straight-line airflow model to account for, on an annual basis, the airflow ch'aracteristics in the St. Lucie site vicinity that affect the atmospheric transport and diffusion conditions. For the St. Lucie coastal site, these conditions consist of sea and land breeze circulations.

The terrain/recirculation correction factors were developed from the ratio of the relative concentrations calculated using the puff-advection model and straight-line model for the meteorological data period of August 1977 through August 1978 (8760 valid observa-tions). Although it is true that the puff-advection model can be run and is more useful with multiple source input, such a run configura-tion is of more importance in areas of complex topography and/or for large distances from the release point. For the St. Lucie application, the one station puff-advection analysis should be appropriate for "distances less than 7.5 miles as the onsite meteorological data will contain the land and sea breeze circulations. Topographic modifica-tions within this range shou1d not be of significance. The appropriate-ness of this application is further supported by the fact that sea breeze ci.rculations have been found to penetrate up to 50 kilometers

~~Revision /ll 8/3/81.~~

inland and that the expected releases from the St. Lucie site are at ground level. Therefore, the data as measured at the onsite tower should, in application in the puff-advection model, be representative of the 7.5 mile radius inland.

Of additional concern is the use of the results of the puff-advection analysis for flows offshore. The fact that the meteorological data are not available over the ocean and on observations of other investigators indicating the slow adjustment of meteorological para-meters to over water trajectory, the application of the one-station puff-advection analysis to the over water trajectories within 7.5 miles is appropriate and reasonable for this site.

The magnitude of the terrain/recirculation factors presented in Table 2.3-102 for large distance from the source are expected and appropriate due to the physical processes involved and the nature of the two models. Because of the lack of major terrain considera-tions and the general persistence of the sea breeze circulations at coastal sites in Florida, a one-station puff-advection analysis may be more appropriate at the St. Lucie location then at others without such ambient meteorological/terrain conditions. But because o the limitation of the puff-advection analysis to the use of one-station, the terrain/recirculation correction value calculated at large distances are more uncertain, but not unreasonable, than the values calculated closer to the source of the meteorological data.

SL2-FSAR d) Emergency Core Cooling System piping e) control rod drive mechanisms f) fuel assemblies and spacer-grids g) reactor internals reactor cavity shield walls secondary shield walls 1.9.4 LOW TEMPFRATURE OVERPRESSURE PROTECTION (LTOP)

Low temperature overpressure protection will be provided via the installa-tion of power-operated relief valves (PORVs) qualified for both saturated steam and liquid relief service ~ The PORVs will be sized to accommodate the pressure transient associated with a Controlled Rod Withdrawal and also (at the low pressure setpoint) to mitigate the pressure transient resulting from either a spurious initiation of safety injection, or a reactor coolant pump start with an excessive temperature difference between the RCS and the steam generator. The final design is described in Subsection 5.2.6. Corresponding transients analyses will be provided in Section 15.8 early in 1981.

1.9.5 HYDROLOGICAL DATA As discussed in Section 2 4. additional information for Hutchinson Island is being evaluated, on the separate sub)ects of further tide data and possible potable well locations'n amendment to Section 2.4 will be filed on or about March 1981 incorporating the relevant information.

1-9.6 UNDERGROUND CABLE REVIEW Kerite insulated power and control cables have been reviewed a~) approved by the NRC for underground environmental qualification 'et/dry lgderground cables is ec'sc i'd+~ s<<4s<4'm 3,il.4.

1 9.7 ENVIRONMENTAL AND SEISMIC QUALIFICATION OF CLASS lE EQUIPMENT In mid-1978 the organized to NRC on Class lE equipment qualification. Sections 3.10 and 3 provide the requested information on seismic

'l issued a letter (10) requesting additional information and have been environmental quali'fication test results. However, at the date of tendering the FSAR several vendors'ualification test summaries and reports of results are still being generated and have not yet been received. Therefore, amendments

~~, to Sections 3.10 and 3.11 will be filed periodically in order to provide the necessary information and also to'provide results of relevant analyses when available-'~~

~~Per a memorandum and order issued on May 23, 1980, (ll) the.NRC has ordered applicants for operating licenses to meet the requirements of~~

~~~ 9-2 'mendment No. 1, (4/81)~~

~~1 0~~

SL2-FSAR integrated radiation exposure combining 40 years normal nperation and the required term of functionality during the post design basis accident (DBA) period (up to 1 year). Tables 3.11-1 present .the design parameters for radiation for each specified envirnnmental condition ~

The normal operations expnsure dnse fnr equipment is either derived more explicity from the design source terms presented in Chapter ll taking account of equipment arrangement and shielding configuration, nr based on the maximum dose rate anticipated for the radiation zone in which the equipment is generally located. See Section 12.3 and the zonal dose maps on Figures 12.3-4 through 12.3-12. For equipment in lower radiation zones (I 6 II) the cumulative 40 year exposure is cnnservatively taken as 10 Rads. For Zone V equipment with a few exceptions, (the CVCS ion exchanger, spent resin tank, spent fuel transfer tube and volume cnntrol tagk) the dose rate is 100 R/hr. For the aforementioned exceptions,,the design dose rate is higher than 100 R/hr.

The DBA exposure dose affecting ESF systems and associated safety related ccmpnnents is dependent nn equipment lncatinn. The DBA considered for the containment, Reactor Auxiliary, Turbine, and Diesel Gener~t.or rruilar.ngs is the pg)uiation of a LOCA in accordance with the recommendatinns nf T?D-14S44 and Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Mater Reactors", June 1974 (R2). The DBA affecting equipment in the Fuel Handling Building is based on the postulation of a fuel handling accident.

~~The few nrganrc materials that exist within the containment are discussed in Subsectinn 6.1.2.~~

The radiation exposure dnse rates given in Table 3.11"1 i s based on gamma radiation exposure. It is recognized that the beta energy release from nnble gases is as much as 2.5 tirIIp~ greater than the gamma energy release within 30 days post accident. However a representative cable geometry inside containment has protective cover sheathing the insulation layer and an overall cover of fire prntective Flamemastic or equivalent.

Therefnre the integrated beta radiatinn dose for a one year post accident perind is less than 10 percent of the integrated gamma radiation dnse over the same 'period. This comparisnn includes the conservative assumption of compari ng effective 2.2 Mev betas with effective 2.2 Mev gammas and assumes a spherical cloud, radius 40 ft, of airbnrne nuclides. Other cnmpnnents inside containment are considered sufficiently shielded from beta radiation since it is effectively attenuated by only a few mills thickness of metal. Therefore based nn the aforementioned discussinn beta radiation is not considered an environmental qualification problem.

3.11.6 SrJBMERGED CABLES Safety related cables located outdoors that could be submerged in water are qualified for nperability under submerged conditions. Pcdb e 'e v,'r~~g

~~@<"Ir'~r~ Q al<~! Ar." p~y M Weri'4.Mp~y m(4.~~

~~W~'M M(~~~

~~~~( ~+, we+/Ay Mud~~ 'km 4u s~L c P.C.( r.~~

~~3.11-5~~

~~0 SL2"FSAR SECTION F 11: REFERENCES (1) D '8 Vassalo (NRC) letter to Dr. R E Uhrig (FPL), "Environmental and Seismic Qualification nf Class IE Equipment" dated July 28, 1978.~~

~~(2) Dr. R E Uhrig (FPL) letter L-78-334 to D B Vassalo (NRC) dated October 16, 1978.~~

~~(3) J J Di Nunno, F D Anderson, R E Baker and R L Waterfield, "Calculation of Distance Factors for Power and Test Reactor Sites," TI0-14844, USAEC,, March 23, 1962.~~

~~4 (4) 1976 ANS Paper: "In-containment Radiation Environments follnwing the Hypothetical LOCA (LWR)."~~

~~3.11-6~~

FLORIDA POWER & LIGHT COMPANY ST LUCIE UNIT 2 DOCKET 50-389 ENVIRONMENTAL DATA FOR UNDERGROUND CABLE. EXPOSED TO WET/DRY ENVIRONMENTS I. T es of Cables Used In Under round Ducts Two cable vendors supply cables for use in underground ducts. They are the Okonite Company and the Kerite Company. Okonite supplies 5KV & 15KV power cable. Kerite supplies 600V power, control and instrumentation cable.

The 5KV and 15KV power cables are insulated with unfilled cross linked polyethylene, wrapped with an extruded layer of semiconducting insulation shield material compatible with the insulation, and covered with a lead sheath and a heavy duty overall neoprene jacket.

~~The 600V power cables are insulated with a high temperature Kerite insulation (HTK) and covered with black heavy duty flame resistant (FR), jacket.~~

~~The 600V control cables are insulated with Kerite flame resistant (FRII) insulation~~

~~~ ~~~

~~and covered with heavy flame resistant (FR) jackets.~~

~~The 600V instrumentation cables consists of twisted P aired shielded and~~

~

unshielded cables. Unshielded cables consist of twisted pairs with Kerite flame resistant (FRII) insulation covered with an extruded polymer layer and having an overall flame resistant (FR) jacket. Shielded cables in addition to the above have a drain wire with each pair in direct contact with a'lumimum mylar tape. Each shielded pair is separated by glass mylar tape.

~~0 II. Test Data Vendor data (Kerite and Okonite) regarding the environmental qualification of their cables exposed to a wet/dry environment are attached for your use.~~

~~In addition'o the above, a procedure was developed on St Lucie Unit 1 to test certain underground cables to confirm their function'ability.~~

The following is a brief synopsis of this Unit 1 procedure. At least once per 18 months, during shutdown, by selecting on a rotating basis at least three (3) cables, one from switchgear to intake cooling water motor, one from switchgear to component cooling water motor and one from switchgear to diesel generator are tested with a 2500VDC megger. Control cables that with each of the above motors and diesel generptors are tested are'ssociated with 1000VDC megger. The three spare cables are DC pro~<tested at 25,000 volts and measured for leakage current at 30 seconds intervals for 10 minutes.

All readings must meet technical specification 4.8.1.1.3. If any installed spare cable fails the Hit Pot test, the NRC will be notified and corrective action take per technical specification 4.8.1.1.3.

~~Attached are copies of actual test data taken at St Lucie Unit 1.~~

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THE Post Office Box 340 Qo ciKciwnE Ramsey. New Jer sey 07446 201-825-0300/Cable: Qkonite NW V'CDMPW a ~ pa4gl 0' i ~ V s '~i JuLy 21, 1981 r'babco Secvmm, 7nc. Two tilo&d T~e, C~m Neul Yak, hl.Y. hPc. George, Aorta,~
~~Subject:~~
SC. Luci.e U~ TT P.O. NY-422574 5 8 15kV PowM Cabee P~ A,. A~uun, TRQ M Co con(~ oall, conversation o$ today helative. Co ouzel, prie&.oufLy dub~ed quaLipc~n documentation on wM and Ay cabLe, ~ru~atiom. Ole, have no obje~n o$ yom sub-n~~g any'p~ o< aLL o( om quaLLPcation rtepo~ W Ae hlRC no< do ue have, any objection W Ae,Ln$ oenatian becoming pubic m$ oenation. Vmy ~iLy yo~, R. A. Guba RAG/mg
'fHH DKANITB CDMPANY Rorrmeg New~ ~ FLORIDA POWER h LIGHT COMPANY ST. LUCIE PLANT
~~~ MEDIUM VOLTAGE POWER CABLE EBASCO SPECIFICATION NO. 211-73, REV. 3 PROS. ID g FLO-2998-291 I~~
~~NUCLEAR QUALIFICATIONREPORT for X-OLENE INSULATED CABLES~~
'This document is The Okonite Company's nuclear qualification report for X-Olene insulated cables. It complies literally with IEEE Standard 383-1974, Section 1. 4 "Documentation>>. Section 1. 4 documents the parameters specified in Section l. 3.
~~Included in this report are seven Appendices which serve to further clar-ify Okonite's test procedures and results. These App'endices are as follows:~~
~~~Aeviix Comparison of Okonite's LOCA Qualification Test to the Suggested Test Procedures and NESCR Sheets as given in the Ebasco 211-73 Specification.~~
Coxnparison of Okonite's LOCA Test Profile to basco's Postulated Event 40- Year Life Detail Document. C Radiation Certification LOCA Autoclave Drawing List of Equipment Elevated Temperature Moisture Absorption The necessary'data to document satisfactory compliance as specified in Section 2. 6 of XEEE 383, Documentation of Type Testing, is provided. in this report. The following cross-reference table illustrates where this
~~~ information can be found.~~
~~THQ DKQNITG COMPANY Rxrrwg New Jomey C9048 APPENDIX 7~~
'OISTURE RESISTANCE Long term'moisture stability is one of the essential factors in selection of an insulation for many applications. It is not un-usual for a power cable to be required to operate in an envi>>
ronment alternately wet and dry. To determine the long term water stability of a cable, a sample insulated with a thin wall dielectric is immersed in water at an elevated temperature to accelerate the deteriorating effects of moisture. Monitoring the electrical properties provides an indication of long term behavior. Based upon actual experience capability of withstand-ing total, water immersion at 90 C should be capable of a life in excess of a generating station's designed life in an environment of 100% humidity. Figure I shows long term 90oC water immersion on a 1/C 814 AWG X-Olene insulated cable. Testing has been performed in accordance with ICEA S-66-524, paragraph 6. 6 "Accelerated Water Absorption Tests " except that (1) the water temperature was 90 C, (2) three 25 ft. samples were tested, (3).the test period was extended for 12 months and (4) a 600 volt ac potential was applied continuously. Even with the more strenuous test parameters, the samples met the requirements of ICEA S-66-524, Section 3. 7. 3. 3. The extended 12 month data demonstrates a large margin of assurance. Alternate wet/dry cyclic testing on this insulation has never been performed. For the purchased 5 and 15 kV cables this ques-tion (as well as the entire question concerning moisture) is academic since the lead sheath would prevent all moisture from becoming in contact with the insulation.
~ r Fhmcmg ~ THE OKONITE COlVIPANY New~ Appendix 7, Page 2 LONG TERM 90 C VOTER IMMERSION TEST Construction: 1/C, gl4 Solid CC, . 047" X-Olene (Ref. 2-18, pg. 240). Continuous Stress, 600 Volts, Ac Avera e of 2 Sam les SIR Measuring Stress M ohms-1pppft. Time Volts /mil Vo PF SIC at-500 V Initial 40 0. 1) 2. 31 &250, 000 80 0. 10 , 2.31 Day 40 0. 10 " 2.09 Z. 09 ~289, 000 1 80 '.13 " " I' 4 ~ 1 %'eek 40 0. 06 Z. 13 ~ 276, 000 80 0. 08 2. 13 2 %'eeks 40 0. 07 2. 15 ~ 263, 000 80 0. 07 2. 15 4 %Pecks 4p 0. 07 2. 15 W 237, 000 80 0. 08 2. 15 2 Months 40 0. 05 2. 15 w 141, 000 80 0. 06 2. 15 3 Months 40 0. 06 2. 15 &257, 000 80 0. 06 2. 15 4 Months 40 0. 03 2. 18 w 257,000 80 0. 04 2. 18 5 Months 40 0. 03 2. 18 ~250, 000 80 0. 03 2. 18 6 Months 4p 0. 05 2. 18 M250, 000 80 0. 05 2. 18 12 Months 40 0. 02 '2. 19 ~257, 000 80 0. 03 2, 19
49 Dey Street Seymour, Connecticut 06483 (203) 888-2591
~~-the kertte company 44ec ssiR~EN r s~~
~~CI ~ ) EJ t j r~ I Ji j,! '~t I'I~~
~
~~t ~ July 1, 1981 E:r ~~~
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~~Ebasco Services Inc. ~t ~ 7$ ,'r'rr "s 2 World Trade Center~~
~~~Lu Z New York, NY 10048 ATTENT I ON: MR. 'W. LUNDGREN SENIOR ENGINEER, ST. LUG IE g2 PROJECT Gentlemen:~~
~~SUBJECT:~~
~~FLORIDA POWER s LIGHT COMPANY ST. LUCIE PLANT UNIT g2 PURCHASE CONTRACT NY-422573~~
~
~~Confirming our conversation of June 29, this is to advise that our~~
~~~ ~~~
~~Engineering Department has released the following documents so that you may copy and submit them to the Nuclear Regulatory Commission:~~
~
Kerite Qualification Documents For: Document Reference Reference 6 Engineering Memorandum 240 Kerite FRII/FR Signal June 6, 1979 and Instrumentation Cable Reference 6 Engineering Memorandum 223 Kerite HTK/FR Power Cable May 4, 1977 Reference 5 Engineering Memorandum 205 Kerite FR/FR Control Cable November 6, 1975 We trust that this will satisfy your requirements. Yours truly, THE KERITE COMPANY From'he office of: E.N. Sleight Assistant Vice President National Generation
~~'Signee: rma H. Dube Administrator-Power Plant Generation NHD:ss s ssssidisfy of HARVEY HUBBELL INOORPORATEO~~
~~
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~
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' 'X UCIE ~ " "'BASCOST. NUCLEAR PLANT SERVICES BlCORPOBATED -.:,FLORIDA POWER AND LIGHT COMPANY
~~~' 'QUALIFICATIONDOCUMENTATION FOR KERITE FB/FB CONTROL CABLES ASSIGNED TO EBASCO SERVICES, INC.~~
~~~ ~ ~ ~ C T. ~ ~ ~ JL ~ w ~ I ~ ~~~
The information in this documentation package is confidential and as such is not to be copied or duplicated in any way, without written permission from The Kerite Company. I REF/dm 4/ao/vv C h APR 22 SV
e the hei.it;e company Apr. 20, 19VV St. Lucie Nuclear Plant Kerite FR/FR Control Cable Other supportive aging data (not referenced in EM 178-A) and not included because of their proprietary nature but available for audit at our plant or upon visitation to other offices are: Four Hour Overload Cycles - Kerite - October 9, 1956 Overload Cycling Test - Kerite - January 23, 195V Product Evaluation Test 49 - Kerite - April 18, 1962 - Oven Aging f High Voltage Lab Test Sheet No '; 3 - Oven Aging Dielectric Strength - PB Insulation.- D, ~ 19VO
-Physical Test Sheet -,,K~+ n Aged - 4 Years at 52 C-April 19, 1962 Miscellaneous Te - -. ~i ite - February 13, 19V3 QL Water Immersion~ ~
~~2 The subject of wate -'ersion is covered in Kerite Engineering Memoran-dum No. 205 (Bef. 5), with the supporting data avaQable for audit at The Kerite Company.~~
. As with the thermal aging, an evaluation technique has been developed by .;: Kerite to compare our later materials to Kerite with its time proven service record. The data developed shows PB insulation capable of operating at a tem-perature level 22 C higher than Kerite. Again, 'this corresponds with the 90 C conductor rating assigned to the PR insulation. Summary sheets (attached to EM 205) cover the points used to develop the plots.
~~In summary, the monitoring of the effect of water on electrical properties showed the XB was the most useful parameter in terms of comparing the later~~
~~...compounds with Kerite insulation. The rate of change of insulation resistance rather than the absolute value of insulation resistance is used. The data~~
! plotted on the charts, unless noted in the summary sheet, is for immersed con-ductors continuously energized at 600 volt AC. A multiconductor i PB/FB control. cable of the type recommended for St..- Lucie i has been immersed in 90 C water with 600 volt DC excitation between
~
conductors and ground (water) for over 39 weeks. Comparison of performance of multiconductor constructions to individual conductors immersed in water
I,,the kerite company April 20, 1977 St. Lucie Nuclear Plant Kerite FR/FH Control Cable shows the benefit of coverings over the insulation (see summary sheet). Continuous water immersion would only normally apply to cables that are used for submarine applications although some underground ducts are almost always Qooded.. Both of these installations also give the alternate wet and dry exposures (i. e., tides, seasonal ground water levels). According to our in-formation, there are several locations in the southeastern U. S. area where Kerite submarine signal cables have been installed and operating since 1926. The Altamaha River'near Everett, Georgia, installed in 1926 and still in ex-
~~'stence in 19VO. Also Satilla River at Woodbine, Georgia; St. Mary's River, Georgia; and Trout River, Jacksonville. This service record can be veri-fied by the railroads if needed.~~
XV. Alternate %let and Dr Kerite Engineering Mew ~y. b a1so states that from our experience, alternate wet and dry is no T~ re than continuously wet and usually much less severe,. depengi0g'~ ~Pe drying temperatures and drying times. Actual supporting test de%<, s eporte8%tarch 16, 1976, ts referenced and attached (Ref. 6). ~ .
~~- ~.'ata i@8eing developed and will be avaQable for audit.~~
~~V. Radiation~~
-.. The FR insulation has been subjected to a number of radiation tests. The 'eport of April 20, 19VO was submitted originally. This report, in itself, con-tains aQ the supportive data necessary to qualify FR/FR control cables for the required total 40 years plus one year emergency integrated radiation dose of 8;5 x 10~ rad, for inside the containment of the St. Lucie Plant.
However, the specific testprogram based upon Par. 2.3.3 of IEEE 383-1974 including pre-aging by The Kerite Company for 101 hours at 150 C, gam-ma irradiation of 50 megarads, and then the electrical integrity verified by iR measurements and high potential withstand tests is covered in Report F-C4020-3 prepared by the Franklin Institute Research Laboratories, entitled "Test of Electrical Cables Under Exposure to Gamma Radiation". The Franklin test was done on single conductor No. 12 AWG, 50 mils FR insulation without the benefit of any outer jackets or coverings (a more severe test). The cable successfully met the requirements. r
EPT P" I 0 ENGINEERING MEMORANDUMNO. 205 November 6, 1975 Supersedes ugus , ~Issue Chart redrawn Oct. 22. 1976 DETERMINING TEMPERATURE 'RATING'F CABLES FOR OPERATION IN ALTERNATE WET AND DRY LOCATIONS Temperature 'rating'f cables for alternate wet and dry 'a locations is estab-lished utilizing the Arrhenius techniques but incorporating reference material
~~'o relate actual field performance of cables to the higher temperature continu-ous water soak data on small wire.. This relationship is then used to predict the~~
. 'water aging'f materials in field service that do not hav'e an extended operating history. Continuous immersion is more severe than alternate wet and dry condi-tions, and a relationship between these 'aging'at s is necessary. Supporting data showing periodic immersion to be no morep ~ re than continuous immer-sion on Kerite and FR insulation is found in -. g test reports or pro-grams: A.Hvizd, Jr.'s project No. ', .Cm er,.196V) - Kerit;e 15-3 Lab Test Sheet Engineering Project No.. '-..O', sulation (July 2V, 1970) R +sulation (in progress) The reference mate j ', ge o% is regear Kerite which has had an extended service history encompas '--. in excess of 100'millions of feet of many con-struction types in all environments and at conductor operating temperatures of VO to V5 C. and cable surface temperatures of 60 to 65 C. The method by which this analysis is performed is described as follows: The basis for comparison between insulations is the "insula-tion resistance". Tests have shown that this electrical para-meter is representative of aging in wet environments. Change in capacitance or dissipation factor, however, is also meas-ured. Engineering Project No. 75-40 covering the electro-endosmosis program with samples energized. with 600 volts AC, 600 volts DC, or not energized showed no significant effect on time to -1/2 IR or approximate doubling of tan delta due to electrification. One further question was whether current loading retarded or accelerated any electrical degradation. A laboratory test to answer this question (15-3 Lab Test Sheet No. 227 dated March 29, 1970)gave no indicatioq that current loading affected
~~'electr icals'.~~
Pg. ~ Having identified the relevant aging factors to be time and. water temperatures, the relationship between materials was selected to be based on the time to reach one-half of the original IR level. Other levels could have been selected; however, the 1/2 IR point was something achievable in reasonable time periods. One other requirement in this analysis is that the 'slopes'f the'aging curves be essentially parallel. This is the case with these materials; i. e.; the slopes of FR insulation and Kerite are parallel. Test points in water for Kerite were at 90, V5, and 52 C. and for FR insulation 90 and V5 C.
~~~D 15-3 Lab Book-B, Pages 1P. ~ '9q.,~~
15-3 Lab Book-B, Samol 1-,. 5",- Chemical Lab Record/ ".~ 75-118 (1971), V5-119 ' (1971), V5-8V 24 (1960), V5-123 (19V1) on this hasis~c s hav hientical 'aging'lopes are expected:-t g~ arly undeh similar environmental service conditions'aqug" ir operating temperatures for equivalent ag-ing'would therefore be relatable. Thus, from the attached chart, the performance of Kerite having a proven service record of more than forty years at insulation surface tempera-tures of 60oC, and higher, it is seen that the equivalent con-tinuous water immersion time at 60 C. to reach 1/2 IR is eighty d'a'ys. Also from the chart, for FR insulation, the water temperature required to reduce the IR to 1/2 the origi.- nal level in eighty days is 82 C. This analysis indicates that FR insulated cables may be rated 22 C. higher than Kerite.- (FR insulation is conservatively rated at 90 C. conductor temperature. )
~
In actual service, cables fully immersed in water will tend to have their surface temperatures approach the temperature of the water. Therefore, attempting to establish a temperature rating for cable (assumed to be dry) may not be as significant as determining what the environmental water temperature will be; however, this analysis provides at least a comparison between
~~' newer materials and service proven materials for general use eon-ditions.~~
~~EM 205 I j', s~~
~~.e insulated conductors or an increase ~~~
~~Theadd>>o o a~ac e t oover e thee cabled in insulation thickness shows significant improvemen o~~
~~~t pe l 15-3 Lab'Bopk-B, Simple I MCB.~~
V r It should be nootede that ssalt water immersion is less severe than tap water. Befer to Product Evaluation No. 1VV, report date d Mare arch 8 1974. insulation in 1966 has given no evidence of field t ire a lications. It Itha s b een an di b d as the insulating jacket on 1/c, 5 EV, 90 C. rated ..h;.ld.d t,'d non-s and dry applications, also without any report d service problems.
~~~e e ..bl,...,t ca~~
~ AH, Jr. /dm copies: Engineering M Sal'es Offices. eLQ~'. V.. P. Hviz, Jr. of Engineering
e EM 205 Determining Temperature Hatings of Fire Hesistant (FB) Cable for Operation in Wet Locations
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~~1 I ~ ~ ~ CONFIDENTIAL. March 16, 1976 To: F. Parsons From.'.A. Hvizd, Jr.~~
~~Subject:~~
~~St. Lucie Nuclear Alternate Wet/Dry Cycle Test - HI-VO Insulation Reference. 'Electrical Lab Report No. 599~~
;~Pur ose II lw Evaluate and compare'the 0 electrical c <<acteristics of HI-VO insulation cycled alternately bebveen 90 C. eater r>,- . temperature air to a con-trol sample continuously immersed i - - ter.
~~I Procedure~~
~~~s Prior to the period@pi-0 of +Cycled test samples from the water bath, measure insulatsch ~q,@tenue 5tisstpation factor and capacitance of all test. samples.~~
~~days in room temp:, .'ir Qp'@iud con 'sts of 3 f/3 days in 90oC. water and 3 /3 (approximately 33oC. ).~~
'am ale Descriotion ZB-1 (control) No. 14 (solid) conductor. . 030" HI-VO insulation. 'B-1 (cycled) No. 14 (solid) conductor, . 030" HI-70 insulation.
ZB-3 (control) No. 14 (solid) conductor, . 050" HI-VO insulation. ZB-3 (cycled) No. 14 (solid) conductor, . 050" HI-VO insulation. Data Elapsed Tin e - 'ays to: 50/0 of 200"/0 of 2GC/0 of Sample Reference Initial IR Initial DF Initial Capac. ZB-1 (control) 30 25. 5 29 ZB-1 (cycled) 58 '51 56 ZB-3 (control) 'r1 82 149 ZB-3 (cycled) 9V 110 p155*
*Test terminated before reaching 200~0 of initial capacitance
~~NFP Mar. 16, 1976. St. Lucio Nucl'car Besults~~
~
1) Tests on samples d to sustained water immersion are more severe than those consPti ., 1ternatesvet/dry cycles.

2) The relation~':, ween s~~les undergoing sustained water im-mersion and those ~yR." frere cycleB is influenced by sample wall thickness.
I
~~. Hvizd, r. .V. P. of Engineering BS/dm 4~~
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i;= k.;rite company 9y
~~. ST. LUCIE NUCLEAR PLANT - UNIT 2 -- "--~~
~~EBASCO SERVICES INCOHPGHATED~~
.FLOPIDA POWER AND LIGHT COIL,iPANY ~UALIFI ATION DOC UivIENT~ON FOH KEHITE FR2/FH SIGNAL AND INSTRUiVIENTATIONCABLES ISSUED . Assigned to: Ebasco Services - copy 1 WAAR Se ~S>0 The information in this documentation package is confidential and as such is not to be copied or duplicated in any way, without written permission from The Kerite Company.
3
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kerite company March 1980 St. Lucie Nuclear Plant - Unit 2 Kerite FR2/FR Si nal and Instrumentation Cables qO SECTION V. HATER IMMERSION The subject of water immersion is covered in Kerite Engineering Memorandum No. 240 entitled "Determining Temperature Rating of FR2 Insulated Cables for Operation in Het and Alternate Het and Dry Locations, dated June 6, 1979 (Ref. 6), with the supporting data available for audit at The Kerite Company. As with the thermal aging, an evaluation technique'has bee'n, developed by The Kerite Company to compare our later materials to Kerite with its time proven service record. The data developed shows FR2 insula-tion capable of operating at a temperature level 30 C higher than Kerite. Again, this corresponds with the 90 C conductor'ating assigned to the FR2 insulation. In summary, the monitoring of the effect of water on electrical properties showed the IR was the most useful parameter in terms of comparing the later compounds with Kerite insulation. The rate of change of insulation resistance rather than the absolute value of 'nsulation is for immersed conductors continuously energized at 600 Volt AC. Continuous water immersion would only normally apply to cables that are used for submarine applications although some underground ducts are almost always flooded. Both of these installations also give the alternate wet and dry exposures (i.e., tides, seasonal ground water levels). According to our information, there are several locations in the southeastern U.S. area where Kerite submarine signal cables have been installed and operating since 1926. The Altamaha River near Everett, Georgia, installed in 1926 and still in existence in 1970. Also Satilla River at Woodbine, Georgia; St. Mary's River, Georgia; Trout River, Jacksonville. This service record can be verified 'nd by the railroads if needed. SECTION V, I ALTE RNATE HET A ND DRY Kerite Engineering Memorandum No. 240 also states that from our experience; alternate wet and dry is no more severe than continuously wet and usually much less severe, depending on the drying temperatures and drying times. SECTION VI I. RADIATION, LOCA AND POST LOCA To cover the requirements'f LOCA and Post LOCA exposure for the St. Lucie Plant, Unit 2, the repor't "St. Lucie Nuclear Plant, Unit 2, LOCA gualification of Kerite 600 volt FR 2 Insulated, FR Jacketed Signal and Instrumentation Cables, dated 3/20/80, (Ref. 7) was prepared.
t
~~~ ~ ~ P ~I EVC1':!:! I',PiC, .'.ll':.li)!!ANDU'..ING. 2'!!) ~ ~ ~ ~ June 6, l"...'s~~
~ ~ DETEI<I'I"G TE:!4IPEIRATt;DE IriXTING ~g } B II O'SULATED
~~~ ~ ~ ~ ~~~
~~KEBITE CA13LES FGB GPEBAgg9ii IN %'ET AND~~
~~~ ~~~
ALTEHVATE ~YET AND DRNLO CATICNS Temperature 'ratin 'f cables for wet and alternate wet and dry locations is established utilizin the Arrhenius technique. but incorporatin a reference material (o relate actual field performance of cables to the higher temperature continuous moisture absorption tests on small insulated wires, This relation-ship is then us d to predict the 'water aging'f materials in field service that do not have an extended operating history. The reference material used is regular Kerite, which has had an extended service history encompassing u>.excess of one hundred million feet of many construction types in all en;ironments and at conductor ooerating temperatures of 70 to ".5 C and cable surface temperatures of 60 to 65 C. The method by which this analysis is performed is described as follows: The basis for comparison between insulations is the "Insulation Resistance". Tests have shown that this electrical parameter is representative of aging in wet environments. Change in capaci-tance and dissipation factor, however, is also measured. Samples energized with 600 volts AC, DC or not energized, showed no significant effect ori the electrical Parameters measured. (Ref. 1). Having identified the relevant aging factors to be time and water temperature, the relationship between materials was selected to be based on the time to reach one-half of the original IR value. Other criteria, could have been selected; however, the one-half IR point was something achievable in reasonable time periods. Test points in water for Kerite were at 90 C, 75 C and 52 0 C, and for FB II insulation (ED-72), 90oC and 75 C were used. On this basis, compounds having essentially identical 'aging'lopes are expected to age siniilarly under similar environmental service conditions and their operating temperatures for equivalent aging would therefore be relatable. Thus, from the attached chart, the perfornnnce of Kerite havin~ a proven service record of more than forty years at insulation surface temperatures of 60 0 C and higher, it is seen that the equivalent continuous water immersion time at 60"C to reach one-half IH is 1950 hours. Also, from the chart for 1 B II insulation, U>e water temperature in 1950 hours is 90 6 C. This analysis indicates that FH II insulation can be rated 30 C higllcr tlian Kerite. This material, however, is conservatively rated at 90 C. (References 1, 2, and 3.)
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\ ~ 6/6/79 aciual service, cablos fully immersed in water will tend to have their surface temperatures ap pro..ch the temperature of the water. 'Therefore, atten:ptin to establish a conductor temperature ratin for cable (assumed to be dry) may not be as significant as determining what ihe environmental water tempc:aiure will be; however this analysis provides a ~ood comparison b bveen newer materials , and service-proven inaterials for general use conditions.
P Laboratory tests to determine the effects of alternate wet and dry environments have also been conducted and indicate no significant difference b tween'"ntinuous water immersion and alternate wet and dry immersion. (Reference 4.) Laboratory References The information pres nteQ above and on the attached plot has b en based on the references given b low. The data has b en collected as part of a continuing vrater absorption pro ram and represents that which is presently availab'le. These references are available for audit at the Kerite Company in the Engineering Depart-ment. Engineering Project No. 75-40. Sample Nos. 97B, 98B, and 99B.
2. Engineering Project No. 75-40 Sample No. 94A.,

3. Chemical Lab Records, Samples 75-118 (1971), 75-119 (1971) 75-87 {1965), 52-24 (1960); 75-123 (1971).

4. Engineering Project No. 75-40, Sample Nos. ZB-29, 30 and 31.
~~Sample No. ZBA, 29, 30, and 31.~~
~~'P H. F. Smith J '.~~
~~Electrical Engineer H FS/lc Attach. cc: Book Holders~~
~~~ ~~~
t APPROVED g()~ J O. Gur(41cL'p . p of Engij>< ering g
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~~0) 000 (ED-72)~~
~ ~ ~ ~ ~ ~ ~ ~ k I ~
~~I I j ~ j j j I I The actual time to I/2 IR for FR II~~
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~~~ A data point is considered conservative ~ ~~~
~~0, 000 since the actual time to 1/2 IR would - :-. be somewhat greater, therefore indi-. t':.:~i~~
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the kerite company ST. LUCIE NUCLEAR PLANT EBASCO SERVICES INCORPORATED FLORIDA POWER AND LIGHT COMPANY U IFIC TION DOCUiiENT TION FOR KERITE HTK/PR POWER CABLES ASSIGNED TO EBASCO SERVICES The information in this documentation package is confidential and as such is not to be copied or duplicated in any @ay without written permission from The Keri,te Company. g REF:mc 5/5/vv
the teel ice company ST. LUCIE NUCLEAR PLANT Kerite HTK/FR Power Cables May 5, 197V Four-Hour Overload Cycles - Kerite - October 9, 1956 Overload Cycling Test - ICerite - January 23, 1957 Product Evaluation Test 49 - Kerite - April 18, 1962-Oven Aging Physical Test Sheet - Kerite - Oven Aged -4 C - April 19, Years't 52o 1962
~~~P Miscellaneous Tests on Kerite - February 13, 19Z3 III. Water Immersion ~,~~
The subject of water immersion is covered No. 223 (Ref. 6),with the supporting da 'tg te Engineering Memorandum r audit at The Kerite Ig Company. As with the thermal aging, @/on ech'niece has been developed by The
~~'erite Company to compa service record. 'The d<~~
~~at a temperature le ~, $"~~
',"jy 0 materials to Eerite vrith its time-proven 'er 'io "d shops HTK insulation capable of operating tiara"Rerite. Again, this corresponds with the 90 C conductor ':. ssigneP4 the HTK insulation.
3s In summary, the monito ing of the effect of water on electrical properties shovred the insulation resistance was the most useful parameter in terms of comparing the later compounds with Eerite insulation. The rate of change of insulation resistance, rather than the absolute value of insulation resistance, is used. The test results for energized samples {either at 600 volts AC or 600 volts DC) versus unenergized samples are, for practical purposes, the same. Also, the comparison of the performance of finish cables versus insulated conductors showed the benefit of the jacket. Continuous vrater immersion would only normally apply to cables that are used for submarine applications, although some underground ducts are almost alvrays Qooded. Both of these installations'also give the alternate wet and dry exposures {i. e., tides, seasonal ground mater levels). According to our information, there are several locations in the southeastern U.S. area where Kerite submarine signal cables have been installed and operating since 1926. The Altamaha River near Everett, Georgia, installed in 1926 and still in existence in 1970. Also SatQla River at Woodbine, Georgia; St. Mary s River, Georgia; and Trout River, Jacksonville. This service record can be verified by the railroads if needed.
s the kerite company ST. LUCIE NUCLEAR PLANT Kerite HTK/FR Power Cables May 5, 19VV XV. Altnerate Wet and Dr t Keiite Engineering Memo No. 223 also states that from our experience, alternate wet and dry is no more severe than continuously wet, and usually much less severe, depending on the drying temperatures and drying times. Supporting data',is available for review and audit at The Kerite Company. V.. Radiation s The HTK insulation has been subjected to a number of radiation tests and is qualified for radiation levels in excess of 200 megarads (more than twice
. the required level for St. Lucie Nuclear 'Plant, Unit 2). Supporting data is presented in the St. Lucie,Nuclear Plant, Unit No. 2, Qualification Test of Kex ite 600 Volt HTK/FR Power Cable Under Simulated Post Accident Conditions Report of 5/3/VV (Ref; V).
~~VL LOCA and Post LOCA~~
~~,To cover the requirements of LOCA and:~~
~~Unit 2, the report~~
~ "St. ': I, e St. Lucis 'Plant, ~
~~Lucis Nuclear ~gag:U 4, Qualification Test of~~
~~~ ~~~
~~Kerite 800 Volt HTK/FB Power tions," (Bef. 7) was prepar.~~
~ jigag emulated Post Accident Condi-one-year Po'st LO n req~gd, from practical time considerations, 'nTheaccelerated test cycl is "accelerated relationship" is developed from the Arrhenius aging analysis in EM 178-A or EM 178-B, using "equivalent aging @mes. " It'hould be noted that the "rate" of aging for HTK insulation is essentially identical, whether the environment is air or water.. The test profile attached to the LOCA,report shows the accelerated test cycle (also described in the report) used to encompass the entire one-year Nuclear Environment. Service Cycle Requirement given in the St. Lucie, Nuclear Plant, Unit 2 Specification.
A requirement in IEEE 323 is that equipment--in this case, cable-- perform under LOCA and Post. LOCA conditions for at least the required operating time. This information was not furnished, and the actual "margins" presented in
~~'these reports may be well beyond the applicable factors suggested in Paragraph 6.3.1.5 of IEEE 323.~~
Vll. Flame Tests ) The HTK/FR power cables meet the fire tests described in IEEE 383. Indi-vidual reports on 1/c, No. 6 (V), 600 volt, HTK insulated and FR jacketed . power cables are as follows:
h ( ~ rs ~ EM 223 ENGINEERING MEMORANDUMNO. 223 Ma 4 19VV Supersedes ZM 223 dated 8-12-V6 DETERMINING TEMPERATURE 'RATING'F HIGH TEMPERATURE KERITE INSULATED CABLES FOR OPERATION IN WET AND ALTERNATE WET/DRY LOCATIONS Temperature 'rating'f cables for wet locations is established utilizing the Arrhenius techniques but incorporating a reference material to relate actual field performance of cables to the higher temperature continuous moisture absorption on small wire. This relationship is then used to predict the
~~'water aging'f materials in field service that do not have an'extended operating history.~~
The reference material used is regular ICerite, which has had an extended service history encompassing in excess of 100 millions of feet of many con-struction tpes in all environments and at conductor operating temperatures of VO to V5 C and cable surface temperatures of,60 to 65 C. The method by . is analysis is performed is described as follows: ~ 4 The bass
~~+$ :.~Am @~'~ajar "insulation re -.agn( '- between insnlations is the y have shown that this electrical parameter environments. +ange 5Q factor,= however, l4 a~ rreasW j~~
~~e'd~gve of aging in wet~~
~~'v@F~d dissipation spies energized with 600 volts AC, 600 foMs DC, 6; - ~~~
energized, showed no significant effect on the electrical p ameters measured. HAY - 6 1977 Having identified the relevant aging factors to be time and water. temperatures, the relationship between materials was selected to be based on the time to reach one-half of the original IR level. Other levels could have been selected; however,. the 1/2 IR point was something achievable in rea-sonable time periods. One other requirement in this analysis is that the 'slopes'f the aging curves be similar, which is the case with these materials. 5 Test points in water for Kerite were at HTK insulation, 90 and V5 C. V5 and 52 C, and for
~~~ ~~~
H
I "l ENGINEEBING NEMOBANDUMNO.'23 Page 2
~ e '6T47n On this basis,. compounds having essentially identical are expected to age similarly under similar environmental 'aging'lopes service conditions and their operating temperatures for equivalent aging would therefore be relatable. Thus, from the attached chart, the performance of Kerite having a proven service record of more than forty years at insulation surface temperatures of 60 0 C and higher, it is seen that the equivalent continuous water immersion time at 60 C to reach 1/2 IR is 1950 hours. Also, from the chart, for HTK insulation, the water temperature required to reduce the IB to 1/2 the original level in,1950 hours is 92 C.
This analysis indicates that HTK insulated cables may be rated 32 C higher surface temperature than Kerite. HTK insulation,, however, is conservatively rated at 90 C conductor temperature. In actual service, cables fully immersed in water will tend to have their surface temperature i=,'p> roach the temperature of the water. Therefore, attempting to estadAiY! J" 'uctor te'mperature rating for cable (assumed to be dry) may notMpgg'qtt,. t as determining what the environmental water temperature wiII'4>> ',gg's analysis provides a good comparison between newer materials an 4 - 'c~ materials for general use conditions. Laboratory tests to determine the effects o ~~'u ate wet and dry environmen s have aiso bee~ conducted and indi'cate that cont "us water immersion is more severe. 4 The addition of a jacket over the individual or the cabled insulated conductors or an increase. in insulation thickness shows significant improvement of IB performance. ~ It should also be noted that salt water immersion is less 5 severe than tap water. 4 I Laborator Beferences The information presented above and on the attached plot has been based on the I references given below. The data has been collected as part of a continuing water absorption program and represents that which is presently available. Engineering Project'No. V5-40. Sample Nos. 22A, 23A, 24A, 228, 238, and 248.
~~2. Engineering Pr'oject No. V5-40. Sample Nos. 22A, 228, 23A, 238, 24A, and 248. Chemical Lab Becords 24, V5-2V, V5-186, V5-202, 75-203, 75-204, and V5-205.~~
(Continued)
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~~~ ~~~
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~~~ ~~~
~~ENGlNEERING MEMORANDUMNO; 223 Page ~ e ~ ~ 3'jv]~~~
~~~ ~~~
~~(Laboratory References - Contin@ed)~~
~~~ ~ j, ~ ~~~
~~S. Engineering Project No. V5-40. Sample Nos. 2MCB, 19A-24A, 1 9B 24B~ VA 8A~ 9A 85A 86Aj 87A~ VBf 8Bp 85Bp 8 6Bp 8VB~~
4. Product Evaluation No. 1VV.

5. Engineering Project No. 75-40, Sample Nos. B-12 - ZB-16.
~,
~~4 '4 R.. mith, Jr. Assistant Electrical Engineer RFSSr:mc APPROVED: Care Chi Electrical Eng e~~
~~~co ies: Book Holders~~
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Cha ter 8.3 SER Item - Isolation Devices The St Lucie 2 design utiliies circuit breakers, fuses, and CT's (4.16KV system on3y) as isolation devices for power and contxol (120,125VDC) circuits This is in accordance with the PSAR, section 8.3 w'hich states: The design ~<ll comply arith the intent of Regulatory GuMe 1.75 (RG 1.7S) and complies in toto with one mception, i.e., the design includes fault current interrupting devices which serve an isolation function. This is required to preserve to the extent practicable the duplicity I of Units 6 Z. CompLiance with this regulatory guide will result in modification of the cable tray system cr the RTGB, or some combination thereof. The exact modi-fications will be delineated during the detailed design. Circuit intorrL>pting devices actuated by fault current (fuses, circuit breakers) are commonly used as isolating devices. Once actuated these devices pre~ant the faulted circuit from influencing the unfaulted circuit in an unacceptable manner. Thus, the design is both compatible llith the duplication concept and is responsive to the intent of RG l.75. The definition of an Isolation device was clearly identified as well & PBAR Section 8.3.1.2.3 c)8 which states: A dw<ce in a circuit which prevents malfunctions in one section of a circuit from causing unacceptable influences in other sections of the circuit or other circuits. Class XE circuit interrupting devices actuated by fault current are considered to be isolation devices. Although this approach was accepted by the HRC as evidence by issue of a con-struction pendt based on the E'SAR, the St Lucia 2 electrical design was en-hanced such that:
a. All cables downstream from isolation devices are fu11y qualified to IEEE 383.

b. A11 cables downstream from isolation devices are subject to the. saic cable dexating, raceway as that af class IE caMe.
~~~~fill flame retardance and splicing restrictions~~~~
~~~~c. The isolation devices axe qualified to the same level of qualification as Hnas XR cixcuit breakers and fuses.~~~~
Other than unique i6ent&icertion and not routing these cables in class IE or associated raceway, the requirements discussed $ a a and b above are in accor-dance with RC 1.15 R 0 paragraph 4.5a for associated circuits. Xt must be considered that all non-class lE cable that share the same raceway as those down stream of isolation devices are of the same quality f..e. envixon-csentally qualified; quality assurance requirements etc as that of cXass IE cables. It must also be considexed that for the 4.16KV and 4SOV systems, single line to ground faults as suggested in RG 1.75 Rev 2 Section C vill not cause the feature tripping of any bus breaker or back up breaker since these systems high xesistance grounding which limits the fault current to 10 to 15 amps, and fs insufficient to trip a bus breaker. Pox 125V DC power circuits, fuses in each leg or n~ pole thermal magnetic breakers assure that in the unlikely event that both lines axe faulted together two interrupting means are provided.
Furthermore, since these cables are routed only with cables that. are <<jualified to IEEE 383 as are the safety related- cables, the 1&ely hood of a cable fire in a non safety cable tray where a cable dcnmstrecm of an iaolation device is routed is no greater than in a safety tray. Should this cable fire cause a three phase fault and the quaXified isolation device fail to clear the fault, it would be considered as a single failure as wouM be the sam event on a safety cable. V Qased upon the previous acceptability by NRC and recognizing the enhancements to the St Xucie 2 design as discussed above, we consider the use of circuit breakers fuses, and CT's as acceptable isolation devices.
0 Tests are per5iodically perform on the onsite safety related electrical dis-tribution system in accordance with the Technical Specifications. Since the two electrical onsite distribution systems (divisions) are physically and electrically redundant and independent, either of the divisions can be tested while the other loadgroup provides power. The diesel generators are tested at 1east once per 31 days and at 18 month intervals during shutdown. During these tests portions of the onsite distri-bution system is also exercised to assure that each safety division is ia a ready state to perfoxm izs intended function. Fox further description of these tests, see the appropriate technical specification. The 4.16KV nndervoltage relays can also be tested through test circuitry provided at the 4.16KV safety related swLtchgear. I Various safety related equipmcnr. is also tested and/or monisored as per the 'ecoaanendations of the equipment. amnufacturer. The CESAR will be rcvfsed accordingly.
Chapter S.3 SEE - YOV Thermal Overload Bypass
'.'afety related 480V motor operated valves that are required to be manually operated during a deoign bases event will have their thermal overload p'o-tection bypassed. This is conoistant with the RG 1.106 "Thexmal Overload Protection for Hlectric Hotoro on HotorOperated Valves.
Por safety related 480V motor operated valves inside the containment, manual or automatic, starter thermal overloads are bypassed. However, because the valve operators are located inside containment., and containment integrity must be maintained at all times, thermal magnetic breakers are utilized in the feeder circuits for these valves. -These-circuit--breakers-are sized such that-they-will-trip between-10 and-20 secceds-of valve-lock rotor-time-;-so ao mot to damage-the. penetration-integrity.
POST ACCIDENT SAMPLING SYSTEM Re: P.A.S.S, Meeting between FPL and NRC S/ll/81 Pursuant to the above referenced meeting, Flordia Power & Light is providing the following concerning the Post Accident Sampling System:
1. Section 9.3.6 is revised to include diluted and undiluted samples. (see attached),

2. Section 9.3.6.2 is "revised by deleting the world "chloride" from 4th paragraph of 9.3.6.2 and from 3rd sentence prior to INSERT "B" (see attached),

3. INSERT "A" of Section 9.3.6.2 is revised to;.include monitoring dissolved 02 in accordance with Regula-tory Guide 1.97, revision 2 (see attached),

4. INSERT "B" of Section 9.3.6.2 will be revised to include a narrative description of how sample activity is correlated to core relative damage.
~~~~Revision to this section will be formally trans-mitted on .or before 9/1/Sl,~~~~
~~~~5. , Flordia Power & Light will also provide the fol-lowing items 4 months prior to 5% power operating license:~~~~
a) An instrument and analysis appliacability test for the acciden't environment, b) Procedures which correlates isotopic con-centration with degrees of core damage.
9.3.6
~ ~ POST ACCIDENT SAt'IPLING SYSTEN ~ .The Post Accident Sampling System (PASS) consists of a shielded skid-mounted sample station and a remotely located control panel.
The PASS provides a means to obtain and analyze pressurized and unpressurized reactor coolant samples .. containment building samples, diluted a'nd u'ndilht8d sampl'es. The Piping and Instrumentation diagrams for the PASS are shown in Figures 9.2.6a and 9.3.6b. Design data is provided in Tables 9.3.10, 9.3.11 and 9.3.12. 9.3.6.1 ~Bi 6 The PASS is designed in accordance with the criteria stated in Section II.B.3 of Enclosure 3 to NUREG 0737. The quantitative design criteria for the PASS are as follows: a) The PASS provides a means to promptly obtain a reactor coolant liquid, containment building sump liquid, and containment build-ing gas samples. The combined time required for sampling and analysis is less than three hours. b) The PASS allows for post-accident sampling with resulting per-sonnel radiation exposure not exceeding the criteria of GDC 19 (Appendix A to 10 CFR Part 50). c) The PASS is capable of accomnodating an initial reactor'coolant radiochemistry spectrum corresponding to a postulated release equivalent to that assumed in Regulatory Guide 1.4, Assumptions Used'or Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Hater Reactors, Rev. 2 dated June, 1974, and Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Ac-cident, Rev. 2 dated September 1976. d) The PASS provides a means to remotely quantify pH and the concen-trations of total dissolved gas, hydrogen, oxygen and boron in the liquid samples.
e) Sample flow is returned to the containment to preclude un-necessary contamination of other auxiliary systems and to ensure that radioactive waste remains isolated within the containment. f) Components and piping are designed to equality Group D (a' defined in Regulatory Guide 1.26) non-seismic requirements. The equipment is located downstream of double isolation valves from safety code systems. 9.3.6.2 S stem Descri tion
. The requirements for post-accident sampling of the reactor coolant and containment building atmosphere are met through the Post-Acci-dent Sampling System (PASS). The PASS provides a means to obtain pressurized and unpressurized reactor coolant samples and containment building atmosphere samples. A reactor coolant sample can be drawn directly from the Reactor Coolant System (RCS) whenever the'CS pressure is between 200 psig and 2485 psig. RCS sample lines are provided with orifices inside containment so as to limit the flow from any postulated break in the sample line.
At pressures below 200 psig, reactor coolant samples can be drawn from a Safeguards System sample line, This pathway also provides a means of sampling the containment building sump during the recirculation mode of Safeguajds system. operatio'n. A containment building atmosphere sample can be drawn with containment building pressure between 10 psia and 75 psia. All sample flow is returned to the containment building to oreclude unnecessary contamination of other auxiliary systems and to ensure that high level waste re-mains isolated within the containment. These sample process path-ways were selected to insure a representative sample under all modes of decay heat removal, The PASS samplinq,flow rates are provided in Table 9.3. 10.
0 t 4
e The PASS consists of a remotely located control panel and a skid-mounted sample station which are designed to maintain radiation exposures to plant personnel as low as 'reasonably achievable (ALARA) and which is located to minimize the length of sample lines. The PASS is interfaced with the
~
~~~~existing reactor coolant and safeguards system sample lines.~~~~
~~~~. Post accident sampling does not require an isolated auxiliary system to be placed in operation.~~~~
The PASS is a totally closed system (i.e., samples taken from are returned to the containment). The grab samples'ontainment are extracted from sample vessels by injection of a syringe through a.septum plug mounted in the vessels. In addition, the PASS sample station skid is provided with a ventilation flowpath that is sized for 333 scfm in air flow from the surrounding room to the ventilation system exhaust. The exhaust air is directed through an activated charcoal filter for iodine removal. The PASS provides the capability for remote chemical analyses of the reactor coolant including total dissolved gas concen-tration, dissolved hydrogen and oxygen concentration", boron concentration and pH. Reactor coolant analysis is provided through the use of. an undiluted grab sample facility. Shielded grab samples of the depressurized undiluted reactor coolant liquid may be obtained. Unshielded, depressurized and diluted grab samples of the degassed reactor coolant liquid, reactor coolant dissolved gas and containment building atmosphere may also be obtained. The operation of the PASS for collecting and analyzing reactor coolant and containment building atmosphere samples may be ca'tegorized as (1) reactor coolant sample purging, (2) reactor
coolant sample gaseous analyses and dilution, (4) undiluted liquid grab sample collection, (5) containment building atmosphere sample purging and dilution, and (6) system flush-
~~~~'ing. An operation description for these categories is provided - below:~~~~
Reactoi coolant sample purging is accomplised by directing the I sample flow through the system isolation valves, the sample vessel/heat exchanger, the pressure reducing throttle valve, and out to the containment building sump. At reactor. coolant pressures of less than 200 psig the containment sump sample flow is purged in the same manner using the safeguards pump discharge connection. Reactor coolant gaseous analysis is performed on a pressurized sample which is collected by isolating the sample vessel/heat exchanger. Total dissolved gas concentration is determined by . degassing the sample. This is accomplished by depressurization and circulation by alternate operation of the burette isolation valve and the sampl'e-circulatio'n pump.- The resulting displace-ment of liquid into. the burette is used to calculate the dis-solved gas concentration; The collected gases, which have been stripped from the liquid, are then di rected through a float valve for moisture separation and"circulated through hydrogen and oxygen analyzers. After recording the hydrogen and oxygen gas
,concentrations, the gas sar>pie vessel, which contains nitrogen, may. be placed on line to dilute the gas volume. This dilution operation reduces the radiation levels such that local samples can be drawn from the gas sample vessel; if desired, by injec-tion of a syringe through a septum plug mounted in the vessel.
Prior to sample withdrawal, additional dilution, which may be necessary for this quantification, may be performed by further nitrogen addition, circulation and venting.
Reactor coolant liquid analyses is accomplished by reinitiattng and directing the, sample flow through the in-line chemistry analysis equipment. The gas residence chamber and float valve downstream of the thrott')e valve allows for automatic venting of gases coming out of solution. This venting is required in order to prevent gas bubble interference with flow rate 'and chemistry measurements in the downstream instrumentation. Boron and pH readings are obtained from the in-line instrumen-tation. A small fixed volume of depressurized liquid sample (collected in a four-way valve) is then drained to the depres-
~~~~.surized liquid sample vessel and a sample is withdrawn in the same manner as described above for the gas sample.~~~~
An undiluted liquid grab, sample for chloride analysis can be collected by directing reactor coolant purge flow through the undiluted depressurized liquid sample vessel. This vessel is provided with a lead shielded container and cart for transfer of sampl'e to the analysis location. The isolation valves for the vessel are provided with stem extensions penetrating the shielding.
~(Vs c'g-7 p gP Containmeni building atmosphere sampling is initiated by opening the containment isolation valves and by using the containment sample pump to purge" the air sample through the system. Purge f')ow is directed back to containment. A sample is manually withdrawn from the containment sample vessel contain-'ng nitrogen. The initial nitrogen volume dilutes the sample to levels 'acceptable for. withdrawal. A containment'air sample may then be withdrawn from the containment sample vessel in the same manner as described previously for the reactor coolant samples.
System flushing of the liquid and gaseous portions is accom-plised by purging with demineralized water and nitrogen, respec-tively, to reduce personnel exposure during withdrawal of the diluted samples and to reduce contamination plateout between P samples.
INSERT A In accordance with item II.B.3 of NUREG-0737 (pg. 3-67, item 4) PASS has 'he capability to monitor total dissolved gases and H2 concentration. The capability of'onitoring dissolved 02 will be in accordance with Regulatory Guide 1.97 rev. 2.
Radionuclide ana')yses are performed on grab samples. These samples are counted in standard radionuclide counting equipment. Grab sample techniques are utilized for analysis. Backuo boron analysis is performed using atomic absorption techniq~es. Containment hydrogen analyzers are described in Subsection 6.2.5. 9.3.6.3 Com onent Description The major PASS components are described in this section. The principal component data summary including design code is pro-vided in Table 9.3.11.
~~~~l. Sam le Station The sample station is a free-standing skid-mounted enclosure.~~~~
The enclosure contai ns the piping, valves, components and in-strumentation necessary to provide the sampling and analysis capability. The enclosure is provided with louvers sized to pass up to 333 scfm from the surrounding room to the ventila-tion system suction connection 'in the upper portion of the enclosure. This air flow precludes any possible buildup of radioactive or hydrogen gas and provides for removal of heat generated by internal components. The enclosure is provided with removable panels on all four sides to ensure accessibility for maintenance.
2. Sam le Circulation Pum The sample circulation pump is a peristaltic type postitive displacement pump. This pump is capable of pumping liquids and/or gases. The pump will be used in the tota'I gas, hydrogen, and oxygen gas analyses operations to strip the gases out of solution in the sample fluid and circulate them through the hydrogen and oxygen analyzers. '
INSERT 8 The estimation of core damage will be done by analyzing gas sample activity (samples from the RCS loop), sump activity, and containment air activity. The total curies availab'le will be determined and'elated to total activity in the "core (based on Chapter 15 data and plant shielding studies) to deter-mine what 'A of the total core activity was released.
3. Sur e Vessel Pum The surge vessel pump is a progressing cavity (helical) pump. The pump is used to pump down the surge vessel con-tents to the containment building sump and is also used, in the calibration operation of the pH in the liquid sample line.

4. Containment Sam le Pum The containment sample pump is a vacuum pump/compressor unit that operates as a positive displacement compressor using a stainless'teel diaphram. The pump is used to collect a containment atmosphere sample and to dilute'he sample via circulation through the containment sample vessel.

5. Gas Sam le Vessel The gas sample vessel is a 12,000 ml sample vess'el initially filled with nitrogen gas. The vessel supplies the gas analysis loop with nitrogen gas to dilute the radioactive gases present in the sample line. The vessel is equipped with a septum plug which allows the operator to withdraw a diluted gaseous sample with a syringe for radiological analysis.

6. De ressurized Li uid Sam le Vessel
~~~~'I ~~~~~
The depressurized liquid sample vessel is a 12,000 mll sample vessel. This vessel col'Iects a liquid sample trapped in the four-way valve located above the sample vessel. The vessel is partially filled with demineralized ~ater before the sample is drained into the vessel. Additional demineralized water is then added to obtain the proper dilution factor so that a liquid sample
~~~~'can be withdrawn for radiological analysis. This vessel is. equip-ped with a septum plug for sample withdrawal using a syringe.~~~~
~~~~7. Containment Sam le Vessel The containment sample vessel is a 12,000 ml sample vessel that is initially filled with nitrogen gas for dilution.~~~~
The containment sample pump draws a sample from con'tainment and ci rculates it through the sample vessel where the ni'trogen gas dilutes the sample so that it can be withdrawn for radio-logical analysis. This vessel iq equipped with a septum plug for sample withdrawal.
8. ~5V The surge vessel has a 10 ga]ion capacity and serves as a vent and drain tank for the depressurized liquid sample vessel and the total gas analysis burette. This vessel can also be filled with buffer solution used to calibrate the in-line pH meter.

9. Sam le Vessel/Heat Exchan er
. The sample vessel/heat exchanger is a vertically mounted, shell and tube type heat exchanger. The heat. exchanger uses component cooling water to cool the reactor coolant, sample flow from,a maximum RCS temperature of 650o to 120oF to allow low temperature sample analysis. The tube side of the heat exchanger serves as a sample vessel for collection of a pressurized reactor coolant sample.
~~~~10. Stainless Steel Burette The stainless steel burette has a 1,000 ml capacity. The burette is used to determine the amount of total gas present in the sample~~~~
/
~~~~fluid by measuring a difference in the fluid level, of the burette upon degassification of -the pressurized reactor coolant sample.~~~~
~~~~11. Strainer The. strainer is designed to remove insoluble particles which cause sample station chemistry instrumentation to become~~~~
~~~~'ay plugged. The strainer can be backflushed with demineralized water remotely by operation of valves at the control pan'el.~~~~
12. Grab Sam 1 e Faci1 it The grab sample facility is designed to obtain a 75 cc undiluted sample of reactor coolant liquid. The facility consisted of a lead shielded sample'vessel and valves mounted on a cart for transport within the plant. The facility is manually operated.

13. Gas Residence Chamber The gas residence chamber is a horizontally mounted lead shielded baffled cylindrical vessel. The chamber is 'used to remove undissolved gases from reactor coolant samples to prevent interference with the in-line process monitors.

14. Charcoal Exhaust Filter The charcoal filter is designed to remove radioactive iodine and particulate 'material from the enclosure ventilation exhaust.
The filter is mounted in a separate housing located on top of the sample skid enclosure. 9.3.6.4 Instrumentation and Control Descri tion The major PASS instruments and controls are described in this section. The on-line process monitor data is provided in Table 9.3.12. I Control Panel The panel .is designed to meet NEYiA-12 requi rements. All sample system non-code isolation valves and pumps are con-trolled from this panel. Indication of all process para-meters and chemistry readouts are displayed on the panel. To facilitate system and~operability all controls and indi-cations are arranged in a mimic-of the system. All process pumps and valves are equipped with hand switches at the control panel.
e
5. ~55 The containment building atmosphere sample piping is heat traced to limit plateout of radioiodine and condensation of containment atmosphere vapor. The heat tracing ensures a representative gas sample.

3. Boron Heter The Boron Heter is a specific gravity measuring device which determines and remotely indicates the concentration of boron present in the liquid sample.

4. ~HNe te r The pH meter determines and remotely indicates pH in the liquid sample.

5. ~il 5 ~A
~~~~. The hydrogen analyzer is a thermal conductivity device that determines and remotely indicates the volume percent of hy-drogen in the gas stripped from the reactor coolant.~~~~
~~~~5. ~DA 'I The oxygen analyzer is a paramagnetic device that determines and remotely indicates the volume percent of oxygen in the gas stripped from the rea'ctor"coolant.~~~~
9.3.6.5 S stem Evaluation The location of the post-accident reactor coolant and containment atmosphere sampling system are in an area of relatively low post-accident background radiation. This ensures compliance with the personnel exposure limits of NUREG 0737 during sampling and . analysis. Additional plant shielding along with selective routing of interconnecting piping to the existing sampling system ensures. that (1) the exposure limits for personnel are not exceeded and (2) the on-site radiochemistry analysis equipment is available for
t
\
post-accident sample analyses. The sample station is also physical-ly separated from safety related equipment such that failure of the associated non-seismic equipment does not cause damage to the II safety related equipment. Cooling water to the reactor coolant sampling system is available during post-accident conditions to enable low temperature sample analyses. Overrides are also available to enable opening of con-tainment isolation valves following a CIAS so that post-accident sampling can be accomplished. Control for the reactor coolant . sampling system return containment. isolation valve is provided in the control room. An interlock is provided to ensure that this valve and the containment sump isolation valve is open before the system inlet isolation valve is open.
*As much as practicable, reactor coolant sampling system connecting piping is pitched downward at least 1Q degrees to prevent settling or separation of solids contained by the sample. Traps and pockets in which condensate or crud.may settle'..are avoided since they may be partially emptied with changes in flow conditions and may result in sample contamination.
9.3.6.6 Testin and Ins ection The sample station skid and control panel are equipped wi th doors for testing and inspection during normal operations. The samp'je station is provided with removable panels on all'our sides for inspection. Each component is tested and inspected prior to in-stallation in the sample system. Instruments are calibrated during initial system installation. Automatic controls are tested for actuation at the proper setpoints'. The system is, operated and tested upon installation with regard to flow paths, flow capacity and mechanical operability.
Period calibration is performed according to the schedule provided in Table 9.3.13. The PASS is designed to function for six months under post-accident conditions without recali-bration. System operability will be tested at a frequency mini-mum of six months, coinciding with the required six-month Emergency Plan sampling exercise. Such operating tests will check the functioni ng of all aspects of the system. ,9.3.6.7 . 0 erator Trainin All FP8L Chemistry Department technicians will be trained both in the classroom and in actual hands-on operations, as a function of the Chemistry Department training program. Operating proce-dures will be developed and they will be consistent with the recom-mendations of the PASS supplier (Combustion Engineering).
Table 9.3.10 Post-Accident Sam lin S stem Flow Rates Nominal Source Flow Reactor Coolant Hot Leg 0.2 - 1.0 gpm Containment Building Sump 0.2 - 1.0 gpm Containment Atmosphere 0.2 cfm
0 Table 9.3.11 Desi n Data for Post-Accident Sam lin S stem Com onents
1. .Sam le Circulation Pum Type Peristaltic Positive Displacement Flui d Post-Accident Reactor Coolant Suction Pressure (max) 5 Temperature (max) oF psig'uction 160 Rated Flow, gpm 1 Rated Head, ft 50 Code Non-Code

2. Sur e Vessel Pum Type Positive Displacement Fluid Post-Accident Reactor Coolant Suction Pressure (max) psig 5
~
Suction Temperature (max) oF 160 Rated Flow, gpm Rated Head, ft 1'85 Code Non-Code t
3. Containment Sam le Pum Type Vacuum Pump/Compressor Fluid Post-Accident Containment Atmosphere Suction Pressure (max) psia 10-75 Suction Temperature (max) oF 300 Rated Flow, cfm 0.2 Maximum Discharge Pressure, psig 95 Code -%on-Code

4. Sam le Vessel/Heat Exchan er Type Shell (cooling); Tube (sample flow)
Tube Sides: Fluid Post Accident Reactor Coolant Piping Design Pressure (max) psig 2485 Inlet Temperature (min/max) oF 120/650 Shell Side:
~
Fluid Component Cooling Mater Piping Design Pressure, psig 150 Inlet Temperature (min/max) oF 65/120 Flow (max) gpm 30 Code Non-Code
Table 9.3.l.l (cont'd) Desi n Data for Post-Accident Samplin S stem Com onents
5. Depressurized Li uid Sam le Vessel Internal Volume', cc 12000 ml Design Pressure, psig 50 Design Temperature, oF 200 Operational Pressure, psig 5 Operational Temperature, F 120 Haterial Stainless Steel 316L Fluid Post-Accident Reactor Coolant'
~~~~'ode Non-Code~~~~
~~~~6. Gas Sample Vessel~~~~
~~~~-- Internal Volume, cc 12000 Design Pressure, psig 50 Desion Temperature, oF 200 .~~~~
Operational Pressure, psig 5 Operational Temperature, oF 120 Haterial Stainless Steel 316L Fluid N2, H2, 02, Fission Products Code Non-Code
~~~~7. Containment Sample Vessel Internal Volume, cc 12000~~~~
~~~~.Design Pressure, ps.ig 50 Design Temperature, F 100 Operational Pressure, psig ~ ..Q.to 20.~~~~
Operational Temperature, oF Haterial 275'tainless Steel 316L Fluid Steam, Air, H2, Fission Products Code Non-'Code
~~~~8. Sur e Vessel Internal Volume, gal. 10 Design Pressure, psi~ 100 Design Temperature, F 200 .~~~~
Operational Pressure, psi~ 5 Operational Temperature, F 120 Haterial Stainless Steel 316L Fluid Post-Accident Reactor Coolant Code Non-Code
Table 9.3.11 (cont'd) Desi n Data for Post-Accident Sampling S stem Com onents
9. Surette Internal Volume, cc 1000 Design Pressure, psig 100 Design Temperature, F 200 Operational Pressure, psig 5 Operational Temperature, oF 120 Haterial Stainless Steel 316L Fluid Post-Accident Reactor Coolant Code Non-Code

10. Strainer Type "Y" Type Hesh Particle Size Retention 250 Hicrons Operating Pressure, psig 2235 Operating Temperature, oF 621,
~
Design Flow, gpm 2 Operating Flow (max) gpm 1 Clean aP (psig 8 gpm) 281 Loaded aP (psig 8 gpm) 1081 Collapse sP (psig 8 gpm) 7081 Gas Pesidence Chamber Design Pressure, psig 130 Design Temperat'ure, oF 350 Operational Pressure, psig 80" Operational'emperature, .oF t20 Volume,cc 4600 Fluid Post-Accident Reactor Coolant Haterial Stainless Steel 316L
~~~~.Code Non-Code~~~~
12. Exhaust Charcoal Filter Type Replaceable Cartridge Type Element Activated Charcoal Design Flow, scfm 333 Operational Flow, scfm 250-333 Operational Pressure Atmospheric Fluid Aux. Bldg., Atmosphere Clean a,P, inches water 8 scfm ( 1'8 333 Loaded aP, inches water 8 scfm 1 8 333 Code Non-Code
Table 9.3.12 Desi n Data for Post-Accident Sam lin S stem Process Instruments Instrument Descri tion Accuracy ~Ran e ,'oron Meter Density Sensor - 100 ppm 0 to 5000 ppm pH Meter Electrode Sensor - 0.05 3 to 12 Hydrogen Analyzer Thermal Conductivity - 2% of scale 0 to 100";, Sensor 0 to 10%
+-
Oxygen Analyzer Paramagnetic Sensor 2% of scale 0 to 25%, Oto5%
Table 9.3.13 Instrument Calibration Fre uenci Component Calibration Maintenance Maintenance or Idehtification ~F ~F C1;1 t,; l Charcoal Filter as req'6 Replace filter when saturated, or when dosaoe is unacceptable (test with freon) Pumps as req'd As required Yalves l8 mos. Functionally test and repair as required Level Instruments 6 mos. Reset zero and span against known vessel levels Pressure Instruments 6 mos. Check accuracy against a standard Pressure Inst'ruments 6 mos. Check pressure setooints with alarm control In functions pH Monitor 6 mos.* Calibrate with buffer solution H~ 8 Op Meters 6 mos.-l yr Set zero and span using standard gases Boron Heter 6 mos.* Check zero, span, and temp. compensator against test boron solution and de-mineralized water Flow Meters 6 mos. Check accuracy against a standard . Panalarm 6 mos. Check alarm function
*Calibration frequency can be extended until instrument malfunctions. or gets unstable readings in a post-accident situation 7,
4 ~ \ ~ SL2-FSAR ST LUCIE FSAR Qu'estion No. In accordance with the FSAR, the St Lucie 2 design incorporates an automatic reactor trip 10 minutes after loss of the component cooling water (CCW),to the reactor coolant pumps (RCP) The 'FSAR also states that the trip is designed to IEEE 279-197l requirements. The RCP's would be tripped manually on loss of CCW. The portion of the CCW system supplying cooling water to the RCP's is not safety grade. Regarding loss of cooling to the RCP, provide the following information: a) State whether the instrumentation that alerts the operators in the control room of the cause of the reactor trip discussed above is safety grade-b) Provide test data or other information to demonstrate that the RCP's can operate without CCW flow for a period of time
~~~~,compatible with operator action to trip the RCP's.~~~~
~~~~c) Assuming the reactor is i'n hot standby with the RCP's tripped, how long will the pump seals perform their function without CCW flow?~~~~
~~~~Response~~~~
a) The reactor trip upon a loss of component cooling water to the reactor coolant pumps is not required for reactor protection. The reactor trip upon loss of component cooling water is delayed for ten (10) minutes after it reaches the preset point. .Four channels of Class IE indication of component
~~~~. cooling water total flow from all reactor coolant pumps is provided on the RTC Board.~~~~
The instrumentation that alerts the operators in the control room of the cause of the reactor trip consists of the following safety grade instruments & control devices. Safety grade isolation devices are also provided to isolate signals generated by safety grade equipment to non-safety grade station annunciators and sequence of events recorder. 410.19-1
SL2- FSAR ST LUCIE FSAR Tag No. Device Function Class Channel
~~~~1. FIS-14-15 Indicator & Indicates CCW Flow IE ma,mb,mc,md A)B)C)D Bistable from RC Pumps &~~~~
~~~~Provides RPS Trip Signal 2 ~ FF-14-15 Sq Root Signal Conditioner & IE ma,mb,mc,md A,B,C,D Extractor " Transmitter Power Supply s~~~~
3. 80XA,b',c,d 10 min; timer Alarms lov CCM flov ma,mb,mc,md instantly & Delay Reactor trip for 10 minutes

4. CS-206- Control Switch Provides testability IE ma,mb,mc,md 1,2,3,4* for Indicator Bistable 10 min.
~~~~timer~~~~
*- Includes set of safety grade test resistors.
b) San Onofre Units 2 and'3 reactor coolant pumps have been operationally tested to demonstrate s'atisfactory seal performance with seal cooling water shut off for 30 minutes with the pump operating. Based on the 30 minute operational test, i.t vas demonstrated that the seals would not lose function (i.es ) gross leakage) but the seal assemblies did require refurbishment following the test. It is the judgment of Combustion Engineering that the. RCP seals vould not lose function following a loss of power two hours in duration. Based on these test results, the simi,larity of these pumps with those of St Lucie Unit 2, and the inf ormation available to the operator (see FSAR Subsection 9.2.2.3.1), the operator is expected to have sufficient time to trip the reactor coolant pumps.. The San Onofre Units 2 and 3 pumps were also operationally tested to demonstrate satisfactory motor bearing performance with cooling water shut off and with the pump operating. The cooling water was shut off for 23 minutes and a post-test examination shoved the bearings to be in excellent condition (i.es) no observable damage) Analysis of test results
~
indicated that the pump motor could run at least 30 minutes without cooling water and remain operable. s 410. 19-2
SL2- FSAR ST LUCIE FSAR The motor bearings for the St Lucie Unit 2 pumps are of the same design as those in the above mentioned test. Therefore, acceptable performance of the St Lucie Unit 2 bearings after a loss of component cooling water was demonstrated by the test of the San Onofre pumps'n addition, there have, been two occurrences of loss of component cooling water at St Lucie Unit 1 (Licensee Event Reports 335-77-23 and 335-80-29). The pump bearings have performed satisfactorily since these incidents, indicating the'cceptable performance of the bearings after loss of component cooling water. c) Tests have been performed to simulate the loss of component cooling water..to the RCPs while at hot standby with the RCPs tripped. After approximately 50 hours at coolant'conditions of 550 F and 2250 psig, the RCP seal cartridge still performed satisfactorily with the pump idle. Some seal damage was observed during the post-test inspection; however, the maximum seal leakage during the test was only 16 gph (
~~~~Reference:~~~~
FP&L letter L-81-107, Harch 10, 1981). No FSAR change required as a result of the above responses. 410.19-3
3.4.2.3.3 Diffuser Each of the 58 ports is mounted on a 14 foot high r1ser, with a four foot inside diameter (Figure.3.4-4). To control marine growth, the 1nside wall of each riser is line8 w'1th NOFOUL, a rubber containing bis-(n-tributyltin) oxide (TBTO). TBTO release rates and its effects on biota are discussed 1n Section 3.6 and 5.3. 3.6.8.4 Bis-(n-Tributyltin) Oxide NOFOUL rubber is a neoprene rubber base with bis-(n-tributyltin) oxide, otherwise known as TBTO, d1ssolved in it. TBTO is toxic at low concentrations to barnacles, snails, tube worms, mussels,- oysters, encrusting byrozoa, algae and other fouling organisms. The antifouling propert1es of NOFOUL are, maintained by the controlled, slow release of TBTO from the rubber. At St Lucie Unit 2, the lining will be 0.5 inch thick with a five percent concentrat1on of TBTO. From estimates made by B F Goodrich(>); the following continuous release rates of TBTO from the NOFOUL liner are expected from the St Lucie plant (total area = 10,950 sq ft; discharge pipe water flow rate ~ 515,000 gpm): 1st year of operation average release rate ~ 0.039 ppb 1st ten years of operation average release rate ~ 0.025 ppb 2nd ten years of operation, average release rate ~ 0.018 ppb The TBTO is released directly to the ocean from the discharge pipe risers. These data are summarized in Table 3.6-1. 5 1.3.2.3 Effects of NOFOUL NOFOUL rubber with TBTO (bis (tri-n-butyltin) oxide) as the act1ve ingredient has been tested as an ant1foulant on coast guard buoys, sonar domes and,recreational boats. Marine paints using TBTO for 1ts antifouling properties have been commercially marketed for the last several years. TBTO is currently registered with the US EPA (Office of Pesticides) for use as an antifoulant. The proposed use of NOFOUL at St Lucie Unit 2 is consistent with the existing registration guidelines.(4) TBTO is released from the NOFOUL lining of the discharge pipe risers directly to the Atlantic Ocean at an estimated average concentration range of 0.018 to 0.039 ppb over the life of the plant (see Section 3.6.8.4) . The expected degradation pathway of TBTO in water is(5): trialkyltin form dialkyltin form monoalkyltin form 1norganic tin form tox1c) " (moderately toxic) (geaerally non toxic) 'most where each degradation product is less toxic than TBTO. TBTO is toxic at low concentrations to barnacles, snails, tube worms, mussels oysters, encrusting bryozoa, algae and other fouling organisms. Results of acute, subacute and chr'onic toxicity studies for a variety of aquatic species are presented in Table 5.3-3. The lowest concentration of TBTO reported to cause acute effects for any species tested 1s about 10 ppb (50 percent of the pink shrimp d1ed in 96 hours). For longer
0 exposure times, the lowest concentration of TBTO reported to cause death is 0.2 ppb (5 percent of the guppies died after a 30 day exposure) and 0.96 ppb (5 percent of the sheepshead m1nnows died after a 21 day exposure) 'In evaluating the potential toxicity of TBTO to biota offshore of Hutchinson Island, three environmental pathways of TBTO were cons1dered. The first case assumes all the released TBTO remains in the water phase and none is lost to the sediments or degraded to other forms. This is a worst case situation for the water phase with respect to organotin. The second case assumes that the released TBTO may associate with the sediment phase. Since TBTO is known to readily associate with organic material and sediments, this situation is considered more real1stic ~ The third case considers the impact of the inorganic tin form (which is assumed to be the eventual degradat1on product) in.the water phase. This situation assumes complete and rapid conversion to the inorganic form. These cases are considered in more detail below. The first case assumes that all released TBTO is fully mixed in the water phase, with no, loss tb t'e sediments. Under these conditions, the expected concentration of TBTO would not exceed that of the maximum discharge (0.039 ppb first year average) ~ The dilution factor, under stagnant conditions, established from hydrothermal studies at the St Lucie plant (assuming a 28oF discharge temperature rise and a 3.5 F surface temperature increase for St Lucie Unit 2) is eight for a volume of approximately one acre-foots The estimated TBTO concentration after dilution from St Lucie Unit 2 is 0 .005 ppb. Both the immediate discharge concentration and the diluted concentration are below those seen to cause acute or chronic effects in the fish species tested to date. Case two considers the potential partitioning of TBTO between suspended solids in the water column and the water at the St Lucie plant discharge s1te. Calculat1ons of relative TBTO distribut1ons have been based upon Freundlich isotherm equilibrium constant values reported by Slesinger(6) for TBTO adsorption to sediments. Utilizing a Freundlich equation constant K ~ 40 (ml/g) (conservatively based upon TBTO adsorption to sandy loam soil) and a probable maximum concentration of total suspended solids (TSS) measured at the site (see Table 2.4-5), TBTO mass distribution percentages have been calculated. The results indicate that for TSS levels of l00 ppm or less, more than 99 percent of the mass of the released TBTO will remain in the aqueous phase with less than one percent of the TBTO adsorbed onto suspended solid particles. This calculation appears conservative for the sandy sediment material characteristic of the St Lucie site. Therefore, the case two analysis is similar to,the case one situation where adverse impact is not expected to occur. Case three, the addition of inorganic tin to the water phase through degradation of TBTO, was also examined. If all the TBTO,discharged (0.039 ppb first year average discharge) is converted to inorganic aqueous tin concentration of tin would be 0.016 ppb. After dilution tin,'he (dilution factor of eight), the expected concentration is 0-002 ppb. Ambient sea water tin concentration has been reported as 0.8 ppb(7)
with a range of 0.002-0.8 ppb(8). This addition of tin to the ambient concentration is expected to have minimal impact on water quality or biota. These TBTO calculations, including the release rate calculations, are based on a discharge rate of 515,000 gpm. The rate of release of TBTO is independent of the amount of water passing through the pipe. Significant decreases in the discharge rate of 515,000 gpm will result in approximately the same quantity of TBTO released into a smaller volume of water. Thus, higher concentrations of TBTO may be expected in the discharge water at these times. Because of the improved thermal mixing properties of the St Lucie Unit 2 discharge pipeline, this pipeline will be the preferred discharge route for both St Lucie Units 1 and 2. Consequently, operation of either plant will result in normal flow rates through the St Lucie Unit 2 discharge pipeline. Several swimming areas exist hear the discharge pipeline. Although no information is available on potentially harmful affects of TBTO exposure in water, TBTO concentration level are expected to be very low at any swimming area due to the low initial release rate and the dilution that will occur through mixing in the discharge plume. In summary, TBTO release from the St Lucie Unit 2 discharge diffuser during normal operation of the plant is not expected to advers'ely affect water quality or biota as examined in the three cases described above'. The TBTO levels expected in these cases are below those seen to cause acute or chronic effects in aquatic species tested to date.
3~ Written Communication, B F Goodrich to Ebasco Services, Inc. 1981 4 ~ Written Communication, US EPA to B F Goodrich, 1976
~~~~5. Cardarelli,, N, 1977. Controlled Release Molluscicides. Environmental Management Laboratory Monograph, University of Akron, Akron, Ohio.~~~~
6~ Slesinger, A. The Safe Disposal of Organotins in Soil, 1978. in Organotin Workshop Report, M.LE Good, Editor. Sponsored by the Office of Naval Research. 7~ NOFOUL Anti-Fouling Rubber. Technical Background Document, B F Goodrich, 1980. 8~ Riley, J P and G Skirrow, 1965. Chemical Oceanography Vol l. Academic Press, New York. 9 ~ Bowen, H J M, 1979. Environmental Chemistry of the Elements. Academic Press, New York.
Sheet 1 of 2 TABLE 5.3 -3 TBTO TOXICITY Acute Studies Exposure Test Concentration
~Secies Time Condition in ppm Reference Iteterotis hemichromes 120+hr LC5p** 0.03 (1977) (3) 'ardarelli 120+hr ~Ttla ia nilotica 120+hr LD5p* . 0.03 Cardarelli (1977)( ~Tila ia nilotica, 15 days LD7p 0.045 Cardarelli (1977)(3)
Hemichromis sp 15 days LD7p 0.045 Cardarelli (1977)(3) Carassius auratus 24hr LD100 0.075 Cardarelli (1977)(3) (goldifsh) s Lebistes reticulatus 24hr LD100 0.075 Cardarelli (1977) ( (guppy) Salmo gairdneri 24hr LD100 0.028 Cardarelli (1977)( (rainbow trout) Salmo Ssirdneri 48hr LDlpp 0 02 Cardarelli (1977) ( (rainbow trout)
~~~~~Le ernie macrochirus 24hr LD5p ,0'07 Cardarelli (1977)(~~~~
~~~~(blue gill)~~~~
~~~~~Le omis macrochirus 48hr LD5p 0.0405 Cardarelli (1977) (5)~~~~
(blue gill) LD100 0.5 Cardarelli (1977)(5) (fungus) Bacillus mycoides LD100 0.1 Cardarelli (1977) ( (bacterium) Bulinus tropicus LD5p 0.01 Cardarelli (1977)(5) (snail) Bulinus contortus 10P 0.075 Cardarelli (1977) (5) (snail) Common mumm ichog 96hr LC5p 0.024 Slesinger 1979 as noted in refe'rence 7.
Sheet 2 of 2 TABLE 5.3-3 Exposure Test Concentration
~~~~~See in e Time Condition ~in n Reference Pink shrimp 96hr LC5p 0.011 Slesinger 1979 as noted in reference 7.~~~~
Piddler crabs 96hr EC5p 7.3. Slesinger 1979 as noted in reference 7. Subacute and Chronic Studies Lebistes reticulatus 30 day LC5 0.0002 Cardarelli (1977) (5) (guppy) Lebistes reticulatus 60 day LC5 0.0014 Cardarelli (1977)(5) (guppy) ep 21 day LC5p 0 00096 Slesinger 1979 as (sheepshead minnow) noted in reference 7. e 21 day LCp 0.00033 Slesinger 1979 as (sheepshead minnow) noted in reference 7. 177 day LClpp 0.0048 Slesinger 1979 as (sheepshead minnow) noted in reference 7. e ECx ~ estimated concentration which results in mortality to "x" percent of the test organisms "LDx dose, which results in mortality to "x" percent of the test organisms
~~~~"LCx concentration which results in mortality to "x" percent of the test organisms~~~~
~~~~SL2 F SAR TABLE 1.90-3 EVALUATION OF XCC DETECTION INSTRUMENTATION~~~~
. TO ATTACHMENT 1 OF IX.F.2 Item Res onse St Lucie 2 has 56 core exit thermocouples (CETs) distributed uniformly over the top of the core, .Section 3.1.3 has a description of the CET sensors, Figure 1.9B-7 dep><,
~~~~Il<< ii'44)4IL i (j'4~~~~
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For compartment (e.g., reactor cavit y, secondary shield wall area, and pressurizer area) structural design evaluations, provide the design differential pressure and discuss whether the design differential pressure is uniformaly applied to the compartment structure or whether it is spatially varied. R~es onse Design differential pressures are 24 PSI and 14 PSI for secondary shield wall, and pressurizer compartment above elevation 62-00, respectively. The pressures are uniformly applied to the compartment structures. Design differential pressures for the primary shield wall have been spat ially varied from a peak value of 86 PSI with provision for dynamic load factor. See revised FSAR Table 6.2-3 and new FSAR Figure 6.2-2~:.
K SL2"FSAR TABLE 6.2-3 PRINCIPAL CONTAINs~i.NT DESIGN PARA'.DIETERS Parameter De c~in ltargi nl Containment Internal design pressure, psig (LOCA) 44.0 (VSLS) 44.0
~
Shell surface design temperature, F 264 9 44 psig Refer to Figure 6.2"12 Differential design pressure, psid 0.70 > 6.6">> Net free volume, 10 6 ft3 2 '0 )lr t appli-cable Design leak rate, percent free volume 0.5 Nnt applicable per day at 44.0 psig Shield Euilding - ~XI E
~~~~, I.~~~~
~~~~xternal design pressure, psig C))))t )))q I)))')q)g 3.0 Q'g '), 387 Rc).~ r ~'i>< Re=i) )) vs))l /-~~~~~~
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I
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(1) - Hargin (~) desi n value eak calculated value, peak calculated value Actual margin, i.e. the margin between design values sod I ak cal-culated values when using realistic nr median parameter values would be much larger. J 6.2-86
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42
Question
.3e0 Provide analyses to determine the external forces and moments, resulting from postulated hot leg and cold leg ruptures within the reactor cavity, on reactor vessel supports.'f applicable, similar analyses should be performed for steam generator and/or pressurizer compartments that may be subject to pressurization where significant component support loads may result. For each analysis, provide the following information:
(1) Provide and justify the pipe break type, area, and location. Specify whether the pipe break was post-ulated for the evaluation of the compartment structural design, component support design, or both. R~es onse FSAR Table .6.2-13 is a summary of postulated pipe ruptures for the containment subcompartment analysis-. Contained in this table is the pipe break location, description of the break, break area and release rate data and table numbers. Pipe break locations were chosen based upon the hign stress point criteria in accordance with Reg. Guide 1.46 and SRP 3.6.2. Refer to MEB question N2 attached. The peak leads tabulated from each of. the pipe breaks listed on, Table 6.2.13 was applied to the corresponding structure. Xn al'1 cases it has'been found tnat the structural des'.gn was adequate to withstand the differential forces resulting from the break. Table 6.2-3 (see revised table contained in answer to CSB branch question I2) provides a comparison between the peak calculated force and the structural design. The major components of "the RCS are designed to withstand the forces associated with the design basis pipe breaks. These pipe thrust forces at, the break location, resultant subcompart ment differential pressurization forces and internal asymmetry hydraulic forces acting in the reactor internals. A complete description of the evaluation of the plant faulted condition for these'omponents is provided in the response to MEB quests N35 attached. The major portion of the Asymetric Analysis nas been completec-MEB question response N35. contains a table (Table 1), that
~~~~'rovides the current asymmetric loading analysis schedulein- .for botn the major components and the structural design dicating the anticipated completion date for. each item.~~~~
~ ~
It must be demonstrated that St. Lucie plant analysis system para-meters fall within the design envelope of CENPD-)68, Revision ). ~Res onse 2 The system parameters of the St. Lucie 2 plant fall within the design envelope of CENPD-)68, Revision )., See attached proposed FSAR amendment to 3.6.2.).).
3.6.2 DETEP81,'iAT10ti OF BREAK LOCATIO,iS At,'D DYtiAHIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPItJG 3.6.2.1 Criteria Used to Define Break locations for Pipe ~ ~
~~~~~Anal s>s I M~h> //~~~~
2.1.2 High Energy Piping Systems P section provides the criteria used to determine postulated piping failure locations for high energy piping systems both inside and outside /RJ containrent. He
) Reactor Coo1ant System Hain Loop Piping /=sH/2, ) A stress survey of the St. Lucie 2 Reactor Coo1ant System Hain Loop Pipint performed in accordance uith the methods described in CENPD 16BA (Reference 1)
The St. I.ucie 2,Reactor Coolant System geometries and transients were employed i the analysis. The results of this analysis are presented in Figure 3.6-4. In accordance with the criteria specified in Reference (1) circumferential type pipe breaks are postulated to occur at all terminal ends and pipe breaks are p ulated at a1) intemediate locations throughout the, piping system where the of primary plus secondary'stress intensity exceeds 2.4 Sm or the cumulative> usage factor exceeds 0.10. Where all intermediate pipe break locations would be considered unlikely because r the stresses and cumulative usage factors calculated for a particular run of piping betwe n terminal ends are everywhere less than the stress and fatigue limits stated above, the two intermediate locations of highest cumulative usage -'actor are chosen as the most likely break locations for piping runs longer than .' )0 diameters tota1 length", and for piping ruris having more than one change in direction through-out the run. >) The results presented in Figure 3.6-4 confirm the break location and types
~~~~,of Reference (1) for the main loop pipe.~~~~
>) For the partial. area guillotine type pipe breaks at the reactor inlet and outlet'o les 'and the steam generator inlet nozzles, the methods of Reference (I) were yed to calculate the f1ow areas and opening times of the break at these locations. The stiffness values are provided in Table 3.6-2 and Figure 3.6-5.
I The resultant break characteristics are shown in Table 3.6-1. The pipe whip restraint at the reactor vessel inlet is shown in Figure 3.6.3.'ll other llotine breaks have been assumed to open to full area. The break locations for RCS are shown in Figures 3.6C-2.1 and 3.6C-2.2.,
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ues uestion 35 The response of certain reactor coolant system components and their supports to postulated asymmetri0 LOCA loads needs to be addressed in accordance with NVREG-0609. ~Ra8 onse TABLE I provides the status of the evaluation of components, structures, and attachments to the RCS when sub)ected to asymmetric loads. Where the evaluation has been completed, the results have been shown acceptable.
TABLE 1: Assessment of Structures/Asymmetric Loads
~ ~ Oi Component/Structure Assessment Evaluation Reference Status Comments Basis Reactor Pressure Vessel'team Complete Plant Specific Analysis FSAR 3.9.1.4 ' Complete Generators Reactor Coolant Pumps Reactor Vessel Supports
~ Steam Generator Supports Reactor Coolant Pump Supports Biological Shield Well FSAR 6.2.1.2 Steam Gen., R C Pump FSAR 6.2.1.2 Compartment Wall RCS Main Piping Complete Plant Specific Analysis
IO TABLE 1: Assessment of Structures/Asymmetric Loads
~~~~~ ~~~~~
Component/Structure Assessment Evaluation Reference Comments Status Basis ECCS Piping -In Progress :. Plant Specific FSAR 3.9.1.4.5 Preliminary analyses Analysis predict acceptable results. FSAR Amendment Nov. 1981. ECCS Piping Supports & Restraints In Progress CEDMS In Progress FSAR 3.9.1.4.3 Reactor Internals In Progress FSAR 3.7.3.14 Analysis nearly complete FSAR 3.9.2.5 Results to date are acceptable. Fuel In Progress Analyses expected to be completed 3/82.
e i.s I i s t t I I I l ~ ~ ~
ST. LUCIE 2 CEDM ~ GEONETRY AND MOMENT CAPABILITY SIMILAR TO PALO VERDE a PIPE BREAK + SSE HEAD VELOCITIES LOWER THAN THOSE FOR PALO VERDE SINCE PALO VERDE HAS BEEN DEMONSTRATED ACCEPTABLE, ST. LUCIE 2 CEDM ARE EXPECTED TO BE DEMONSTRATED TO BE ACCEPTABLE ~ ANALYSIS IS EXPECTED TO BE COMPLETED SY SEPTEMBER 1, 1983,
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~~ST. LUCIE 2 ECCS PIPING 'PRELIMINARY CALCULATIONS INDICATE THAT LINES lA AND 1B ARE THE MOST SEVERELY LOADED ~ COMPARISON OF 'INPUT MOTIONS WITH OTHER ECCS LINES PREVIOUSLY ANALYZED INDICATE THAT~~
1) PLASTIC ANALYSIS IS REQUIRED.

2) RESULTS ARE ANTICIPATED TO DEMONSTRATE ACCEPTAB I LITY.
~ ANALYSIS. IS EXPECTED TO BE COMPLETED BY SEPTEMBER 30, 1981
A ca'fcu1atfon pf the ref'oreatfon of the St. lucfe f 2 RC p$ pfng Ken subjected to tgte aexftram rionent a11mad by Section Ill. Nb 3652 was pilaf'omed. The ,bteached mteria1 sh~s the vesu1ts of that, ca1culatkon.
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0 3.9.1.4 Consideration for the Evaluation of the Faulted Condition 3.9.1.4.1 Seismic Category I NSSS Items major components of the reactor 'coolant system (RCS) are designed to
'he withstand the forces associated with the design basis pipe breaks discussed $ n Section 3.6.2, in combination with the forces associated with the Safe Shutdown Earthquake and normal operating conditions. See Sections 3.9.1.1 and 3.9.3 r~r-discussion of loading combinations. The forces associated with, the postulated pipe breaks include pipe thrust forces at the break location, resultant subcompartment differential pressurization forces, and internal asymmetric hydraulic forces acting on the reactor internals. The pipe break thrust forces are determined by the methods discussed in Section 3.6.2. 6.1. The time and spatially dependent asymmetric hydraulic loads acting on the reactor internals are determined by the methods discussed in Section 3,9.2.5.
A dynamic non-linear time history analysis was performed to generate reactor vessel loads and motions due to the forces associated with the partial area pipe 0 breaks at the reactor inlet and outlet nozzles and the steam generator inlet nozzles (See Section3.6.2.1.1.3). The analysis used the DAGS code to perform a direct integration of the coupled equations of motion, in which the system characteristics are updated at each integration step to account for local non-linearities., These non-linearities include initial gaps and preloads at system restraints or local plastic response which may occur following a pipe break. The FORCE code post-processes DAGS response output in order to provide the loads and motions at pre-specified locations.- The analysis used a lumped parameter model including details of the reactor vessel and supports, major connected piping and components,and the reactor internals (Figures 3.9-19 through 3.9-22). This mathematical model provides a three-dimensional representation of the dynamic response of the RCS major components subjected .to the simultaneous time varying pipe break forcing functions. This aedel is defined mathematically in terms of the ICES STRUDL II computer code to develop appropriate matrices for the elements of the three-dimensional space frame model.
The results generate reactor vessel and support loads and time history motions of ACS piping at,ECCS piping juncture points, and RV shell motions at internals and CEDN support points. These motions provide input excitations.for the pipe break analyses of the reactor internals, fuel, CEAS, CEDARS and ECCS piping. The component and support loads for the Steam Generator, Reactor Coolant Pump, and Pressurizer were determined by equivalent static analyses. A load factor equal to 2.0 on the calculated thrust, jet impingement, and subcompartment pressure loads is employed to account for the dynamic response of the structure. The model employed for static analysi is shown in Figure 3.9-18 The system or subsystem analysis used to establish, or confirm, loads which are specified
~
~~~~for the design of components and supports is performed on an elastic basis.~~~~
~~~~~ ~ ~~~~~
Mhen an elastic system analysis is employed to establish the loads which act on components and supports, elastic stress analysis methods are also used $n the design calculations to evaluate the effects of the loads on the components and supports. In particular, inelastic methods such as plastic instability and limit analysis methods, as defined in Section III of the ASHE Code, are not used in conjunction with an elastic system analysis. Analyses of the reactor coolant system components (reactor vessel, steam generator, reactor coolant pump, pressuriier, . and reactor coolant piping) and their supports have been performed in accordance with the methods described above. For each component and support member, the calculated loads, in combination with the seismic loads, are below the loads specified for design, and the stresses (piping rupture 4n combination with SSE) are below those allowed by Section III of the ASNE BKPV code for Service level 0.
3.9.1.4.2 Reactor Internals See Sections 3.7.3.14 and 3.9.2.5 '.9.1.4.3 Control Element Drive Mechanisms (CEDMs) The capability of the control element drive mechanisms {CEDMs) to withstand the effects of design basis pipe breaks in combination with safe shutdown seismic (SSE) loadings is evaluated by analysis. This dynamic loading is experienced by the CEDNs via the motion of the reactor vessel head. The reactor vessel head/CEON motions due to pipe rupture and seismic loadings are calculated using the models described in section 3.9. 1.4. 1.
3.9.1.4.3. I
~~~~~ ~ ~ ~ ~ Method of Analysis ~~~~~
Previous studies on other CF plants (Reference I) have'indicated that the reactor vessel asyornetric load aspects of a hypothetical guillotine break produce motions which result in stresses which exceed the ASIDE Code'evel 0 allowable stresses for elastic calculation. Elastic plastic dynamic analyses have demonstrated for those plants that the structural. integrity Of the CEDNs is not impaired by these loadings and that the ASIDE Code Level D allowable limits for elastic plastic calculation are not exceeded. In order to demonstrate that*the integrity of the CEDHs are not impaired by pipe break and SSE loads, elastic-plastic dynamic analyses are performed. In the elastic plastic analysis, the motions of the RV are input to the finite element model of the CEDN. Moments and deformation are computed as a function of time during the event. The moment to cause plastic instability of the most severely loaded section is computed by elastic plastic static analysis. The actual moments during the dynamic event are then compared to the plastic instability moment in order to evaluate integrity. 3.9.1.4.3.2 Models,, Dynamic analysis finite element models are prepared for CEDMs near the center of the RV head and near the outer edge. The models are made up of beam type elements., Yhe model of the calculation of the plastic insta6ility load is made up of shell elements in order to consider the effects of ovalization of the cylindrical section. The nozzle at the RV head is usually the most severely loaded section. 3.9.l.4.3.3 Material Properties Recently the material properties necessary'or elastic plastic analysis have been developed by the CE Metallurigical and Materials Laboratory. Yhese properties are available for all of the materials at all of the temperatures that the CEON normally experiences.
~~~~.9.1.4.3.4 Loading. ~ ~ ~ ~ ~~~~~
The effects of pipe break and SSE are transmitted to the CEDN by the ~ motion of the reactor vessel head resulting from the analysis of Section 3:9.1.4.1..
A response spectrum is calculated for the motion of the reactor vessel head resulting from the primary system dynamic analysis for pipe break
~
loads.~ This response spectrum is combined with the SSE response spectrum ! by taking the square root of the sum of the squares (RSS) of the ordinates of the two spectra. An artificial time history of motion is then developed from the combined acceleration spectrum and used as the input to the dynamic GEOM analysis. Acceleration spectra resulting from pipe rupture at the RV inlet nozzle, the RV outlet nozzle, and at the steam generator inlet nozzle are compared in order to determine the most severe loading condition. If one loading condition can be identified as the most severe case, only that loading condition is used in the dynamic CEDH analysis. Other loadings are also used 4f they are not clearly enveloped by the most severe one. 3.9.1.4.3.5 Response The models, material properties and RY head motion history are used in the NRC finite element program fot analysis. The ANSYS program may also be used. The results of the dynamic analysis include moments, strains, stresses and deformation as a function of time. These results are presented graphically for critical regions of the CEDM. The same material properties are used in the static analysis for the plastic instability moment. 3.9.1.4.3.6 Evaluation 3.9. 1.4,3.6.1 Acceptance criteria The CEOMs are not required to operate for safe shutdown after a loss of coolant event resulting from the design basis pipe breaks. In order to comply with existing ECCS analysis methods, however, the integrity of the CEDMs must be maintained and leakage must be prevented. The ASME Boiler and Pressure Vessel Code Section III Division 1 Appendix F lists a number of cHteria which assure that the. pressure boundary will not be violated. These criteria include an instability limit for comparison to elastic plastic analysis results. The integrity of the pressure boundary is assured
~~~~$ f the applied loads do not exceed 70" of the plastic instability 1oad.~~~~
3.9.1.4.3.6.2 Evaluation of Integrity The results of each dynamic analysis are compared to the results of the static plastic instability moment analysis. Integrity of the CEDMs is azured if the, acceptance criteria are satisfied. Based on Reference (1) studies, it is expected that results of these analyses will demonstrate the integrity of the CEDMs. Results will be submitted in a November, 1981 amendment. REFERENCES
~~~~1. "Reactor Coolant System Asymmetric Loads Evaluation Program Final Report", Combustion Engineering, Inc., July 1, 1980.~~~~
~~~~3.9.1.4.4 The components not covered by the ASME Code but which are related to plant safety include: (1) fuel, (2) non pressure boundary portions of control~~~~
~
element drive mechanisms (CEDMs) and (3) control element assemblies (CEAs). Each of these components is designed in accordance with specific criteria to insure their operability as it relates to safety.
3.9,1.4.5 EMERGENCY CORE COOLING SYSTEH+ECCS) PIPING AND SUPPORTS The capability of the emergency core cooling system (ECCS) piping and supports to withstand the effects of design basis pipe breaks are evaluated by analysis. The capability of the ECCS piping and supports to withstand the combined effects of pipe ' break and safe shutdown seismic (SSE) loadings are also evaluated. Pipe rupture loadings are experienced by the ECCS piping via the motion of the primary system piping, and the SSE loadings are experienced by the ECCS piping via the motion of the primary system piping and the ECCS piping supports. The primary piping motions due to pipe rupture loadings are calculated using the models described in section 3.9.1.4.1. The seismic loadings are provided from the code stress'nalysis of the ECCS lines. 3.9.1.4.5. l Method of Ana~lsis
~~~~~ ~~~~~
~~~~Previous studies on other CE plants (Reference 1) have indicated that the motion f the ~ primary system piping at the~~~~
~
~~~~ECCS injection nozzle due to pipe rupture~~~~
~
~~~~loads contains frequencies which are in the range~~~~
~
of the natural frequencies of the ECCS piping.. The ECCS piping response, therefore, is sensitive to small geometry and input frequency changes. Because of this sensitivity the analysis of a pipe system may require either elastic or elastic plastic analysis. Each ECCS pipeline to be evaluated will be analyzed by traditional dynamic elastic analysis and evaluated according to appropriate elastic stress limits for ASHE Level B and Level 0 conditions. For pipelines where Level 0 limits are not satisfied, a detailed elastic plastic analysis to demonstrate integrity and functionability of the piping will be performed. 3.9.1.4.5.2 Nodels The elastic @namic analysis will be performed by using distributed mass models and the appropriate ECCS nozzle motion history. The NRC finite element program ill be used for the elastic dynamic analysis for pipe rupture loads. The program will determine the motion history of the ECCS pipeline and the loads in the supports by performing the time history analysis.
Elastic plastic dynamic analysis, if required, will .also be performed with the HARC finite element program. A detailed analysis of a typical pipe elbow and a typical straight section will be performed to determine the moment carrying capability, or plastic instability moment, of the elbow and pipe. This analysis also provides an elastic plastic stiffness of the elbow to be. used in the pipeline dynamic analysis. The finite element model used for the elastic plastic dynamic analysis is riade up of pipe elements with modified stiffness at elbows to incorporate the ovalization effects observed in the detailed plastic elbow analyses. The stiffness and load carrying capability of the supports input to the analysis is computed by elastic or elastic plastic analysis. 3.9.1.4.5. 3 Materi al s The material used for the 'ECCS piping is ASNE SA376 GRT316 stainless steel. 'The elastic properties required for analysis will be taken directly from the ASHE Code, The elastic plastic properties will be established by scaling stress strain data available from previous CE tests to the specified code yield and 1 ultimate stress values. 4 3.9. 1.4.5.4 ~Loadin The effects of primary, system pipe breaks are transmitted to the ECCS piping by the motion of the primary piping. For the evaluation of pipe break loads "only, the displacement time history of the primary piping (at the ECCS'injection nozzle) Mill be applied directly to each dynamic ECCS pipeline analysis. The displacement time history is obtained from a dynamic analysis of the reactor coolant system for Postulated pipe breaks at the vessel inlet, outlet nozzles and steam generator inlet nozzle. 3.9.1.4.5.5
~ ~ ~ ~Res onse The natural frequency of all ECCS pipelines will be determined. The r esults of the Primary system dynamic analysis for pipe rupture at; the reactor vessel inlet nozzle Mill be compared to the. pipeline frequencies to determine which hot leg injection
~~~~and which~~~~
~
~~~~intact cold leg injection line is loaded~~~~
~~~~~ ~~~~~
~~~~most severely. The most verely'loade4 pipelines are analyzed for cold leg pipe rupture loads.~~~~
~
~~~~The resul'ts of the primary system dynamic analysis for pipe rupture at the reactor~~~~
~~~~~ ~~~~~
vessel. outlet nozzle and steam generator inlet nozzle will also be compared to the pipelibe .frequencies. This will enable determination of the cold leg injection line which is loade4 most severely. The most severely loaded cold leg injection line and the intact hot leg'injection line will be ana1yzed for the most severe hot leg pipe rupture loads. The analyses will result in motions and stresses in the piping an4 pipe support loads. Elastic-plastic analyses will in addition, result in plastic strains and deformation in the pipe and elbows. 3.9.1.4.5.6 Evaluation 9.1.4.5.6.1 Acceptance Criteria The integrity and functionability of the ECCS piping must be demonstrated. Integrity and functionability are assured if the Level 8 (upset condition) limits of the ASME Boiler and pressure.Yessel Code Section III, Division I, are not exceeded. If the Level 8- limits are excee4ed, then Level D or faulted limits may be used to demonstrate that'ntegrity is .maintained. Functionability may be assured by demonstrating that the deformations of the piping are acceptable. 3.9.1.4.5.6.2 Evaluation of Integrity and Functionability The evaluation of the effects of pipe break loads and SSE. loads combined when both loadings produce only elastic stresses is by the comparison of the square root of the sum of the squares of the stresses caused by the two loadings with the elastic stress allowable. The elastic dynamic stress results will be compared to the Level 8 stress'imits f the ASME Code. In the event that these stress limits are not satisfied, Level limits will be compared for demonstration of integrity. If Level D elastic limits are met, functionability will be evaluated by assessing the extent of deformation of the pipe.
0 The evaluation of the effects of pipe break loads and SSE loads combined in the case where significant plasticity exists in the pipe is conducted by computing the sum of the strains due to the two loa'dings and comparing the sum to the strain at .70% of the plastic instability load. Integrity is demonstrated if the applied maximum moment is less than 70% of the plastic instability moment or correspondingly if the applied strain is less than the strain at 70K of the plastic instability moment. Functionability will be evaluated by comparing the extent of deformation at the maximum loading to the deformation required to significantly affect ECCS 1I flow. Results will be submitted in a November l981 amendment. REFERENCES
~~~~1. "Reactor Coolant System Asymmetric Loads Evaluation Program Final Report, Combustion Engineering, Inc. Duly 1, 1980.~~~~
e h l
3.S.2.5 ~Dnamic S stem Anal sis of the Reactor Internals Under Faulted Conditions ynamic analyses are performed to determine blowdown loads and structural responses of the reactor internals and fuel to postu1ated LOCA loadings and to verify the adequacy of their design. A brief description of these methods is provided below. The LOCA maximum stress intensities in the reactor internals are det'ermined using the combinations of lateral and vertical LOCA time-dependent loadings which result in maximum stress intensities. The maximum LOCA stresses and the maximum stresses resulting from the SSE are then combined using the root sum square method to obtain the total stress intensities. 3.9.2.5.1 D namic Anal sis Forcin Functions The hydrodynamic forcing functions during a postulated LOCA result from transient pressure, flow rate, and density distributions throughout the primary reactor coolant system. 3.9.2.5.1.1 H draulic Pressure Loads The transient pressure, flow rate and density distributions are computed for the subcooIed and saturated portions of the blowdown period during a LOCA. e computer code utilized is based on a node-flowpath concept in which ntrol volumes (nodes) are connected in any desired manner by flow areas (flowpaths). A complex node-flow path network is used to model the Reactor Coolant System {RCS). The modeling procedure has been compared to a large scale experimental blowdown test with excellent agreement. The laws of conservation of mass, energy and momentum along with a repre-sentation of the equation of state are solved simultaneously. The hydraulic transient of the reactor is coupled to the thermal response of the core by analytically solving .the one-dimensional radial heat conduction equation in each core node. Pre-blowdown steady state conditions in the RCS are established through the use of specified input quantities. The blowdown loads model uses a nonequilibrium critical flow correlation for computing the subcooled and saturated critical fluid discharge through the break. 3.9.2.5.1.2 ~DL d A break in the primary coolant system will result in large local pressure 'ifferences across 'various reactor vessel internal components and an accel-eration.of the local fluid velocity in various regions. The acceleration of the local fluid velocity can result in higher component drag loads than occur during steady state reactor operation.
e
~g h .9.'2.5.1.3 Core'oads he total instantaneous load across the core is given by the sumnation of the pressure and drag forces acting parallel to the flow. The loads are obtained using a control volume approach utilizing an integrated fluid momentum equation.
The drag forces are represented by the fluid shear term in this equation and consist of both frictional and form drag. 3.9.2.5;1.4 CEA Shroud Loads During normal operation, the reactor coolant flow axially through the core into the upper guide structure. Hithin the upper guide structure, the coolant flow changes direction so that it exits radially through the hot leg nozzles. During a t.OCA, the transverse flow of the coolant across the CEA shroud gives rise to loads which induce deflections in these shrouds. The transverse drag forces were determined from flow model experiments which were geometrically and 'dynamically similar to the full-scale upper guide structure design. The measured experimental model forces were scaled-up to represent the actual forces on the upper guide structure using the computed transient flow rate and density information. 3.9.2.5.1.5 Results of Blowdown Loads Anal sis alysis was performed of a postulated pipe break at the reactor vessel inlet ozzie. The transient pressure differences throughout the vessel are evaluated and used in the structural response calculation described below. The pressure difference across the core is also evaluated for the break. A postulated pipe break occurring at the reactor vessel outlet nozzle was also analyzed. The pressure difference throughout the vessel is calculated. The decompression in the annulus is syometric early in the transient because the pressure wave must tiavel through the core barrel internals to reach the lower plenum from where the wave propagates uniformly up through the downcomer. The .axial pressure diffe'rence across the core was also calculated. A postulated pipe break .occurring at the steam generator inlet nozzle was also analyzed. The pressure difference throughout the reactor vessel was calculated. The axial pressure difference across the core was also calculated.
3 9.2.5.2 Structural Res onse Anal ses I
-dynamic LOCA analyses of the reactor internals and core determine the shel beam and rigid body motions of the internals, using established computerized structural response techniques. The analyses consist basically of three partsi in the first part, the time-dependent shell response of the core support barre to the transient loading is calculated using the finite-element computer code, . ASHSD. The second part of the analysis evaluates the buckling potential of the core support barrel for hog ]eg freak conditions using the finite-element computer code, SANNSOR-DYNASOR<1'~'2>. In the third part, the nonlinear dynamic time history responses of the reactor internals and core to vertical and hor-5zontal loads resulting from hot and cold leg breaks are determined with the CESHOCK code, which is further described in Reference $ 10).
3.9.2.5.2.1 Shell Res onse of the Core Su ort Barrel A cold leg break causes a pressure transient on the core support barrel that varies circumferentially as well as longitudinally. The ASHSD finite-element computer code is used to analyze the .shell response of the CSB to the pressure transient from a cold leg break. The CSB is modeled as a series of shell elements joined at their nodal point cles as shown in Figure 3.9-1. The length of the elements in each model is cted to be a fraction of the shell attenuation length. A damped equation of.~wtion is formulated for each degree of freedom of the system. Four degrees of freedom, radial displacement, circumferential displace-ment, vertical displacement, and meridional rotation are considered in the analysis. The differential equations of motion are solved numerically using a step-by-step integration procedure. The circumferential variation of the pressure time-history is considered by representing the pressure as a Fourier expansion. The pressure at each elevation
~~~~$ n the model is determined by linear interpolation. Thus,. a complete spatial time load distribution compatible with the ASHSD computer program is.,obtained.~~~~
Each load harmonic is considered separately by ASHSD. The results for each har-nenic are then added to obtain the nodal displacements, resultant shell forces and shell stresses as a function of time. V 3.9.2.5.2.2- D namic Stabilit Anal sis of CSB 'A hot leg break causes net external radial pressure on .the coi e 'support barrel. A stability analysis of the CSB is performed using the finite-element computer code, SANNSOR-DYNASOR. The effects of an initially imperfect shape based on actual out-of-roundness measurements are included in the analysis. e CSB is modeled as a series of shell elements, es.shown in Figure 3.9-2. iffness and mass matrices for the barrel are generated utilizing the SAMMSOR pal t of the code. The equations of motion of the shell are solved in DYNASOR using the Houbolt numerical procedure.
n initial imperfection is applied to the core support barrel by means of a pseu-load for each circumferential harmonic considered. The actual pressure tran-ent loading generated by the outlet break is uniform circumferentially but varies longitudinally. The response is obtained for each of the imperfection harmonics. Appendix F, Section III of the ASME Boiler and Pressure Yessel Code requires that permissible dynamic external pressure loads be limited to 75K of the dynamic instabi15ty pressure loads, or alternately, the dynamic instability loads mvst be greater than 1.33 times the actual loads. Consequently, this analysis is repeated with the imperfection applied in the critical harmonic and the pressure loadinp is i cr ased beyond 1.33 times the actual loads in order to demonstrate the stabi iity ot tie core support barrel. 3.9.2.5.2.3 ~Di ~ 2 I 2 t I I't 2 t I
~~~~'0 t t ~~~~~
Dynamic analyses are 'performed to determine the structural response of the reactor internals to postulated asyrhnetric LOCA loading {including reactor vessel motion effects) and to verify the adequacy of their structural design. The postulated pipe breaks result in horizontal and vertical forcing functions which cause the internals to respond to both beam and shell modes. Detailed structural mathematical models of the reactor internals are developed based on the geometrical design. These models are constructed in terms of lumped sses connected by beam or bar elements, and inclvde nonlinear effects such as pacting and friction. The models are developed for input to the CESHOCK code ich solves the differential equations of motion for lumped parameter models by a direct step-by-step numerical integration procedure. The model definitions employ the procedures established in Combustion Engineering Topical Report CENPD-42 and, in additiog include hydrodynamic coupling effects and a detailed representation of thecore supportbarrel to upper guidestructure to reactor vessel interfaces. Separate models are formulated for the horizontal {fig. 3.9-3) and vertical {fig. 3.9-4) directions to more efficiently account for structural and response differences in those directions. The models for the'or'izontal directions are developed in terms of lumped masses connected by beam elements. The stiffness values for the beam elements are gen- . erally evaluated using beam characteristic equations. The lumped-mass weights are based upon the mass distribution of the internals structvres. Local masses such. as plates and snubber blocks are included at appropriate nodes;--The effect"of the surrounding water on the dynamics of the internals for horizontal motion is accounted for by hydrodynamically coupling the components separated by a narrow annulus-the vessel, core barrel, core shroud, lower support structure cylinder, and upper guide structure cylinder. The clearance between the core support barrel and the
. reactor vessel snubbers as well as the clearance between th'e core shroud guide lugs and the fuel alignment plate is simulated by nonlinear springs which account for the loads generated should impacting occur. A representation of the core in re is included in the internals models which provides appropriate inertial and impact feedback effects on the internals response.
~~~~The vertical model stiffness values are generally calculated using bar character-~~~~
~
~~~~stic eqvations.~~~~
~~~~~ ~~~~~
~~~~Nonlinear. couplings are )nclvded between components to account~~~~
~~~~~ ~~~~~
~~~~( 'nd or structural interactions such as those between the fuel and core support plate>> between the core support barrel and upper guide structure upper flanges. pre-~~~~
~~~~~ ~~~~~
~~~~1oads, which are caused by the combined action of applied external forces, dead~~~~
~
Weights, and holddowns are also included. friction elements are used to simulate the coupling between, the fuel rods and spacer grids.
,j s y~+I~eg~>~~ ~~~~~~ - ~ .
~g~~~~~Q~~P} i ~ '<< " ~ ~ ~~~~sh~s'~> 4" I > M~A I
s I A reduced model of the reactor vessel internals (Fig. 3.9-5) is developed for', incorporation into the reactor coolant system model. The detailed nonlinear '., orizontal and vertical internals (plus core) models are condensed and combined nto a three-dimensi'onal model compatible with the reactor coolant system model and the computer programs through which the latter model is analyzed. The purpose of this reduced internals model is to account vr the effects of the internal LOCA loads on the reactor vessel support motion and the structural loading interaction between the internals and the ~essel. The reduced internals model is developed so as:to produce reactor vessel support motions and loadings equivalent to those produced by the detailed internals models. The dynamic responses of the reactor internals to the postulated pipe breaks are determined with the CESHOCKcode utilizing the detailedmodels. Horizontal and ver-tical analyses are performed for both hot and cold leg breaks to determine the 1ateral and axial responses of the internals to the simultaneous internal fluid forces and vessel motion excitation. The vertical excitation of the internals is calculated by the LOAD2 computer code using the control volume method. In this method, the reactor internals are divided into volumes containing both structure and fluid or structure alone.'he momentum equation is then applied to each volume, and a resultant force. is calculated which is distributed over the structural nodes within the volume. This method takes into consideration pressure, fluid friction, momentum changes, and gravitational'forces acting on each volume. The resulting load time histories are in a form consistent or CESHOCK code input. order to achieve an initial (prior to the pipe break) equilibrium, the initial static deflections and'aps are calculated. The resulting initial conditions and
~~~~'load time histories are input to the CESHOCK code and the dynamic response of the model is calculated.~~~~
The horizontal input excitations resulting from a cold leg break are the core support barrel force time history and the vessel motion time history determined from the reactor coolant system analysis. The core support barrel forces are obtained by representing the asymmetric pressure distribution time history as a Fourier expan-. sion. The two terms (sine and cose) which excite the beam mode of vibration -are
~~~~-then integrated over the core support barrel and transformed into nodal force time histories.~~~~
The horizontal input excitations resulting from a hot 1eg break are the CEA shroud crossflow load time histories and the vessel motion time history determined from the reactor coolant system analysis. The forces applied to the shroud mass points are determined directly from the blowdown pressure time history and include the
'drag force and forces due to the pressure differential on the shrouds. 'he results from these analyses consist of time-dependent member forces, and nodal displacements, velocities and accelerations. The load and displacement responses are used in the detailed stress analyses of the internals.
Preliminary results of reactor internals analyses indicate, on a load comparison basis. that the adequacy of the structural design of the internals will be confirmed the detailed stress analyses. Results of the stress analysis will be submitted a later amendment in December 1981. (
~ II 31.~ "LOAD2 - A computer Code to Calculate Vertical Hydraulic Loads
~
~~~~on Reactor Internals Using CEFLASH-4B Data As Input",~~~~
~
~~~~Calculation No.~ 79-STA-003, G. Garner, August 24, 1979.~~~~~
~
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~~~~'l 7A'L2-PSAR 3!9.5.3 ~Desi n Loadin Cata orion The design loading conditions are categorized below:~~~~
3.9.5.3al Normal'perating and Upset The normal and upset category includes the combinations of design loadings consisting of normal operating temperature and pressure differentials, loads due to flow, weights, reactions, superimposed loads, vibration, shock loads including operating baris earthquake, and transient loads not re-quiring shutdown. 3.9.5.3.2 Faulted, C The faulted category consists of the mechanical loading combinations nf Subsection 3,9.5.3.1 with the exception that the safe shutdown earthquake (SSE) (in piace of the operating basis earthquake) and the loads resulting fran the Loss-of-coolant accident (LOCA) are included. 3.9.5.4 3.9.5.4.1 'eactor XnternaLs The stress limits to which the reactor internals are designed are listed in Table 3.9-14. No emergency condition has been identified for the applicable components, therefore, no appropriate stress criteria are provided. X'g~ 8 psr~g~atagoziae-end~C ~
.~ are e in The maximum stress intensities in the reactor internal components are de-termined utilizing the most conservative combinations of the lateral LOCA time-dependent loadings in the structural analysis. These and'ertical ~
maximum stresses and 'the maximum stresses resulting from the SSE are then combined absoluteLy to obtain the total stress. intensities. To properly perform their functions, the reactor internal structures'are designed to meet the deformation limits listed below: a Under design loadings plus operating basis earthquake forces, de-flection is limited so that the control element assemblies (CEAs) can function and adequate core cooling is preserved. I b) Under normal;operating loadings, plus SSE forces, plus pipe 'rupture loadings resulting fran a break equivalent in size to the largest kine connected to the Reactor Coolant System piping, deflections are limited'o that the core is held in place, adequate core cool-ing is preserved, and aLL CEAs can be inserted. Those deftectinns which would influence CEA movement are limited to less than 80 per-cent of the deflections required to prevent CEA insertion. 3.9-54
Reactor internals are designed according to Subsection NG of the ASME Code, Section III, with the exception of stamping and a code stress report.
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3. Emoter enc Conditions Five Cycles of complete loss of secondary pressure. This transient would follow a steam line break. A steam line break is not considered credible in forming the basis for design of the Reactor Coo1ant System. However, system components will not fail structurally in the unlikely event that does happen.
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~~~~4. Faulted Conditions The loading combination resulting from the combined effects of the design basis earthquake and normal operation at full power are categorizedas faulted condition .~~~~
~~~~The loading combinations resulting from the design basis earthquake,~~~~
~~~~'ormal operation at full power and pipe rupture conditions are categorized as faulted condition. Oesign basis earthquake and pipe rupture loadings are combined by the SRSS method.~~~~
S. Test Conditions Ten cycles 'of system hydrostatic testing at 3110 psig and at a temperature not less than 60 F above the'ighest component reference .temperature (RTgpT ) or 100 F above the highest component section (RT >) value. This is based on one initial hydrostatic test plus .a majIII repair every four years for 36 years which includes equipment failure and normal plant cycles. 200 cycles 'of leak testing at 2235 psig and at a temperature not less than 60 F above the highest component reference temperature (RTR0T) or 100 F above the hijhest pipe section RTR0 . This is based on normal p'lant operation involving five shutdowns for head emoval or valve repair per year for 40 years. 3.9-64
H SL2-FSAR 4 The fuel assembly is designed,to be capable of vithstanding the axial loads vithout buckling and vithout sustaining excessive strcsscs. 4.2.3.1.2.2 Safe Shutdovn Earthquake (SSE) The axial and lateral loads and deformation sustained by the fuel assembly during a postulated SSE have thc same origin as those discussed above for thc OBE, but they arise from initial ground accelerations tvicc those assumed for the OBE. The analytical methods used for the SSE are identical to those used for the OBE. 4.2.3.1.2.3 Loss of Coolant Accident (LOCh) Xn the event of a'arge break LOCh, there vill occur rapid changes in pres-sure and flov vithin the reactor vessel. hssociated vith the transient are relatively large axial and lateral loads on the'uel assemblies. The response of a fu'el assembly to thc mechanical loads produced by a LOCA is considered acceptable if the fuel rods are maintained in a eoolable array, k,e., acceptably lov grid 'crushing. The methods used for analysii of-combincd seismic and LOCA loads and s resses is described in Reference 50. To qualify the complete fuel assembly, full scale hot loop testing vas'con-ducted. The tests vere designed to evaluate fretting and vear of compo-nents, refueling procedures, fuel assembly uplift forces, holddovn perfor-mance and compatibility of the fuel assembly vith interfacing. reactor in-ternals, CEAs and CEDHs under conditions of reactor vater chemistry, flov velocity, temperature, and pressure. The test assembly vas a 16 x 16 five guide tube design.'he test vas run for approximately 2000 hours ~ The tests 'results demonstrated the acceptability of the design. llechanical'esting of the fuel assembly and its components is being per-formed to support analytical means of defining the assembly's structural characteristics. The te'st program consists of static and dynamic testa of spacer grids i'nd static and vibratory tests of a full siae fuel as-sembly. 4.2.3.).2.4 Combined SSE and LOCh lt is not considered appropriate to combine the stresses resulting from the SSE and LOCh events. 'evertheless, for purposes of demonstrating margin in the design, the maximum" stress intensities for each individual event vill be 3 Combined by a square root of sum of the squares (SRSS) method. This vill be performed as a funct'ion of fuel assembly elevation and posit'ion, eg, the maximum 490.. stress intensities 'for the center guide tube at the upper grid elevation {as determined in the analysis discussed in Subsections 4.2.3.1.2.2 and 4.2.3.1.2.3) vill bc combined by the SRSS method. Zt is expected that the results vill demonstrate that the allovable stresses described in Subsection 4.2.1.1 are not exceeded for any position along'thc fuel assembly, even under the added con-oervatism provided by this load combination. 4.2.3.1.3 Spacer Grid Evaluation The function of the spacer grids is to provide lateral support to fuel and burnablc poison rods in such a manner that the axial forces arc not suffi- ~ \ ge
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uestion ~ 3~ Provide analyses to determine the external forces and moments, resulting from postulated hot. leg .and cold leg
~ 'ruptures within the reactor cavity, on reactor vessel supports. Xf applicable, similar analyses should be performed for steam generator and/or pressurizer compart-ments that may be subject to pressurization where, sign-ificant component support loads may result. For each analysis, provide the following information: . ~2~ For each compartment, provide a table of blowdown mass flow rate and energy release rate as function of time for the break which was used for the component support .evaluation.
~~~~Response~~~~
FSAR Table 6.2-13 is a summary of postulated pipe ruptures for containment subcompartment analysis. The last column in this table "Release Rate Data Table Numbers" will refer to, for each compartment, a table o f blowdown mass theflow rate
~~~~'and energy release rates as a function of time for. break which was used for the component, support, evaluation.~~~~
0 question ~ .. Provide analyses to determine the external forces and moments, resulting from postulated hot leg and cold leg ruptures within the reactor cavity, on reactor vessel supports. If applicable, similar analyses should be performed for steam generator'nd/or pressurizer compartments that may be subject to pressurization where significant component support loads may result. Far each analysis, provide the following information: (3) Describe and justify the nodalization sensitivity studies performed for the major component supports evaluation (if different from the strucutural analysis model), where transient forces and moments acting on the components are of concern. Where component loads are of primary interest, show the effect of noding variations on the transient forces and- moments. Use this information to justify the nodal model selected for use in the component supports evaluate.on. s onse The analysis performed for the 'mayor component supports does not differ from the structural analysis model. As discribed in FSAR subsection 6. 2. 1. 2. 3 divisions between subcompart-ment are determined by the, physical flow restrictions within each compartment. A flow .restriction is defined by the presence of an object in the flow path that changes the flow area in that direction, with the subdivision defined at, the point of minimum flow area. This minimum flow area becomes the j unct ion flow area used in the RELAP 4 analysis
. For the models constructed for the reactor cavity and second-ary shield wall area flow restrictions included the pre-sence of steel and concrete supports, doorways, vent shafts and gratings, as well as large equipment such as the reactor vessel, primary piping, the steam generator, reactor coolant. pumps and the pressurizer. By choosing node boundaries at the various physical flow restrictions, a method consistent with the lumped-parameter calculation model used by RELAP 4 and described above, calculated differential pressures and consequent support loads are realistically maximized. The nodalization sensitivity study performed for the Shearon Harris PSAR (Docket 50-400, 401, 402 and 403) shows that the peak calculated differential pressure is very sensitive to an increasing number of nodes until that, number. equals the number de-fined by physical flow restrictions. Increasing the subdivision of the compartment is unwarranted and can lead to unrealistic results if these "fictitpo'us junctions" <
~~~~are modeled. The subcompartment models discussed below take account, of-all physical flow restrictions present in a manner identical to that shown to be optimum by the sensitivity study.~~~~
~
~~~~Table 6.2-25 presents the overall results of the sub-~~~~
~~~~'compartment analyses. The reactor cavity, Secondary Shield Wall and Pressurizer Area Design evaluation is FSAR Subsection 6.2.1.F 3 '~~~~
0 uestion 3~ Provide analyses to'determine the external forces and moments, resulting from postulated hot leg and cold leg ruptures within the reactor cavity, on reactor vessel supports. If applicable, similar analyses should be performed"for steam generator and/or press-urizer compartments that may be subject to pressuriza-tion where significant component support loads may result. For each analysis, provide the following= in-formation: (4) Graphically show the pressure (psia)'nd differential pressure (psi) response as functions of time for a representative number of nodes to indicate the spatial pressure response. Discuss the basis for establishing the'ifferential pressure on components. ~Res onse FSAR Table 6.2-25 list the Results of the Subcompartment Analysis. In this table the peak node pressure, and peak differential pressure is listed. Along with these valves a figure is referenced for both of those valves. The component and support loads for the Steam Generator, Reactor Coolant Pump, and Pressurizer were determined by
'quivalent static analyses. A load factor of two on the calculated thrust, jet impingment, and subcompartment pressure loads is employed. to account for the dynamic response of the structure. The model employed for static analysis is shown in Figure 3.9-18.
e Question , Provide analyses to determine the external forces and moments, resulting from postulated hot leg and cold leg ruptures within the reactor cavity, on reactor vessel supports. If applicable, similar analyses should be performed for steam generator and/or pressurizer component support. loads may result. For each analysis,
~~~~.provide the following information:~~~~
(5) Provide the peak. and transient loading on -'the major components used to establish the adequacy of the support design. This should include the load forcing functions (e g, f<<(t) fp) f (t) ) and transient moments (e.g.,
~
Mg(t) I My(t), M>(t) as resolved about, a specific identified coordinate system. The centerline of the break nozzle is recommended as the X-axis and the center line of the vessel as the 2 axis. Provide the projected area used to calculate these loads and identify the location of the area pro-jections on plan and section drawings in the selected coordinate system. This information should be presented in such a manner that confirmatory evaluations of the loads and moments can be made. ponse Refer to FSAR Tables 6.2-25 and 6.2-26 for a discussion on the peak and transient loading on the major components used to establish the adequacy of the support design. The mass and energy release data that was utilized for the structural design is identical to that used for com-ponent support design verification. Therefore, the peak and transient forces provided in FSAR Figures 6.2-23 thru 6.2-30 were utilized for both the structural and component design, where applicable. The analysis of the RCS com-ponents (i.e., Reactor Vessel, Steam Generator, RC Pumps, Pressurizer and RC Piping) due to the asymmetric pressure loadings is provided in revised FSAR Section 3.9.1.4.1. A tabulation of results and comparison with the appropriate allowables is also provided.
Question
~~~~4. Figure 6.2-71,'regarding containment isolation valves, should be revised to show the containment isolation valve arrangements for each containment penetration.~~~~
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uestion 5 FSAR Sections 6.2.4.1.'1 and 7.3.1.1.4 indicate that either a high con-tainment pressure signal- or a high containment radiation level signal will generate a containment isolation actuation signal. However, SRP Section 6.2.4 also recommends that a high radiation signal should not be considered one of the diverse containment isolation parameters. Therefore, we reauest that the safety injection actuation signal should be used as one of the parameters for the initiation of containment isolation, and the above cited FSAR sections should be revised accordingly. Res onse 5 The Safety Injection Actuation Signal (SIAS) will be used as one nf the parameters for initiation of Containment Isolation.
e stion
~~~~6. FSAR Section 6.2.4.4 indicates that the following penetrations~~~~
( will not-be considered possible sources of bypass leakage and, therefore, will not be subject to Type C leak rate testing: a) Main steam (Penetrations 1 and 2); b) FeeQwater (Penetrations 3 and 4); c) Steam generator blowdown (Penetrations 5 and 6); and d) Steam generator blowdown samp'ling (Penetrations 30 and 49). In order for us to determine the acceptability of this, discuss the conditions that will exist or the action to be taken to assure that outleakage will not occur after a LOCA for a period
of 30 days. In this regard, discuss the pressure response of the steam generators relative to the containment pressure, in the short term, and the feasibility of reflooding the steam generators, in the long term, to preclude outleakage.
~Res ense The Main Steam System, Main Feedwater System, Steam Generator Blowdown and Blowdown Sampling System are connected to a closed seismic Category I, Quality Group B system inside containment and are therefore classified as GDC 57 systems in accordance with 10CFR50 Appendix A. These systems are main-tained at a temperature and pressure cond&ion that is higher than the containment, atmosphere during normal plant operations. During accident conditions, the Main 5'team and Feedwater Isolation valves will close upon receipt of a MSIS (high con-tainment. pressure or low S.G. pressure) while the Steam Gener-a/or Blowdown and Blowdown Sampling Isolation valves will close upton receipt of a CIAS (hign containment pressure or high con-tainment radiation) . FSAR Figure 6.2A (attached to CBS question I l) provides the containment pressure response for the worst bieak scenario and illustrates that the containment atmesphere rie to a peak of 58.4 psia and reduces to atmosphere pressure within the first day post-LOCA. Therefore, the Steam Generator inventory that existed prior to the accident will be available post-LOCA . and will act as a steam seal at, the onset of the accident. Subsequent to the decay of the steam, generator pressure and level, the Auxiliary Feedwater System will automatically maintain the steam generator level to guarantee the pressure of a water seal and thereby preclude bypass leakage.
stion FSAR Section 6.2.4.2 indicates that only one isolation valve outside 'containment is provided for the isolation of each of the containment emergency sump suction lines. For this type of isolation valve arrangement, the piping between the containment and the valve should be enclosed in a leak-tight. or controlled leakage hou ing (as des-cribed in SRP Section 6.2.4) leakage housing. If, in lieu of a housing, conservative design of the piping and valve is assumed to preclude a break of piping integrity, the design should conform to the requirement of SRP Section 3.6.2. Also, design of the valve and/or the piping compartment should provide the capability to detect leak-age from the valve shaft and/or bonnet seals and terminate the leakage. Therefore, discuss the design of the con-tainment 'emergency sump suction penet'rations (Penetrations 32 and 33), and the leakage detection and control provisions.
~~~~Response~~~~
The emergency sump suction penetrations process lines are enclosed in a leak-tight housing, (i.e., carbon'teel guard pipes) which extend from the sump inside containment to the Containment Isolation Valve located outside contain-ment. Each guard pipe is directly welded to a steel con-tainment vessel nozzle and acts as an extension of the containment in both directions. Passing through'ach guard pipe is the stainless steel sump suction line. These lines are welded to the guard pipe in the sump so that. water cannot enter the annulus formed by the concentric pipes. Outside containment the suction lines are sealed to the guard pipes by means of a stainless steel bqllows to allow for thermal movement. FSAR Figure 3.8-6 provides a detailed description of this type IV penetrations. The containment isolation valves are located 'in the Reactor Auxiliary Building pipe tunnel which is a controlled leakage area. Leakages from these systems are directed to the ECCS room sump which is provided with safety grade, seismic Category I level indications. A backup seismic Category I level indicator is also provided in each ECCS room sump to alert the operator of any abnormal condition. The ECCS area is also provided with two safety related radiation monitors to measure the airborne effluent. A complete description of these monitors is provided in FSAR Section 11.5.2.2.10.
Table 6.2-52, "Containment Penetration and Isolation Valve Information," should be revised to designate the fuel transfer tube (Penetration 25) and charging line (Penetration 27) as direct bypass leakage paths. FSAR Table 6.2-52 will be revised to designate the fuel transfer tube (Penetration 25) as a direct bypass leakage path. The charging line (Penetration 27) is not considered a credible source of bypass leakage following a LOCA. Charging'umps 2A and '2B are automatically started following receipt of a Safety Injection Actuation Signal .(SIAS) and are powered by the emergency diesel generators. Thus, after an accident flow is directed into containment through this penetration precluding bypass leakage by establishing a water seal. If the pumps were not operating radioactive contaminants are prevented from reaching the environment by a minimum of three seismic-ally qualified, Safety Class 2 check valves in series. These design features virtually eliminate any possibility of bypass leakage.
1 uestion ~
~
I
~~~~9. Provide the information as required by NUREG-0737 concerning the following TMI Action Plan items:~~~~
a) IX.E.4.2 Containment Xsolation Dependability; b) XX.F.1.4 Containment Pressure Monitor; and c) IX.F.1.6 Containment Hydrogen Monitor ~Res ense The information as required by NUREG 0737 concerning the following TMI Action Plan items has or will be incorporated into the,.St Lucie 2 FSAR.
~~~~') is contained II.E.4.2 Containment Isolation Dependability information in Appendix 1.9A and is attached for your use.~~~~
b) XI.F.1.4 Containment pressure monitor information is attached to the question/response. This information will appeax in Amendment, 5 to the St Lucie Unit 2 FSAR to be issued August 17, 1981. c) II.F.1.6 Containment Hydrogen Monitor information is contained in Appendix 1.9A item II.F.l(c) from there we refer you to Subsection 6.,2.5.2.1 in which we completely describe the Containment Hydrogen Analyzer Subsystem. This is also attached for your information and use.
SL2 "FSAR 1 EMERGENCY POWER SUPPLY FOR PRESSURIZER HEATERS' sufficient number of pressurizer heaters and associated controls to necessary to maintain natural circulation at hot standby condition are l3 provided with power supply from either the offsite power source or the 0 emergency power source (when offsite power not av'ailable) ~ Each redundant group of heaters has access to only one Class lE division of power supply b) Any changeover of the heaters from normal offsite power to emergency onsite power is accomplished manually in the control room. (See Subsection 8.3.1.1.1) c) Procedures and training will be-established to make the operator aware of when and how the required pressurizer heaters are connected to the emergency buses. The procedures will identify a) which engineered safety features loads may be appropriately shed for a given situation, b) manual operation of the heaters and c) instrumentation and criteria to prevent overloading a diesel 'generator. The time required to accomplish the connection of the necessary )0 l3 number'f pressurizer heaters to emergency buses is consistent with the timely initiation and maintenance of natural circulation. I o 5 fo Pressurizer heater motive and contgol power'nterfaces with emergency buses are through devices which arg<qualified to safety grade 3 requirements. Safety grade circuit breakers are provided to protect this Class lE interface's per the St Lucie Unit 2 commitment to Regula-tory Guide 1.75, "Physical Independence of Electric System" 1/75(R1) in Section 8.3. f) Being non-class lE loads, the pressurizer heaters are automatically shed from the emergency power sdurce upon occurrence of a SIAS ~ IleE ~ 4~1 DEDICATED HYDROGEN PENETRATIONS As discussed in Subsection 6.2'5, redundant internal hydrogen recombiners are provided. Therefore this requirement is not applicable to St Lucie Unit 2. II.Eo4 ~ 2 CONTAINMENT ISOLATION DEPENDABILITY The following items address corresponding NRC positions contained in NlJ REGW7 37:
1) As discussed in Subsection 7.3.1.1 the containment isolation actua-tion signal (CIAS) is initiated upon high pressure or high radiation inside the containment. Therefore, the CIAS gomplies with the recommendation in Standard Review Plan 6.2 ~ 4 "Containment Isolation System" (Rl) with respect to diversity in the parameters sensed for initiation of containment isolation.
1,9A-7 Amendment No. 3, (6/81)
SL2-F SAR 0" Using the APCSB 3-1 definition in Appendix A to (ll/24/75) (attached to the the Branch Technical Position Standard Review Plan 3.6.1), essential system and components are defined as those systems and components required to shutdown the reactor and mitigate the conse-quences of an accident. Table 6-2-52 identifies the essential penetrations as ESP penetrations's indicated in Subsection 6.2.4, all containment penetrations associated with nonessential systems are either administratively locked closed or automatically isolated upon a CIAS. Penetrations for systems like post accident monitoring instrumentation and RCS sampling however are provided with manual override of the CIAS to enable the operator to open the containment isolation valves and activate the systems as necessary.
3) The St Lucie Unit 2 containment isolation system complies with General Design Criteria (GDC) 55, 56 and 57- A CIAS is used to isolate nonessential systems. GDC 57 permits the use of one con-tainment isolation valve located outside containment which is capable of automatic or remote manual operation and does not require closure on a CIAS- The penetrations that fall into this category are main steam and feedwater which are automatically isolated upon receipt of a MSIS. However, with the diversity of high containment pressure or low steam generator pressure, a MSIS is generated and isolates the main steam isolation valves and Main Peedwater isola" tion valves The component cooling water lines to and from the reactor coolant pump fall under the requirements of GDC 56. An SIAS isolates these penetrations and is initiated by diverse parameters, 1ow pressurizer pressure or high containment pressure.

4) The present design of control systems for automatic containment isolation valves are such that resetting the isolation signal does not result in the automatic reopening of containment isolation valves. Certain valves (eg, post accident sampling, containment radi'ation monitoring~instrument air) which are required to open during an accident are provided with the capability of manually overriding the automatic isolation signal- Reopening of these containment isolation valves requires deliberate operator action, ~
~~~~and can be accomplished only on a valve-by-valve basis. The con-tainment isolation design does not utilize "ganged" control switches for containment isolation valves.~~~~
5) The CIAS, MSIS and SIAS containment pressure setpoint is selected to account for the normal operating pressure inside containmentf equipment uncertainty, setpoint drift and associated instrumentation time delay- The pressure setpoint selected is far enough above the maximum expected pressure inside containment during normal operation so that inadvertent containment isolation does not occur during normal operation from instrument drift or fluctuations due to the inaccuracy of the pressure sensor.
1 ~ 9A-8 Amendment No. 1, (4/81)
SL2- FSAR
6) The containment purge valves will comply with the operability criteria provided in Branch Technical Position CSB 6-4 (Rl) and the staff inte'ri'm position of October 23, 1979. The 48" purge valves are administratively closed during normal plant operation and only opened when the reactor is in cold shutdown or refueling mode. The 8" continuous containment purge valves will be able to close under the DBA pressure and flow condition loading (time dependent) within the required valve closure time limit.
~~~~The 48"'urge valves are verified to be closed at least every 31 days.~~~~
~~~~7) The continuous containment purge valves close on a CIAS which, as stated in Item 1, is initiated upon a high radiation or high pres-sure inside containment.~~~~
II+F 1 ADDITIONAL ACCIDENT MONITORING INSTRUMENTATION In order to minimize the potential for operator error, display panel controls added to the control room as a result of this action item will undergo a human factor analysis. a) The containment pressure measurement and indication capability will be upgraded to four times the design pressure of steel containment. A continuous indication of containment pressure will be provided in the control room, in addition to recording. b) A continuous indication and recording of water level in the reactor cavity sump will be provided in the control room. The following will be provided:
1) A permanently installed narrow range reactor cavity sump level instrument will cover the range from the bottom of the reactor cavity sump to elevation 0-0 ft inside the containment.

2) Permanently installed redundant wide range containment water level instrument will cover the range from elevation -1.0 ft to the elevation on equivalent to 600,000 gallons inside the containment.
c) Redundant physically separate safety related hydrogen analyzers are presently provided with a measurement range of 0'to 10 percent hydrogen concentration. The analyzers are manually operated from the control room and readings are continuously displayed in a panel meter and recorded on an analob~ strip chart in the control room., As indicated in Sections'.10 and 3.11 the analyzer system are seismic Category I,- meets the seismic qualification of IEEE 344-1975, and environmental qualifi-cation of IEEE 323-1974. The power is supplied from Class 1E emergency bus with automatic loading onto the diesel generators. Provisions are made for periodic testing. Subsection 6-2.5.2.1 provides a detailed description of the hydrogen analyzers-1 'A-9 Amendment No. 1, (4/81)
0
.5.3 TMX RELATED ADDITIONAL ACCXDENT MONITORING INSTRUMENTATION .3.1 TMI Containment Pressure Monitors In Compliance with NUREG 0737 permanently installed wide range containment pressure monitors are provided for post accident monitoring of containment pressure.
7.5.3.1.1 Design Bases a) Measurement and indication capability is provided over a range of -5 psig to four times. the containment design pressure (175 psig) b) Safety related redundant instrumentation channels are provided to meet the single failure criteria. c) The redundant containment pressure monitoring instrumen-tation channels are energerized from independent class XE power sources, and are physically separated in accordance with regulatory Guide 1.75 "Physical Independance of Electric Systems" January 1975 (Rl) d) The containment pressure monitoring instrumentation is qualified in accordance with XEEE 323-1974 for the . design bases accident environment in which they operate. e) = The containment pressure monitors are designed seismic category I and qualified per the IEEE .344-1975 criteria. Continuous indication and recording of conthinment pressure, is provided in the control room. g) Each instrument covers the entire pressure range. h) The monitoring instrumentation inputs are from sensors that directly measure containment pressure and provide input only to the containment pressure monitors. i) An instrumentation channel is available during normal operation prior to an accident as specified in plant technical specification. Testing and calibration requirements are specified in j) plant technical specification k) The instruments are specifically identified on the control panels so that the operator can easily discern that they are intended for use under accident conditions. '7 ~ l. 2 Design Description The containment pressure detectors are electronic trans-mitters (Rosemount 1153GB7) mounted outside the Reactor
Containment Building'. The detectors utilize independent sensing lines which penetrate the containment. A normally open fail closed solenoid valve with remote manual control operated from the control room is provided for containment ksolation for each loop. The redundant containment. pressure monitoring channels are provided with indicators in the control xoom and one of the channels is recorded in the control zoom. Instrument loop accuracy, provided in Table 7.5-1 7.5.3.1.3 Safety Evaluation The TMX containment pressure monitors are designated seismic category I and designed to the Quality Group B standard; Two more channels of containment. pressure monitoring instrumentations with a range of 0 to 60 psig " are provided as post. accident monitors (refer to Table 7.5-1). Hence in the unlikely event when the two redundant TMI containment pressure monitor displays disagree the operator has available to his disposition these other monitoring channels for verification purposes as described in the plant technical specifications, Channel calibration and channel check are performed periodically. 5.3.2
~ ~ TMI Containment Mater Level Monitors Xn compliance with NUREG 0737, permanently installed=narrow and wide range containment water level monitors are provided for post accident monitoring. The narrow range instrument covers the range from the =bottom to the top of the. reactor cavity sump. The wide range instruments cover the range fx'om the bottom of the containment to the evelation equiv-alent to 600,000 gallon capacity.
7 '.3.2.1 Design Bases a) Safety related, redundant. wide range water level monitors are provided to meet the single failure criteria. The wide range monitors are designed to seismic Category I requirements. b) The redundant wide range water level instrumentation channels are energized from independent class IE power sources and are physically separated in accordance with Regulatory Guide 1.75 "Physical Xndependence of Electric Systems" January 1975 (Rl) ~ c) One narrow range containment water level monitor is provided. d) Both the narrow and wide range containmenh water le vel monitoring channels are qualified to IEEE 323-1976 Q for post accident environment in which they also operate Seismic qualification per XEEE 344-1975 is provided. e) Continuous..indication and recording of containment water level is provided in the control xoom.
0, f) Adequate overlapping of the ranges of narrow and wide range monitors are provided. g) Signals from the associated sensors are only used for monitoring the containment water level. h) The availability requirement. of the wide range containment water level monitors is specified in plant technical specification. i) Testing and calibration requirements are specified in plant technical specification. The instruments are specifically identified on the control j) panels so that the operator can easily discern that, they are intended for use under accident conditions. 7.5.3.2.2 Design Description The wide 'and narrow range containment level transmitters are located 'inside +he containment. The narrow range monitor measures discrete level points from the bottom of the reactor cavity sump (elevation -7ft.) to the top of the sump (elevation Oft.). The wide range monitors measure discrete level points from elevation -1 ft. to elevation 26 ft. of the containment. The electronics portion of each of the sensors are located outside the containment and converts the discrete point measurement to a continuous lkvel indication in the control rooms. .The two channels of wide range level monitors are indicated in the control room, one ch'annel is recorded. The narrow range level monitoring channel-is both indicated and recorded in the control room.. 7.5.3 '.3 Safety Evaluati'on The redundant wide range water level monitors are safety related and designated seismic, Category I. They are qualified for 'the design basis accident environment in which they operate per IEEE 323-1974, seismic qualification .is per IREE '344-l975. These mo itors are provided strictly for monitoring purpose. Hc 0'etymo3ated-operator-action-ks
~~~~~ed~~ formalism-pk~dedMyMhis-instrument; The narrow range water level instrument is primarily used during normal operation and does not serve any safety related function post accident.~~~~
SL2-PSALM In addition to the redundant CGCS, the Continuous Containment Purge/Hydrogen Purge System is available for fission product removal and hydrogen purge following a LOCA. 6.2.5.2 S stem Desi n 6.2.5.2.1 'Containment Hydrogen Analyzer Subsystem The Containment Hydrogen Analyzer System consists of two redundant subsystems as shown on Figure 6.2-62, consisting of the sample and return piping, associated valves, hydrogen analyzer, grab sample cylinder, sample pump, moisture separator, cooler, instruments, calibration gas line and reagent gas line. Each of the redundant subsystems is physically separate and operates in-dependently of the other, and is powered from an independent onsite power source. No single failure can result in a total loss of hydrogen concen-tration measurement capability. Failure of one train is annunciated in the -control room. Components of the system are accessible for periodic inspection and main-tenance. The system is designed to permit local calibration at periodic intervals with a reference hydrogen gas standard (span gas) and a zero hydrogen content reference gas. The system is independent of any system used during normal plant operation, so that plant operation does not impose restrictions on such testing. The Containment Analyzer System is designed to seismic Category I and appli-cable Quality Group B requirements.'omponents at the hydrogen analyzer system, including pumps, val ves and tubing are specified to ASME Code Section III, Code Class 2. Instrumentation'nd -controls and electric i-equipment associated with the system are Class 1E. Conformance to appl cable IEEE Standards is discussed in Chapter 7, Sections 3.10 and 3.11. The system is initiated by manual operator action from the control room. No action outside the control room is necessary for system operation. However calibration can be done only at the 1 ocal panels Once initiated, the system draws a continuous air sample from one of the sample points inside containment. Sampling valves can be manually controlled to analyze any samp1 e point. The air is passed through the detector, analyzed, and pumped back into containment. Analyzer readings'are recorded in the control room, and an alarm is actuated if concentration is above three percent. Alarm is also provided for low flow and high temperature of the sample gas. Design and performance data f'r the analyzer is listed in Table 6.2-54. The system is designed for 40 years of normal and one year post-LOCA environmental.'ondition and the components are qualified to operate under the applicable environmental conditions as described in Section 3.11. The operating princip] e of the hydrogen analyzer is thermal conductivity of the sample Air samples are drawn from any of the following samp]e points 6.2-63 Amendment No. 0, (12/80)
SL2-FSAR inside containment: a) Containment dome c) Pressurizer enclosure d) Vicinity of reactor coolant pump (RCP) 2A1 e) Vicinity of reactor coolant pump 2A2 f) Vicinity of reactor coolant pump 2Bl g) Vicinity of reactor coolant pump 2B2 These points provide broad coverage of the containment for hydrogen monitor-ing and constitute a redundant independent H2 Samp) ing System. Sampling l ines originating from the containment dome, pressurizer, RCP 2Al and RCP 2A2 areas constitute one independent train of the H Sampling System. The other train consists of sampling lines originating from the upper contain-ment, RCP 2Bl and RCP 2B2 areas. Each train of the sampling lines has a common header inside the containment and penetrates the containment in a separate penetration assembly. As discussed in Subsection 6.2.2.2, there is adequate mixing of containment atmosphere so that local stratification or pocketing of hydrogen does not occur. The analyzer cubicles are located at elevation 19.5 ft of the Reactor Auxiliary Building (RAB). The analyzer, system control panel is located in the control room. A grab sample chamber located at elevation 19.5 ft of the RAB is provided to permit hydrogen concentration measurement independent of the containment 'ydrogen analyzer detector. 6.2.5.2.2 Containment Hydrogen Recombiner Subsystem The containment hydrogen recombiners control hydrogen in containment by using heat to cause recombination of liberated hydrogen with free oxygen in the air to form water. The hydrogen pcombiner system is described in Westinghouse Topical Report WCAP 7709-Li i and shown on Figure 6.2-63. Supplement 1 through 4 of WCAP 7709-L were accepted by NRC on May 1, 1976. It is designed seismic Category I and Quality Group B requirements. Each recombiner consists of a thermally insulated vertical metal duct with electric resistance metal sheathed heaters provided to heat a continuous flow of containment air to a temperature which is sufficient to cause a reaction between the hydrogen and the oxygen in the air. The recombiner is provided with an outer enclosure to provide protection from water spray coming from the Containment Spray System. The recombiner consists of an inlet preheater section, a heater-recombination section, a mixing chamber, and a cooling/exhaust section. Mixing of containment air is by the con-
~~~~6. 2- 64 Amendment No. 0, (12/80)~~~~
Q35 ST. LUCIE UNIT 2 STEAM GENERATOR SUPPORT LOADS COMBINED LOCATION SPECIFICATION LOCA + N.Op. + SSE Upper keys (ea.) Z1 1.51 2.172 Z2 2.00 2.172 Snubbers (ea.) 0.22 0.55 SLIDING BASE Vertical pads Y1 1.71 5.974 Y2 2.33 3.588 Y3 2.23 2.458. Y4 1.72 2.586 Anchor bolts Y1 1.85 2.716, (per pair of bolts) Y2 1.72 2.856 Y3 0.58 2.086 Y4 1.73 2.948 Lower stop X3 5.648 7.085 Lower keys Z11 3.28 3.755 Z12 1.06 2.772 Units - millions of pounds
$ 35 ST. LUCIE UNIT 2 RCS COMPONENT NOZZLE LOADS RSS MOMENTS NOZZLE LOCATION COMBINED SPECIFICATION LOCA + N.O . + SSE R V Inlet 3.47 9.93 R V Outlet 14.01 42.49 S G Inlet 6.73 21.75 S G Outlet 6. 20, 7.79 RCP Suction 3.90 4.45 RCP Discharge 3.98 5.42 Units millions of pounds
stion Provide analyses to.determine the external'orces and moments, resulting from postulated hot. leg and cold leg ruptures within the reactor cavity, on reactor vessel supports. Xf applicable, similar analyses should be performed for steam generator and/or pressurizer compart-ments that may be subject to pressurization where sign-ificant component support loads may result. For each analysis, provide the following information: For each compartment, provide a table of blowdown mass flow rate and energy. release rate as function of time for the break which was used for the component support evaluation.
~~~~Response~~~~
FSAR Table 6.2-l3 is a summary of postulated pipe ruptures for containment subcompartment analysis. The last column in this table "Release Rate Data Table Numbers" will refer you to, for each compartment, a table of blowdown mass flow rate and energy release rates as a function of time for the break which was used for the component. support evaluation.
question 5 Provide analyses to determine the external forces and moments, resulting from postulated hot,leg and cold leg ruptures within the reactor cavity, on reactor vessel supports. If applicable, similar analyses should be performed for steam generator and/or pressurizer compartments that may be subject to pressurization where significant component support loads may result. For each analysis, provide the following information: Describe and justify the nodalization'sensitivity studies performed for the major component supports evaluation (if different from the strucutural analysis model), where transient forces and moments acting on the components are of concern. Where component loads are of primary interest, show the effect of noding variations on the transient forces and moments. Use this information to justify the nodal model selected for use in the component supports evaluation.
~~~~Response~~~~
Divisions between subcompartment are determined by the physical flow restrictions within each compartment. A flow restriction is defined by the presence of an ob-ject in the flow path that changes the flow area in that direction, with the subdivision defined at, the point of minimum flow area. This minimum flow area becomes the junction flow area used in the RELAP 4 analysis. For the models constructed for the reactor cavity and second-ary shield wall area flow restrictions included the pre-sence of steel and concrete supports, doorways, vent shafts,and gratings, as well as large equipment such reactor vessel, primary piping, the steam generator, as'he reactor coolant pumps and the pressurizer. By choosing node boundaries at. the various physical flow restrictions, a method consistent with the lumped-parameter calculation model used by RELAP 4 and described above,.calculated differential pressures and consequent support loads are"realistically maximized. The nodalization sensitivity study performed for the Shearon Harris PSAR (Docket 50-400, 401, 402 and 403) shows that the peak calculated differential pressure is very sensitive to an increasing number of nodes until that number equals the number de-fined by physical flow restrictions. Increasing the subdivision of the compartment is unwarranted and can lead to unrealistic results if these "fictituous junctions" are modeled. The subcompartment models discussed below
~
take account of all physical flow restrictions present in a manner identical to that. shown to be optimum by the sensitivity study. Table 6.2-25 presents the overall results of the sub-compartment analyses. The reactor cavity, Secondary Shield Wall and Pressurizer Area Design evaluation is described in FSAR Subsection 6.2.1.2.3.
( Question Provide analyses to determine the external forces and moments, resulting from postulated hot leg and colds leg ruptures within the reactor cavity, on reactor vessel supports. If applicable, similar analyses sh'ould be performed for steam generator and/or press-urizer compartments .that may be subject, to pressuriza-tion where significant component support loads may result. For each analysis, provide the following in-formation: Graphically show the pressure (psia) and differential pressure (psi) response as functions of time for a representative number of nodes to indicate the spatial pressure response. Discuss the basis for establishing the differential pressure on components.
~~~~Response~~~~
FSAR Table 6.2-25 list the Results of the Subcompartment Analysis. In this table the peak node pressure, and peak differential pressure is listed. Along with these valves a figure is referenced for both of those valves. The component and support loads for the Steam Generator, Reactor Coolant Pump, and Pressurizer were determined by equivalent static analyses. A load factor of two on the calculated thrust, jet impingment, and subcompartment pressure loads is employed to account for the dynamic response of the structure. The model employed for static analysis is shown in Figure 3.9-l8.
8 Figure 6.2-71, regarding containment isolation valves, should be revised to show the containment isolation valve arrangements for each containment penetration. In addition, the isolation valve arrangements shown in this figure should be consistent with the valve arrange-ments as shown in the system flow diagrams.
~~~~Response~~~~
The attached figures show the containment isolation valve arrangement for each containment penetration. These figures will be placed in the FSAR via Amendment 6.
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SL2-FSAR inside containment: a) Containment dome c) Pressurizer enclosure d) Vicinity of reactor coolant pump (RCP) 2Al e), Vicinity of reactor coolant pump 2A2 f) Vicinity of reactor coolant pump 2B1 g) Vicinity of reactor coolant pump 2B2 These points provide broad coverage of the containment for hydrogen monitor-ing and constitute a redundant independent H Samp) ing System. Sampling lines originating from the containment dome, pressurizer, RCP 2Al and RCP 2A2 areas constitute one independent train of the H Sampling System. The other train consists of sampling lines originating from the upper contain-ment, RCP 2Bl and RCP 2B2 areas. Each train of the sampling lines has a common header inside the containment and penetrates the containment in a separate penetration assembly. As discussed in Subsection 6.2.2.2, there is adequate mixing of containment atmosphere so that local stratification or pocketing of hydrogen does not occur. The analyzer, cubicles are located at elevation 19.5 ft of the Reactor Auxiliary Building (RAB). The analyzer system control panel, is located in the control room. A grab sample chamber located at elevation 19.5 ft of the RAB is provided to permit hydrogen concentration measurement independent of the containment hydrogen analyzer detector. 6.2 '.2.2 Containment Hydrogen Recombiner Subsystem The containment hydrogen recombiners control hydrogen in containment by using heat to cause recombination of liberated hydrogen with free oxygen in the air to form water. The hydrogen jqcombiner system is described in Westinghouse Topical Report WCAP 7709-L~ ~ and shown on Figure 6.2-63. Supplement 1 through 4 of WCAP 7709-L were accepted by NRC on May 1, 1976. It is designed seismic Category I and Quality Group B requirements. Each recombiner consists of a thermally insulated vertical metal duct with electric resistance metal sheathed heaters provided to heat a continuous flow of containment air to a temperature which is sufficient to cause a reaction between the hydrogen and the oxygen in the air. The recombiner is provided with an outer enclosure to provide protection from water spray coming from the Containment Spray System. The recombiner consists of an inlet preheater section, a heater-recombination section, a mixing chamber, and a cooling/exhaust section. Mixing of containment air is by the con-
~~~~6. 2- 64 Amendment No. 0, (12/80)~~~~
492.10 With regard to the Analog Core Protection Calculator, provide a listing of the algorithms used, discuss their verification and evaluation. ~Res ense The algorithm for the Thermal Margin/Low Pressure Limiting Safety System Setting (LSSS) has been discussed in the answer to question 492.9. The "algorithm" for the Local Power Density (LPD) LSSS results in a trip limit-line of power vs. axial shape index as shown in the attached figure. FSAR Figure
~~~~7. 2-'16 is the LPD trip functional diagram.~~~~
The verification and evaluation of the LPD trip limits are discussed in CENPD-199-P "CE Setpoint Methodology". As noted in the answer to question 492.9, C-E is cur-rently updating this report for final NRC review and approval.
p-; '(Vz.io -1 ST, LUC1E 0i'11T 2 CPC-2 LOCAL POh'EB DENSITY TRIP
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SL-2 Round One uestions 440.25 Provide a detailed analysis on the consequences of a RCP shaft (15.3.3) seizure event. Justify selection of limiting single failures. The time at temperature studies which justify your claims of peak clad temperature beind limited to 1300"F are not accepted by the staff. In assessing fuel failures, any rod which experiences a DNBR of less than 1.19 must be assumed failed. Confirm that the results of the analysis meet the acceptance criteria of SRP 15.3.3.(2). Provide your assumptions on flow degradation due to the locked rotor in the faulted loop, and reference appropriate studies which verify these assumptions. Also provide a similar analysis for the locked rotor event presented in section 15.3.4.1, and show that acceptable consequences result. ~Res onse: The justification for the selection of limiting single failures was presented in the response to NRC guestion 440.9. For the one pump resistance to forced flow with a loss of offsite power as a result of turbine trip event, the percent of fuel pins with CE-1 DNBR less than 1.19 should not be used to determine fuel failure since; (1) a CE-1 ONBR less than 1.19 does not mean thatagiven,fuel pin will experience DNB, and (2) DNB does not necessarily result in fuel failure. For. these reasons, the approach proposed by NRC for calculation of fuel failures is unduly 'conservative. A more reasonable, yet still. conservative, method of calculating fuel failures, presented in CENPD-183, was submitted to NRC in July 1975. Using this method for St. Lucie Unit No. 2 results is postulated DNB and assumed failure of 134 of the fuel pins as presented in the FSAR. The percentage should be used in::eval-uating the consequences of this accident; The description and justification of the C-E method is provided in the response to NRC guestion 440.11. The flow. coastdown which was used in the analysis of the one pump resistance to forced flow is presented in Figure 440.25-1. This figure shows the variation of core flow fraction with time. The seized shaft is assumed to instantaneously stop at time 0.0 with the seized. rotor acting only as a resistance to flow. This coastdown was generated using the COAST code as documented in CENPO-98 (see Reference 440,25-1).
~~~~Reference:~~~~
~~~~l. "Coast Code Description", CENPD-98, April 2, 1973.~~~~
~~~~~P/fhflg A change to the FSAR, Sec+i~15.~.~., accompanies this response.~~~~
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~~~~~ St. Lucie.2 Seized Shaft IU'."~~~~
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'40.28 Your FSAR i rdicates that operational procedures allow detection of a (15.4.2) boron dilution event 15 minutes prior to criticality. This is not acceptable. The staff >>ill require that alarms be available to alert the operator to a. boron dilution transient 15 minutes prior to crit-icality (30 minutes when in refueling mode). Show that, the plant is protected for all postulated boron dilution events assuming the worse single active failure. In particular, consider the failure of the first alarm. If a second alarm is not provided, show that the con-sequences of the most limiting unmitigated boron dilution event meet the staff criteria and are acceptable. Also, indicate for all six modes, what alarms would identify to the operators that a boron dilution event was occurring. Confirm that the results of these analyses meet the acceptance criteria for these events per SRP 15.5.1.
~~~~~Res ense:~~~~
~~~~SRP 15.4.6 requires that at least 15 minutes is available from the time~~~~
*the operator is made aware of an unplanned boron di,lution event to the time a loss of 'shutdown margin occurs during power operation (auto-matic control and manual modes), startup, hot standby, and cold shutdown.
For NODES 1 through.6 any of several alarms and/or indications will provide the operator with at least 15 minutes (For NODE 6, 30 minutes) to terminate the event before the shutdown margin is lost. The indications and/or alarms available to alert the'operators that a boron dilution event is occurring in each of the operational modes are outlined below.
l. The'following control room indications and corresponding pre-trip alarms are available for NODES 1 and 2: a high power or, for some set of conditions, a high pressurizer pressure trip in fODE 1 or a high logarithmic power level trip in MODE 2. Furthermore, a high TANG 'alarm may also occur prior .to trip.

2. In MODES 3'and 4 with CEAs withdrawn, the high logarithmic power level trip and pre-trip alarm will provide an indication to alert the operator of an inadvertent boron dilution.

3. In MODES 3, 4, and 5 with CEAs fully inserted and in MODE 6, 'a high neutron flux alarm on the startup flux channels will provide indication of any boron dilution event. Limiting boron dilution events in subcritical operating modes will be analyzed to establish the startup channel alarm setpoint and reset time. The times to complete loss of shutdown margin, and hence reactivity insertion rates, and neutron flux responses at the startup channel excore detectors will be determined such that the startup channel alarm
*setpoints based on these responses satisfy the requirements of SRP 15.4.6.
~~~~alarm will~~~~
~~~~'his be powered by an onsite power source in the event the offsite power is lost.~~~~
w The times to loss of shutdown margin calculated for the postulated boron dilution event represent the fastest credible dilution rates and, therefore, the shortest time for each mode. Conside'ration of additional single failures would not increase the dilution rate, and therefore, would hot reduce the time to loss of shutdown margin. The only failure of significance involves the loss of the indications-that alert the operators to a boron dilution. In HODES 1 and 2, there are no single active <-ailures that result in the loss of any of the RPS alarms used to alert the operators that a boron dilution is, in progress. In HODES 3, 4, 5, or 6, in case one or both startup flux channel alarms become inoperable, the operators would be required to implement operational procedure guidelines which would assure detection of a boron dilution event. In HODES 3, 4, and 5; the guidelines are based on determining the RCS boron concentration by either boronometer or RCS sampling at frequencies which depend on the mode of operation. No single active failure can eliminate more than'one of the methods of monitoring or determining the RCS boron concentration. In NODE 6, the boron dilution event is pre-cluded becasue the manual isolation valve (V 2183) in the makeup. water line and the primary makeup water supply to char ging pump isolation valve {V 2180) are normally, locked closed in this mode.'+ 7 7aI A change to th e FSAR, Sections15. f.2 /&accompanies this response. 4
SL2"PSAR 7,7.1.1.10.3 Turbine Runback The following inputs cause a.turbine runback: a) One main feedwater pump tripped b) Two heater drain pumps tr.ipped The runback input causes e contact to close in the DEll runback circuitry. The turbine runs baclc at a predetermined rate until the contact opens at which time the runback is stopped. In the case of the heater drain pumps, the runback is stopped at 70 p rcent of full load as determined by first stage pressure, and in the other case the runback is stopped when feed-water flow and steam flow are equal. XNSBh' 88 7.7.1.2 Design Com arison The design differences between the control systems in'the St I.ucie Unit 2 design scope and the control systems provided for the reierence plant are discussed in this section. 7.7.1.2.1 Reactivity Control Systems The RRS is functionally identical to that supplied for St Lucie Unit 1 (HRC Docket 50-335). The CED!JCS combines the Control Element Drive System (CEDS) and the coil power programmers (CPP) into one integrated system thus reducing the inter-facing required between the previous two separate subsystems. The'EDlfCS is functionally identical to the CEDS/CPP of St Lucie Unit 1 with the following changes: The CEAs axe controlled in subgroups consisting of four or five CEAs located symmetrically about. the core; All timing functions within the CEDHCS are performed using digital tech-niques to increase the accuracy and flexibility of the integrated system; The CEA withdrawal prohibit (CWP) is effective in all modes, and CWP can be bypassed at the operator's module; . While the CEDNCS is in the automatic sequential mode, either or both part-length CEA (PLCEA) groups can be inserted or withdrawn, but motion of individual CHAs or PLCEAs is not possible; While in the automatic sequential mode, the CEA motion inhibit (CHX) cannot be bypassed and the system can handle up to 91 CEAs, 7.7.1.2.2 Reactor Coolant Pressure Control System The reactor coolant pressure. contxol system is functionally identical to that sup'plied for. St Lucie Unit (NRC Docket 50-335). 1 7.7-9 Amendment No. 0, (12/80)
S3.2"FSAR 7.7.1.2.3 Prc. surizer Level Control System The Pressurizer 3.evcl Control Syrtcm is functionally identical to that supplied for St Lucic Unit 1 (NRC Docket 50-335). 1 7.7 '.2.4 Feedwatcr Regulating Sy"tern The Fcedwater Regulating Syst"m is functionally identical to that supplied for St Lucie Unit 1 (NRC Dockbt 50-335). 7.7.1.2.5 Steam Dump and Bypass Control System The Steam Dump and Bypass Control System is functionally identical to that supplied for St Lucie Unit 1 (NRC Docket 50-335). 7.7 '.2 ' Analog Display System The Analog Di. play System is functionally identical to the metrascope
~
supplied for St Lucie Unit 1 (NRC Docket 50-335). 7.7.1.2.7 . Boron Control System. The boronometer is functionally identical to that supplied for 4'aterford Steam Electric Station Unit 3 (NRC Docket 50-382). The only difference is that the recording range is switch selectahle for 0-1250 ppm and 0-5000 ppm. For any di.fferences in the control of boration and deboration see Subsection 9.3.4.* Incore Instrumentation The Incore Instrumentation System. is similar to that supplied for Arkansas Nuclear One-Unit 2 (NRC Docket 50-368). The difference being 44 detector assemblies vs 56 on St Lucie Unit 2. 7.7.1.2.9 Excore Neutron Fl ux 3fonitoring System The start-up and control channels of the Excore Neutron Fl'ux Honitoring System are functionally identical to that supplied on System 80 (NRC Docket STH-50470F). The safety channel's are of a new design but based on System 80 circuitry. 7.F 1.2.10 Digital Data Processing Syst: em The Digital Data Processing System is functionally identical to that sup-plied for St. Lucie Unit 1 (NRC Docket 50-335). The only difference is that the Unit 2 system has redundant computers. 7.7-10 Amrndmcnt No. 0, (12/80)
t nsert BB 7.7. 1. 1. 11 Boron Dilution Alarm System Reactivigy control in the reactor core is affected, in part, by soluble boron ih 'reactor'oolant system. The Boron Dilution Alarm System (Figure 7.7-8) utilizes the startup channel nuclear instrumentation signals to detect a possible inadvertent boron dilution event while in Modes 3-6. There are two redundant and independent channels in the Boron Dilution Alarm System (BDAS) to ensure detection and alarming of the event. The BDAS contains logic which will detect a possible inadvertent boron dilution event by monitoring the startup channel neutron flux in-dications. 1<hen these neutron flux signals increase (during shut-down) to equal or greater than the calculated alarm setpoint, alarm signals are initiated to the Plant Annunciation System, The alarm setpoint will only follow decreasing or steady flux levels, not an increasing signal. The current neutron flux indication and alarm setpoint (per channel) are displayed". There is also a reset to allow the operator to acknowledge the alarm and initialize capa-'ility the system. The BDAS will be powered from an offsite power source with an onsite backup power source. Insert CC 7.7. 1.2.11 Boron Dilution Alarm System The Boron Dilution Alarm System is an addition to the St. Lucie Unit.'2 design. There is no functional comparison to St. Lucie Unit I (NRC Docket 50-335).
e FIGURE 7.7-g
, BOROtl D I LUT ION ALARl1 SYSTEtl S It)PL I F I ED~ BLOCK DI AGRlN Reset Boron Dilution Startup Channel Current Flux 8 Nuclear Alarm System 'nstrumentation Signal Setpoint Display Logic Alarm Signal to the Plant Annunciation ~
System Note: Only one of two identical channels is shown.
SI,2- 1'SAR lh,d,2.S l.imitinr~lo.".s nf Shutdow~nhsr la fvaat " Slow Po'litivr. Rrsc"
~tivit lnsnrtion 15 ~ 4 ~ 2.4 ~ 1 Identificat:inn of Event and Causes The Infrequent .event groups from the Reactivity and 1'ower Distribution lhnomalies event type and the Infrequent event coiobinations shown in'able J.5,4.2-1 were compared to find. the event: combinar ion res>>)ting in the closest approach to the complete loss of shutdown margin. Tire Slow Posi" tive Reactivity Insert:ion was identifit:d as tlute most limiting event because no other Infrequent event affects shutdown margin.
The.event:groups and event combinations evaluat:ed and the significance of the approach to the loss of shutdown margin,acceptanc'e guideline for each are indicated in Table I 5.4 2-1 a a The slow positive reactivity insertion may occur due to a closure of a boron flow cont:rol valve or a m lfuncticn of the makeup controller which causes a boron dilution. The most limiting initiating event'resulting in a slow positive reactivi y insertich is a malfunction of the makeup controller mode selector swi'ch in
. the di'lute mode. This may occur. by a failure in the boron control system which causes continuation of the makeup operation after a planned dilution has been completed. This failure results in the maximum possible dilution rate.
The other initiating event which can cause a slow positive reactivity in-I sertion is the failure of thc solenoid in the. boron flow control valve in the boric acid line with the boron control syst: em in t: he autorlatic or borate modes, This failure result:s in termination of the boron flow to t: he kCS and thus'ould approach the loss of shutdown margin at the same rate as makeup mode selector swit:ch malfunction, \lowever, this event yields a low flow alarm in the boric acid line which alerts the operator at the ini" tiation of the event:. The makeup mode. selector switch malfunction would not produce an immediate alarri. and therefore is more 1" miting than the in-advertent closure of the boron flow contxol .valve. Analysis of a slow positive x'eactivity insertion event initiat:ed during each of the six operational modes defined in the Technical Specifications was performed. These analyses show t:hat bode 5 (cold shutdowr.) results in the learnt t:ime ava-lable for detection and t:ermination of the event. This is because the shutdorv~> margin requirement which will be speci fied oy the Technical Specification" ia "mallest in Hade 5 (i.cat tvo percent dp auhcritical). 'ln either. Hode five orl,tode -tx, and utch the RSS 'eral lowered, administ:rative procedures governing the frequency of boric acid sam ling will preclude reaching criticality. 15a4a2e4a2 Sequence of Events and Systems Operations Table l5.4.2.4-1 presents a chronological list and t:iming of system actions which occur fo11owing a boron dilution eventa Refer to Table 15a4a2a4"l. wl>ile reading this and the following section. The success patlrs referenced are'hose given on t: he sequence of. events diagran (SED), Figure 15.4 2.4-I ~ ~ 15.4-53 Anlen(lllrent No. 2, (5/81)
e SL2-) S/;It This f i->>r~p t.'u"-thrr with Table 1'p 0-6, whici> contains a glossary ot Sl;D
~
mbols and acronyn<s, may br usrd to trace thr actu;<tion and interaction of systems uord to mitigate th>> <<ouse<)u!nces oE t)iis event, The timings Table 15 4 s 2 s4 1 <nay be used to determine when, af ter the init iat i ng eve<tt, each action occurss The sequence nE event;s and systrms operastiono described below represents the way tn w)tic)t'th>> plant was stszunad tn respcnd to the event initiator,
))any plant responses are possible, however, certain responses are lin<iting wit)t reaps><<t to the acceptancr. guidon'nes for this sections Of" th" li<nit-ing responses, the most: likely one to bc followed was selecteds Table 15,4,2,4-2 contains a inatrix which describes the 'extent t:o <kich nor<nally operating plant systems are assuned to function du<in'he tran-sients The operation of these systems is consistent with the guidelines of Subsection 15.0s2.3.
Table 15'4s2s4-3 contains a matrix which describeo the extent to which safety systems arr. assumed to function during the transient, 'The success paths in the sequence of event:s dia< rams, Figure 15,4,2,4-1, are as follows:
~~~~~ ~4tAP~~ $Lv ~ wJ cJ~.~~~~
~~~~React:ivity Control:~~~~
~
~~~~J x. e operator is al erted to a decrease in Reactor Coolant System (PCS) boro~~~~
~
concentration e!.tner throughVsempling,'."boronoineter indications~ or-by- 'I Arl Gs~Q etartup-eflux-channel inaicationeo. Vie t;urns off the cha'ging pumps and closes the letdown control valves. in order'to halt further dilut'ion, The operator then turns off,the primary makeup pu.np tnd closes the primary makeup isolation valve to stop the flow of primary makeup water to the charging pumps, .)text, he increases the RCS boron conc'entration by opening the boric acid gravity feed line from th'e boric acid makeup tank to the charging pump suction and restarting the charging pumps to provide borated water to the RCS, Letdown flow may be diverted to the flash tank to in-crease the rate. of boration, FP<L is rsvicwinp a proposed uethod of providins rtdurdant indications of boron dilution t'bat utilis: control rosie indicstron and RCS samplin< at varying frequencies (depending nn plant operating node), FP&L will advise t:>e NRC of the results of this review, s 15d4d2s4s3, . Analysis of Ef frets and Consequences a) Hathematical l)odel Complete mixing of boron in the RCS and equal letdown and cnarging Elowrates arr. assumed. The <at:e nf change of boron concentration
~~~~~ during a diLution in which <<atcr without boron is added and coolant at the tine deprndent RCS boron concentration is re ioved is described by the following differentia1 equation:~~~~
15.4-54 Antendment No. 4, (6/81)
t Sl.?-FSAR
4) Complete mixing of: thc boron in t:hc RCS is a ss~nncd because of the large RCS mass circu1ation by a minianmi of one low prcssure safety injection pump operating in thc:hutdown cooling mode, compared to thc 'relatively small mass added through the charg-ing pumps.
The critical boron concentration at cold "hutdown wi.th a11 CEAs in is 845 ppm including,uncertainties. The inverse boron worth is 55.8 ppm/% lb'hich includes uncertainties. Applying 2th'/6~7 i~i ~ uncertainties to this number in the most conservative direc-
~tion I'be initial subcritical boron concentration for the cold mode is found by adding the product of the inverse .'hutdown boron worth and the minimum shutdown margin required (i.e., two percent hp ) to the critical boron concentration. The result-ing minimum initial boron concentration'n Hode 5 is .956.6 ppm.
~~~~The parameters discussed above arc summarized in Table 15.4.2.4"4. c) Results~~~~
-The conservative parameters listed in Table 15.4.2.4-4 are used in Equation 3 to calculate the time to criticality during a Slow Posi-tive Insertion.. The minimum possible time to dilute from two~ rcent hs sub-cri"ical to criticality is 62 minutes. fOperational procedures periodic monitoring of the startup flux channels allow detection of the event with at least 15 minutes available to terminate the event before criticality is reached.
The 61Ota PO" itiVa ('skc d... CcLK ReaCtiVity InSertiOn in7.fcdR=F dcaS bbT rbranir 'iu paar . d... rue.vc, any pressure or temperature perturbations in the RCS because the event gg terminated before criticality is reached. Other principal 1'CS and secondary system parameters are not perturbed by this event. As the RCS. boron concentration is reduced by the dilution, positive reactivity is inserted. This causes the two percent Ap subcriticality margin to be reduced but the core does not become cr tical. 15.4,2.4.4 Conclu'sions II This evaluation shows that the plant response to a Slow Positive Reactivity Insertion vill produce results within the acceptance guideline for Infre-quent events in Table 15.0-4, f r
~~~~15. 4- 56 Amendment No. 2, -(5'/8l)~~~~
~~~~S),2-1'SAR~~~~
~~~~,SESQUINCI'. OF )>V)'.HTS, COB)l) SPOKE'))ihC Tii(I'.S AND SU)li~)h)LY OF Rl',SU].TS FOR SI.O';) )>OS'CATV); )<I',hG'1')VJTY 1HSI.'I<T10N Success Pat)cs 0~~~~
4J oj 0> C>> 4J Sw M C! 0 Q Q Lt r. C0 Analysis Q 4J cs r c> 0 0 4J0 C> f-g CO C> 4J Set Point C> E:C C> ~ Time c> o 0 C 4 C oz'alue S4 O C>>> C>> C>> W Rc >-C Sec Event 0 Makeup mode selector switch malfunction, 'RCS boron concen-tration, ppm<< 1800 Operator'dc.tects,event through operating.'procedures ZtuZG)~7'~ 0 Operator turns off charging +78 37zo pumps to terminate the event~ QGS-boron-ee~n trw t.ia~pm-Amendment No, 2, (5/81)
Inserts to 15.4 '.4
~~~~~%7 3 The indications and/or alarms available to alert the operators that a boron dilution -event is occurring in each of the operational modes are outlined below.~~~~
~~~~1. The following control room indications and corresponding pre-trip alarms are available for NODES 1 and 2: a high power or, for some set of eon-~~~~
~~~~'itions, a high pressurizer pressure trip in MODE 1 or a high logarithmic power level trip in MODE 2. Furthermore, a high TAVG alarm may also occur prior to trip.~~~~
2. In MODES 3 and 4 wi'th CEAs withdrawn, the high logarithmic power level trip and pre-trip alarm will provide an indication to alert the oper-ator of an inadvertent boron dilution.

3. In NODES 3, 4, and 5 with CEAs fully inserted and in MODE 6, a high neutron flux alarm on the startup flux channels will provide indication of any boron dilution event.

4. In NODE 5 with the RCS par tially drained for system maintenance, the startup flux channel. alarm will provide indication of. any boron dilution event. In this plant condition, administrative controls would allow operation of only one charging pump at a maximum rate of 44 gpm. Plant operating procedures will require that the power to the other two char-ging pumps be removed and their breakers locked out.
This drained down case is less limiting than th'e MODE 5 event presented above. The operational procedure guidelines, in addi tion to these indications and/or alarms, will assure. detection and termination of the boron dilution event before the shutdown margin is lost in accordance with the requirement of SRP 15.4.6. gp5EW . The critical boron concentration with CEAs withdrawn (All Rods Out);. the inverse boron worth, and the net rod worth for the cold shutdown conditions are 984.5 ppm, 55.8 ppm/Khp, and 2.5Ahp respectively, including uncertainties. The critical boron concentration value of 845 ppm was obtained by subtracting the product of the inverse boron worth and the net rod worth from the critical boron concentration with all rods out. ~5~7 3. A high neutron flux alarm on the startup flux channel will assure detection of' boron dilution event with at least 15 minutes prior to criticality as'er the requirements of SRP 15.4.6. 2820 High neutron flux alarm on the startup flux channel alerts operator . to a boron dilution event.
e Discuss the provisions and precautions for assuring proper, system filling
~~~~~ ~~~~~
~~~~40.38 and venting of ECCS to minimize the potential for water hammer.and air~~~~
~
~~~~(6.3)~~~~
~
~~~~binding. Address piping and pump casing venting provisions and sur-~~~~
~~~~~ ~~~~~
veillance frequencies. ~Res onse The ECCS system is provided with sufficient drainage capability on the piping low points and system vents on the piping high points to assure that air will not be entrained in the system. The ECCS components are provided with vent and drainage capability. The HPSI pumps, LPSI pumps, and shutdown heat exchangers are provided with component vents and drains as shown in Figure 1.2-34. The piping vents and drains are shown on Figure 6.3-1a'nd 6.3-1b. Prior to system operation, the ECCS piping and components will be adequately vented in order to minimize the potential for water ha+acr and,air binding. Administrative procedures will be written to ensure that the ECCS piping and components are properly drained and filled.
40.39
~~~~~ Identify all ECCS valves that are required to have power locked out; confirm they are included under the appropriate Technical Specifications, ~~~~~
~~~~(6.3)~~~~
~
~~~~with surveillance requirements listed.~~~~
~
~Res ense The ECCS valves that are required to have power locked out are listed below. The Technical Specification section of the St. Lucie-2 FSAR is'urrently being gen-erated. Surveillance requirements for these valves will be listed.
1) V-3550, Y-3551 - Hot Leg Injection Isolation Valves. "Power rack out required to motor during plant power operation".

2) V-3614, V-3624, V-3634, V-3644 - SIT Isolation Valves. "Power rack out to motor required when pressurizer pressure greater than 700 psig."

3) V-3613, V-3623, Y-3633, V-3643 - SIT Vent Valves. Power to those valves is removed in the control room during normal operation.
~~~~Identify the plant operating conditions under which certain automatic~~~~
~~~~~ ~~~~~
~~~~40.41~~~~
~
~~~~safety injection signals are blocked to preclude unwanted actuation of~~~~
~
these systems. Describe the alarms available to alert the operator to a failure in the pi y ~d available to mitigate the 1 ill consequences t'ai Pt of such f 0 ti an accident. Ith ti
~Res onse Mhile the plant is in power operation, the safety injection signals may not be blocked. During the interim phase, while RCS pressure, is being reduced to re-fueling mode, it becomes necessary to partially block the SIAS.
A safety injection block is provided to permit shutdown depressurization of the Reactor Coolant System (RCS) without initiating safety injection. This block is accomplished manually after pressurizer pressure has been reduced and a per-missive signal is generated by the Engineered Safety Features Actuation System.
~
This blocking proce'dure is under strict administrative control; block and block permissive is annunciated and indicated in the control room. It is not possible to block above a preset pressure: if the system is blocked and pressure rises above that point, the block is automatically removed. The block circuit com- ~ plies with the s'ingle fail'ure criterion in IEEE 279-1971. he SIAS block removes only the pressurizer pressure signal from the SIAS trip logic. The high containment pressure transmitters still remain in direct con-nection with the trip logic. Should an event occur whereby the containment pressure is sufficiently raised, high containment pressure alarms sound on RTG B-206 and the SIAS is initiated automatically, regardless of the pressurizer signal block. The Technical Specifications will permit blockage of the SIAS in plant modes 5 and 6, while the'hutdown cooling system is in operation. In these modes pro-tection against overpressurization of the Reactor Coolant and Shutdown Cooling System, due to a spurious actuation of the HPSI, is provided by relief valves Y-3666 and V-3667 in the SDC suction lines. FSAR Tables 7,5-1 and 10.4-5 in-dicates the display instrumentation and their alarms<h~~h<'available to the operator to establish primary and secondary system conditions. During cold shutdown or r'efueling {modes 5 and 6) should a loss of coolant occur,
. level guages in the containment and cavity sump and the safeguards room sump with alarms would alert the operator of such an accident. During the plant cooldown, operator action is required to continually monitor the S.G. secondary water 3evel and feedwater flow. Because of this the operator is aware of the secondary system conditions.
During a refueling, for specific maintenance tasks, it is'xpected that some instrumentation will be inoperable. Administrative procedures will assure that the operator will be able to assess the status of the primary and secondary sys-tems for the specific situations,'
40.44 A reported event has raised.a question related to the conservatism of NPSH calculations with respect to whether the absolute minimum avail-able NPSH has been taken by the staff as a fixed number the applicant by either the architect engineer or the pump supplied'hrough manufacturer. Since a number of methods exist and the method used can affect the suitability or unsuitability of a particular pump, i't is requested that the basis oh which the required NPSH was de-termined be branded (i.e., test, Hydraulic Institute Standards) for all the ECCS pumps including the testing inaccuracies be provided.
~~~~~Res on'se The required NPSH of the St. Lucie Unit 2 ECCS pumps is confirmed by test. The high pressure safety injection pumps are supplied by Bingham-!li llamette Co.~~~~
These'pumps are tested in accordance with the ASNE power test. code 8.2.(cen-trifugal pumps), Each of the St. Lucie Unit 2 HPSI pumps were tested for the NPSH required at the runout flow. Similar pumps were also supplied for St. Lucie 'nit l. Each of the St. Lucie Unit 1 pumps were also tested for the NPS)I re-quired. The results show (see following table) little variance between pumps for s,imilar flow. The LPSI pumps are supplied by Ing rsol-Rand. The NPSH characteristic is eon-firmed by test. Both of the St. Lucie Unit 2 LPSI 'pumps were tested. The ydrualic Institute Standards were used for the tests. NPSH;,TEST RESULTS FOR ST. LUCIE UNITS 1 AND 2 1 St. Lucie Unit 1, HPSI. Pum s , GPN ~NPSH ft
~~~~¹200113 640'40~~~~
~~~~19. 7~~~~
¹200114 19.9 ¹200115 640 19.6 St. Lucie Unit 2 HPSI Pum s ¹14210014 (spare pump) 640 19.9 '14210015 631 19.0 ¹14210016 639 19.4 St. Lucie Unit 2 LPSI Pum s ¹1076149 3000 13.0 ¹1076150 3000 -11.0 The NPSH vs. flow curves for the St. Lucie Unit 2 HPSI and LPSI pumps are shown in'igures 6. 3-3a, 6. 3-3b, 6. 3-4a, and 6. 3-4b.
I 440. 51 In the event of early manual reset of the safety injection actuation 6.3) signal (SIAS) fo1'lowed by a.loss of offsite power during the injection phase, operator action may be required to reposition ECCS valves and restart some pumps.- The staff requires that operating procedures specify SIAS manual reset not to be permitted for a minimum of IO minutes after a LOCA..Provide the administrative procedures to ensure correct load application to the dies@ generators in the event of loss of offsite power following an SIAS reset. ~Res ense The SIAS can only be reset when the initiating signal has been removed; i.e. normal conditions have been reestablished. If the signal that generates an SI'AS is still present, the SIAS cannot be reset. Following a loss of offsite power subsequent to an SIAS manual reset, the safety injection pumps and valves will not load onto the diesels if the con-ditions that require automatic safety injection are not present. However,. if the conditions that require automatic safety injection are present after the manual SIAS reset followed by loss of offsite power, the, safety injection pumps and valves will sequence onto the diesels automatically. No operator action is required. uring low pressure operation of the safety injection system, during shutdown ooling; the. operating procedures will require the operator to manually load the low pressure safety injection pumps onto the diesel generator following a loss of offsite power. The required actions that would provide SIAS when the pressurizer pressure signal is locked out (during depressurization for shutdown) are given below. I An SIAS is initiated by a low pressurizer pressure signal or a high containment pressure signal.. There are four independent pressure transmitters each for the containment and the" pressurizer. In order to allow depressurization of the pressurizer (i.e. system) a safety injection block is provided by manually block'ing only the pressurizer transmitter's signals to the SIAS trip logic. The containment pressure transmitters remain in direct connection with the SIAS trip logic. Therefore, an inci'dent which would raise the containment pressure sufficiently >>ill automatically initiate an SIAS (no operator action is required). If necessary, the 'operator can manually initiate an SIAS, as described in FSAR Section 7.3. 1. 1. 1. Should the situation be evaluated as requiring less than full actuati'on, the operator can align the safeguard pumps'n a component basis to provide makeup water for the reactor coolant system.
e e
C/Ielgi Question 440.54 Describe the means provided for ECCS pump protection including instrumentation and alarms available to indicate degradation of ECCS pump performance. Our position is<<that suitable means should bc provided to alert the operator to possible degradation of ECCS pump performance. All instrumentation associate'd with monitoring the ECCS pump, performance should .be operable without offsite power, hand should be able to detect conditions of low discharge flow. Describe dur'ing post-LOCA operation (injection mode and recirculation mode). t Response 440.54 Below are listed instrumentation used in conjunction with the Low and High Pressure Safety Injection (LPSI and HPSI) pumps for use in determining pump performance:
1. P-3314 and P-3315 are used for LPSI 2A and 28, respectively. They are used to determine pump discharge pressure. They have indicators on local panels.

2. P-3316 and P-3318 are used for HPSI 2A and 28, respectively. They determine pump discharge pressure, and have indicators on local panels.

3. F-3301 and F-3306 determine total flow (minus any miniflow) for, LPSI 28 and 2A, respectively. They indicate, record, and control the flow. The recorder. is usable as'n indicator by the operator.
~~~~There is an indicator display on t'e Hot Shutdown Panel.~~~~
4. F-3312 (LPSI A), F-3322 (LPSI'A),'-3332 (LPSI 8), and F-3342 (LPSI 8) are used to determine flow through the various LPSI flow branches they have indicator displays in the control room.
~~~~S. F-3317 and F-3327 determine total SDC flow from HPSI 2A and 28, respectively. The results are recorded in the control room.~~~~
~~~~6. F-3315 and F-'3325 determine total SDC flow from HPSI 2A and 28, respectively. The results are displayed on an indicator in the control zoom.~~~~
~~~~7.'-3313, F-3323, and F-3343 determine branch flow from HPSI pumps A and B. The flows determined are recorded in the control room.~~~~
~~~~8. F-3311, F-3321, and F-3341 determine branch flow from HPSI pumps A and B. The results are displayed on an indicator in the control room.~~~~
~~~~, 9. Low flow alarms are being added'to the LPSI and HPSI pumps. These alarms will have emergency power.~~~~
Question 440.58 0 List all ECCS valve operations and controls that are'ocated below the maximum fIood level following a postulated LOCA or main steam line break. If any are flooded, evaluate the potential consequences of this flooding both for short and long-tenn ECCS functions and containment isolation. List all control room instrumentation lost following these accidents. Response 440.58 The maximum flooding event, which results from a large LOCA, will cause the water level inside containment to reach an elevation of 26 feet. This conservatively assumes that the entire contents of the Reactor Coolant System drains and that the Refueling Mater Tank was at its overflow level at the time of the accident. The operation of safety related equipment in a post-LOCA, potentially submerged, environment will be addressed in accordance with NUREG 0588 Appendix E and will be submitted by November 30, 1981. As stated in FSAR section 3.11.6 this study will confirm that no essential equip-ment will be'ost as a result of the maximum postulated post accident containment water level.
r ,ct
440.59 it (If) is our position that the SIS hotleg injection valves should be (6;3) locked closed with power removed during normal plant operation in order to prevent premature hotleg injection following a LOCA. ~Res ense The hotleg SIS injection valves (V-3540, V-3523, V-3550, and V-3551) do not have power removed during normal plant operation because there are two (redundant) valves in each line. Administrative procedures ensure that these valves are locked closed'"the control room. In addition, each set of valves is provided with an open/closed status indication in the control room.
Question 440.61 During our reviews of license applications we have identified concerns related to the containment sump design and its effect on long term cooling following a Loss of Coolant Accident (LOCA). These concerns are related to (1) creation of debris which could
~~~~'potentially block the sump screens and flow passages in the ECCS and the core, (2) inadequate NPSH of the pumps taking suction from . ~~~~~
the containment sump, (3) air entrainment from streams of water or steam which can cause loss of adequate YPSH, (4) formation of vortices which can cause loss of adequate NPSH, air entrainment and suction of floating debris into the ECCS and (5) inadequate emergency pro-cedures and operator training to enable a correct response to these problems. Preoperational recirculation tests performed by utilities have consistently identified the need for plant modifications. The NRC has begun a generic program to resolve this issue. However, more immediate actions are required to assure greater reliability of safety system operation, Ve therefore require you take the following actions to provide additional assurance that long term cooling of core can be achieved and maintained following a postulated the'eactor LOCA.
1. Establish a procedure to perform an inspection of the cootainment, and the containment sump area in particular, to identify any materials which have the potential for becoming debris capable of blocking the containment sump'when required for recirculat'ion of coolant water. Typically,'hese materials consist of: plastic bags, step-off pads, health physics instrumentation, welding equip-ment, scaffolding, metal chips and screws, portable inspect,ion lights, unsecured wood, constuction materials and tools as well
~~~~'as other miscellaneous loose equipment. "As licensed" cleanliness should be assured prior to each startup.~~~~
2. Institute an inspection program according to the requirements of Regulatory Guide 1.82, item 14.. This item addresses inspection of the containment sump components including screens and intake structures.

3. Develop and implement procedures for the operator which address both a possible vortexing problem (with consequent pwnp cavitation) and sump blockage due to debris. These procedures should address all likely scenarios and s3iould list all instrumentation" available to the proper operator (and its location) to aid in detecting problems which may arise, indications the operator should look for, and operator actions to mitigate these problems. '
~ Pipe breaks, drain flow and channeling of spray flow released below or impinging on the containment water surface in the area of the sump can cause a variety of problems; for example, air entrainment, cavitation and vortex formation-Describe any changes you plan to make to reduce vortical flow in tlute neighborhood of the sump. Ideally, flow s3iould approach 'n i iformly f rom all d rect fons.
~~~~0 e Question 440, 61 (Cont'd)~~~~
5. Eva]uate the extent to which the containment sump(s) in your plant meet the requirements for each of the items previously identified; namely debris, inadequate NPSll, air entrainment, vortex formation, and.operator actions.
The following additional guidance is provided for performing this evaluation. (1) Refer to the recommendations in Regulatory Guide 1.82 (Section C) which may be of assistance in performing this evaluation. (2) Provide a drawing showing the location of the drain sump relative to the containment sump. (3) Provide the following information with your evaluation of debris: (a) Provide the size of openings in the fine screens and compare this with the minimum dimensions in the pumps which take suction from the sump (or torus), the minimum dimensions in any spray nozzles and in the fuel assemblies in the reactor core or any other line in the recirculation flow path whose size is comparable to or smaller than the sump screen mesh size in order to show that no flow blockage will occur at any point past the screen. (b) Estimate the extent to which debris could block the trash rack or screens (50 percent limit). If a blockage problem is identified, describe the corrective actions you plan to take (replace insulation, enlarge cages, etc.) (c) For. each type of thermal insulation used in the containment, provide the following information:. (i) type of material including composition and density, (ii) manufacturer and brand name, (iii) method of attachment, (iv) location and quantity in containment of each type, (v) an estimate of the tendency of each type to form particles small enough to pass through the tine screen in the suction lines. (d) Estimate what the effect of these insulation particles would be on the operability and performance of all pumps used for recirculation cooling. Address effects on pump seals and bearings.
~~~~Response~~~~
l. St Lucie wi11 institute an inspection program to verify that the containment is free of debris that .may lead to blockage of or damage to the ECCS sump. These inspections will assure that the containment and sump are in the "as licensed" state of cleanliness prior to each reactor startup.
2~ The sump inspection program will include an examination of sump structures, such as intakes and screens, as outlined in Regulatory Guide 1.82. 3~ Long term cooling operating procedures will require periodic veri-fication of system performance to insure safe operation under re-circulation conditions.
4. The dynami.c effects associated with pipe whip and jet impingement of all high energy lines in the vicinity of the sump have been evaluated. In no case would any high or moderate energy piping failure compromise the functional capability of the ESF sump when .
it is required. Therefore, the sump model test outlined in the response to question 440.60 will not simulate spray flows im-pinging on the containment ~ater surface. However, various screen blockage tests will be performed to simulate worst case approach flow and flow channeling conditions.
5. 1. The St Lucie Unit 2 sump consists of one large full capacity res-ervoir which physically separates the redundant ESF suction lines by approximately 15 feet. The sump design, described in detail in FSAR section,6.2.2.2.3, meets all the requirements of Regulatory Guide 1.82 with the exception that only one sump is provided. The literal intent of this last requirement is satisfied by the use of fine screen to separate the suction lines.

5. 2. The relative locations of the Reactor Cavity and containment sumps can be seen on FSAR figure > ~ 2 5.3. a) The sump design incorporates a ninety (90) mil fine mesh filter screen to protect the suction piping from entrained particles.
These screens are sized to eliminate all particles too large to pass through the reactor fuel assemblies whiih is the most restrictive flow path in the system. Particles smaller than this will pass through all system components including reactor, pumps, heat-exchangers and spray nozzles. b) Debris generated inside containment as a result of an accident will be confined between the primary and secondary shield walls. Large debris generated here is prevented from reaching and possibly damaging the sump by the Seismic Category I trash racks located at the secondary shield wall openings. Although it is believed that this design vill minimize debris at the sump, model tests will be performed assuming blockage of half of the vertical screens and all of the horizontal screens.
Response 440.61 (Cont'd),
~~~~~.~. (Cont'd) c) A description of the various types of insulation expected to be used inside containment and an estimate of the quantities appear in FSAR Table 6.2-40.~~~~
~~d) As stated previously, particles small enough to pass through the fine, screens can pass through the systems without dele-terious effects. Pump operability is not expected to be impaired.~~
~~~~6. The Reactor Drain Tank, located in the sump, is designed to remain in place following an accident. The effect of uplift loads re-~~~~
~~~~suiting from the submergence of an empty tank have been analyzed and found to be well within the capabilities of the holdown bolts.~~~~
A Containment Isolation Signal (CIS) isolates the tank and stops the drain pumps. esponse 6 address addxticnal NRC concern expressed in the review 'meeting f July 23 and 24, 1.98l regardi ng Reactor Drain Tank.
The submittal for the LOCA analyses does not address the effects of steam generator tube plugging. The effect of a decrease in steam generator tube flow area is an increase in the peak cladding temperature (when the peak occurs during the reflood portion of the transient). If the analyses provided are considered to support generators with plugged tubes, describe the extent of the plugging the analyses support and the method used to account for the plugging. If steam generator tube plugging was not considered, the appli-cant will be required to perform additional ECCS analyses prior to operation with plugged generator tubes. In either case, the applicant is required to include an interface requirement on the validity of the LOCA analyses (acceptance criteria of 10CFP50.46) and the Technical Specification .limit for the number (or percentage) of allowable plugged s.earn generator tubes. Res onse to uestion 440.62 The St. Lucie Unit 2 ECCS analysis does not assume any steam generator tubes are plugged. The effect of tube plugging has been treated on an as needed basis for C-E operating. plants and to date tube plugging has been'inimal. In one example, an ECCS analysis w'as performed assuming 500 tubes per SG plugged which, represents approximately 6~~ of the unplugged total. The pre-dicted ECCS performance changed very little and the allowable peak linear heat generation rate remained unchanged from the case with no SG tubes plugged. The method of analysis for the assessment of ECCS performance with a portion of the SG tubes plugged is provided in the Reference. Since the NSSS design utilized in the referenced calculation is similar to the St. Lucie Unit 2 design, a similar conclusion is anticipated for this plant. Presently, St. Lucie Unit 2 has 47 steam generator tubes which have been plugged, which represents approximately 0.6i. of the unplugged total. This is significantly below the 6~ plugged analysis which demonstrated minimal change in ECCS performance and no change in the allowable peak linear heat 'eneration rate. Based on this, C-E feels that the current ECCS perfo~iiance analysis, which does not consider steam generator tube plugging, remains applicable and no new analysis is required unless tube plugging becomes more significant. t FS~g 5'cc-k'~s 4 3 7 Q 3 ~ 0 G 3 3 3 s ha~a. Lco~ ~od AcJ. 0o s@kc M+>> Mre ~$ ~~+ 'fo ~ p(u pp g.Q ~
s SL2"FSAR pump flow is credited. The actual delay time will not exceed 30 sccon<" fol-lowing a SIAS. In the large break analysis, no operator action has been
~
assumed 6.3.3 '.3 Core and System Paramctcrs The significant core and sy. tern parameters used in thc large break calcula-tions are presented in Tablc6.3-7. Thc Peak Linear Heat Generation Rate was assumed to occur in t)ie 'top of the core, the conservative location as identified in Section IV.A.4 of Reference 2. A copservativc beginning-of-life moderator temp.rature coefficient (80.5 x 10 'p/F) was used in all large break cases Q zi~srp7 >1'.
'1r Tl>e initial steady state tool rod conditions sere determined as a function of burnup using 'the FATES computer program. The limiting condition for ECCS performance was determined to occur for a hot rod average burnup of 620 tI'liD/SITU. A parameter study was performed which demonstrates that f/g~'.3.3.2.4.
clad temperature and oxidation were maximized at this exposure. The results of this study are presented on Figure 6.3-13. Containment Parameters Subsection 6.2.1.5 discusses in detail the containment parameters assumed in the ECCS analysis. The values for these parameters'ere chosen to minimize containment pressure such that a conservative determ1nation of the core reflood rate was made. Pressure suppression equipment startup times
~ were selected at their m1nimum values corresponding to offsite power being available 6.3.3.2.5 Break Spectrum In'general, all 'possible break locations are considered in a LOCA analysis.
However, as demonstrated 1n other Appendix K LOCA calculations (References 8 and 9 for example), hot leg .ruptures and cold leg ruptures on the suc-tion side of the reactor coolant pump, yield clad temperatures substantial-ly lower than those observed for cold leg ruptures on the discharge side of the pump. Pump discharge leg ruptures are:limit1ng due to the minimiz-ing of blowdown core flow and reflood rate for the break location. Thus, only these breaks need to be considered in order to identify that rupture which results in the h1ghest clad temperature or largest amount of clad oxidation. Since core flow is a function of break size, calculations have been performed for both guillotine and slot breaks over a range of breal sizes up to twice the flow area of the cold leg. A list of the breaks exam1ned appears in Table 6.3-8 which refer to Figures 6.3-5 through Figures 6.3-11. 6.3.3.2.6 Results and Conclusions The important results of this analysis are summarized in Table 6.3-9 and the transient behav1or of important VASSS parameters is shown in the fig<ircs listed in Tables 6.3-10 and 11 which refer to Figures 6.3-5 through 6.3"11 and Figures 6.3-9 respectively. Peak clad temperature vs. brcak size 1"
~~~~6. 3<<17 Amcndmcnt No. 0, (12/80)~~~~
SL2-FSAR Based on these assuniptions, the following credit is taken for injection flow in the smal1 break analysis. For a discharge leg break: of the flow from one.1IPSI pump IcI 75% 50% 100% of of the the flow flow from from one LPSI pump three safety injection tanks 50% of the ilow from one charging pump and ~ for breaks in other. local.ons: 100% of'the flow from one HPSI,pump 100% of the flow from one LPSI pump 100% of the flow. from four safety injection tanks 100% of the flow from one charging pump'able 6.3-12 presents the high and low pressure safety injection pump flow rates assumed at each of the four injection points as a function of reactor coolant system pressure. 6.3.3.3.3 Core and System Parameters The significant core and system parameters used in the small break calcula-tions are presented in Table 6.3-13. The peak linear heat generation rate of 15.0 kw/ft was assumed to occur 15 percent from the top of the active beginning-of-life moderator temperature coefficient
~~~~~ ~ ~~~~~
~~~~core. A conservative of +0.2xlo ha/oF wss used.~~~~
~
~~~~gtj to4R~~~~
~~~~~met'-T ()A The initial steady stake fuel rod conditions were obtained from the ~ ~ ~~~~~
~~~~FATES ( ) computer program. -The'small break analysis assumed the same~~~~
~~~~~ ~ ~~~~~
~~~~hot rod average burnup as was found limiting in the large break analysis described in Subsection 6.3.3.2. 11owever, since the small. break analysis~~~~
'conservatively used a higher PL11GR than did the large break analysis (15.0 kw/ ft" vs 13.0 kw/ft) .the fuel rod parameter values given in Table 6.3-13 differ from those in Table 6.3-7, ' s 6.3.3.3e4 Containment Parameters The small break analysis does not credit any rise in containment pressure.
~~~~Therefore, other than the initial containment pressure, 'which is assumed~~~~
~~~~.to remain constant, no containment parameters are employed for this analysis. The initial containment pressure was assumed to be 0.0 psig.~~~~
6.3.3.3.5 , Break Spectrum Five breaks were analyzed to charac(erize the small break spectrum. Four breaks, ranging in size from 0.5 ft to 0.015 'ft were 'postulated to occur in the pump discharge leg. The 0.5 ft .freak was also analyzed for the lar'ge.break ~~~ctrum (Subsection 6.3.3.2) and is defined as the transi-tion break size . One. break representing a fINlly open pressurizer relief valve with an equivalent area of 0.008 ft was postulated to occur in the top of the pressurizer. The equivalent area is based on the design. flow. requirements qf the relief valve as utilized in the C-E small 'break evaluation model( ~. Table 6.3-14 lists the various break sizes and locations examined for this analysis.
~~~~.6 '-18a ,Amendo>cnt No. I, (4/81)~~~~
Insert AA The ECCS performance analyses, as pe'rformed, do not account for steam generator tube plugging which may occur during the plant's lifetime.
August 18, 1981 Kbasco Services, Inc. Agents for Florida Power 6 Light, St. Lucie 2 world Trade Center, 83d Floor Hew York, NT 10048 Attention: R. Ragbavan ~
~~~~Subject:~~~~
Florida Power and Light Co St. Lucia Unit No. 2 32" MSXV Ebasco PO 422528 RockMell S.O. No. 36-11000 Xn response to your TVX dated July 31, 1981 Rockwell is providing the following Interim response re1ative to "as do11vered" valve bonnet, and main disk thicknesses for pressure retaining purposes: Sa ed on "on-going" iterative finite element analysis the bonnet thick-ness is at. least 30 percent greater and the main disk thickness is at least 25 percent greater than minimum thickness required for pressure retention purposes. The exact 'percentages and other considerations for thickness .allowances for corrosion and mechanical loads vi11 be identified in our final report scheduled for cosapletion this month. Prospect Bngineering Supervisor Rockwell International cc: N. Hangieri - FL-9 R. D. Hordea
~~~~3. R, Black S. 3, Mum B. B. Hildreth~~~~
~~~~( EBASCO SERYtCES tNCORPORATED BY~CHl~ll >6 oAT(. '8 IZ.S I NElh'YORK 0 ll j~~~~
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~~~~TH PLASTIC METHODS'F STRUCTURAL ANALYSIS.~~~~
~~~~. B. G. Neal hl.A., Ph.l)., M.l.C.l'... hl.l.Mech.E.,~~~~
~~~~A.M.I.Slcuct.E. Pr%seer of Applied Science fmperioi Collrgr of Scirnce~~~~
~~~~~ nd Trchnofo~~~~~
CHAPMAN & HALL LTD and SCIENCE PAPERBACKS
i
\
LOAD DEFLECTION RELATLONS FOR BEAM ESTIXATES OF DXFLECTION8 have indicated that the beam of rectangular cross-section. 'To determine the bending which are discussed in Section 6,2, moment for a given curvature, and thus a given linear'distribu-either a small or zero drop -material of rolled steel joists exhibits tion of strain across the section, it was only ncccssnry to add the feature of the assume of stress at yield. However, the particular is that it includes bending moment due to thc rectangular.web to the bending-stress-strain relation which is of especial interest momcnt due to the flanges, which is equal to thc product of thc strain-hardening range. Thc strain
'hc the yield stress in a flange, its area, and the depth of the beam. At higher ha rd ening co commences is 0 018 whereas the strain cat 0 018 curvatures, a process of step-by-step integration became necessary.
of c, to cs is 0 0011 or 16 4. Four values of the ratio of total flang area to wcb area werc point is 0 0011, so that the ratio considered, namely, 0, 05, 10 nnd 15. The value zero cor-the analysis is that of homo-A further implicit assumption in the fact responds to the case of a beam of rcctangiilar cross-section, and geneity This geneity. is is ra er difficult to justify in view-of rather the other values cover the range of standard I-sections. The results were presented in the form of curves, and for the purpose Initial aihph, r r of accurate calculation were tabulated for the case in which this ratio is 1 0, so that Ar A. Thc bending moment-curvatiire
~~~~~ r .l 40 625 Aipa phr ~. In.~~~~
relation for a beam of I-section of this type, as dcrivcd froin these g30 I I tabulated results, is shown in Fig, 5,8(b). The results are plotted I > 0nhht Pf Ptrain- non-dimensionally, the ordinates being the ratio of the bending o4 I I harChning
~~~~~20 <~~~~
6ippc, JQ000Inps pe'~.nr. I I moment to the yield moment and the abscissae being the ratio of the curvature to the curvature at'yield. It is readily verified I I I I I I Q 10 I w I I I that for such a cross-section the shape factor ct is 1 125, so that vl lg I MP = 1125 Mr. It will be seen from the figure that strain-0 40 PPh P.P2 6train 0 PJ OPS JP 20 JP hardening commences when ~164, this being t)ie ratio of Ks ta> fb) Sh tO Cs ~ (o) Strees-strain relation. (b) Bending moment~rvaturs Further results were also tabulated which enable the load-relation! AI ~ Ah deflection curves of statically determinate beams and simple Fig. 5.8. Bcadingnunncnt~in'c rchtion for I.section tsith statically indeterminate beams and frames to be derived. Some strain-hardening (after Hrennikoff). applications of this work will be given in Sections 5.8 and 5.4, tensile specimens For a more general treatment of the problem of determining that the stress. strain relations obtained hem bending moment-curvature relations from any assumed form of rolled steel joists are known to cut from various positions in stress-strain relation, the work of Kadai r may bc consulted. 6.2. vary widely, as pointed out in Section that the thickness of Several comparisons with experimental results for light alloy To simp lify th e analysis it was assumed with thc depth o e t beams have been made, for instance by Rappleyca and Eastman the flanges vias negligible in comparison by Dwight, s and the case of a light alloy beam of rectangular
'nd be regarded as concentrated beam, so that each flange area could section which is subjected to hendi>>g moments about axes other axis. With this assump-at a constant distance from the neutral relation depends than the principal axes has been discussed by tion thc form of the bending moment-curvature Barrett.'.3 ratio of the total flange only on a single parameter, namely, thc Load-deflection relations for simply supported beams curvature at which strain-area Ar to the web area A. Up to the For a beam resting on two simple supports thc bending moment thc analysis was a hardening develops in the outermost fibres, distribution in the beam for a given loading is known from has just been given for a simple extension of the theory which 178 172
r.sTIM~TEs or DErLEcTIoNs LOhD-DEFLECTION RELATION8 roR REawa considerations of statics alone. Once the bending moment-curva- reaches the value Mo The corresponding ture relation is specified the curvature at any section is known, and value of the load, JJ>> is therefore given by the equation the deAected form of the beam can then be found by integration. In the first place beams of rectangular cross-section will be con- If the M~)JVl... load is increased to a value W greater
. 5.9 sidered, with the bending moment-curvature relation of equation than JJ', the yield moment Mwillbe attained at some distance 5.6; subsequently, some of thc results'btained by Hrennikolf s a from the sup-ports, as shown in the figure. In thc central for joists with the bending moment-curvature relation of Fig. portion of the beam where the bending moment exceeds M>> yield occurs, 5.8(b) will be given. ity spreads inwards towards the neutral axis. The and plastic-general form of thc plastic zones thus crcateil is shown in the Beam of rectangular cross-section Ioith central conccntratcd load figui.c; a deriva-tion of the shape of the elastic-plastic bounilary will bc given Consider a uniform beam of rectangular cross-section, breadth b later. Eventually, collapse occurs wlicn thc central bcniling and depth h, which is simply supported over a span J, as shown in moment reaches the value M~, so that plasticity has spread right down to the neutral axis at thc centre of the beam. The cor-responding collapse load JV, is given by the. equation 81~ ~ )JYP Using equation 5,9 it follows that JV, N JVN since the shape factor for a rectangular beam has the value From statical considerations it follows that 1 5.
~~~~1lfv = JJJa Combining this equation with equation 5.9, it is found that~~~~
~
JYg. 6,4 Simply capportcd rectangular scctims beam arsA ~ Since the slope ccatrnt conccntratcd toad. at the centre of the beam is zero the central deflection 8 is seen to be given by symm t by the integral. Fig. 5.4. It will be assumed that the relation between bending moment and curvature in the yielded regions is the relation given by equation 5.6, and for simplicity it will be assumed in the first where a; is measured from the left-hand 0 I a'e dz .. ~ . 5.11 pla'ce that f~ ~ f~, so that support. For 0 < e < a, the beam is elastic, so that the curvature ic is equal to -~,.tc~ ~ For or M-~<- 8 " o o o 5.8 a < ai < ,, the beam has partly yielded, so that the relation This relation corresponds to the assumption of the ideal-plastic between bending moment and curvature is given by equation 5,8. stress-strain relation of Fig. 1.4(b). Solving this equation for e, it is found that The bending moment diagram for the beam is as shown in Fig. 5.4, the central bending moment being HAJJ'l. Yield first .,fora (a < , occurs at the centre of thc beam when this bending moment 1V4
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