ML20138P331

Forwards Response to 851101 Request for Addl Info Re App R Safe Shutdown Model.Steps Included in Fire Protection Shutdown Procedure for Operators to Properly Position motor- Operated Valves

ML20138P331
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/20/1985
From:	Walker R PUBLIC SERVICE CO. OF COLORADO
To:	Berkow H Office of Nuclear Reactor Regulation
References
P-85488, TAC-54373, NUDOCS 8512260055
Download: ML20138P331 (38)

Text

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s PUBLIC SERVICE COMPANY OF COLORADO P.

O.

BOX 840 DENVER.

COLORADO 80201 R. F. WALKER

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December 20, 1985 Fort St. Vrain Unit No. 1 P-85488 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Attn: Mr. H.N. Berkow, Project Director Standardization and Special Projects Directorate Docket No. 50-267

SUBJECT:

Response for Additional Information on the Appendix R Safe Shutdown Model

REFERENCE:

NRC Letter Butcher to Lee Dated November 1,1985 (G-85459)

Dear Mr. Berkow,

Attached is Public Service Company's responses to the

-16 questions submitted per the above referenced letter.

If, you~ have any further questions regarding the matter, please contact Mr. M. H. Holmes at (303) 480-6960.

Very truly yours,

'flj J0-1%~4~7 AO' R. F. Walker, Chairman, President & CEO RFW/FT:pa 00b Attachment i

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ANSWERS TO NRC QUESTIONS FOR FORT ST. VRAIN NRC OUESTION 1.

There are situations in which a single failure can defeat a function common to both trains proposed for forced circulation safe shutdown.

a.

The service water strainer (F-4201) and the associated flow valves (HV-4257, HV-4225, HV-4221 - 1, and HV-4221-3) are required to be operated in both proposed trains. What provisions have been made to provide for continued system function if a fire disables or precludes positioning of these components?

PSC RESPONSE l.a Valves HV-4257, HV-4225, HV-4221-1, and HV-4221-3 are motor operated valves with manual operators.

These valves are located in the Turbine Building; the strainer is located in the Service Water Pump Building. The service water strainer (F-4201) and these valves are water-filled mechanical components and, as such, are not considered damaged by a fire, per specific criterion Ill.l. of the PSC letter dated August 17, 1984 (Ref. P-84281). Also, per that some criterion, mechanical components with manual operators are considered to be manually operable within one hour offer the start of a fire.

Steps have been included in the fire protection shutdown procedure, " Safe Shutdown and Cooling with Highly De-graded Plant Conditions" (SSCHDPC), for operators to properly position these volves.

Criterion 111.1 is consistent with the approach utilized at other plants and accepted by the NRC.

DC-85-l84 1

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NRC QUESTION b.

Have all other situations where single failures could disable the function been evoluoted?

PSC RESPONSE l.b. The fire protection shutdown trains were defined so that a fire induced failure of a component would not cause loss of function of both shutdown trains.

In addition to the service water strainer and valves discussed above, other situations where shutdown equipment is common to both trains have been evaluated as follows:

1)

Control Rods - The analysis described in Report No.1, Section 2.4 demonstrates that fire damage cannot prevent rod insertion and reactor shutdown.

Note: The Reserve Shutdown System is also avail-able, and can be manually operated.

2)

Any common piping or manually operable valves -

Per specific criterion Ill.I, there is no fire damage and valves are operable after one hour.

3)

Condensate Storage Tank - Per specific criterion 111.1, there is no fire damage.

4)

Decay Heat Removal Exchanger - Per specific cri-terion 111.1, there is no fire damage.

Thus, components common to both shutdown trains are not subject to fire damage.

NRC OUESTION 2.

Review of the tables contained in Section 2.1 of Report I (as revised) shows that not all of the required volves are listed.

DC-85-184 2

a.

What criteria was used to determine if a component should be included in the table (e.g., Table 2.1-3, sheet 3 of 5, lists V75595 to be closed, but does not indicate that parallel volves V75600 and V75605 would need to also be closed; also, other valves off the same line would need to be properly positioned, but are not listed).

PSC RESPONSE 2a. The following summarize the criteria for inclusion of components in Table 2.l.

1)

Manual valves in the flow path or needed for isolo-tion have been tabulated in Section 2.1 of Report I if the " normal position" is not the desired position for the shutdown model.

2)

Powered valves in the flow path or needed for isolation have been tabulated because of the need to operate them and/or because of potential " spurious operation." Powered valves that must function (i.e.,

powered volves normally opened whose desired posi-tion is closed, valves normally closed whose desired position is opened, and flow control valves in the flow path whose desired position is opened) have been tabulated in Section 2.1.

Valves that could spuriously change position and jeopardize shutdown functions have been tabulated separately in other sections. Note: Spurious valves are addressed in Report No. I, Section 2.2 (potential spurious valves),

Report No. 2, Section 2.4 (spurious valve power sup-plies), and Report No. 3, Section 6.3 (spurious valve corrective actions).

3)

Manual valves which can prevent the 'need for

" spurious" considerations or multiple valve opera-tions have been tabulated in Section 2.1 of Report 1.

DC-85-184 3

4)

Check volves used for isolation have not been tabu-lated in Section 2.1 of Report I unless they are "Stop Check Valves."

5)

Other components have been tabulated in Section 2.1 of Report I where functioning is required for operation of the system (e.g., pumps, circulators, etc.).

6)

Large possive components (tanks, heat exchangers, sumps, steam generators, etc.) were included to better reflect the flow path being used.

In developing the shutdown procedures for the fire protec-tion shutdown models, a verification was performed of the flow path to confirm incorporation of required compon-ents. This has resulted in the identification of certain additional volves, as reflected in further page changes that will be provided in a future revision to the Appendix R Evoluotion Reports. This includes the additional volves referenced in the above question.

NRC QUESTION b.

Do the implementing procedures require checking the position of oli volves in the flow path? (e.g., A volve in the flow path that is designed to fail close and is required to be close should be verified to be closed; HV-2223 should be verified closed.)

PSC RESPONSE 2b. Shutdown procedures "Sofe Shutdown and Cooling with Highly Degraded Plant Conditions" (SSCHDPC) and "Ser-vice Water and Domestic ' Noter Systems" (SOP-42) have been revised and issued in October,1985 to reflect the required operator actions for the fire protection shutdown DC-8S-184 4

models. These procedures require checking of all powered volves, whether required to change position or whether changing a position is not desired but could occur due to spurious operation.

This. is done irrespective of the volve's failure position since a spurious signal could cause a valve to go to the nonfail position. which may not be the desired position for the method of operation in the fire protection shutdown model. These shutdown procedures also identify manual valves in the flow path that must be changed from their normal position in order to assure desired flows for the fire protection shutdown model, and to isolate divergent flows that could impede proper accomplishment of shutdown functions.

Normal align-ments of manual valves are controlled through System Operating Procedures (SOPS). These are implemented and volve alignments confirmed prior to placing a system back into operation. Technical Specifications establish limiting conditions for operation and action statements that in-clude limitations on system outage time.

The above approach for the Appendix R shutdown procedures is consistent with that applied to other Ft. St. Vrain emer-gency procedures, and is also consistent with the approach taken to emerger.cy operating procedures throughout the industry. It should also be noted that the fire protection shutdown model incorporates various instrumentation that serves as a further check to confirm proper system operation.

t With respect to valve HV-2223, this is a stop check valve in the main steam system whose closure would be desired, but not necessary, in flooding the EES section and divert-ing flow to the decay heat removal exchanger (E-4202).

An alternative to closing volve HV-2223 is to trip the EHC pumps to close turbine stop valves SV-1 and SV-2.

DC-85-134 5

This alternative could result in filling of the steam lines with water; however, this condition has been analyzed for the steam lines and steam supports. Accordingly, shut-down procedure SSCHDPC does incorporate steps for operators to confirm that volve HV-2223 is closed, and to trip the EHC pumps if HV-2223 is not. closed, for fire protection shutdown Train A.

NRC QUESTION 3.

The condensate storage tank is common to both safe shutdown trains.

a.

What is the minimum storage capacity necessary to complete the assumed function (s).

PSC RESPONSE 3a. Condensate Storage Tanks - Minimum Capacity Although the Condensate Storage Tanks are shown com-mon in both Trains A and B, the CST's in Train "B" are used only as a " piping system".

That is, the tank overflows convey the return water to the service water return sump so that the cooling water will be returned to the Main Cooling Tower Basin. No inventory in the CST's is required.

In using a condensate pump for driving a circulator in Train "A", the condensate storage tanks " float" on the system. Level will rise and fall based on cycling of a turbine water rernoval pump.

Since leakage from the system is minimal, the maximum inventory required would be less than 5,000 gallons based on the volume of the turbine water drain tank and assuring the tank is near the low level seyting at the time of the fire. The CST's are 100,000 gallon tanks that are operationally controlled at DC-85-184 6

40,000 gallons each. It is not credible that Ft. St. Vrain would be operating with a combined CST inventory of only 5,000 gallons. Accordingly, Technical Specification limits are not believed to be required.

b NRC QUESTION b.

This minimum copocity should be incorporated as a Technical Specification limit, as should all equipment for which credi is taken in the shutdown models.

PSC RESPONSE 3b. PSC is currently reviewing the extent to which operability requirements for equipment for which credit is taken in the fire protection shutdown models, and for use in fire protection, will be covered in the Technical Specifica-tions, in conjunction with industry wide efforts in this area. PSC is presently considering the development of a separate fire protection program, which has been discus-sed with the NRC Ft. St. Vrain Project Manager. This is considered to be an ongoing licensing issue, which will be the subject of future separate correspondence.

NRC QUESTION 4.

The shutdown models for forced circulation cooldown, for a fire in a noncongested cable area, do not contain provisions for maintaining PCRV liner cooling. In light of the acceptance criterio B.2.b contained in the PSC letter dated August 17, 1984, provide an evaluation of the necessity of maintaining liner cooling.

PSC RESPONSE 4.

The acceptance criterion B.2.b., contained in PSC letter No. P-84281 of August 17, 1984, sets a performance goal for safe reactor shutdown /cooldown for a fire in non-congested cable areas of " maintaining the PCRV liner integrity and PCRV structural and pressure containment integrity."

DC-8S-184 7

The loss ' of PCRV liner cooling is addressed in FSAR Sections:

5.9.2.5, 9.7.5, 14.4.2.2, D.2.2, and 1.2.1.2.2, under various conditions of primary coolant flow.

In particular regard to this question, FSAR Section 1.2.1.2.2, f

states "... If any of the forced circulation cooling modes are available for core cooling after shutdown, the PCRV con then withstand a complete and permanent loss of liner cooling without producing a hazard to the public."

A recent study by. GA Technologies in April,1985 (GA letter GP-2504, Attachment GA No. 907935), focused on the we> st case of an interruption of forced circulation coolinc, for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, followed by the use of the boosted fire water to operate one water-turbine driven circulator, l

and fire water to provide cooling water flow to the steam generator tubes.- Liner cooling was assumed to be lost r

throughout this event. It was found that the absence of liner cooling had no significant effect on maximum fuel or orifice volve temperatures. The results are summarizsd below and shown on the attached Figures 3,4,5, and 7.

if Calculated Limit c

Parameter Max. Fuel Temperature 2775 F 2900 F Max. Average Core Temperature 1422 F (Not Specified)

Max. Upper Plenum Temperature 1214 F 1500 F-Max. Cover Plate Temperature 1047 F 1500 F Max. Liner Temperature 239 F 1500 F Max. Orifice Valve Temperature 1647 F 2000 F lt can be seen that' the results of this study meet the performance goals of B.2.b.

Thus, liner cooling is not necessary to meet this criterion, and was not selected as part of the shutdown model for fires in non-congested cable areas.

V

' DC-85-184 8

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NRC QUESTION S.

Provide un evaluation of the need of a main cooling tower fan in safe shutdown model Train B.

PSC RESPONSE 5.

On the basis of conservative calculations, it was conclud-ed that the Train B heat generated could be removed by using all ten tower sections and natural draf t generated by the bouyant effect of the heated, wet air column in the cooling tower relative to the cooler air outside the tower.

A conservative approximation of the natural draft, air flow was made using the square root of the estimated draw divided by the rated fan draw. In the absence of specific tower fill design data (since specific fill design data is proprietary to Morley Corp.), a very large air / water enthalphy difference provides the confidence that this approach is reasonable.

To assure maximum tower draf t under most meterological conditions, the desired method of tower operation is to establish the highest reoscnable tower discharge tempera-tures.

Maintaining the discharge temperature as near constant as possible by observation of tower return water temperature and reasonable flow control based on the temperature provides means for this type of control.

The most extreme meterological conditions for gravity generation of air flow are those which result in extreme reductions in the weight of the ambient atmosphere. High operating temperatures reduce the effect of such events.

As is customary in all adiabatic saturation operations using wet bulb cooling, air flow rate is critical and some cooling assistance may at times be desired. At such times there are several options:

DC-85-184 9

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Make-up water may be added to over flow the tower basin. The source can be domestic supply, wells, or storage ponds at the Ft. St. Vrain facility.

2)

The Fan Power Sources for two (2) tower cells are diversely fed via on underground duct system used by the B train cables.

These cooling tower fan cables are not routed through fire areas containing Train A cables. Thus, although cooling tower fans are not projected to be used in this shutdown model, they are available should operators choose to use them.

Based on the above, odequate heat removal con be provid-ed by the cooling towers without fans; but fans are available for Train B as an enhancement should these be desired.

NRC OUESTION 6.

A list of required ACM fire protection shutdown compon-ents is provided in Table 3.2 of Report 1.

a.

Provide o listing of the equipment identification numbers and the normal power supplies for these components.

PSC RESPONSE 6a. Table 6A, attached to this letter, provides a list of required ACM fire protection shutdown components, their equipment identification numbers, and their normal power supplies.

A future revision to Appendix R Evaluation Report No. I will include a modified Table 3.2 showing Equipment Identification Numbers corresponding to the components listed. This is consistent with the information previously provided in Table 3.2 of Report No. I and Table ! of SOP 48-01, included as Appendix B to Report No. I.

DC-8S-184 10

f NRC QUESTION b.

Provide a description of all changes made to the "ACM shutdown system" since its use was opproved in License Amendment Nos.14,18, and 21.

PSC RESPONSE 6b. The ACM shutdown system has had only a few changes to it since its use was approved in License Amendments Nos.

14,18, and 21.

Amendment No. 22 contained a change to Tech. Spec.

4.2.17 concerning the ACM operational fuel supply, which was approved by the NRC on poqe 5 of the NRC Safety Evaluation Report on that Amendment.

This change modified the basis to read, "The 10,000 gallons of fuel provides for 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> operations (instead of "one week of operation") of the generator with full ACM load, which is adequate time for obtaining additional fuel from offsite sources."

The following change notices have been completed or have work in progress after June 6,1979:

CN DESCR!PTION CN 0904 Incorporates system 48 description and design criteria into the documentation system.

CN 0912, A Replaced V-48104 with a h" gate volve to prevent fuel oil leakage. Reissue otso replaces V-48204.

CN 1072 The vendor, Woodward Governor, rebuilt the ACM diesel generator to the current standards to increase equipment reliability.

CN ll69 Relocated the ACM batteries from the diesel bottery compartment to the turbine plant HVAC building for o more suitable environment.

- DC-85-l84 II

k CN DESCRIPTION CN I169A -

Perform on evaluation of hydrogen out-gosing of the batteries during charging.

CN 1751 Upgrades the cable from 480 VAC tur-bine plant HVAC bus to 480 VAC load center switchgear ACM.

CN 181l Replacement of the ACM diesel batteries because of a battery failure.

CN 18ll A installs spacers to the battery rock to improve containment.

CN 1906, A Replacement of a bearing support kit in the blower assembly due to a

manufacture's maintenance information notice. The original CN was reissued as a safety related CN.

CN 1962 Document update of the data bases to include the indentifier that the ACM equipment items be designed as safety-related.

The following change notice is in the design and production phase:

CN 2004 Per Appendix R Evaluation, the circuit breaker interlocks will be removed to enable back-feeding ACM diesel genera-tor power to the 480 VAC essential buses.

- From this, it con be seen that the design, loads and intent of the ACM System has not been modified significantly since its use was approved in License Amendments 14,18, 21, and 22. The License Amendment Safety Evaluation Reports on the ACM are still valid.

Additionally it should be noted that the ACM fire protec-tion shutdown components will allow performance of a liner cooldown using the some ACM shutdown system approved in License Amendment Numbers 14,18 and 21.

DC-85-184 12

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1 NRC QUESTION 7.

Drawing PI 31-1 shows a removable spoolpiece between the fire water system and the line from the main feed-water pumps.

a.

Is th' spoolpiece normally installed? If not, describe Imw its storage and installation are controlled.

(In similar applications at other facilities, the spoolpiece is a Techni-col Specification controlled item.)

PSC RESPONSE 7a. The spoolpiece between the main feedwater pumps and the fire water system shown on Pl 31-1 is not normally installed; however, as noted below, this spoolpiece is not required for the fire protection shutdown models.

NRC OUESTION b.

Describe all other spoolpieces which are being used in the shutdown models and the controls used.

PSC RESPONSE 7b. Certain spoolpieces are shown on P&l drawings for systems that support the fire protection shutdown model; however, the fire protection shutdown flowpaths were selected so that use of these spoolpieces are not required.

The following summarize the spoolpieces shown on such P&l's, and provide the basis for not requiring their use in the fire protection shutdown model.

Spoolpiece Dwg.

Function I) M-3129 31-1 Fire Water Back-up to Emer-gency Feedwater

2) M-21834 22-1 Emergency Water Booster Pump to Circulator Water Drives

3) HTFA's 23-1 Reactor Plant Cooling Water (Sys. 46) to A-2301 and A-2302 DC-85-184 13

Spoolpiece M-3129 - Not used in fire protection shutdown model. Alternate flow used is through the Emergency Condensate supply via valve V-4525.

Spoolpiece M-21834 - Not used in fire protection shut-down model. Alternate flow used is through the emer-gency water booster pump cross-tie via valve V-211567.

Spoolpieces for HTFA's - Not used in fire protection shutdown model for forced cooling shutdown.

ACM shutdown does use them (reference SOP 48-01, paragraph 6.0). CN-2142 has replaced these spoolpieces with "hard pipe," and isolation volves (manual).

Accordingly no special controls of spoolpieces are required.

NRC QUESTION 8.

Since provisions are not included in the shutdown models for its operation, previde on evoluotion of the necessity of the Buffer Helium Sy.. err. for circulator operation.

PSC RESPONSE 8.

The fire protection shutdown model for a postulated fire in a non-congested cable area takes no credit for the buffer helium system for operation of a helium circulator.

The fire protection shutdown model does rely on availa-bility of the bearing water system to cool and lubricate the bearings of the operating helium circulator. This is consistent with the licensing basis of FSV whereby the buffer helium system is not required to be functional for helium circulator operability (Technical Specification LCO 4.2.2).

FSAR Section 4.2.2.3.7 states "Although buffer helium is not required insofar as containued opera-tion of the circulators is concerned, the pressure boundary DC-85-184 14

h integrity of the buffer helium system is required to maintain primary coolant pressure in the bearing water surge tanks, insuring proper operation of the bearing water surge tanks, insuring proper operation of the beor-ing water system". FSAR Section 4.2.2.3.2 states that

" loss of the normal make-ep supply of helium from the helium purification system will not impair the operability of the helium circulators, but may allow trace amounts of p

water vapor to diffuse through the helium buffer labyrinth L

j seals into the primary coolont system".

k To address concerns-associated with the potiential for leakage of bearing water into the PCRV when a helium l

circulator is operated with no buffer helium flow, tests were performed in 1975, to assess the acceptability of this mode of circulator operation. GA conducted tests on helium circulator operation with no buffer helium flow at their Son Diego test facility. The results of these tests were published in GA Report No.1975-3, GA-Al3393, Rev. I and Addendum I to this report, dated 12/10/75.

This report was prepared under US ERDA Contract No.

E(04-3)633. Addendum I to this report indicates that in each case tested, where no buffer gas flow was supplied, inleakoge of bearing water stopped after the helium circulator shaf t speed reached or exceeded 280 rpm.

Bearing water rotates the helium circulator at about 400 rpm by itself, with no steam or water supplied to the turbine drives ("self turbining").

Special Tests T-23 and T-31 were conducted at FSV in 1975 to study in-plant operation cf a helium circulator with no buffer helium supply. The tests determined that lack of buffer helium flow to the operating helium circu-lotor did not result in ingress of water to the primary coolant system.

DC-85-184 15 0

s The fire protection shutdown model for a postulated fire in a non-congested cable area provides for a closed system. Buffer helium will equalize in pressure with the primary coolant system, the high pressure separator and the ollage in the Bearing Water Surge Tanks. Since some primary coolant will dissolve in the bearing water supplied to the operating helium circulator, the bearing water system will become contaminated with radioactivity.

Therefore, operation of a helium circulator with no buffer helium flow is an acceptable mode of operation following a fire in a non-congested cable area.

NRC OUESTION 9.

The shutdown models for forced circulation cooldown consist of providing cooling water to the economizer-evaporator-superheater section of one of the steam gen-erators via a low-pressure pump (condensate or fire water). Since the steam generators will be at relatively higher temperatures (greater than 10000F at shutdown) and are helical wound tubes with no storage capacity, provide an evaluation of the effectiveness of this mode of cooldown. This evaluation should include a study of the possibility of damage, caused by occurrences such as water hammer or over pressurization, during such a cooldown.

PSC RESPONSE 9.

The shutdown model for a postulated fire in a non-congested cable area is similar to Safe Shutdown Cooling, described in FSAR Section 10.3.9 and 14.4.2.2, in regards to the transient experienced by the steam generators.

The fire protection shutdown model selects a 90 minute interruption of primary and secondary coolant flow to be consistent with Safe Shutdown Cooling. For Fire Protec-tion Shutdown Train A, primary coolant flow is resumed DC-85-184 16

4 by supplying condensate from small condensate pump P-3106 to the water turbine drive of helium circulator C-2l01 and secondary coolant flow is provided to the economizer - evaporator - superheater section of the loop i steam generator, again from P-3106. For Fire Protec-tion Shutdown Train B, primary coolant flow is resumed by supplying boosted fire water (P-450lS in series with P-2110) to the water turbine drive of helium circulator C-2l03 and secondary coolant flow is provided to the economizer - evaporator - superheater section of the loop 2 steam generator using fire water.

In Safe Shutdown Cooling, primary coolant flow is resumed at 90 minutes by supplying boosted fire water to any one of the four helium circulator water turbine drives and secondary coolant flow is provided at 90 minutes to either the economizer -evaporotor - superheater section or the reheater section of the steam generator in the associated loop using fire water. Safe Shutdown Cooling, utilizing fire water for helium circulator motive power and secondary coolant flow, is one of the FSV licensing bases for cooldown following a Design Basis Earthquake, a Maximum Tornado and a worst case steam line rupture accident. The conclusion of these analyses (FSAR Section D.2.5.3)is that sufficient cooling would be provided to the steam generator tubes and other pressure parts of the steam generators so that they would not be the part most susceptible to damage if the circulators are restarted after several hours.

PSC recently contacted the individual who was the design section leader of GA's steam generator branch at the time the FSV steam generators were designed. He recalled that a task was performed which analyzed the structural DC-85-184 17

effects of introducing relatively cold fire water to the

.c economizer - evaporator - superheater section of a hot steam generator. The analysis concluded that the steam generator _ would retain - its structural integrity and perform its safety function throughout at-least one of these emergency transients. This analysis was the basis for reliance on the steam generator's capability to perform their safety function in the Safe Shutdown Cooling scenario, which is part of the licensing basis for FSV.

To date, PSC has been unable to recover documentation of this particular analysis. In light of this documentation problem, PSC has initiated a task that is

-Intended to reconfirm the results of the original analysis.

The task will analyze the fire water cooldown method (which is more conservative than condensate) to determine the impact of introducing cold fire water to a steam generator following a 90 minute interruption of forced circulation cooling from 100 percent power. ~The following concerns will be addressed in the analysis:

1)

Will fire water flashing to steam in the steam generator's tubes result in a vapor lock condition that could prevent establishing the required fire water flow through the steam generators?

2)

Will water hammer caused by collapsing steam bub-bles adversely affect the integrity of the steam generator tubes or their heat removal capacity?

3)

Could on overpressure condition exist which would threaten the integrity of steam generator tubes?

This task will demonstrate the effectiveness and integrity of the FSV steam generator modules following i

DC-85-184 18

introduction of cold water from the plant's fire protection system.

PSC does not expect undue problems with NRC's concerns because enough discharge flow paths can be established to prevent buildup of steam pressure which would inhibit cooling water flow, existing relief valves would protect the steam generator tubes from overpressure and water hammer effects can be minimized by controlling the fire water flow rates into the hot steam generators.

The results of this task will be iransmitted to the NRC upon completion of the analysis.

NRC QUESTION 10. The capacity of the diesel fire water pump is indicated on drawing PI-45 to be 1500 gpm; Figure 2.1-9 of Report I indicates that this pump will provide 155 gpm for a Helium Circulator and 1050 gpm for a steam generator.

Provide an evaluation which verifies that adequate capa-city is available to perform the safe shutdown functions in addition to providing sufficient fire suppression water flow.

PSC RESPONSE

10. The FSV fire pump / water supply system consists of two 100% capacity fire pumps, one electric motor driven and one diesel engine driven. Each pump is rated at 1500 gpm at 125 psig. The pumps take suction from independent sumps which are connected to the main cooling tower basin. Water for the fire protection systems is provided from two-13 million gallon capacity ponds via a circulat-ing water make-up pump. For a postulated fire event in a safety-related area, creas protected by automatic sprinkler systems will require starting of one of the fire DC-85-184 19 I

dis' overy.

pumps within several minutes after fire c

Manual hose streams will be used as a backup to the

" automatic sprinkler systems or as a primary means of protection in non-sprinkler creas.

The heaviest water flow demand is expected to occur in this period af ter'the

. fire brigade begins its initial extinguishment activities. In general fire water demand will be greatest in the time period following reactor trip and prior to starting forced circulation functions.

Consistent with NRC policy, PSC believes that any fire event will be controlled in a reasonable time period, thereby allowing re-entry to the fire area within one-hour of fire discovery. This was also noted as part of the evaluation criteria in the August 17, 1984 PSC letter (P-842812). After this time period, it is anticipated that the automatic sprinkler system would be stopped and that one or two lYz" hose lines could still be needed for final extinguishment and cleaning up of any remaining hot spots. The flow requirement of 200 gpm for a Helium circulator and -1050 gpm for a steam generator would not occur until approximately one and one-half hours after initiation of the fire / shutdown event. This would allow up to 250 gpm for residual manual fire fighting if only one fire pump is started.

In addition, the electric motor driven fire pump can be manually loaded onto the ACM diesel bus or on emergency diesel generator providing an additional 1500 gpm for fire fighting activities.

(Note that the helium circulator demand of 155 gpm shown on Figure 2.1-9 of Report No. I is no longer correct in that the Emergency Water Booster Pump capacity is now 200 gpm. This will be revised in a future revision to Appendix R Evaluation Report No.1).

DC-85-184 20

NRC QUESTION I1. Provide a description of the testing program which will be implemented to verify the operability of the proposed safe shutdown models.

PSC RESPONSE II. Component functional tests will be performed as part of plant modifications being installed for Appendix R. Addi-tionally, the resolution described in the response to item 3.b above with respect to Technical Specifications will address preparation of appropriate surveillance require-ments to confirm operability of required components from the shutdown models.

NRC QUESTION 12a.There are numerous operations contained in the Tables of Section 2.1 of Report I which require the operator to "De-energize and open or close" manually (HV-3133-1 and HV-3133-2 in Table 2.1-3) or " Remove power and open or close" locally (HV-4221-2 of Table 2.1-6).

Operations such as removing fuses and, in most cases, opening power supply circuit breakers are considered repair operations and are not allowable for the train required for hot shutdown (lli.G.1).

PSC RESPONSE 12a. Attachment C, "NRC Staff Positions on Post Fire Shut-down Capability", to SECY 83-269 states that:

"Section Ill.G of Appendix R states that repairs are permitted to provide the cold shutdown capability.

Additionally, Section Ill.L indicates that procedures for these repairs must be developed and materials needed for the repairs stored on site. To establish consistency in the plant designs, the staff issued the following guidelines concerning repairs. (memoron-dum R. Mattson to R. Vollmer, dated July 2,1982)

DC-85-184 21

"Section Ill.G.I of Appendix R states that one train of systems needed for hot shutdown must be free of fire damage. Thus, one train of systems needed for safe shutdown has to be operable during and follow-ing the fire.

Operability of the hot shutdown systems, including the ability to overcome a fire or fire suppressant induced maloperation of hot shut-down equipment and the plant's power distribution system, must exist without repairs. Manual opera-tion of valves, switches and circuit breakers is allowed to operate equipment and isolate systems and is not considered a repair.

However, the removal of fuses for isolation is not permitted. All manual operations must be achievable prior to the fire or fire suppressant induced ma! operations reaching an unrecoverable plant condition." (from page C-2 of SECY-83-269, July 5,1983)

For the Ft. St. Vrain fire protection shutdown models, removing of fuses is not relied on.

Opening of power supply breakers is relied on to remove potential faults, or effects of spurious signals; such opening of breakers is allowed per the above referenced guidance.

NRC QUESTION 12b. Provide a description of what actions are necessary to accomplish the operations indicated in the Tables of Section 2.1 along with an evoluotion of the time required to perform these actions.

i PSC RESPONSE 12b.

Actions for Post-Fire Forced Circulation Cooling The general method used in the fire protection shutdown model is water drive of a Helium circulator using either DC-85-184 22

i, j'

boosted fire water or o condensate pump, bearing water

[:

. for circulator lubrication and cooling, and steam genera-for cooling from o condensate pump or fire water. An extensive analysis has been conducted to ensure the i.

- avallobility of one train of equipment to perform these functions including supporting emergency power, service water, ventilation, and instrumentation. The opproach in the development of procedures to utilize the fire protec-tion shutdown.model was first to have the procedure capable of oddressing conditions from very little fire damage' to conditions of extensive fire domoge. With little fire damage, the operator would continue power operation or shutdown as equipment and power (electrical) conditions permitted. With greater levels of fire damage, shutdown would be initiated and proceed with ovailable equipment. With extensive fire domoge, it is possible that only the equipment described in the fire protection shut-

- down model will be available for shutdown.

Post-fire conditions of little to moderate domoge are addressed by the system operating procedures (SOP's).

With greater levels of domoge, the operators could utilize the procedure titled, "Sofe Shutdown and Cooling with Highly Degraded Plant Conditions (SSCHDPC)."

l Post-Fire Shutdown Sequence "A" Approximate Procedure Time Required i~

Action Required Reference (Minutes) 1.

Reactor scram l

c.

Scram from control room EP B-l 0

DC-85-184 23 i

2.

Provide emergency power a.

Attempt to start emergency EP F-4 0

diesel generators (assumed unavailable) b.

Manual start of ACM diesel APR-B 20 generator aligned to back-feed 480V Bus I and 2 through the 4160 switchgear c.

Verify power to 480V bus, SSCHDPC 0-10 re-energize buses per Table la of procedure 3.

Verify service water flow

SSCHDPC, 5-30 (for major SOP 42, valves; equip.

SOP 48-01 isolation as conditions warrant per procedure) a.

Alion valves b.

Start service water pump (P-4201) c.

Start cire, water make-up pump (P-4 l 18-S) d.

Start service water cooling tower fan (C-4201 X) 4.

Cine up bearing water supply

SSCHDPC, 5

Option E (normal lineup) 5.

Line up bearing water make-up

SSCHDPC, 15 supply Option E 6.

Line up the turbine water

SSCHDPC, 15 drain tank to the condensate Option E pump suction 7.

Line up the condensate pump

SSCHDPC, 75 to the emergency condensate Option E header

'c 3

DC-85-184 24

8.

Line up the emergency condensate

SSCHDPC, 85 header to the circulator PW Option E and the EES of the steam generator 9.

Line up the steam generator out-

SSCHDPC, 85 let to the decay heat removal Option E exchanger 10.

Verify necessary ventilation to EP APP 0 (not required operating equipment in the IY2 hr.

time interval)

Total 340 Post-Fire Shutdown Sequence "B" Procedure Action Required Reference Time 1.

Reactor scram a.

Scram from control room EPB-l 0

2.

Provide emergency power a.

Attempt to start emergency EP F-4 0 (from diesel generators control room) b.

Verify power to 480V bus, SSCHDPC 10 re-energize buses per Table Ib 3.

Verify service water

SSCHDPC, 5-30 (for major SOP 41, valves; equip.

SOP 42 isolations as conditions warrant per procedure) a.

Align valves b.

Start service water return pump (P-4204S) c.

Start cire, water pump (P-4104 DC-85-184 25

m 4.

Line up bearing water supply

SSCHDPC, 5

- Option I (normal lineup) 5.

Line up bearing water make-up

SSCHDPC, 15 supply Option 1 6.

Line up the turbine water drain

SSCHDPC, 5

tank to the Reactor Building Option I sump 7.

- Line up fire water to the emer-

SSCHDPC, 75 gency condensate header Option 1 8.'

Line up the emergency condensate

SSCHDPC, 85 header (fire water) to the Option I circulator PW and the ESS of the steam generator 9.

Line up the steam generator out-

SSCHDPC, 85 let to the decay heat removal Option I exchanger 10.

Verify necessary ventilation to EP APP S 0 (not required operating equipment in the IY2 hr.

Iime inIervol) i Total 310 t

The time required to complete the various actions for Trains A and B were determined by reviewing the prece-dural steps required (per the Procedure Reference listed) with the PSC operations personnel who would be imple-t menting the actions. The above times represent the sum total to perform all manual actions for fires in various areas. It should be noted that not all of the above actions will be required for o single fire. Thus for a single fire the time required to perform manual actions will be significantly reduced.

l 1

L I

DC-85-184 26 l

t

NRC QUESTION 13.

Provide an evaluation that the Technical Specification

required, onshift, crew size is sufficient to perform the actions proposed for the various shutdown models without reliance on Fire Brigade members.

PSC RESPONSE 13.

The minimum onshift crew required by Technicat Specifications (Table 7.1-1) consists of the following personnel:

One Shift Supervisor One Senior Reactor Operator Two Reactor Operators One Equipment Operator One Auxiliary Tender Additionally, administrative controls require one additional equipment operator, one lead security officer, and one health physics technician to be available on each shift, for a total of 9 personnel.

Administrative controls on staffing levels have been accepted at other plants for Appendix R shutdown actions and the Fire Brigade.

For Sequence "A"

actions, 340 minutes is the conservative maximum time required from 12 above.

This equates to 5.67 manhours of time.

Since forced circulation cooling must be re-established within 1-1/2 hours per Report No.

1 Section 2.5, this is the time window in which the Sequence actions must be accomplished.

With five of the nine onshift personnel dedicated to the Fire

Brigade, the remaining four personnel are available for shutdown activities.

This results in approximately 1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per individual for Sequence "A".

Based on the conservative accounting of required actions and estimated time for actions, an adequate staffing level is maintained.

One building will not be accessible for one hour.

DC-85-184 27

NRC QUESTION 14. It is the stoff's position that monitoring of core flux provides the only direct indication of the reactor shut-down condition and therefore provisions for postfire source range flux monitoring are necessary to meet Section Ill.L.2 of Appendix R and, therefore, A.3.a and B.2.b of the PSC letter dated August 17, 1984.

The position stated in Section 2.3 of Report I, that, since o fire connot credibly prevent control rod insertion, neutron flux monitoring is not required, has not been odequately justified. Include provisions in all shutdown models for source range (startup chonnels) monitoring or provide on evoluotion of alternatives available to monitor core re-octivity conditions.

PSC RESPONSE

14. Appendix R, Section lil.L, " Alternative and Dedicated Shutdown Capability" requires as a performance goal for the shutdown functions that process monitorhg be provid-ed, as do the criteria specified for Ft. St. Vrain contained in the August 17, 1985 PSC letter (P-84281). IE Informa-tion Notice 84-09, (also SECY 83-269) provided the mini-mum monitoring capability which the NRC staff considers necessary to achieve safe shutdown. Source range flux monitoring was included in the list of instrumentation for PWR's; however, it was not included in the list of instro-mentation for BWR's. (A list of instrumentation was not provided for an HTGR.) The justification for source range flux monitoring for PWR's is provided in internal NRC staff memorandum from L. S.

Rubenstein to R. J.

Mattson, dated March 16, 1983. Per the staff memoron-dum, the need for source range monitoring is to detect changes in reactivity resulting from potential fire-induced boron dilution. For PWR's, shutdown margin is provided by a combination of control rods and boron concentration in the reactor coolant system. Whereas BWR's (and Ft.

DC-85-184 28

w St. Vrain) provide sufficient shutdown margin by control rods alone. The staff memorondom does not mention a need to verify control rod insertion. Therefore, source range flux monitoring is not needed to verify rod insertion

.and thus, BWR's and Ft. St. Vrain would not need flux

. monitoring capability. Further support of this position is provided by Generic Letter 85-01 which states that for P

control room fire considerations, a manual scram may be assumed to occur prior to leaving the control room.

~ For Ft. St. Vrain, on analysis was performed as described in Section 2.4 of Report I demonstrating that a fire connot credibly prevent control rod insertion. Based on this analysis. and the above described staff guidance, source range neutron monitoring is not required. How-ever, further steps have been taken beyond the guidance described above to assure reactivity control in the Appen-dix R shutdown model. As noted in Section 2.3 of Report

- I wide range neutron monitoring instrumentation has been identified, and can be used if not damaged by a fire. The redundant channels are generally widely separated; how-ever, they.do not meet the separation / protection criteria of Appendix R in the vicinity of PCRV. In accordance with Appendix C (Poge C-2) to Report 4, steps have been added to procedure SSCHDPC to monitor neutron instru-mentation following a reactor trip. If following a fire the neutron monitoring instrumentation is lost or indication appears to be erroneous, the Reserve Shutdown Procedure will be activated.

Based on the above referenced guidance, the similarity of Ft. St. Vrain reactivity control to a BWR, the control rod drive evaluation performed for Ft. St. Vrain, and the additional steps token to assure reactivity control by the DC-85-184 29

Reserve Shutdown System, startup range neutron monitor-Ing need not meet Appendix R Section Ill.G criteria.

NRC OUESTION 15. Provide on explanation of how the reactor decoy heat removal function will be monitored following a fire Tn a noncongested cable area (Criterio B.2.c. and d.) without maintaining Primary Helium temperature instrumentation operable.

PSC RESPONSE

15. Reactor decay heat removal I as two components: core heat removal and secondary heat removal. In PWR's core Delta T (THot and T old) are monitored to provide an C

indication of core heat removal since flow rates under natural circulation are so low.

To confirm core heat removal (i.e., transfer of heat to the helium), instrumen-tation is provided to confirm helium flow. It is a given physical condition that heat will be transferred to the helium as it passes through the core.

To confirm secondary heat removal (i.e., that heat is transferred to the water in the steam generator EES section), two instrument variables are required. Monitor-ing of S/G feedwater flow confirms that water is passing through the EES section, and in the process transferring heat from the primary helium fo the service water system through the downstream decay heat removal exchanger.

To confirm that sufficient feedwater flow is provided, temperature indication is available monitoring EES outlet temperature.

Constant or decreasing temperature is indicative of adequate feedwater flow, and confirmation of decay heat removal.

DC-85-184 30

Accordingly, the combination of primary helium flow, s

feedwater flou, and steam generator EES outlet tempera-

-ture instrumentation provide confirmation of reactor decay heat removal. All of this instrumentation will be separated / protected to satisfy Section Ill.G of Appendix R or is covered by on exemption request.

NRC QUESTION 16. Provide an explanation of how the reactor pressure control function will be monitored for a fire in a congested cable area (Criteria A.3.b. and d.) without maintaining Primary Helium pressure instrumentation operable.

PSC RESPONSE

16. Criterion A.3.b states that "The pressure control function shall be capable of achieving depressurization (if required) through the helium purification system." This is accomp-lished via SOP 48-01, Section 8.4, through the timed opening of two manually operated volves. These valves are orificed to achieve the desired depressurization in the' proper time. Accordingly instrumentation is not required for this function; however, the process can be raonitored (Criterion A.3.d) via a local mechanical pressure indicator, PI-23162, as indicated in Report No. I Table 2.3-3.

t

+

DC-85-184 31

TABLE 6A REQUIRED ACM FIRE PROTECTION SHUTDOWN COMPOENTS Component ID Normal Power Supply Diesel-Driven Generator K-4804 Power Source (2500 kW)

Electrical Equipment ACM Bus Transfer Switches 4160V to 480V N-4867 ACM Diesel Transformer Stock Effluent Radia-PING-1 480V ACM Bus

• tion Monitor (PING-1)

Firewater Pump (Motor P-4501 480V Essential Bus No. I Driven)

Service Water Pump P-4201 or 480V Essential Bus No. I P-4202 480V Essential Bus No. 2 Service Water Tower C-420lx or 480V Essential Bus No. I Fan C-4202x 480V Essential Bus No. 3 Service Water Return P-4203 or 480V Essential Bus No.1, TB MCC l Pump P-4204 480V Essential Bus No. 2, TB MCC 2 PCRV Liner Cooling P-4601 or 480V Essential Bus No. I Pumps (2)

P-460lS &

480V Essential Bus No. 2 P-4602 or 480V Essential Bus No. 3 P-4602S 480V Essential Bus No. 2

^

Circulating Water P-4118 or 480V Essential Bus No. 3 Makeup Pump P-4118S 480V Essential Bus No.1 Reactor Plant Exhaust C-7301 or 480V Essential Bus No. I, Rx MCC I A Fon C-7302 480V Essential Bus No. 2, Rx MCC 2 Diesel Oil Transfer

- P-4803 or 480V ACM Bus

• Pump P-4804 480V ACM Bus

• Helium Purification P-4701 or 480V Essential Bus No. I, Rx MCC I A Cooling Water Pump P-4702 480V Essential Bus No. 3, Rx MCC 3

TABLE 6A REQUIRED ACM FIRE PROTECTION SHUTDOWN COMPOENTS (Continued)

Component

_ID Normal Power Supply Normal Lighting Cabinets Selected Plant Lighting Firewater Pump House C-7521 or 480V Essential Bus No. I, TB MCC I Vent Fans & Louvers C-7522 480V Essential Bus No. 3, TB MCC 3 Motor Operated Valves (2)

HV-2301 or 480V Essential Bus No. 2, Rx MCC 2 HV-2302 480V Essential Bus No. 3, Rx MCC 3 Reserve Shutdown Rocks 1-21 A, instrument Bus No. I and 2 or manual System I-218,1-2 l C, 1-21D Startup Battery for 480V ACM Bus

• Diesel Generator and D.C. Control Liner Cooling Water TI-4629 and N/A - Local Mechanical Temperature indicators TI-4630 Helium Purification PI-23162 N/A - Local Mechanical System, Helium Pressure Indicator Reactor Building Exhaust PDI-7323-1 &

N/A - Local Mechanical Fan Delta P PDI-7339-1 Liner Cooling Pump PI-46334, N/A - Local Mechanical Discharge Pressure PI-46335, PI-4663, &

PI-4664 Service Water PI-4214, N/A -Local Mechanical Instruments PI-4216, PDIS-4226, PI-4204 &

PI-4206 Fed from 480V Turbine Plant HVAC non-Essential bus during normal operation.

Several. One for each.