ML20205C845

Forwards Info Supporting 850830 Request to Operate at 8% Power to Facilitate Core Dryout for 45 Days,Per 850826,0903 & 04 Telcons.Moisture Removal Needed to Maintain Conditions Prescribed in FSAR & Tech Specs

ML20205C845
Person / Time
Site:	Fort Saint Vrain
Issue date:	09/10/1985
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
P-85312, TAC-42527, NUDOCS 8509230253
Download: ML20205C845 (10)

Text

.

2420 W. 26th Avenue, Suite 1000, Denver, Colorado 80211 h PublicService s.,m.

Companyof Colorado September 10, 1985 Fort St. Vrain Unit No. 1 P-85312 Regional Administrator ,

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Region IV -

U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 i SEP l 6 E Arlington, Texas 76011 Attn: Mr. Dorwin Hunter 1 Docket No: 50-267

SUBJECT:

Request for Low Power Operation

REFERENCE:

(1) NRC Letter, Martin to Lee, dated July 19, 1985 (G-85288)

(2) PSC Letter, Warembourg to Hunter, dated Au ust 30, 1985 (P-85302

Dear Mr. Hunter:

In Reference 1, Public Service Company (PSC) received authorization from the NRC to operate Fort St. Vrain at a power level no greater than 15 percent until certain equipment qualification issues are resolved. In Reference 2, PSC submitted a request for NRC concurrence to operate Fort St. Vrain at power levels up to 8 percent power for a period of time not to exceed 45 days. As a result of telephone discussions between the NRC and PSC on August 26, September 3, and September 4, 1985, PSC is submitting, in Attachment 1, additional information requested by the NRC.

Reference 2 indicates PSC's desire to operate FSV for a period of up to 45 days at 8 percent power in order to facilitate core dry-ou t.

PSC does not intend for this operation to extend beyond November 30, 8509230253 850910 pm awcn eagg7 eb 9 6 i

Page 2 1985. Concurrently with low power operations, PSC will continue to aggressively resolve the issues associated with 10CFR50.49. Plant operations during the early phases of the environmental qualification work will not have any adverse impact on the work in progress.

PSC also summarized in Reference 2 the reasons for the need to operate at low power for moisture removal as opposed to allowing the moisture to remain in the PCRV for an additional period of time.

After the extensive refurbishments that PSC recently completed, PSC considers it most prudent to continue to reduce moisture in the PCRV.

Present removal techniques are limited by the availability of auxiliary steam and, at this time, a moisture removal rate plateau has been reached. Minimizing the amount of moisture in the PCRV reflects PSC's desire to maintain the FSV reactor in conditions described in the FSAR, design documents, and as prescribed by Technical Specifications. An additional benefit to moisture removal during low power operation is the opportunity for PSC to restart FSV more readily and efficiently at a future point in time, thus providing a substantial benefit for the company.

Following a high energy line break, PSC has verified that the liner cooling system can be established totally by manual actions and no qualification of electrical equipment in harsh environments is required. Procedures are being developed which provide for: manual initiation of the liner cooling system; verification of equipment operation and valve line up following a high energy line break; and periodic verification of equipment operation and valve line up thereafter. Operators will receive training concerning possible erroneous indications or misleading information and the need for independent verification of proper corrective actions.

In conclusion, PSC believes the information contained in Attachment 1 to this letter, the materials submitted in Reference 2, and related discussions held with the NRC provide the justification needed to operate FSV at 8 percent power for a period not to exceed 45 days.

If further technical discussions are needed to address this issue, the PSC staff will be available to meet with the NRC at a time convenient for the NRC.

Very truly yours, b W % ds D.W. Warembourg Manager, Nuclear Engineering Division DWW/DG/jt Attachment

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

Attachment 1 to P-85312 Page 1

, PSC RESPONSES TO NRC INQUIRIES

1. NRC Question: Why is nuclear heat required to remove moisture

from the FSV core?

PSC Response: Nuclear heat / steam has two primary advantages over auxiliary boiler steam. First, nuclear steam can drive the circulators at a higher speed and j increase primary coolant flow. Second, increased helium temperatures increase the moisture carrying capability of the helium. Past experience has shown that this higher primary coolant flow coupled with the higher helium temperatures is very significant in reducing the time involved in removing the moisture from the PCRV liner insulation.

2. NRC Question: Will both auxiliary boilers be on line at 8 percent power?

PSC Response: Both boilers would likely be on line at 8 percent

- power. Insufficient steam, however, is available from these sources alone to provide the helium j circulator rpeeds required to increase the .

moisture removal rate.

! 3. NRC Question: Has PSC considered bringing in an additional j boiler, perhaps truck mounted, to give more steam

output?

4 PSC Response: Due to the complexities of connecting a temporary

boiler to the existing steam piping systems in

accordance with ASME boiler code, PSC does not i consider this to be a viable option. ,

- 4. NRC Question: What are the predicted maximum and average fuel temperatures at 8 percent power? What would these 4

temperatures be if there were a reactor trip and only liner cooling available?

4-

! PSC Response: Computer simulations of the FSV plant were run i with the RECA 2/EE computer program to calculate the maximum and average fuel temperatures as a result of a steam line break occurring from 8 4

percent power with loss of forced circulation cooling. These runs were made both for the case i

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Attachment 1 to P-85312 Page 2 with the liner cooling system operating (one-loop only) and without any liner cooling. An initial core average temperature of 640 degrees Farenheit was assumed for both cases at time of trip. The maximum fuel temperature was also assumed to be 640 degrees Farenheit, i.e., initial uniform temperature distribution was assumed. The results of these runs are:

A. One Loop of Liner Cooling in Operation - The maximum fuel temperature is calculated to be

-938 degrees Farenheit. This maximum temperature is reached at 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and then slowly decreases. The maximum core average temperature was calculated to be 726 degrees Farenheit at about 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br />. The maximum liner temperature was 140 degrees Farenheit.

B. No Liner Cooling in Operation - The maximun fuel temperature at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the loss

of forced circulation cooling was calculated to be 1034 degrees Farenheit. At this point it was still increasing at a rate of about 0.5 degrees Farenheit per hour. The core average temperature after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> was 850 degrees Farenheit. The faulted condition limit for the concrete temperature is 400 degrees Farenheit. The. liner temperature is computed to reach approximately 400 degrees Farenheit after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> into the transient.

Since the concrete temperature will be less than the liner temperature, the time available to manually establish PCRV liner cooling is conservatively estimated at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. This is ample time to gain access to the Peactor Building and manually establish PCRV liner cooling.

The above computations were performed for equilibrium core conditions at 8 percent power and therefore no credit was taken for the maximum of 45' days of power operation at 8 perccnt power.

5. NRC Question: In : Reference 2, PSC letter P-85302, PSC indicated seventy hours could elapse prior to the establishment of forced cooling. What is the

Attachment 1 j to P-85312 1 Page 3

]

1 basis for this seventy hour time? l PSC Response: This seventy hour calculation was. made on an l interim basis to establish a conservative estimate '

prior to the computerized calculations given in l Question 4 above. The purpose of this calculation l was to establish an outer bound for the time limit to restore forced circulation cooling after it has been lost from an initial power level of 8 percent which is equivalent to the 1 1/2 hour time limit '

at 100 percent power.

4 The decay heat rate curve for 100 percent equilibrium power was integrated over the 1 1/2

hour period following shutdown to determine the total heat required to raise the fuel, and operating

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surrounding materials, from the conditions to the point where the maximum fuel temperature approached 2900 degrees Farenheit .

The decay heat rate curve for the case of 8 percent power operation for 45 days was integrated out to a time "T" after shutdown: This total heat input, which is a function of "T", was equated to the total decay-heat derived above. The resulting equation was then solved to yield a value of "T" (70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br />) which is the minimum time available before forced circulation cooling must be established for the 8 percent power case.

This calculation was conservative in comparison with the RECA run for the 8 percent equilibrium case with no liner cooling. In this condition, the RECA results indicate that the maximum fuel temperature is far below 2900 degrees Farenheit after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (1034 degrees Farenheit in Question 4 above).

, 6. NRC Question: In P-85302, PSC indicates that 1 1/2 hours has been established as the time required before forced circulation cooling at 100 percent power must be resumed to prevent fuel failure. Would PSC elaborate on this statement?

l PSC Response: For the design basis earthquake or " maximum tornado" discussed in FSAR Section 10.3.9 (assuming 100 percent reactor power),

i

i Attachment 1

- to P-85312 Page 4

! l l

reestablishment of forced circulation core cooling is accomplished by' using the Safe Shutdown equipment items. In this scenario, the fire water system provides the coolant and motive power for the circulators and secondary coolant for core

. heat removal. Under this accident condition, non-Class I systens are postulated to fail. It is conservatively assumed (FSAR Section 10.3.9) that approximately 60 minutes are required to inspect the critical boundary valves and to manually close these if necessary. An additional 25 minutes is

, assumed to be required to align the fire water for motive force to drive a circulator and provide cooling to a steam generator. For these time-conditions, approximately 1 1/2 hours is assumed before forced circulation core cooling can be restarted. FSAR Section 14.4.2.2. contains the analysis for the cooldown on one boosted fire water driven circulator assuming the 1 1/2 hour a

delay in restart. Using various computer codes 1 (e.g.,RECA),this analysis indicates that the 3 ,

maximum fuel temperature peaks at the 1 1/2 hour

restart of forced circulation core cooling and

then again at 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> into the transient. Both peaks (maximum of 2750 degrees Farenheit) remain below the 2900 degree Farenheit value where fuel deterioration is postulated to occur (see FSAR, Figure 14.4-6).

7. NRC Question: Please discuss the availability and reliability of

the liner cooling system in the event of a steam line break at 8 percent power , especially as it ^

relates to equipment subject to a harsh environment.

PSC Response: The normal supply for liner cooling water is the '

Reactor Plant Cooling Water System (System 46) pumps. These pumps are located on Level 8 in the

. Reactor Building and east of the 4A wall.

Therefore, with the relatively low Reactor Building temperature profiles at 8 percent power, it is expected that these pumps would not be affected during the steam line rupture. The remainder of System 46.is also expected to survive 2 the accident due to the relatively low temperature profile.

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Attachment 1 i -

to P-85312-Page 5 However, even if all electrical. items in System 46 fail, fire water can be manually valved into the liner cooling system. This would be a once through system with the water being supplied by the fire water pumps located outside of the building. In this case, no qualification is required of any electrical items. Portions of the liner cooling system exposed to a harsh '

i environment can be manually operated.

8. NRC Question: Does PSC expect to have to rely on any cooling system other than the liner cooling system if a break were to occur at 8 percent power?

PSC Response: No. Using the scenario described in Question 7 with the liner cooling being supplied by the fire

water pumps, no electrical items inside the

. hazardous environment are required. The fire j water pumps are located outside of the Reactor and

Turbine Buildings and all items in these buildings

, required to line up liner cooling via fire water {

are manually operated.

9. NRC Question: What codes were used for the design and i installation of the reheat steam piping? To what

codes has PSC performed ISI and IST for the reheat i steam piping?

PSC Response: The original piping code used in design and installation of the reheat steam piping was USAS (currently ANSI) B31.1.0-1967. PSC has no formal ISI/IST program for the reheat steam piping.

During the course of maintenance or modification work on the reheat steam piping, inspecting and l testing of these areas is performed.

l' 10. NRC Question: Please discuss the effects of other potential high energy line breaks such as main steam and feedwater.

I PSC Response: In the Reactor Building, a cold reheat steam line break is still the worst case line break since all other ruptures are automatically isolated by the existing steam pipe rupture detection system. ,

In the Turbine Building, the hot reheat steam line

! rupture is still the worst case. Main steam and i

I l

Attachment 1 to P-85312 Page 6 feedwater at 8 percent power are both normally solid water. After a rupture of either line, the fluid flashes to steam and results in a saturated mixture. Therefore, the steam released in either 4 case would be about 204 degrees Farenheit which would result in lower atmospheric temperatures than the Reference 2 hot reheat case which resulted in air temperatures up to 207 degrees Farenheit.

11. NRC Ouestion: Would PSC explain why the hot reheat steam line break in the Turbine Building and the cold reheat steam line break in the Reactor Building are the worst case scenarios?

PSC Response: In the ' Turbine Building, at 8 percent power, hot reheat steam temperature is 634 degrees Farenheit, pressure is 25 psig and enthalpy is approximately 1352 BTU per pounds mass. For main steam in the Turbine Building, temperature is 590 degrees '

Farenheit, pressure is 1737 psig, and enthalpy is 601 BTU per pounds mass. As demonstrated above, hot reheat is considered the worst case due to enthalpy and the fact that main steam is below saturation point.

In the Reactor Building, cold reheat steam temperature is 410 degrees Farenheit, pressure is 135 psig and enthalpy is 1227 BTU per pounds mass.

Main steam temperature is 590 degrees Farenheit, pressure is 1737 psig, and enthalpy is 601 BTU per pounds mass. In this case, cold reheat has a higher enthalpy and is not automatically isolated .

using the steam pipe rupture detection system.

4 Main steam and hot reheat steam are isolated and therefore their blowdown time is much shorter.

l NOTE: In both cases, the auxiliary boilers'are assumed to be operating.

l 12. NRC Question: Would flooding or submergence result at FSV from a feedwater or main steam line break?

PSC Response: Submergence of safe shutdown equipment in the Reactor Building is addressed in Section 1.4.6(d) of the FSAR. Based upon the automatic isolation of feedwater/ main steam piping in the Reactor i

a s

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. Attachment 1 to P-85312 Page 7 Building and the large capacity of the Reactor Building Sump (approximately 345,000 gallons),

submergence and flooding are not deemed to be problems in the Reactor Building for safe shutdown equipment, other than the four valves previously identified to the NRC in our August 20, 1985 letter (P-85293).

Flooding of equipment in the Turbine Building is addressed in detail in Appendix I.6.1 of the FSAR.

Even for a worst case scenario of manual isolation of full feedwater flow (4612 gpm) in four minutes, the capacity of the condenser pit and sump (approximately 110,000 gallons) is sufficient to contain the amount of liquid from the postulated leak.

13. NRC Questions: What is the status of PSC's walkdown?

PSC Response: PSC has completed the field walkdown efforts and is presently evaluating the field walkdown findings. Based on the results of our evaluations, PSC will determine the necessity of i expanding the scope of the walkdown efforts.

14. NRC Question: Please describe the corrective action system being used in the environmental qualification program field walkdown.

~

PSC Response: Inspection findings from the field walkdown are recorded in field walkdown verification sheets, which are, in turn, evaluated and processed into the Station Service Request (SSR) system. The SSR system (which has been revised to address EQ concerns) is utilized as the controlling document for initiating work. The SSR is a recognized corrective action vehicle within the FSV procedural system and provides for:

A. Daily technical review for reporting requirements and compliance to Technical Specification.

B. QA/QC reviews

i . Attachment 1 i -

to P-85312 Page 8 i

i C. Identifying procedures and/or controlling i i

documents. ,

l The SSR system also provides for a technical i review of potential generic concerns-(such as tape splices, rust in junction boxes, etc.). PSC intends to review the generic applications and take appropriate corrective actions. It is important to note that in terms of the disposition i of corrective action items, PSC intends to utilize j the liner cooling systen to recover from . design basis events when operating at the 8 percent power level.

15. NRC Question: If PSC is not taking credit for automatic operation of Safe Shutdown equipment can the r failure of this equipment affect the liner cooling system?

PSC Response: As explained in the answers to questions 7 and 8, i the fire water can be used to provide liner

! cooling water. Since all operations to line up i the system with fire water are manual, no ,

i automatic actions located in the hazardous i environment are required. Also, any of the liner j cooling valves that may receive inadvertant

! signals to close can be manually overridden to [

j continue cooling water flow. [

! 16. NRC Question: In the case of a steam line break, would the e cooling medium for the System 46 heat exchangers still be available?

i PSC Response: The cooling medium can be service water, i circulating water, or fire water. It is

! reasonable to assume that these systems can be

manually lined up, if necessary. However, it is j important to note that fire water cool down is not

a closed loop system and is not dependent upon j heat exchanger cooling water.

i 4

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