ML20138C811

Forwards Response to 851011 Request for Addl Info Re 850924 Request for Schedule Extension for Environ Qualification Program Under 10CFR50.49 from 851130 to 860331

ML20138C811
Person / Time
Site:	Fort Saint Vrain
Issue date:	10/21/1985
From:	Lee O PUBLIC SERVICE CO. OF COLORADO
To:	Butcher E Office of Nuclear Reactor Regulation
References
P-85378, TAC-59787, NUDOCS 8510230052
Download: ML20138C811 (24)

Text

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P.O. Box 840 October 21, 1985 Denver, co 80201 0840 Fort St. Vrain Unit No. 1 OSCAR R. LEE P-85378 VICE PRESIDENT Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Mr. E.J. Butcher, Jr., Acting Chief Operating Reactors Branch No. 3 Docket No. 50-267

SUBJECT:

Additional Information to Support Schedule Extension, 10CFR50.49

REFERENCE:

1) PSC Letter dated 9/24/85, Walker to Chilk, (P-85334)

2) NRC Letter dated 10/11/85, Butcher to Lee, (G-85416)

Dear Mr. Butcher:

Reference 1) submitted the Public Service Company of Colorado's (PSC) request for a schedule extension for the Fort St. Vrain (FSV) environmental qualification program under 10CFR50.49 from November 30, 1985 to March 31, 1986.

.In Reference 2) the NRC submitted a list of questions to PSC regarding the PSC schedule extension request.

The attachment to this letter provides PSC's responses to the NRC questions submitted in Reference 2).

If you have any further questions regarding PSC's schedule extension request, please contact Mr. M.H. Holmes at (303) 571-8409.

Very truly yours, 0.R. Lee, Vice President l

Electric Production ORL:DM:jrp cc: Ken Heitner 8510230052 851021 PDR ADOCM 05000267 P

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NRC QUESTION #1

.What is the capability of the reactor building confinement and louvers to withstand a steam line break and still function ~ effectively to reduce the radiological consequences of Design Basis Accident-1 (DBA-1)?

4 PSC RESPONSE The capability of the reactor building confinement and louvers to withstand a steam line rupture and still function to reduce the radiological consequences of DBA-1 is under evaluation..The results of this' evaluation will be provided to the NRC as soon as they are available.

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NRC QUESTION #2 What is the effect of the moisture released 'by the break on the effectiveness of the reactor building exhaust filters?

PSC RESPONSE The reactor building exhaust filters are designed to operate under abnormal conditions with an inlet temperature of 170 degrees F,

26" w.g.

pressure and relative humidity of 100%. The filters include a moisture separator to remove entrained water droplets from a steam-air mixture and prevent blinding of the downstream water repellent HEPA filter. The charcoal filters have a minimum efficiency of 99% for removal of elemental iodine at the above abnormal design conditions.

Tests also show that the efficiency of impregnated charcoal to remove i

organic halides such as methyl iodine is better than 98% at a relative humidity below 70% and better than 85% up to 97% relative humidity.

(FSAR Section 6.2.3.2.2)

Building temperature and relative humidity profiles are currently being developed for a steam line break isolated by the Steam Line Rupture Detection / Isolation System (SLRDIS).

PSC will assess the effect of moisture released by the break on the effectiveness of the Reactor Building Exhaust Filters once these profiles are defined.

PSC is also evaluating the need to increase the setpoint of the charcoal adsorber bed fire protection water spray system to preclude actuation of this system following a steam line rupture in the Reactor Building. The current setpoint of 200 degrees F is well below the minimum ignition temperature of 656 degrees F specified for the charcoal adsorbers. The decision on whether to increase this setpoint is also contingent on the results of the building temperature profiles currently being developed based on the SLRDIS system.

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NRC QUESTION #3 l

What are the safety benefits from operating the plant at partial pcwer during the period November 30, 1985 to March 31, 1986?

PSC RESPONSE PSC has, in the recent past, performed evaluations for 15% and 8% power operation for the purposes of obtaining NRC permission to dry out the core in preparation for full power operation. The lowest partial power operation between November 30, 1985 and March 31, 1986 that would produce any economic benefit would be 35% since this is the minimum power at which electricity could be generated.

Steam temperature and pressure conditions at 35% power are similar to those at 100% power. Thus, the harsh environment resulting from a steam line break at 35% power will be about the same as that for a stean line break at 100%

power.

For the interim

period, November 30, 1985 to March 31, 1986, liner cooling with fire water still has to be assumed as the only method of cooling.

Detailed calculations have not been performed by PSC to determine the effects on either the PCRV internals or the radiological consequences to the health and safety of.the public for partial power operation at 35% and above.

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NRC QUESTION #4 How many equipment items ar.d total ccmponents will be qualified as of November 30, 1985? How many additional equipment items and total components will be qualified by March 31, 19867 PSC RESPONSE PSC has committed to the qualification of equipment necessary for both the SLRDIS and liner cooling, using fire water, by November 30, 1985 or prior to plant restart for 100% power operation. PSC is concentrating its efforts on this equipment and will have this set. of equipment qualified as previously committed. Based on the current FSV EQ master equipment list, this set of equipment includes a total of 51 equipment items comprising 438 components.

(Note:

Several components may constitute one item; for example, if there are six valves of the same type, the valve type would be considered one item and the six individual valves would be considered six components).

By March 31, 1986, PSC will have qualified all equipment necessary for i

forced circulation cooling.

Based on the -current FSV EQ Master Equipment List, this set of equipment includes 156 items representing a i

total of 1246 components. These totals include some of the equipment numbers stated above for liner cooling.

PSC is concentrating its efforts primarily on the equipment necessary l'

for the SLRDIS and liner cooling using fire water.

Because of the number of identical items in both sets of equipment identified above, and due to the aggressive efforts underway, PSC is confident that a large number of the forced circulation cooling equipment-will also be qualified by November 30, 1985.

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NRC QUESTION #5

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I What actions would be required to return Fort St. Vrain (FSV) to power operation from a permanent loss of forced circulation cooling at partial power?

PSC RESPONSE As discussed in the response to Question #3, PSC has not performed any detailed calculations on the effects of a steam line break at partial powers of 35". and above.

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NRC QUESTION #6 i

Identify and discuss the reasons for the equipment used at FSV being j

considerably different from the equipment used in Light Water Reactors (LWRs)

(e.g.,

transmitters, valve operators, temperature sensors,

.etc.).

l PSC. RESPONSE i

With a few exceptions, PSC does not consider the electrical equipment required for safe shutdown of FSV to be significantly different than the safety related equipment found in LWRs. The lack of EQ information applicable to the FSV HTGR is due to the higher temperatures that i

result from the FSV HELB, assuming a four minute isolation time. See j

the PSC response to NRC Question #7 for further information.

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• NRC QUESTION #7 Allowing 4 minutes for leak termination, the accident environmental profiles for FSV are not exceptionally higher than those seen in Pressurized Water Reactor (PWR) containments (400 degrees F - 500 degrees F). What is the basis for stating that Equipment Qualification (EQ) information from LWR industry is not available or applicable to FSV?

PSC RESPONSE While some Westinghouse PWRs may experience temperatures as high as 430 degrees F following a HELB, the electrical equipment that is exposed to this accident is fairly limited.

The vast majority of the safety related electrical equipment in Westinghouse PWRs have been qualified to the combined PWR/BWR profile in IEEE-323-1974 (Peak temperature 340 degrees F). This has been confirmed by Westinghouse.

PSC has assumed that c'urrent industry practice does not allow use of the Arrhenius method to demonstrate qualification when the test profile does not envelope the peak accident temperature.

Thus, the 340 degrees F peak profiles used to oualify most LWR equipment could not be used to qualify the FSV equipment where peak temperatures range from 400 degrees F to 650 degrees F for a 4 minute isolation.

Section 1.4 of the FSV FSAR provides details of the temperature profiles calculated previously for 4 minute operator terminated steam line breaks.

l In the development of the EQ program, it was PSC's understanding that while thermal lag calculations are in certain instances an acceptable means of qualifying components with a non-enveloping accident profile, PSC did not believe this to be an acceptable basis for the entire FSV EQ progran. Use of industry qualification data in lieu of actual test data for FSV profiles would require thermal lag calculations on most all of the FSV safe shutdown equipment. Thus, the entire program would have to be based on analyses which were not consistent with our original program bases.

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NRC OUESTION #8 How does the acceptability or unacceptability of the 4 minute isolation time impact the following specific. deficiencies in the FSV EQ program?

Aging effects on equipment?

Post-accident operability time?

Field walkdown for equipment verification?

Identification of the equipment to be included in the qualification program?

Review and evaluation of the completeness of EQ documentation and i

files?

1 PSC RESPONSE i

PSC did not mean to imply in the 9/24/85 letter that the 4 minute j

isolation time solely impacted any deficiencies in the FSV EQ program.

However, consideration of this issue at the same time as other issues (e.g., aging, operability times, etc.) caused many areas of FSV's EQ program to be impacted.

With these thoughts in mind the following provides an item by item response to your question:

Aging effects and Post-accident operability times.

i The 4 minute isolation time has no effect on the necessity of providing aging or operability studies.

However, once the i

need for these studies was determined the associated calculations are impacted. That is, the accelerated aging that occurs during the 4 minute isolation is very much different from that which occurs during a faster isolation time.

PSC has committed to install a SLRDIS to resolve the operator response time issue and, in addition, has coninitted i

to aging and operability time studies. One area cannot be r

divorced from the other with reference to the overall EQ program requirements.

Field walkdown for equipment verification and identification of equipment to be included in the qualification program.

j The 4 minute isolation issue did not necessitate the plant walkdown and systems review efforts undertaken by PSC. These efforts were initiated as a result of PSC's composite considerations of all the issues, i.e.,

operator response, aging, operability time, files and docurentation, which were identified by the NRC in early 1985.

We agree that these efforts resulted in the identification of previously unknown concerns that cannot be directly related to a four minute isolation time.

It was not our intent however, to infer any i

such direct relationship in our September 24, 1985 letter.

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.g-Review and evaluation of the completeness of EQ documentation and files:

The 4 minute isolation time by itself did not necessitate an overall review and evaluation of the FSV EQ files.

All EQ i

files, however, will be affected and must be revised because of PSC's connitment to install the SLRDIS to resolve the 4

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minute response time issue. As' a result,. new equipment items are now being introduced and harsh environment conditions i

have changed.

In addition, the EQ files would have had to be revised to incorporate aging and operability time.

Again, the direct relationship and impact of the 4 minute isolation time is difficult to demonstrate, but the indirect relationship and impacts are clearly evident.

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NRC QUESTION #9 l

Does the installation of an automatic leak detection / isolation system f

adversely impact any aspect of the present assumptions of the EQ j

program regarding resulting accident environment?

(i.e.,

do the resulting accident environmental conditions become more severe than i

presently estimate?)

i pSC RESPONSE The installation of SLRDIS will not adversely impact any aspect of the l

present assumptions of the EQ program regarding the resulting accident environment. On the contrary, by reducing the amount of steam escaping l

following the steam line rupture, the resulting temperature profiles will be lower.

These lower temperature profiles will have a favorable impact in the areas of aging and operability time.

However, it will 4

take additional tine to generate and evaluate these new profiles and to -

incorporate them into the overall EQ program.

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NRC QUESTION #10 Provide sufficient detail for the NRC to evaluate the adequacy of the proposed leak detection / isolation system or the schedule for its submittal.

Provide assurance that the system will be fully EQ'd at time of installation.

PSC RESPONSE The design of the SLRDIS is in the final stage. The present schedule indicates that the onsite review committee (PORC) review and the offsite review committee (NFSC) review of this system will be complete by the end of October 1985. The design details and a safety evaluation of the SLRDIS will be submitted to the NRC during the first week in November 1985.

The SLRDIS will be environmentally qualified for the harsh environment to which its equipment will be subjected.

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NRC QUESTION #11 l

Redundant, safe shutdown, forced circulation cooling water systems using fire water were considered qualified prior to early 1985.

Assuming a 4 minute isolation time is acceptable, identify the requirements that were modified that led the PSC to conclude that now "not all of this eouipment can be considered fully qualified to the requirements of 10CFR50.49".

PSC RESPONSE As stated in past correspondence (e.g.,

PSC letter 0.R.

Lee to E.H. Johnson, P-85103, dated March 25, 1985), PSC considered the redundant, safe shutdown, forced ' circulation cooling water systems i

using fire water to be qualified to the FSV EQ program which was i

described to the NRC in correspondence several times during the period between 1980 and 1983.

(See Attachment 1 to PSC letter R.F. Walker to S.J.

Chilk, P-85334, dated September 24, 1985).

In 1985 the NRC 1,

questioned and found unacceptable, several elements of this initial FSV EQ program.

Specifically, in NRC letter E.H. Johnson to 0.R. Lee dated May 7, 1985 (G-85178), the NRC staff took the position that aging effects should be evaluated and operability times need to be i

established for FSV EQ progran equipment.

In this same NRC letter, the l

NRC staff took the position that PSC's 4 minute operator response time had not been adequately demonstrated and that PSC should evaluate fully i

automating the isolation of all steam pipe ruptures.

PSC's initial determination and treatment of aging and operability time wTs made in accordance with the provisions of the NRC Order for M ification of License Concerning Environmental Qualification of Safety Related Electrical Equipment for FSV dated October 27, 1980.

This order required PSC to apply the D0R Guidelines for FSV equipment qualification "to the extent applicable to a gas cooled reactor".

Paragraph (k) of 10CFR50.49 states that licensees are not required to requalify electrical' equipment to 10CFR50.49 if the NPC has previously i

required oualification of that equipment in accordance with the D0R Guidelines. Whereas PSC initially treated aging and operability times i

as not being of concern to a gas cooled reactor, and described that i

i treatment and justification to the NPC in writing several times between 1980 and 1983, the NRC redified these initial FSV EQ program requirements in May 1985 by taking the position that aging and operability time are of concern to a gas cooled reactor.

As noted in PSC's EQ schedule extension request dated September 24, 1985 PSC has never received from the NRC a comprehensive 4

Technical Evaluation Report (TER) or Safety Evaluation Report (SER) i including required corrective actions, in response to the initial FSV I

E0 program.

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- The above considerations, as well as the equipment qualification and re-qualification effort (including revised documentation) related to the unacceptable 4 minute isolation time and the fully automatic isolation of all steam pipe ruptures, led PSC to conclude that "not all of this equipment can be considered fully qualified to the reouirements of 10CFR50.49."

r NRC QUESTION #12 In a letter dated August 20, 1985 PSC stated that completion of all l

major aspects of EQ program review would be accomplished by September / October 1985 with the exception of steau detection / isolation system installation.

The schedule is now being revised to March 1986.

i What assurance is there that the new schedule is realistic?

PSC RESPONSES i

i PSC is not certain what dates are being referred to out of the j

August 20, 1985 letter.

PSC did indicate that various reviews and studies were going to be completed in various time frames, but there was no inference that the EQ program would be completed in a September / October 1985 time frame.

In fact, PSC indicated in the August 20, 1985 letter that we would be filing for extension beyond i

November 30, 1985. Many of the schedules set forth in the August 20, j

1985 letter were met. A few could not be completed because of the impact of other evaluations.

l The projected completion dates transmitted to the NRC are always based on the best information that is available at the time they are given.

j Unforeseen difficulties that arise during any project of this magnitude are impossible to accurately predict when schedules are formulated.

All of the necessary PSC evaluations, reviews and studies have not as i

yet been completed, and to this extent there are some unknowns in the overall scheduling process.

PSC has tried to accommodate these unknowns in the scheduling process and will make every effort to complete the environmental qualification program by March 3l,1986.

To c

this end, PSC has committed large resources, both manpower and money, to avoid delays.

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^ NRC QUESTION #13a The request for a schedule extension for meeting the requirements of 10CFR50.49 is based, in part, upon relying solely on the FSV PCRV Liner Cooling System (LCS) for decay heat renoval (using fire water as the cooling medium) until such time as EQ concerns are resolved on those systems and components used for forced He circulation decay heat removal. The following comments pertain to this mode of operation:

a.

While decay heat removal via the LCS is analyzed in the FSV FSAR

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as DBA-1, the licensing t' asis of the plant assumes several lines of defense (steam drive and water drive for the He circulators) prior to reaching a LCS only decay heat removal situation.

Therefore, it appears that the request to rely solely on the LCS for decay heat removal assumes these lines of defense are not necessary, which would appear to be beyond the licensing basis of the plant and not in accordance with Technical Specification (TS) requirements. Justification to support such operation needs to be provided.

PSC RESPONSE The FSV FSAR describes the large variety of systems and equipment available at FSV to perform safety functions, particularly the reactor

-core cooling safety function.

These systems and equipment are available to perform safety functions during normal operation, and could be utilized under accident conditions. The majority of these systems and equipment are in service during normal plant operation, thereby providing a high degree of assurance that the systems and equipment would also be available to perform the required safety functions under accident conditions. While this variety of systems and equipment could, and in reality most probably would, be available to respond to accident situations, only a limited portion of these systems and equipment are relied upon to perform the required safety functions under postulated accident conditions.

For the postulated high energy line breaks and primary coolant pressure boundary accident conditions which could create a harsh environment in either the FSV reactor building or turbine building, the same variety of systems and equipment would in all probability be available to respond, with the single exception of the system or equipment directly associated with the postulated failure.

However, not all of the remaining systems and equipnent are relied upon to respond to the DBA-1 accident in the FSAR accident analyses.

Existing FSV energency procedures for responding to the postulated accident conditions, including harsh environment accident conditions, call for the progressive application and use of all of the available lines of defense (steam and water drives for the helium circulators; feedwater, condensate and fire water to the steam generators; all of the available flow paths, pumps and heat exchangers; etc.).

For the proposed interim period of operation from November 30, 1985 to March 31, 1986, only the PCRV liner cooline system using fire water will be relied upon.

However, many other reactor cooling systems and equipment which are not directly affected by the postulated failure

will in all likelihood be available and be utilized as provided for in PSC procedures.

The PCRV liner cooling system is adequate to protect the health and safety of the public during the interim period of operation while the re-qualification and documentation of the forced circulation, safe shutdewn systems is completed.

Even though the forced circulation cooling systems are not fully qualified and documented during the interim period, this does not mean that they are unavailable and would not be utilized.

PSC considers the probability of a harsh environment accident during the interim period to be small, and PSC also considers the probability of a total sustained loss of forced circulation in the event of a harsh environment accident to be small.

Even if both of these conditions were to occur, the public health &nd safety would be fully protected and the total exposure to the public would only be a e

small fraction of 10CFR100 limits.

PSC considers the risk to our investment in the FSV plant to be minimal during the proposed interim j

period of operation, and PSC is confident that the public health and safety will be protected even in the highly unlikely event that the liner cooling system must be utilized during a postulated harsh 1

1 environment accident condition.

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• NRC OVESTION #13b b.

Section 14.10 and Appendix D of the FSV FSAR indicate that reliance on the LCS alone for decay heat removal could lead to plant damage and fuel failures at decay heat levels associated with high power operation. Therefore, the potential for plant

damage, fuel failure and fission product release should be i

d'scussed as part of the extension request.

In addition, the extent of plant cleanup or repair after such an event should also be discussed.

PSC RESPONSE i

The use of liner cooling with fire water following a steam line rupture.

with a postulated loss of forced circulation cooling will lead to fuel damage and damage to PCRV internals, including the liner thermal barrier. Outside the PCRV the amount of damage will be small, apart from any local effects from the steam line rupture. The balance of j

plant cleanup and repair would be relatively minor.

l Because of the economic costs of replacing the PCRV, it is extremely l

unlikely that the plant would ever. be restarted following the 1

postulated steam line break and failure of the forced circulation i

equipment.

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The fission product release from within the PCRV would be the same as

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that for DBA-1, based on a 0.2% leakage rate from the PCRV.

Although r

i this has been the historical number used for the FSV PCRV, it has been j

documented in correspondence between PSC and the NRC (PSC letter dated i

April 1, 1980, Warembourg to Grimes, P-80066) that, due to a mathematical error, this 0.2% leakage rate is high by a factor of about 4

200.

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NRC QUESTION #13c l

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The FSAR presents only an analysis of LCS decay heat removal using i

the Reactor Plant Cooling Water. System; however, the PSC 10CFR50.49 extension request is based upon LCS decay heat removal using fire water cooling.

Analysis of LCS performance j

(temperatures, adequacy of cooling water inventory, etc.) should be provided to' demonstrate that operation of the LCS in the fire water mode is within the bounds of the FSAR DBA-1 analysis.

PSC RESPONSE l

The FSAR states in Section 14.10 that the normal method of removing decay heat from the reactor following a DBA-1 is by a closed loop i

circulation of water using the Reactor Plant Cooling Water System's pumps to circulate the water through the system's coolers and the PCRV liner cooling tubes. FSAR Section 9.7.2 states that fire water can be j

used as a backup to this system.

r Engineering Evaluation EE-EQ-0009, is scheduled to be submitted to the NRC during the first week of November 1985 for staff reviews.

This evaluation analyzes the adequacy of fire water and the capability of its supporting mechanical and electrical equipment to safely cooldown i

the reactor and maintain it in a safe condition with radiological consecuences to the public only a small. fraction of the 10CFR100 limits.

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• NRC QUESTION #14 j

The 10CFR50.49 extension request indicates that only the LCS with fire water cooling and a Steam Line Rupture Detection / Isolation System l

(SLRDIS) need be qualified to support continued operation of the plant I

and stay within the bounds of the FSV FSAR DBA-1 analysis. Our review of the DBA-1 analysis contained in Section 14.10 and Appendix D of the FSV FSAR indicates that there are many other plant systems and i

components which are assumed to operate in the DBA-1 analysis whose i

performance could affect the outcome of this event. These are listed i

below:

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j Equipment assumed operating in the FSV DBA-1 analysis:

i At least one He purification system train for depressurization.

i this includes:

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He and charcoal cooling systems Reactor Building ventilation system. This includes:

1 One air handling unit Moisture separator i

Two out of three HEPA and charcoal filters j

ii Filter exhaust fans 1

Particulate, iodine, and noble gas exhaust monitors At least one Icop of the Liner Cooling System. This includes:

)

Control room actuation of PCRV water flow control Control room actuation of LCS cover gas pressure increase i

I Reactor plant cooling water system.

l Reserve shutdown system. This includes:

3 Control and pressurization systems The status of the qualification of all the above systems and components under steam line or feedwater line break conditions should be j

described.

This also includes their attendant AC power, DC power,

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instrument air supply, instrumentation, cooling, etc.

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For any of the above which are not qualified, the impact on the DBA-1

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analysis and conclusions should be described.

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~ PSC RESPONSE PSC's request for EQ extension identified the SLRDIS and PCRV Liner Cooling System (LCS) as the intended overall procedural method.for cooling of the FSV reactor plant during the interim period of operation (11/30/85 through 3/31/86). This method of cooldown follows the FSAR DBA-1 scenario, outlined in FSAR Section 14.10 and Appendix D, except that it minimizes the number of electrical components and/or. automatic operations required to align, control, and monitor the cooling following a HELB.

Following actuation of SLRDIS to minimize the harsh environment and to permit early access into the reactor or turbine building, this method relies on manual initiation of once-through fire water flow through the PCRV's liner cooling tubes, as mentioned in FSAR Section 9.7.2.

The detailed operating procedure will utilize one or more of the available cooling paths described in FSV's emergency procedures and " Safe Shutdown and Cooling with Highly Degraded Plant Conditions".

Engineering evaluation EE-EQ-0009 lists the systems (or portions of systems) and their function which will be relied upon to perform the required safety function, including reactor cooling with fire water.

The listed systems are:

SYSTEM FUNCTION t

SLRDIS/PPS Isolate steam line break, initiate reactor scram, and permit building access within one hour.

He Circulator Brake & Seal Isolate reactor coolant pressure i

boundary.

Hydraulic 011 Close hydraulic steam valves, f

He Purification Train Depressurization and purifi-cation flow path.

Liquid' Nitrogen Storage LTA cooling during depressur-ization.

Reactor Building Exhaust Filter gases vented through Ventilation exhaust stack.

i Reserve Shutdown Assure reactivity shutdown margin.

i Fire Protection Supply fire water to LCS, Hi h Temperature Filter Adsorber HTFA),

i Helium Purification Coolers HPC),

and fuel storage wells.

Reactor Plant Cooling Provide PCRV liner cooling j

Water System flowpath.

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SYSTEM FUNCTION Service Water Fire water outlet flowpath.

Cooling for hydraulic oil i

system and standby diesel l

generators.

Circulating Water Makeup water supply to fire pumps.

Electrical Power Operate fire, service and makeup i

(Standby Diesel Generators) water pumps, reactor-building exhaust fans, hydraulic pumps, solenoid and motor operated valves.

Electrical equipment that is required to perform the above functions and is located in a harsh environment created by a steam line break will be environmentally cualified to 10CFR50.49 prior to any power operation during the interim operating period. This equipment will be qualified based upon the enveloping temperature profile resulting from various line break scenarios. Essential electrical equipment that is not located in the harsh environment may be relied upon to function during the extended period of shutdown cooling and needs not be qualified to 10CFR50.49.

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2 NRC QUESTION #15 q

Does the Alternate Cooling Method (ACM) diesel generator provide power p

to:

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LCS instrumentation?

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Reactor Plant Cooling Water System?

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He Purification System and LCS valve operations and controls?

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He Purification System He and Charcoal Cooling System?

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PSC RESPONSE i

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PSC is not relying on Alternate Cooling Method diesel generator power i

for any harsh environment accident scenario. Normal offsite power or the two standby emergency diesel generators will provide the power required for all electrical equipment that will be used to mitigate the t

consequences of this accident. The standby emergency diesel generators i

are located within a mild environment.

Any component required for operation of the standby emergency diesel generators that is located in a harsh environment will be qualified by November 30, 1985 or prior to plant startup.

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NRC QUESTION #16 Are there procedures in place to support decay heat removal with the LCS on fire water cooling? Have the operators been trained in this mode of operation? Is all equipment required to perform in this mode of decay heat removal periodically tested?

PSC RESPONSE Normal and abnormal operating procedures were in place to support decay heat removal with the liner cooling system on fire water cooling prior to the environmental issue.

For purposes of providing a concise and explicit procedure to manually lineup the associated systems, parts of the System Operating Procedures and Safe Shutdown Cooling Under Highly Degraded Conditions procedures were combined and issued as Operations Order No. 85-17.

The procedure contained in this Operations Order assumes that no AC power, DC power, instrument air supply, or instrumentation is available to assist in the system lineups and continued operation.

All Operations personnel, including licensed i

operators and Technical Advisors have completed training on this mode of operation.

The addition and potential actuation of the SLRDIS establishes a new basis for plant operator response to a steam line break accident.

SLRDIS emphasizes isolation of the steam line break and thus ensures reactor or turbine building personnel access within one hour.

Also, during the interim-period of operation, November 30, 1985 to March 31, 1986, the electrical equipment listed in response to Question

1. 14 will be environmentally qualified.

Prior to November 30, 1985 or plant restart, procedures will be in place, and operators trained in these procedures, to-respond to these new conditions.

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Since all valve lineups are manual operations, periodic testing of these supporting valves is not deemed necessary.

Technical Specification limiting conditions of operation and surveillance tests 4

of essential equipment associated with this interim mode of operation assumes that this equipment will be ready and operable. The two fire water pumps will be surveillance tested for their rated flow and t

discharge head prior to startup for power operation during the interim operating period, i.e., their rated capacity surveillance test will be beyond March 31, 1986, i

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