ML20102B224

Fort St Vrain Station Annual Rept of Changes,Tests & Experiments Not Requiring Prior Commission Approval Per 10CFR50.59, for Period 910123-920122

ML20102B224
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/22/1992
From:	PUBLIC SERVICE CO. OF COLORADO
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ML20102B223	List: ML20102B221 ML20102B224
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NUDOCS 9207280269
Download: ML20102B224 (22)

Text

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-u 94 4 4-a' i PUBLIC SERVICE COMPANY OF COLORADO -

FORT SAINT:VRAIN STATION .

7 Ji t

ANNUAL REPORT OF CHANGES, TESTS, AND EXPERIMENTS ' L: '

' NOT REQUIRING PRIOR COMMISSION APPROVAL PURSUANT l- .

u 4 TO 10 CPR 50.59.' '

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TABLE OF CONTENTS Section Iitle. Page I NTR O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1.0 C H A NG E NOTICES (CN) . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 2.0 DOCUMENT CHANGE NOTICES (DCN) . . . . . . . . . . . . . . . . . 13 3.0 SETPOINT CHANGE REPORTS (SCR) .................. 14 4.0 SPECIAL TESTS .(T-TESTS) ........................15 5.0 P R OC E D U R ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

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INTRODUCTION This report is submitted to comply with the requirements of Part 50.59(b) Title 10, Code of Federal Regulations, as they apply to Fort St. Vrain Station, Unit No.1.

It includes the period January 23,1991 through January 22,1992.

The following defines certain activities conttined in this report:

Change Notice (CN) - A document containing installation, inspection and testing requirements, design background information, and design document updating requirements which specify the design control requirements applicable to a plant modification and authorizes changes to "as. built" plant design documentation.

Document Change Notice (DCN)- A document which authorizes a change to design documents. As a minimum, it contains a design input statement, a design analysis statement, a document update list and the document update information.

Setooint Change Report (SCR)- A document which authorizes setpoint changes which do not constitute an alteration to the design of the affected equipment.

T Tests - Special tests proposed and confected by Public Service Company of-Colorado.

The following is a list of standard abbreviations used at Fort St. Vrain:

AC - Alternating Current ACM - Alternate Cooling Method

_CRD - Control Rod Drive DCCF - Document Change Coordination Form EME. - Electro Motive Force EQ - Environmental Qualification FHM - Fuel Handling Machine FPPP - Fire Protection Program Plan ESAR - Final Safety Analysis Report :

ESY - Fort St. Vrain HELB . - High E.,ergy Line Break-ahi_h.-_.A-.-___.- _ . . . _ - _ -

HVAC -

licating, Ventilating, and Air Conditioning LCQ -

Limiting Condition for Operation ISESI -

Independent Spent Fuel Storage Installation LFJ1 -

Licensee Event Report LTA -

Low Temperature Adsorber hiCC -

Motor Control Center pal -

Piping and Instrument Drawing PCRV -

Prestressed Concrete Reactor Vessel EES - -

Plant Protective System ,

RERE -

Radiological Emergency Response Plan SQE -

System Operating Proce<!ure SR -

Surveillance Requ*rement The following defines terms used in safety evaluation summaries contained in this report:

Enhanced Ouality Items for which quality program requirements have been identified, but which are not safety related. This includes non safety related fire protection.(System 45 excluding safety related portions), portions of the Independent Spent Fuel Storage Installation (ISFSI), Secudy (System 78 excluding Gai-Tronics), and. packaging and transportation of radioactive materials.

Safety Related items Those plant systems, structures, equipment, and components which are identified by the FSAR, and as detailed and supplemented by applicable P&I, IB and IC diagrams;  ;

E and E-1203 schematic diagrams; the Cable Tab; SR-6-2 and SR-6-8 lists to include.

the following:

(' a) Class I per Table 1,4-1 of the FSAR.-

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4 b) Safe shutdown components per Table 1.4-2 of the FSAR.

c) Alternate Cooling Method (ACM) system.

EXCEPTION: The ACM system is exempt from requirements for seismic and environmental qualification, d) Interface circuits (IC) within the Environmental Qualification Program.

Safety Significant Change Changes to the facility, systems, components, or structures as described in the FSAR -

that may do any one (I) of the following:

a) - Affect their capability to prevent or mitigate the consequences of accidents--

described in the FSAR.

b) Could result in exposures to the plant personnel in excess of occupational limits.

Changes in the safety related systems which involve the addition, deletion, or repair of components, structures, equipment, or systems such that the original design intent is changed (i.e., changes in redundancy, performance characteristics, separation, circuitry logic, control, margins of safety, safe shutdown, accident analysis, or any _

change that would result in an unreviewed safety question'or require a Technical Specification change).

Unreviewed Safety Ouestion Any plant modification-or activity that is deemed to involve an unreviewed safety question as defined in 10 CFR 50.59.

a) The probability- of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased.

b) The possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created.l c)- The margin of safety as delimed in the basis for any Technical Specification is reduced.

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JULY 1992

' 10 CFR 50.59 ANNUAL REPORT

Background:

The followmg is a brief discussion of the changes, tests, and experiments affecting

) the Fon St. Vrai'i Station in the time period from January 23,1991, to January.22, J l' 42, that has not been previously reported to the Nuclear Regulatory Commission C m,C). It should be noted that many of the activities discussed in this report are-e directly relatui to the permanent shutdown condition of the FSV reactor, the defueling and decommitioning preparations of the plant.

1.0 CHANGE NOTICES (CN)

CN-2113 System 79frechnical Support Center CN-2213 was initiated to provide air monitoring capabilities in the Technical Support Center (TSC) utilizing an insta' 2. monitor, RIT 7937. Previously, room air in the TSC was monitored by . 'able unit which produced high noise levels and was unreliable. RIT-79 n be aligned to sample either outside air drr.wn in by the HVAC system or room air within the TSC. The TSC room air moniaring is accomplishe<1 by manual actmtion of solenoid

-valves in sample lines.

This activity was net safety related or mesy significant, aad did not involve a an unreviewed safety question.

CN-2854 and CN-2854A System 52/ Turbine Steam CN-2854 changed raanual valve V-5288, Bypass Flash Tank drain system valve, from a gate velve to a globe valve. .. The gate valve had been used as.

a throttling valve which resulted in excessivt intemal crosion. The globe valve was iuily qualified for the intended -service and better suited for the specific application. The modification was considered Enhanced Quality since

- V-5288 was r. component of Appendix R Fire Protection Shutdown /Cooldown

~ Tiain A. The change from a gate valve to r. globe valve was evaluated and determined to have no affect on the heat imoval rate of Train A.

This activity was not safety related or safety significard, and did not involve

. . an unreviewed safety question.

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CN-2956 and CN-295M System 23/ Helium Purification System CN-2956 mechanically and electrically removed Helium Purification Regeneration components from the System 23 regeneration pit. The modification allows use of the pit as a loading port ior defueling elements.

As previously reported, CN-2953 structurally modified the pit for use as a loading port for defueling elements.

CN-2956A removed associated control and power supply cables from panels located in the Control Room, and cable trays located throughout the plant.

This activity was class +.cJ safety related, but was not safety signifier and did not involve an unreviewed safety question.

CN-2965 and CN-2965A System 14/ Fuel Storage Facility CN-2965 provided an additional seismically qualified Reactor Isolation Valve '

(RIV) and a seismically qualified spacer to improve defueling operations. The spacer acts caly as a seismically qualified permanent storage pedestal. The new Rh and s,meer satisfy the requirements of Technical Specification LCO 4.7.2 in providing a seismically qualified location to store the Fuel Handling Machine when it is not attached to the Reactor Building werhead cranc.

CN-2965A was issued to qualify the new RIV as safety related for use over the reactor vessel and fuel storage wells, and provide a document update  ;

package. Th. RIV, as originally purchased, was not supplied vith safety related seals and/or the scal (s) failed v.ak ' tests as performed by -the manufacturer Qualified seals were placed on the RIV and successfully tes'xd. f This 3ctivity was classified safety related, but was c.ot safety significant and did not involve an unreviewed safety question. J.

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CN-3009 System 84/ Auxiliary Boiler and Heating System CN-3009 was written to replace the existing backup auxiliary boiler (S-840lS) with a lower steam generating capacity boiler. The new backup auxiliary boiler has a rating of less than 15,000 lbm/hr and 650 degrees F. Meeting these capaci' requirements ensured a harsh environment, as descriSed in 10 CFR 50.49, could not 1,e created in the event of a steam line rupture at Fort St. Vrain. During defueling of the FSV Reactor, a greater steam capacity wa not needed and environmental qualification in accordance with 10 CFR 50.49 was not acqui.ed.

The auxiliary boiler (S-84GI) was down rated by CN-3006 to meet the environmental qualification requirement; . .f 10 CFR 50.49. This activity was reported in the 10 CFR 50.59 report submitted to the NRC in 1991.

Replacement of the backup auxiliary boiler (S-840lS) was classified safety related, but was not s.fety significant and did not involve an unreviewed safety question.

CN-3024 System 36/ Independent Spent Fuel Storage Installa' ion System 83/ Communication System CN-3024 constructed a new asphalt road from the Reactor Building to the north-east corner of the Protected Area fence. The road provided a route a the Independent Spent Fuel Storage Installation (ISFSI) to allow transport of spent nuclear fuel to the ISFSI. Gates were installed in the inner isolation zone fence and the Protected Area fence. Communication cables were also installed for the in-plant system and the telephone system.

This act y g cg604rdi'agN4 LNYev,iyitv.gp Ewed d7;6 9.e not safety related or .ufety significant, and d T

CN-3037 System 11/ Prestressed Concrete Reactor Vessel:

System 93/ Controls and Instrumentation -

CN-3037 provided the design. engineering and analyses to support the addition of the Startup Channel-nuclear instrumer.t inputs to the plant-data logger;.

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This required the addition of an acceptable means ofisolation between safety-related signals and non-safety related equipment to prevent system degradation - i fer postulated failures, j The startup channels monitored ' core reactivity while the reactor was shut---

down and.being defueled. ' Since the data logger system'is non-safety related .  :.

and the startup channel inputs'_were safety related, the isolators prevented a.

starup channel failure resulting from any postulated failure in the' data logger :

system.-

This activity was classified safety related, but was'not safety significant and' did not involve an unreviewed safety question.

_C_N.-3038 .-

System 21/Primsry Coolant System:  !

-t CN-3.038 installed piping and-valves __between the Emergency Condensate - '

a Header and the Loop 2_ Helium; Circulator Beadng Water l Surge Tank,iT '

2105. This provided a permanent. water' supply to the tank during shutdown--

conditions and reducut weat and maintenance on the Bearing Water ~ Makeup : .

i Pumps. The modification closed out a temporary configuration which utilized 1

a flexible hose.

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did not involve an unreviewed safety-question. ;j mi ,

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C&3041 System 31/Feedwater and Condensate 4-CN-3041 positively isolated lines to 1,revent the entry of water and to assure a dry layup of the feedwater heaters. The feedwater heaters were equipment items not required to provide core cooling or support defueling operations, and for preservation purposes, were isolated, drained, and placed in dry layup. As is discussed in FSAR Section 14.14.2.7, reactor decay heat

[ removal could be accomplished with the PCRV Liner Cooling System, System 46, only.

This activity was not safety related or safety significant, and did not involve an unreviewed safety question.

_CN-3046 System 11/ Prestressed Ccacrete Reactor Vessel CN-3046 was initiated to obtain approximately six samples of PCRV concrete.

The samples were obtained by core drilling into the PCRV outer surface. .The core drills were a nominal diameter of two inches and a nominal depth

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(length) of four. inches. The ' samples were used ' to determine the-concentration ofimpurities in the concrete to validate assumptions made in the neutron activation analysis, in support of decommissioning FSV! The samples -

were taken from elevations corresponding to various elevations of the active .

core. PCRV wall thickness in this area .is approximately nine feet. The holes .

were refilleo with grout of comparable compressive strength to-that'of the original PCRV concrete.

This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

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.CN-3050 System 11/ Prestressed Concrete Reactor Vessel System 93/ Controls and Instrumentation CN-3050 disabled the three Wide Range Channels (WRCs) of the nuclear instruments by placing modified bistable cards into the Plant Protective System (PPS) modules in the Control Room. The modification also disabled the Six 1.inear Power Channels. The two Startup Channels were not adverseiy nTected by the modification.

The WRCs provided two. protective functions: (1) a Rod Withdrawal Prohibit, and (2) reactor scram. With the FSV reactor permanently shut-down, neutron flux well below the operating range of the WRCs, and with controls in place to prevent any approach to :riticality during defueling, the protective actions associated with the WRCs'were no longer necessary.--

This activity was classified safety related, but was not safety.significant and did not involve un umtviewed safety question.

CN-3051

' System 46/ Reactor Plant Cooling Water System

. System 75/ Turbine Building CN-3051 installed piping and valves to collect condensate from the heating.

coils of the Loop 1 and . Loop 2 System 46; surge tanks and deliver' that condensate to either the Miscellaneous Drain Tank,L T-7505, or the Drain Collector Taak, T-7507. - The normal drain path was to the _ Reactor Building _

Sump and was retained as a backup. l Benefits of the new drain IV included condensate recovery and enhnced water quality.- _

This activity was classified safety related., but was not safety significant and did not involve an unreviewed safety' question.

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CN-3052 Systen' 34/ Auxiliary Boiler and Heating System CN-3052 installed a new deacrator unit, complete with its own feedwater pumps, in the auxiliary boilers' feedwater system. Excessive boiler down time was believed to be due to high dissolved oxygen content in c cr.bmation with low feedwater temperature. The auxiliary boiler (s) will continue to be needed throughout the decommissioning phase of Fort St. Vrain.

This a:.ivity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

CN-3058 System 22/ Secondary. Coolant System i

CN-3058 removed the main steam ring header from steam generator module -

B-1 and portions of the main steam ring heade.t hom module B-1-2.'

Extensive cracking in the main steam ring headers of several of the 12 steam

- generator modules lead to the premature permanent shutdown of Fort St.

Vrain in August,1989.' The ring headers were delivered to EG&G Idaho, for thorough metallurgical examination. The effort was directed at determining the root cause of tne ring header failures in order to support the Department -

of Energy's interests in applying FSV experiences to any new HTGR design .

(i.e., technology transfer).

This activity was classiGed safety-relatM, but was not safety significant and did not involve an unreviewed safety _qustion.

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2.0 DOCUMENT CH ANGE NOTICES (DCN)

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SCR 91-002

System 33/ Water Treatment

[ Following permanent shutdown of FSV-in August,1989, water chemistry

limits were necesnrily evaluated and. changed. System-temperatures and contaminant levels were quite different for the shutdown' plant and the use of-different chemicals :was also considered.' The steam / water ; combined.

chemistry sample recorder and alarm unit R-33200 was reset to alarm at the ,

designated level and gcorder pointse based on the wate. chemistry unit's : i aralysis of shutdown er;ndiO:ns and available chemicals to control shutdown - r j

,#ut t hemistry parameters. -!

' Thi? abtivby was not safety relate'd or safety significant, and did not involve-an unreviewed safety question; g SCR 92-0Q1 4

q System 46/ Reactor Plant Cooling Water System '

b SCR 92-001 lewemi the setpoints of TSL-4637 and TSL4638. The.two -

F. temperature swiiches provide'a low-temperature alarm for each of the two- 'l

loops of ik PCRV .;ooling water circuits. . The alarms were set a few ' degrees

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(approxinstely 4 degrees F)/above-;t,ieLlow limit specified;in Technical!

Specificati6n (TS) LCO 4.2.15;, Amendment N.o. 83 to the FSV TS approved o lowering :the temperature Amits' of LCO 4.2. F-basedf on: the permanent shutdown and defuelin); Londition.of the reactor.i SCR 92-001 implemented l TS Amendment No. 83, _which lowered the allowable operating temperature.-

band by:15'F.

This activity _ was not safety related'or safety significant, and did not involve 'l-

- an unreviewed safety question. ,

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4.0 SPECI AL TESTS (T-TESTS)

I_421 System 13/ Fuel Handling Equipment

Purpose:

T-423 was written to support the Engineering Evaluation Test

. Series that was performed in 1989 prior to defueling. The: ,

purpose of the test was to exercise each of the Fuel Handling Machine (FHM) drive systems and the giapple head sensors while monitoring and recording the analog and digital signals with a PC-based data acquisition system (DAS). The recorded

. signals were to provide baseline data against which the planned

upgrades of the FHM could be judged for performance evaluation.

Results: The DAS has been used on numerous occasions and has been permanently installed by CN-3049. The DAS was also instrumental in making cther upgrades associated with the _FHM successful. This test proved the DAS, and the DAS enhanced the reliability of the FHM.

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This activity was classified safety related, but was not safety significant and did not involvc an unreviewed safety question.

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T-441

- System 13/ Fuel Handling Equipment

Purpose:

T-441 was written to assist in learning how to effectively use the FHM DAS It provided a means by which the process of determining h,w to make the physical connections to the FHM -

circuitry, implementing thcsc plans, and performing the system monitoring desired could be extended in time without having to remove the connections and reconnect each day. T-441 was used for initial assessment of a problem before pennanent connections were made by a design modification, r for general system monitoring.

Result:: As with T-423 (previous page), the most critical connection points associated with r-441 have been selected and installed via CN 3049 for further enhancement' of FHM operation and reliability.

This activity was not safe +y related or safety significant, and did not involve an unreviewed safety question.

I:8 14 4 System 11/ Prestressed Concrete Reactor Vessel l

Purpose:

T-454 was initiated to remotely obtain and record the radiation dose rates at various elevations' within a helium circulator-penetration. Dat.: collected wih assist personnel involved with

.the decommissioning.of Fort St.-Vrain in the assessment for-  : t manned access to penetration primary flanges.

Results: T-454 data.was not requi ed to supp<wt any conclusions,' but simply to allow habitability assessment. . Elevationr4 monitored were from approximately'4768 feet to approximately 4792 feet. Dose rates ranged from 0 mr/hr at approximately five feet into the penetration l:

'up to-72 mr/hr at approximately 24 feet intoLthe penetmtion and lower portion of the PCRV (i.e.', O cpm tc 1.6K - 2K cpm).

. Thia a::tivity was not safety related or safety significant, and did not involve -

an unreviewed ' safety question.

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- 1 T-456 System 11/ Prestressed Concrete Reactor Vessel

Purpose:

T-456 was initiated to remotely access a Loop 1 steam generator module hot reheat pipe to obtain and record radiation dose rates at various elevations within the niodule. Data 3 '

collected will assist - decommissioning personnel in their assessment- for mantied access to various steam generator components and elevations.

Results: T-456 data was not required to support any conclusions, ut simply -

to allow habitability assessment. Elevations monitored were from approximately C68 feet to approximately 4815 feet. Dose rates ranged from less than 5 mr/hr at the .irst measurement point _ up to approximately 300 mr/hr at 42 feet into the steam generator module hot reheat pipe (i.e.,10 to 30 cpm to approximately 7.5K cpm).

This activity was classifkd safety related, but was not safety significant and did not involve an unreviewed safety question.

O I-457 System 21/ Primary Coolant System - .

Purpose:

T-457 was initiated to help determine operability of C-2103, the helium circulator 1C brake system, and potential cause(sl of the brake problem. The test was divided into two parts.

Part I applied 1000 psig to the brake supply line which, with -

no motive power to drive the circulator, would normally stop it from' self-turhining. _ Part 2 of the; test actuated - the pressurizing line several times and monitored the pressure drop -

in the supply bottle.

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}; Results: Part I did not stop the circulator, indicatmg no brake pad' contact.

with the circulator shaft. Part 2 indicated that thbe (3) actuationst

- reduced -bottle pressure 20 psi. - When compared to actuating; requirements for another helium circulator, the Ldifference was negligible and indicated no 1 internal blockage :of; the brake pressurizing line existed.1 At that time it was detern'ined that only circulator removal and visual inspection of the brakes ceuld resolve -

brake status. >

This actMty was classified safety related, but was not safety significant and i did not involve an unreviewed safety question. ,

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T-458 System 93/ Controls and Instrumentation

Purpose:

T-458 was written to determine the effects of permanent de-energization of Wide Range Ch.anels (WRCs) III, IV, and V.

During reactor operations prior to beginning defueling, the WRCs provided two functions: (1) Rod Withdrawal Prohibit, and (2) reactor scram. The interlock / trip functions related to the WRCs were not required during defueling:, and only served to prclong defueling by 'mnecessay actuations of the Plant Protective System.

Results: The test was successfully performed with all requirements met and no abnormal conditions observed. CN-3050 (this report) permanently disabled the WRCs.

This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

T-459 System 13/ Fuel Mandling Equipment

Purpose:

T-459 was written to gather preliminary data concerning loose surface contamination that would be created during transfer of

. irradiated fuel elements into an Independent Spent Fuel Storage Instal!ation (ISFSI) fuel storage canister. The contributin;;

components were the Reactor Isolation Valves' (RIVs) gates during opening and closing, and the fuel elements themselves.

Based on the design of the ISFSI, significaat loose surface contamination is not acceptable on a fuel storage canister during storage.

Results: Although contamination levels were less than anticipatcd, they were still higher than allowa' ale limits specified in the ISFSI Technical-Specifications. The test showed . that the RIV used did not contribute any detectable amount of contamination during normal operations. However, loose surface contamination on the top of a

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fuel storage canister from actual fuel element ~ movements was v concern. T-459 showed that further testing was needed and ,

decontami.1ation of the ISFSI fuel storage canisters used for the test would not be difficult.

This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

W T-460 System 13/ Fuel Handling Equipment M

Purpose:

T-460 was : written to determine- the best method (s) of mitigating loose surface contamination caused by handling irradiated fuel elements ' (see T-459 - previout page) in preparation for defueling the FSV reactor to the ISFSI.

Results: T-460 utilized several mitigation methods in the contrcl of loose.

surface contamination. The best results were obtained using the maslin wrapped ISFSI spacer ring test. The ISFSI spacer is used -

between a RIV and the fuel storage canister. - This leaves a gap of about 0.080 of an inch between the outside diameter of the_ spacer ring and the inside diameter of the lower portion of the RIV seal .

ring. During fuel transfer, loose surface contamination could pass: l through the. gap and contaminate the-outer sutface o." the fuel storage canister. The maslin wrap was effective in mitigating this -

source of contamination to the outside of the fuel storage canister. -

This activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

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T-461 System 46/ Reactor Plant Cooling Water System

Purpose:

T-461 was initiated to determine. the effects'of a reduced water -

- flow in one loop of System 46 with the other loop shut down.

Flows were reduced to appioximately 25% of normal in the PCRV lower barrel, bottom head, and core support floor areas while the upper barrel and top head flows were reduced to (

approximately 75% of normal. During defueling of the FSV reactor, decay heat generation was low and in order to maintaie System 46 water temperature within Technical Specification LCO 4.2.15 limits (i.e.,100 to 120 degrees F) auxiliary heating had to be provided. -The amount of auxiliary heat required we.s : very - small and, therefore, not L easily-compatible with the heat source; The intent,was to be able to shut down -the auxiliary heat source for a financial and manpower savings.

Results: Typically, PCRV liner. coolir.g water temperatures rose apptrr..mately: one. degree as' measured by each subheader outlet:

the .ocouple. A-few tubes in the upper barrel and top head area

.o. approximately four to five degrees'F above subheader inlet water temperature. In summary:

1. Changes in System 46 flow and temperature did not appear to have a significant affect on liner cooling tube j differential temperatures at the flow rates used in this '

test.

2. There was some heatup of the liner cooling tubes in the 'i

- top head and upper barrel areas.

3. Reactor coolant flow and core temperatures had little,-

if any effect on liner cooling tube temperatures.

4. Reactor building ambient temperature had a significant .

effect on operating liner cooling tube temperatures.

This activity was classified safety related, but was not safety significant andL did not involve an unreviewed safety question.

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T-462 i

System 84/ Auxiliary Boiler and Heating System System 92/ Accessory Electrical Equipment P

Purpose:

T-462 was written and. performed to verify that three underground storage tanks at Fort St. Vrain were not leaking and did not present any potential for product. (diesel fuel) release to the environmeni. The test utilized a vacuum technique whereby a slight vacuum was pulled on each tank tested. Sensitive instrumentation w:a used to inonitor for bubble formation (indicating air ingress), or a change in water. .

level (indicating water ingress). L

'4 Results: - Testing indicated no apparent leakage in any of the t tee tanks.

This activity was classified safety related, but was net safety significant and i did not involve an unreviewed safety question.

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P 5.0 PROCEDIURES DSOP 84-02. Inuc_2 System 84/ Auxiliary Boiler and Heating Following completion of CN-3052, which installed a new deaerator unit for the auxiliary boilers,-DSOP 84-02 was revised. This procedure provides detailed instructions for startup, operation, and shutdown of the auxiliary boilers. DSOP 84-02 also identifies appropriate precautions necessary to limit operations to one beiler at a- time, and- steam flow and temperature requirements '

within the limits established for the FSV environmental qualification program (10 CFR 50.49). Refer to CN-3009 and CN-3052 for addidonal information related to the auxiliary boilers.

Ti a activity was classified safety related, but was not safety significant and did not involve an unreviewed safety question.

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