ML20054H142

Forwards Safety Evaluation Supporting Work Performed to Cap Leak in Sys 46 (Liner Cooling Sys) & to Ensure Functional Capability of Reserve Shutdown Sys.Reactor Vessel Repressurized

ML20054H142
Person / Time
Site:	Fort Saint Vrain
Issue date:	05/07/1982
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Kuzmycz G NRC
References
P-82135, TAC-67896, NUDOCS 8206230024
Download: ML20054H142 (21)

Text

. s foM public Servlee Company of Cohndo VM 16805 WCR 19 1/2, Platteville, Colorado 80651 May 7, 1982 Fort St. Vrain Unit #1 P-82135 Mr. George Kuzmycz U. S. Nuclear Regulatory Commission 7920 Norfolk Ave.

Bethesda, MD 20034

SUBJECT:

Fort St. Vrain Unit No. 1 Operational Conditons

Dear Mr. Kuzmycz:

In the above referenced letter we outlined our operational problems concerning System 46 (liner cooling system) and the need to investigate the functional capability of the reserve shutdown system.

This letter is being forwarded to update the status of these systems.

System 46 test work was completed and the leak in System 46 was isolated to one core support floor tube. This tube (F4T21) is in Loop I and is on the top plate of core support floor with the inlet and outlet of the tube being on either side of center weld for the two halves of the support floor liner. (See attached sketch). We have evaluated the leak with reference to the FSAR commitments, and we have effected the necessary repairs. The tube has been isolated from System 46 and double capped on both the inlet and outlet side with the interspace between the double caps being pressurized with purified helium. Our safety evaluation which supports the work performed is attached for your information.

With reference to the reserve shutdown system, the control rod drive assembly from Region 19 was removed and replaced. An actual release test of boron balls from this control rod drive hopper was successfully conducted in the hot service facility. Since the control rod drive assembly in Region 19 appeared to be the one assembly most seriously affected by high moisture levels in the primary coolant, we feel the successful test on the control rod drive hopper is indicative that the reserve shutdown system is functional and the conditions set forth by Amendment 13 have been satisfied.

8206230024 820507 f PDR ADOCK 05000267 P PDR

-g-The reactor vessel has been repressurized and we have returned to power operation (approximately 10*o power). We intend to continue with the rise-to power program.

Very truly yours,

/h~7Yiku Don W. Warembourg

. Manager, Nuclear Production Fort St. Vrain Nuclear Generating Station DW/skd cc: John T. Collins (Region IV) -

PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCl. EAR GENERATING STATICN 6 CR / SCR / PC / TR NO. IS SAFETY EVALUATION PAGE I*E 8J TVPE:

E CH ANGE NCTICE 'CN) OVERALL CCHANGE NOTICE (CN) SU8MITTAI. CPROCEDURE CHANGE (FSAR)

C TEMPORARY CONFIGURATICN CSETPCINT CHANGE CTEST RECUEST CLASSIFICATION: ARE THE SYSTEMIS) EQUIPMENT OR STRUCTURES INVOLVED OR DOES THE ACTIVITY AFFECT:

CLASS I E YES C NO ENGINEERED SAFEGUARD C YES E NO '

SAFE SHUTDCWN E YES C NO PLANT PROTECTIVE SYSTEM C YES E NO SAFETY RELATED E YES C NO SECURITY SYSTEM C YES E NO 1

REMARKS EVALilATIGR uSE ADDITIONA4. SHEET 5 IF MEGuaHEO.

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I 1 OETERMINE WHETHER CA NOT THE ACTIVITY INVOLVEQ IS 10ENTIFIED IN THE F S A R CR TECH SPEC.

PPUCA8LE SEC9CNS 9EVIEWED The des 19n features and coeratino orocedures cro- i l uST mEded vi Tor dealing with lea,Kage or tne core succort floor casing and a PCRV 6iner cooling system tuce are ident1rleo and evaluated in .-5AR 5ections '

3.3.2.2, 5.9.2.4, 11.1, Ouestions 3.2 and 11.7, 0.2.2 and Tecnnical ..i Saecification LCO 4.2.14 All FSAR Sections and Technical Soecifications d1C" '

were reviewed. LCO 4.8.1 is also certinent to this activity. L% w. r JC )

% f r.a t & Lee 4J./A b & '%c.L % e sw n lla url%O 4 Er w '

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j 2 CGES THE ACTIVITY RECulRE THAT CHANGE (Si SE MACE TO THE F $ A R CR TECH SPEC 7 4 C YES 3 NO l UST SECTIONS TO SE CHANGED AND THE CHANGES TO SE MACE. _

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l 3 OETERMINE AMETHER CR NOT THE ACTIVITY INVOLVED IS AN UNREVIEWED SAFETY CUESTICN UTIUZING THE FOLLCWING GulOEUNES.

(A) HAS THE PRCBASIUTY CF CCCURRENCE CR THE CCNSEQUENCES OF AN AC01 CENT CR MALFUNCTICN CF ECulPMENT IMPORTANT TO SAFETY PREVICUSLY EVALUATED IN THE F S A R SEEN INCREASED 7 C YES G NO STATE SASIS: Ihe orobabilitV of OCOUr'*ence and the cons @0uencas of leakage of the core supoort floor casino and a 2CRV liner coolino svstam tube have been thorouchly evaluated, with acceotable casults. in the :SAR Sections cited under 1, above. This chance notice imolenents the remedies described in the FSAR and thus does not chance any of the evaluations cra-sented therein. (See attachment) 1 (B) HAS THE POSSIBluTY OF AN ACCIDENT CR MAL.8UNCTICN CF A CIFFERENT TYPE *HAN ANY EVALUATICN PREVIOUSLY IN THE F S A R 8EEN CREATED 7 C YES 3 NO STATE 3 ASIS: 'he tVoe Of malfunctions associated with this activity are thorouchiv evaluated in tne F5AR. This cnance notice imolements remedies described theroin, thus, does not create any new tyoe of accident or malfunction.

tC) HAS WE MARGIN OF SAFETY. AS oEFINED IN THE 3ASl3 FOR ANY TECHNICAL SPECIFICATICN OR IN THE FSAR 3EEN REQUCED? C YES 3 NO STATE SASIS: Dis chance confo Ms to conditions allowed hy the FSAR for full ocwer coeration of the clant and  ;

coes not excead anv coerstino condition cemitted under exis-ino tachnical J soecifications. l 3CES *HE AC 1VITY APPEAR TO: INVCLVE AN UNREVIEWED SAFETY CUESTICN [ YES 3 NO

/ 3E SAFE *Y SIGNIFICANT E VES *10

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S_afety Evaluatio,n I. Background Increasing pressure has been observed in the PCRV Liner '

Cooling System (Sys. 46) surge tank during power operation.

Radiochemical analysis of the system fluid revealed the j presence of fission products. The cause of these conditions i has been investigated and it has been concluded that a small t

flow path exists between the primary coolant and liner .

cooling tube F4T21 located on the concrete side of the top surface of the core support floor. The rate of primary coolant leakage has been conservatively calculated to be 1.3

) pounds per hour based on the observed rate of pressure increase in the Loop 1 surge tank.

XI. Discussion i

The possibility of core support floor casing leakage i and/or liner cooling tube leaks has been provided for in the design of the plant and has been thoroughly evaluated relative to operability and safety of the plant.

The space inside the core support floor is vented and drained to maintain the pressure in the core support floor i

at 60 psia or less. The vent and drain cavities were fomed during the concreting of the core support floor and are connected to tubes pemanently cast in the concrete and routed through the floor, down the core support columns and

to the radioactive gas waste system. (FSAR Sections 3.3.2.2 and 5.9.2.4)

The radioactive gas waste system (FSAR Section 11.1) is 4 designed to collect, monitor and control the release of radioactive gases in confonnance with 10 CFR 20. provision is made in this system to collect any primary coolant leaking into the core support floor (FSAR Section 11.1.2.3).

The system has sufficient capacity to permit processing a continuous leak of 14 lbs/hr of primary coolant from that source in addition to the maximtsn gas stream from other expected sources, and in the absence of other flows, could handle as much as 53 lbs/hr. The actual flow is monitored and recorded in the control room by a flow recorder (FR-6351). (FSAR Section 11.1.3.4, Reference Design 50-63).

The holdup capacity of the radioactive gas waste system is provided by two 700 ft 3 surge tanks, capable of operating at 450 psig. Wi th a 14 lb/hr leak, the tanks provide a total of 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> hold up capacity. Beyond 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />, the leakage would have to be vented from the plant through the reactor plant ventilation system filters. After the first tank is filled, its decayed activity would be released while the other tank is filling, and so on. However, this mode of

CN O operation is not contemplated and, in any caso, the quantity BY N Md of makeup helium available would not be sufficient to PAGF 7 of N~

continue plant operation for more than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> with a leak rate of this magnitude. The normal procedure would be to "* 1'"Y[II shutdown the reactor in an orderly manner and depressurize the main loop if such a leak should develop. Even if the reactor is assumed to remain at full power for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and the leakage during the 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> between 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> is vented directly from the plant the whole body gama (WBG) dose at the exclusion area boundary would be less than 5 millirem (worst short tenn dilution, design primary coolant activity, 5 m/s wind speed, elevated release with '

downwash) (FSAR Question and Answer 11.7). At current primary coolant levels the dose would be a small fraction of that value or coversely, a leak rate considerably greater could be allowed under 10 CFR 20 limitations. Considering that current equilibrium primary coolant activity at full

power is about 1/60th of the design activity level; the allowable leak rate could theoretically be increased by the inverse proportion or 60 times without exceeding 10 CFR 20 limits.

The possibility of a core support floor casing leak occurring in close proximity to a liner cooling tube leak and leading to helium bypassing the vent system and leaking into the affected cooling water loop has also been considered in the FSAR (Question and Answer 3.2). That evaluation was based on the leakage resulting in high pressure ( 140 psig) in that loop and authomatic isolation of all 13 subheaders in the loop. The portion of the liner cooling system between the inlet and outlet subheader block valves is designed for PCRV reference pressure (845 psig) and a safety valve in each loop return header protects the rest of the system fran over pressure by relieving water and heltum to the gas wasta vacuum tank, where separation l occurs.

III. Descriotion of the Modification The proposed modi fication consists of plugging the affected liner cooling tube supply and return (F4T215 and F4T21R) at locations near the outer surface of the PCRV.

The plug design utilizes two separate caps to provide a high degree of integrity. The plug design and anaylsis is based on ASME Code,Section III (1965 Edition) for Class A vessels. Purified helium pressurization is provided to the space between the caps from the Helium Circulator Penetration Interspace Pressurization Supply for leak

! moni toring in the same manner as provided for PCRV interspaces. The existing flow alann, set at 16.7f/hr, on the supply header will alann in the control room if plug leakage exceeds that amount. The purified helium supply piping and valves are 1-M-2 Class 032 - Purified Helium (845 psig and 750 degrees F) and the arrangement and support is in accordance with the requirements for Seismic Class I.

In addi tion, the ends of the plugged tubes extending beyond the PCRV concrete will be provided with missile protection.

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The schematic of the proposed modification is shown in CN / 4%

Figure 1. By ## Jd as PAGE 4 cE YJ IV. Evaluation ,m m.a.-

A. PCRV Liner Cooling System To protect the core support floor from damaging heat, the outer surfaces of the structure are lined with a themal l barrier. The limited quantity of heat which passes through the thennal barrier is removed by a system of cooling tubes welded to the concrete side of the steel plate casing. .

During normal reactor operation the thennal barrier and liner cooling system together maintain the casing a mean temperature of 130 degrees F, and :naintain the adjacent concrete wi thin 150 degrees F, except in identified localized areas. The liner cooling system consists of two independent loops. Al ternate tubes are connected to different loops and the water flows in opposite directions.

The design heat removal capaci ty of each loop is approximately twice the total expected heat flux.

The loss of one cooling tube in any set of six on the core support floor top casing results in acceptable liner and concrete temperatures within the limiting temperature of 250 degrees F and reactor operation at full power is allowed. (FSAR Section 5.9.2.4). The FSAR (Section 5.9.2.4) also states that if a tube cannot be restored to service , the adjacent tubes will be monitored continuously and an alann will be set to indicate to the operator if blockage of an adjacent tube has occurred. If the blockage is confirmed and can not be rectified within a few hours the plant would have to be shutdown under existing Technical Specifications (LCO 4.2.14). These requirements will be implemented by plant administrative procedures.

Furthermore, the ability of.the core support floor to support the core during a postulated loss of forced circulation (LOFC) accident compounded by complete loss of cooling water to the core support floor has been comfirmed (FSAR Appendix 0.2.2.3).

Thus, the probability of occurence and consequences of plugging a single PCRV cooling tube have been described in the FSAR and have been found to be acceptable. This notification does not affect the previous analyses in that regard.

Since the possibility of having a small number of PCRV liner cooling tubes out of service has been previously described in the FSAR, (Section 5.9.2.4) this modification does not create the possibility of a new type of accident.

The FSAR and Technical Specification margins of safety applicable to this system consist of redundant loops, excess design heat removal capaci ty and the prohibi tion of operation if two adjacent tubes are out of service. The plugging of a single tube does not reduce these safety ma rgins.

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The radioactive gas waste system is designed as PAGE < af JJ described in the Discussion above to collect, monitor and ' " "o. m m m control the release of primary coolant leakage through the - ~ ~ - ' - -

core support floor casing. Since the observed flow rate and primary cool:nt activity are well within the design criteria of the gas waste system, there are no new safety considerations with recard to this system.

The probability and consequences of primary coolant leakage through the core support floor casing have been .

analyzed and found to be acceptable. This modification has no effect on that aspect of plant operation.

Since the possibility of leakage through the core support floor casing has been described in the FSAR (Sections 11.1, 3.2 and Questions and Answers 3.2 and 11.7),

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the occurrence of the described condition does not create

! the possibility of any new type of accident or malfunction.

Technical Specification LCO 4.8.1 defines the conditions necessary to keep radioactive releases from the plant vent and liquid waste system as low as practicable and in any event within the limits of 10 CFR 20 and, thus, establishes the margins of safety relative to operation of the Gas Waste System. This modification does not propose operation outside the limitations imposed by that Technical Specification and, thus, does not reduce the margin of safety as defined therein.

C. Core Suocort Floor Vent and Orain System The core support floor vent and drain system is designed, as described in the Discussion above, to collect leakage into the interior of the core support floor structure and route it to the gas waste system. The existence of such leakage within the operational capability of the gas waste system to process it has no increased or new safety implications.

D. Containment The proposed modification does not affect the existence of concurrent leaks in both the core support floor casing and PCRV liner cooling tube F4T21. Therefore it is necessary to consider the possibility that the plugged ends of the tube , which extend a short distance (10 inches) beyond the PCRV concrete into the Reactor building atmosphere, may contain primary coolant. Although the plugged tube ends are well shielded fran missiles by the protective covers , a redundant pressure boundary is not provided. This is acceptable on the following basis:

helium sampling lines containing primary coolant do not have a redundant pressure boundary. The FSAR identifies the rupture of one of the 1/4 inch helium sampling lines in Section 14.7, which results in a primary coolant leak rate of 18 lbs per hour. Leak rate test data taken on the leakage flow path from the PCRV through the core support floor casing into the PCRV liner cooling tube F4T21 u _ ___

indicates tha leak rata to be much less than 18 lbs per hour. Several leak rate tests have detennined actual SY f 4 J.M leakage to be 1.3 lbs per hour. Conservative calculations PAGE 7 a[8J indicate that at 100% power, with normal operating PCRV = % me.m.

pressure (700 psia) leakage flow rate wuld be less than 3.0

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1 lbs per hour, assuming a complete shear of PCRV liner cooling tube F4T21 at the PCRV exterior surface. The consequences of a single failure at either end of the plugged PCRV liner cooling tube, postulated to be a gross rupture or shear, would thus not exceed the consequences of a ruptu re of a primary coolant helium sampling line identified in FSAR Section 14.7. The probability of leakage -

or rupture of the PCRV liner cooling tube F4T21 plugged pipe stubs external to the PCRV is less than the probability of rupture of a primary coolant helium sampling line. The probability of failure of a primary coolant sampling line, ,

though low, is somewhat greatar due to its greater length.

No new accidents relating to containment are created by this modification. Primary coolant leaks are discussed in l FSAR Sections 14.7 and 14.8. The maximum credible accident, shear of the 2 inch diameter helium purification regeneration line coupled with failure of an interlocked valve to close followed by failure of operator to close the valve upstream of the leak, results in an initial primary coolant leak rate of 3.4 lbs per second (12,240 lbs per hour). This accident results in doses orders of magnitude below the guidelines of 10 CFR 100 even assuming design level primary coolant circulating activity. Postulated shear of a plugged stub of PCRV liner cooling tube F4T21, resulting in a primary coolant leak rate of less than 18 lbs per hour, does not prevent shutdown and cooldown of the

, reactor, nor would it result in an unacceptable dose rate at l the exclusion area boundary.

l ihe margin of safety associated with containment is not reduced since any primary coolant leakage resulting from a failure of PCRV liner cooling tube F4T21, external to the PCRV will not exceed leakage resulting frcm a rupture of a primary coolant helium sampling line (18 lbs per hour). It is very unlikely that the crack in the core support floor casing, which serves to limit any primary coolant leakage to PCRV liner cooling tube F4T21, will increase in size. The core support floor casing was constructed from the same material used in the PCRV cavity liner. FSAR Section 5.7.2.2 states "The liner material s have an initial nil ,

ductility transition temperature (NOT) of at least minus 60 degrees F wnich is 160 degrees F below the minimum operating temperature. The 160 degrees F value allows for a shift in the NOT of 100 degrees F and provides for operation above the fracture transition elastic temperature (FTE NDT + 60 degrees F). This provision will ensure that crack t

propagation in the liner at any tensile membrane stress up l to yield stress would be incredible ...

Based on the extemely low probability of crack propagation in the core support floor casing, it is highly unlikely the primary coolant helium leak rate through this casing crack, into PCRV liner cooling tube F4T21, to atmosphere (assuming gross failure of a plug and cap assembly or F4T21 tube wall external to the PCRV) would exceed 18 lbs per hour in an

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accidant. Thus tha double closure, plug and tube cap is BYygM j adequate to insure safety in the short term. In the long tem, a set of double isolation valves will be installed on PAGE l d 88 both ends of tube F4T21, with associated piping. This will 'm =. m. -

provide the capability for monitoring leakage flow rate from the core support floor casing crack into the PCRV liner cooling tube F4T21 to insure the primary coolant leakage flow rate (which would exist if tube F4T21 failed by gross rupture or shear) remains less than 18 lbs per hour. This installation will pemit leak monitoring to identify any long tem degradation of the core support floor casing leak l and liner cooling tube F4T21 pressure boundary. ,

10 CFR 50.55a, Footnote 2, provides an exemption from Code requirements for components connected to the primary coolant pressure boundary provided that in the event of postulated failure of the component during nomal operation, the reactor can be shutdown and cooled down in an orderly manner. According to General Atomic Company, the current large HTGR design criteria makes use of this exemption for reactor coch:.c lines one square inch in flow area and i smaller and only requires single pressure boundary and isolation in such cases. Analysis of Fort S t. Vrain has shown that the reactor can be shut down and cooled down in an orderly manner with continuous discharge of primary coolant through a one inch pipe.

E. Core Suocort Floor Integrity Leakage of primary coolant through the core support floor casing, within the limit described in the FSAR, i.e.,

up to 14 lbs/hr, has no significant damaging effect on the core support floor. As previously discussed, the vent system is designed for just such an eventuality and full power reactor operation is permitted without increasing the probabili ty or consequences of any accidents, creating different types of accidents or reducing the margin of safe ty.

Likewise, the plugging of a single liner cooling tube has been evaluated in the FSAR and found to be acceptable for full power operation without increasing the probability or consequences of accidents, creating di f ferent types of accidents or reducing the margin of safety. In fact, FSAR analysis of a loss of Forced Circulation (LOFC) accident compounded by the complete loss of water cooling to the core support floor confims the ability of the core support floor l to support the core, wnich is its only safety function.

(FSARSection0.2.2)

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