ML20042B323
| ML20042B323 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 03/18/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20042B313 | List: |
| References | |
| NUDOCS 8203250216 | |
| Download: ML20042B323 (10) | |
Text
.-_ ______ - _-______ ___ _______
f erg)q UNITED STATES 2
o NUCLEAR REGULATORY COMMISSION 3
j WASHINGTON, D. C. 20555
FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 1.0 Introduction Fort St. Vrain, a 330 MWe high temperature gas-cooled reactor (HTGR),
was designed by the General Atomic Company (GAC) and is operated by the Public Service Company of Colorado (PSCo) near Platteville, Colorado.
PSCo was i.ssued a contruction permit on September 17, 1968 and submitted the Final Safety Analysis Report as Amendmant 14 to its application for a construction permit and operating license for the Fort St. Vrain Nuclear Generating Station (FSV) on November 4, 1969.. A Safety Evaluation Report dated January 20, 1972 and a first supplement which was issued on June 12, 1973 concluded that FSV can be operated, as proposed, at power levels up to 842 MWt, full 100 percent power, without endangering the health and safety of the public.
l 2.0 STEAM GENERATOR PENETRATION INTERSPACES
2.1 INTRODUCTION
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Since mid-1980, the Fort St. Vrain plant has been experiencing purified I
helium leaks in the Loop 2 steam generator penetration interspace system.
I The leakage path was internal to the Loop 2 penetrations and occurred between the purified helium interspace and the cold reheat steam piping internal to the penetration.
In October 1981, during rise-to-power test-ing, the -leak rate increased and in order to complete the rise-to-power j
testing, temporary relief was granted to operate steam generator module B-2-3 interspace at below primary coolant pressure but above cold reheat
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steam' pressure (See Amendment No. 24). The purified helium header pioin modified, and the rise-to-power testing successfully completed to 100% ~ g was power in November 1981, with module B-2-3 interspace maintained at slightly greater than cold reheat steam pressure. Additional testing, which was somewhat quantitative, indicated that the largest penetration i
interspace gas leak to the reheat steam system was in module B-2-3, but i
that smaller leaks existed in modules B-2-2 and B-2-6.
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8203250216 820318 PDR ADOCK 05000267 p
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2.2 DISCUSSION 2.2.1 Modified System Descriotion The PCRV auxiliary piping system for the steam generator penetration interspaces has been modified as shown in Figure 1.
Pressure control valves PCV-ll379 and PCV-ll380 have been added on the inlet side for each steam generator loop. This permits the penetration interspaces for either or both steam generator loops to be maintained above cold reheat steam pressure or above primary coolant pressure as the situation requires.
Sample lines have been added to each module penetration interspace, and in turn routed to activity monitors 2263 for Loop 1 and 2264 for Loop 2.
Each sample line contains an individual flow control / block valve. Normally, these activity monitors will sample all penetration interspaces that are being maintained below primary coolant pressure, but above cold reheat steam pressure. However, the interspace gas from each penetration can be separately directed to the activity monitor for diagnostic testing. The discharge flow from the activity monitors will be routed to the gas waste system. All other features of the PCRV auxiliary piping system for the steam generator penetration interspaces remain unchanged by this modifica-tion.
2.2.2 Potential Environmental Imoact and Limitations When either or both steam generator loop penetration interspaces are bei.ng operated below primary coolant pressure, but above cold reheat steam pres-sure, there exists a potential for primarv coolant leakage across the pri-mary closure into the penetration interspace. Assuming an interspace gas leak to the reheat steam system (which would be the reason for operating the interspaces below primary coolant pressure), the potential exists for primarily noncondensible noble gases to be removed by the condenser air ejector and discharged at the plant stack. This potential requires primary closures for the steam generator penetrations to be leaking.
If there is
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no primary closure le'akage and none are currently known to exist, then there would be no off-site environmental impact.
Pressurized water reactors (PWRs) commonly experience primary coolant leakage at the steam generators. Primarily noble gases in the secondary coolant are removed at the condenser air ejector and discharged to the environment from the plant stack.
In a survey of five operating plants, NUREG-0017, " Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code), "
reported that the monthly ave age primary / secondary leakage varied from 0 to 830 pounds per day with an average for the five plants of 98 pounds per day over a 3-1/2 year period. The PWR-GALE Code specifies that 100 l
pounds per day for primary / secondary leakage be used for this one effluent path in demonstrating compliance (in PSARs and FSARs) to 10 CFR 50, Appendix I criteria. However, 10 CFR 50, Apoendix I was prepared specifically for-LWRs and has not been applied to Fort St. Vrain.
Instead, the Fort St.
Vrain Technical Specification in the Basis for LCO 4.8.1, Radioactive Gaseous Effluents, states the design objective for the plant's radioactive gas releases is 4160 curies per year. This design objective, in conjunc-tion with Fort St. Vrain meteorology, would also satisfy 10 CFR 50, Appendix I criteria as applied to LWRs.
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It is proposed in the amended LC0 4.2.9 to limit the release of noble gases vis this one effluent pathway to 10% of the Fort St. Vrain plant design' objective of 4160 curies per year. This equates to 1.4 curies per day based upon a plant capacity factor of 0.8 (292 days operation per l
year). Based upon plant operation to date, all gaseous releases, including th'e upper limit proposed for the condenser air ejector pathway, will remain well within the plant design objective of 4160 curies per year. The only detectable gaseous effluent path is the gas waste system and releases have been less than 400 curies per year.
2.2.2.1 Allowable Primary Closure Leakage The quantity of primary closure leakage consistent with the above release limit, which in turn escapes to the reheat steam system and is ultimately discharg'ed to the environment by the condenser air ejector, will vary i
l directly with the magnitude of the primary coolant activity. PSC calcula-tions were performed using " design" circulating activity for 105% of rated i
power.
Following are the bases used by PSC to calculate the primary closure leak rate consistent with the release of 1.4 curies per day, and the resultant whole body gamma dose at the exclusion area boundary.
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(1) Annual average dilution factor at the exclusion area boundary:
1.37 x 10-6 sec/m3 (Reference FSAR, Section 14:12.8.3).
(2) Annual average wind speed: 3.67 meters per second (Reference FSAR, Section 14.12. 8. 3 ) ".
(3) Primary coolant activities: 879 MW(t) design values (Reference Table 3.7-1 of FSAR).
(4) Primary circuit helium inventory:
700 pounds of helium (Reference l
FSAR, Section 4.2.1).
(5) Average gamma decay energies and half lives: Table 1,.GA-Al2499 (LTR-4).
(6)
Oose rate equation: DR i = 0.25 E X (Reference Regulatory Guide 1.3). 3t j and DRaj = 0.23 5gjXi y
(7) Distance to the exclusion area boundary: 590 meters (Reference FSAR, Table 14.12-1).
(8) Plant capacity = 0.8 (292 days operation per year).
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. The noble gas nuclides and their decay products for " design" circulating.
1 activity are shown in Table 1.
The total primary coolant activity for noble gases is 17014 caries. With a primary coolant helium inventory of 7370 pounds each pound of helium contains 2.3 curies. To maintain the offsite releases -
due to this one effluent pathway within the proposed limit of 1.4 curies per day, up to 0.61 pounds per day could leak across the primary closure into the reheat steam and ultimately be exhausted out the condenser air ejector. Correspondingly, at current circulating activity, which is about 1% of " design," about 61 pounds per day could be released via this same pathway. The resultant annual exclusion area boundary whole body gamma dose and beta skin dose, based upon the release of 1.4 curies per i
day and a plant capacity factor of 0.8, is 0.14 millirad and 0.06 millirad, respectively.
The thyroid dose consis' tent with the leakage of 0.61 pounds per day primary i
coclint at " design" activity and a plant capacity fgetor of 0.8 was also l
investigated. The annual thyroid dose is 5.2 x 10-o millirads.
It can be seen from the preceding calculations that the offsite doses for this one effluent pathway are small, and that a much more stringent requirement on leak tightness of the steam generator primary closures is imposed. With the absence of this potential effluent pathway, the limit on steam generator primary closure leakage for each loop is 400 pounds per day at a differential pressure of 10 psi. With the existence of the effluent pathway, the leakage limit is reduced to not exceed about 60 pounds per day with a differential pressure across the primary closure which will always be at least 50 psi. Therefore, the accident consequences of a secondary closure failure as discussed in the Basis for LCO 4.2.9 are not compromised.
2.2.2.2 Secondary Coolant Activity Fort St. Vrain Technical Specification LCO 4.3.8 limits secondary coolant activity level to 0.009 microcuries per cubic centimeter of I-131 and 6.8 microcuries per cubic centimeter of tritium.
The basis for these is to limit offsite doses in the event of an accident involving loss of outside power, main turbine trip, and failure of one diesel generator to start (FSAR, Section 10.3.2).
In that event, about 52,000 gallons of secondary coolant would be vented to the atmosphere as steam.
Calculations performed by PSC indicate that these secondary coolant activity limits will not be approached for the limiting case of 0.61 pounds per day leakage of primary coolant at " design" activity levels entering the reheat steam system.
For I-131 and neglecting any removal by the demineralizers or condenser air ejector,1064 pounds per day of primary coolant -
leakage at " design" activity would have to cccur to reach the LCO limit.
The corresponding leakage rate for the tritium LCO limit t=>uld be 48,060 pounds per day.
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5; 2.3 Monitoring There are a number of plant radiation monitors which potentially could be utilized to monitor effluent releases of primary coolant leakage into the reheat steam system. These are the loop header radiation monitor channels 93250-10, 93251-10, 93252-10, 93250-11, 93251-11, and 93252-11; loop header activity monitor channels 2263 and 2264 (also known as " loop header condensate monitors"); air ejector monitor channel 31193; and the ventilation exhaust gas monitor 7324-1. These monitors are discussed in FSAR, Section 7.3.5.
The loop header radiation monitors are part of the plant protection system and are designed to stay on scale at primary coo.lant leak rates of 1 to 3 pounds per second with the primary coolant at " design" activity levels.
These monitors will not be useful for monitoring releases at the imposed low limit of 1.4 curies per day.
The remaining radiation monitors all have about the same degree of sensitivity, but their capability to monitor will vary with the dilution of the sample.
The piping for activity monitors 2263 and 2264 has been modified as previously described in the modified system description. They will be primarily used to monitor the activity of the steam generator interspace gas.
The capability for monitoring the het reheat header activity will be retained. With an instrument range of 10-5 to 10-1 3
uCi/cm they will be quite adequate for this service.
The air ejector monitor has an instrument range of 5 x 10-' to 5 x 10 2 3
pCi/cm The ventilation exhaust monitor has an instrument range of 1 x 10 I to 1 x 10 -2 pCi/cm. The ventilation exhaust flow rate is typi-3 cally about 32,000 cfm. The air ejector flow rate is 15 to 30 cfm, depending on whether one or both air ejector trains are in operation. The air ejector monitor will, therefore, be more sensitive by about three orders of magni-
- tude due to less dilution of the sample. The plant ventilation monitor is not suitable for monitoring the low release rate of up to 1.4 curies per day because of greater dilution. This monitor does have the capability of detecting 001 MPC based upon the annual average dilution factor.
It would, therefore, serve as a backup should there be a sudden release in excess of 1.4 curies per day.
The air ejector monitor, assuming maximum dilution with both air ejector trains operating and a fluid transport time of 1.5 minutes, has a lower 1.evel of detectability of 0.014 pounds per day primary coolant leakage at " des-ign" activity level into the reheat steam system. At current primary coolant activity levels which are only 1% of design activity level, the lower level of detectability is 1.4 pounds per day. The monitor, which has a range of four orders of magnitude, will remain on scale at the limiting release level of 1.4 curies per day.
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. 2.4 Evaluation 2.4.1 System Modification The primary. purpose of the steam generator interspace as part of the PCRV Auxiliary Piping System is to monitor the primary coolant leakage through 4
the primary seal and to facilitate testing the primary and secondary seals on the steam generator penetrations.
The modifications to the 'pCRV auxiliary piping system for the steam generator penetration interspace do not impair the capability to nonitor the interspace for any possible primary coolant leaks. The piping modifications performed increase monitoring capability in that they permit routing and sampling of each interspace penetration separately for potential primary coolant leakage by the existing activity monitors.
The piping. system, as modified by Public Service Company of Colorado, continuously monitors both loops for primary coolant leakage.
If activity is detected on either loop 1 or loop 2 monitor. the indvidu31 interspaces can be valved such that they are separately monitored to deter-mine which primary seal is leaking.
The staff has reviewed the modified system piping and determined that the modifications do not inhibit monitoring of potential primary seal leakage.
Therefore, we find the modifications acceptable.
2.4.2 PCRV Pressurization Two separate safety valve installations protect the primary-secondary inter-spaces on the steam generator penetration modules from overpressure. The
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six steam generator modules in Loop 1 are grouped together and protected by a single safety valve installation, as is each module in loop 2.
Each in-sta11ation contains two safety valves and rupture discs in parallel to fa-cilitate maintenance, testing and replacement.
Interlocked block valves isolate each of the two parallel safety valve systems and are connected so that one safety valve is available at all. times.
The four safety valves, two in each loop, are each set to open at a nominal pressure of 475 psig.
This permits a 370 psi drop in a safety valve inlet line when the valve is relieving l
at. nameplate capacity and prevents the penetration pressure from exceeding its design pressure of 845 psig.
In order to prevent continuous helium leak-age past the safety valve seats, a rupture disc is located upstream of each safety valve and it acts as the primary barrier. The rupture discs have a maximum burst pressure of 840 psig which, including the maximum downstream static pressure of 5 psig, prevents penetration pressure from exceeding 845 psig.
The staff has evaluated the possibility of PCRV overpressurization as a result of the proposed technical specification revision to LCO 4.2.7-PCRV Pressuriza-tion Limiting Conditions for Operation and found the proposed change acceptable, since operating the steam generator interspaces in the proposed mode does not contribute to the potential for PCRV overpressurization.
Because of the existing leakage of the steam generator interspace, reducing the interspace pressure from above primary coolant pressure to above cold reheat steam pressure reduces the driving force for helium leakage.
Reducing the amount of purified helium in the cold reheat steam permits a condenser vacuum such as required for normal plant op3 ration. However, this also per-mits a possible primary coolant leak across the primary closure and into the reheat steam system.
To preclude possible adverse environmental consequences.
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i more stringent leak tightness has been specified for the steam generator primary closures when operating in this mode.
204.3 PCRV Closure Leakage PSC proposed to operate the steam generator penetration interspaces at above cold reheat steam pressure and below primary coolant pressure. By operating below primary coolant pressure, there exists a potential pathway of primary coolant helium across the primary closure and into the reheat steam system.
The helium, along with primarily noble gases, would be removed by the con-densor air ejector and released out the plant stack. Based upon a plant capacity factor of 0.8, 292 days of operation per year, and the LCO 4.8.1 Limiting Condition of Operation for gaseous release of 4160 curies per year, PSC proposed to limit the release of noble gases by means of this pathway to 10% of the LC0 4.8.1; this results in a limit of 1.4 curies per day.
Based on PSC calculations, using the above limits results in an annual exclu-sion area boundary whole body gamma dose of 0.14 millirad, and a bets skin dose of 0.06 millirad. The annual thyroid dose corresponding to 0.61 pounds per day of primary coolant at design activity is 5.2 x 10-6 millirads.
The staff has reviewed PSC calculations in light of our own more conservative bases. We are also cognizant of the fact that to date the circulating activity is about 1% of the design value.
Our evaluations have determined that, with our own more conservative models, the originally approved limits as stipusated
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in our Safety Evaluation Report as well as the limits of 10 CFR 50 Appendix I are not exceeded by the proposed change to the Techn'ical Specifications.
2.5 Conclusions The staff has reviewed the requested change to technical specifications LCO 4.2.7-PCRV Pressurization Limiting Conditions for Operation and LCO 4.2.9-PCRV Closure Leakage Limiting Conditions for Operation as presented in letter dated January 8, 1982. We have determined that:
(1) the modifications to the piping system do not inhibit monitoring of potential primary seal leakage (2) the operation of steam generator penetration interspaces at below primary coolant pressure but above cold reheat steam pressure will not overpressurize the PCRV, but will however permit a pathway for primary coolant to the reheat steam system, if a primary seal leak exists.
(3) imposing the closure leakage limits as stipulated will not exceed previously established environmental limits and consequences.
We have further determined that'the proposad changes to the technical spect-fications will not endanger the health and safety.of the public and we therefore find the changes acceptable.
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. Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement er negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public
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will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: MAR 1 8 1982 Principal Contributors:
George Kuzmycz m,
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TABLE 1 879 M'J(t) Noble Design Activities - f rom FS AR Tab. 3.7-1 Primary Buildup +
Coolant Decay (1)
E AE E (MeV)
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'"' I Huclide llalflife (C1)
(C1)
(McV)
(C1 MeV)
(C1 MeV)
Kr-83m 112m 1780 1751 4.14-2 72.5
=-
K r-85m 4.4h 2080 -
2065 1.86-1 384 0.229 472.9 Kr-85 10.3y 4
4 2.20-3
.01 0.229 0.9 Kr-87 78m 2810 2744 7.64-1 2096 1.375 3773 Kr-UB 2.8h 4200 4154 2.03 8433 0.307 1275.3 I}
Hb-ll8 17.8m N/A 436 6.77-1 295 2.00 872 Kr-89.
3.2m 1240 694 2.11 1464 1.279 887.6 I)
Rb-89 15.4m N/A 113 2.25 254 0.896 101 Xc-133m 2.3d 30.1 30.1 2.33-1 7
Xe-133 5.27d 726 726 8.2-2 59.5 0.1005 73 Xe-135m 15.3m 1100 974 5.26-1, 512 Xe-135 9.13h 1500 1498 2.62-1 392 0.303 454 Xc-137 3.9m 686 426 1.88-1 80 1.813 772.3 Xe-138 17n 858 769 2.18 1676 0.457 351.4 E-17,014 E-16,384 E-15,727 E-9034 (1) Based upon time to reach EAR of 161 sec 5904a EAR distanc 161 sec
=
3.67 m/sec wind speed (2) Not applicable sinco activity of this nuclide in the primary coolant which enters the secondary will be negligibly stripped from the steam by the air ejector.
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