ML20080T813
| ML20080T813 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 02/24/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080T803 | List: |
| References | |
| NUDOCS 8403020061 | |
| Download: ML20080T813 (7) | |
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o UNITED STATES 8
g NUCLEAR REGULATORY COMMISSION o
t E
WASHINGTON, D. C. 20555 SAFETY ~EVALUAi10N BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 95 TO FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDfUli N0. 94 TO FACILITY OPERATING LICENSE N0. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Introduction By letter dated September 13, 1983, as supplemented October 6, November 30, December 19, 1983, and January 18 and 25,1984, the Virginia Electric and Power Company (the licensee) requested. amendments to the Operating License Nos. DPR-32 and DPR-37 for the Surry Power Station, Unit Nos.1 and 2, in the form of changes to the Technical Specifications. The proposed changes would reduce the boron concentration in the Boron Injection Tank (BIT). Specifi-cally, the proposed changes to the Technical Specifications will eliminate the minimum boron concentration requirement in the BIT, ano reduce the boron concentration requirement in the boric acid system from 11.5 wt% to 7.0 wt%.
Background
Westinghouse incorporated a boron injection tank (BIT) into the Surry Safety Injection (SI) system to mitigate the consequences of postulated steam line break (SLB) events by purging highly concentrated boric acid solution (20,000 ppm B) into the RCS. The licensee has submitted a request for
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Technical Specification changes including reduction of the BIT boric acid concentration from 11.5% (20,000 ppm B) to 0%, and reduction of the minimum boric acid concentration in the boric acid tanks (BATS) from 11.5% to 7%
(12,000 ppm B). The minimum specified BAT temperature would be reduced fron' 145 F to 112 F, and the BIT temperature specification would be deleted. The licensee's proposed Technical Specification changes include increasing the minimum allowable BAT inventory associated with each unit from 4200 gallons to 6000 gallons. This would preserve the capability for cold shutdown at any time in core life with the most reactive control rod assembly withdrawn from the core.
The licensee has stated that the requested change would reduce maintenance problems and associated personnel radiation exposure (136 man-rem savings) by reducing leakage due to corrosion and decreasing heat tracing circuitry failures..The latter 'can cause line plugging and flow. restrictions as the temperature decreases and precipitation of the concentrated boric acid
. solution occurs. The ~ potential for corrosion of carbon steel components and supports due to leakage would also be reduced. The physical modifications involved in the proposed change would include cutting the recirculation between the BIT and the BAT, welding the ends closed, and removing the 8403020061 840224 DR ADOCK 05
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electric power to the re' circulation line isolation valve, BIT heaters and recirculation line heat tracing.
The BIT would remain in place.
Evaluation The design basis accident for which the BIT was designed is the Main Steam Line Break (MSLB). The licensee therefore submitted revised analyses which include the MSLB accident with and without offsite power as well as a smaller
. break equivalent to the capacity of a single steam dump valve or safety valve.
The Surry steam generators are equipped with integral flow restrictors at the generagor outlet which serve to reduce the largest effective break area to 1.4 ft.
Each main steam line has a fast closing trip valve, designed to close in less than 5 seconds, and a non-return valve. These valves prevent blowdown of more than one s+eam generator even if one valve fails to close.
The energy removal due to the MSLB causes a rapid reduction of reactor coolant system (RCS) temperature and pressure. This results in an increase in reactivity and decrease in shutdown margin. The analysis assumes conservative initial conditions, including hot shutdown with all but the most reactive control rod inserted at end of core life. These assumptions maximize the positive reactivity insertion resulting from conidown. The single most restrictive failure of the Engineered Safety Features is also assumed, resulting in operation of only one high pressure Safety Injection (SI) pump.
The time delay required to sweep unborated water in the BIT and associated SI piping prior to delivery of borated water of 2000 ppm from the RWST has been included in the analysis.
The MSLB analysis was performed by the licensee using the RETRAN computer code. The analyses indicate return to criticality for all 3 cases, with the highest peak heat flux of 23.7% achieved for the MSLB with offsite power available.
In the determination of the critical heat flux at which burnout could occur during a steam lire break, the W-3 correlation was used. This correlation is generally considered valid between pressures of 1000 to 2300 psia. The
.resulting RCS pressures in the three steam line break ca.aes which were reanalyzed were below 1000 psia and ranged from 959 to 733 psia. Although the staff has not approved the use of the W-3 correlation belaw pressures of 1000 psia, the calculated DNBRs were appreciably higher than the W-3 1.30 limit, and the staff concludes that there is sufficient conservatism in the Surry steam line break calculations to assure that DNB will not occur.
Elimination of the Vron concentration requirement in the BIT could affect the containme:it press 6re and temperature response under MSLB accident conditions through changes in the mass and energy release rates. The licensee has performed senr.itivity studies to address the impact of reducing the BIT boron
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concentration on early MSLB energy release, and has concluded tnat the current equipment qualification temperature envelopes for the Surry plants are adequate. Sincg LOCA conditions dominate the containment functional design considerations, the licensee used the LOCA temperature profiles for post-accident equipment qualification in lieu of MSLB temperature profiles.
We have also made comparisons with similar reviews for the Beaver Valley and
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. North Anna plants. The boron concentration reduction programs at these plants were previously found acceptable by the staff. Based on a review of the information submitted by the licensee, and because of the similarity of the licensee's request to other staff actions, we conclude that the licensee's proposal to eliminate the minimum boron concentration requirement in the BIT will not adversely affect the containment functional. performance.
The licensee evaluated three scenarios with an MSLB and its-associated cooling of the primary system. For each, the licensee calculated a worst-case return-to-power transient.
Of these, the scenario with the greatest peak power after. scram was that which included assumptions of zero power at the break, the availability of offsite power, and the largest possible break.
(Other licensee assumptions to make the scenarios worst-case.with respect to return to power included a minimum reactivity shutdown margin, end-of-life core moderator temperature coefficient, the highest-worth control rod assembly stuck in its fully withdrawn position, and the failure of one high-pressure coolant injection pump.) We also assumed coolant iodine equilibrium and spiking consistent with SRP section 15.1.5.
This scenario is, however, not the worst case with respect to radiological consequences.
For example, the zero-power. initial condition assumption implies that there is negligible decay heat, thus easing the long-term plant cooldown. Also, the availability of offsite power increases the chances of better mitigating the accident by using the condenser to retain iodine from a potential leak of primary coolant to the secondary side of the unaffected steam generators. We were concerned about the potential for prolonged plant cool-down and primary-to-secondary leakage caused by additional energy from the return to power, which might be in addition to normal decay heat.
However, we found that significant return-to-power would not cause-energy input to the primary system that would be additive to decay heat, since the large return-to-power cases involved zero power initial conditions. Therefore, we determined that the worst-case accident (assuming there is no additional fuel failure) is for full power initial conditions, together with loss of offsite power.
Additionally, we assumed no further fuel or cladding defects because of the finding that the worst-case return to power reported by the licensee would not result in further fuel or cladding failure (beyond the assumed minor cladding defects associated with the assumed iodine spike).
Further, the proposed amendment does not affect the validity of the original staff MSLB dose evaluation in the Surry SER. A calculation was performed, using current methods, to confirm that the postulated MSLB doses are within the dose guidelines of 10 CFR 100.11 and appropriate fractions thereof, as defined in the acceptance criteria of SRP 15.1.5.
We have determined that the proposed amendment does not exceed or detrimentally affect our radiological consequence guidelines.
The calculated.MSLB doses are shown in Table 1, for both the 0-2 hour dose at the Exclusion Area Boundary and for the duration of the accident (judged to be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) at the outer boundary of the Low Population Zone. The assumptions are given in Table 2.
The acceptance criterion for the preaccident spike case is 100% of 10 CFR 100.11 guidelines, or 300 rems thyroid and 25 rems whole body; for the concomitant spike case, it is 10%, or 30 rems thyroid and 2.5 rems whole body.
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The licensee has stated that lowering the BIT boric acid concentration to O ppm eliminates the need to maintain BIT heaters and heat tracing on the associated SI. piping'and recirculation lines.
We requested information from the licensee on whether the normally stagnant section of the SI piping between the charging pump normal discharge piping and the closed isolation valves upstream-of the BIT can contain concentrated boric acid in the event of operator error or equipment failure (e.g., valve leakage), and, if this could occur, to explain how precipitation of boric acid and consequent possible pipe blockage would be avoided.
The licensee responded that the charging headers and stagnant SI piping upstream of the. BIT inlet valves are not currently heat traced, since they are located inside buildings.
Operating experience has indicated that boron precipitation in the charging header and stagnant lines has not been a problem.
The stagnant SI piping between the main charging header and the BIT inlet valves is flushed monthly in accordance with Technical Specification -4.1.E.
Also, the licensee has not deleted the Technical Specifications requirement to have two channels of heat tracing available for the SI flow paths.
We conclude that the licensee's response on this subject is acceptable.
Several letters have been received from the licensee since these requests for amendments were published.in the Federal Register on November 22, 1983.
These letters are dated November 30 and December 19, 1983, and January 18 and 25, 1984.
None of these letters changed the Technical Specifications proposed by the application dated September 13, 1983, nor the substance of'the application
.in a significant manner and were of a nature of providing additional clarification and details to the staff.
These letters are summarized in the following paragraphs.
The November 30, 1983 letter responded to staff questions related to the containment response following a postulated design basis main steam line rupture with a reduced boron concentration.
The licensee compares the Surry analysis to that performed for the Beaver Valley Power Station and concludes that the Beaver Valley calculations are bounding.
The December 19, 1983 response provides additional details of the reactivity feedback model and mixing coefficients utilized in the MSLB analysis, clarification of heat tracing requirements and discussion of offsite doses.
The January 18, 1984 letter provides an estimate of the man-rem reduction that the proposed change would effect.
This letter was provided for the interest of the staff and was.not the basis for the review.
The' January 25, 1984 letter provides a tabulation of data of MSLB Accident Statepoints which confirm that DNBR is well above 1.3.
An insignificant change was made to Table 4.1-1 in that the line item related to Boron Injection Tank Level-has been deleted instead of being shown as not applicable.
(This was not in the licensee's submittal tat was discussed by telephone.)
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Based'on our' review, we conclude that the proposed Technical Specification changes will not result in unacceptable consequences in the event of the design basis accident, and are acceptable.
Environmental Consideration-We have determined that the amendments do 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 amendments involve 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 or negative decleration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the r oposed manner, and (2) public such.
activities will be conducted in compliance with the Commission's regulations and tne-issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date:
EB 2 4 -1984
. Principal. Contributors:
J. Guo B. Mann P. Easley.
D.-Kovacic p
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Table 1 Radiological Consequences of a Postulated Main Steam Line Break Description Thyroid dose, rems
-Whole body dose, rems Pre-accident iodine spike 0-2 hour, Exclusion Area Boundary F.2 less than 0.1 0-8 hour, outer boundary of Low Population Zone 0.4 less than 0.1 Concomitant iodine spike (causedbyaccident) 0-2 hour, Exclusion Area Boundary 7.8 less than 0.1 0-8 hour, outer boundary of Low Population Zone 2.3 less than 0.1 That these doses are lower than the concomitant iodine spike doses is atypical compared to other plants, and is caused by the low Technical Specifications on the short-term maximum coolant iodine concentration.
Y Table 2 Assumptions Used in Estimating Doses From a Main Steam Line Break 1.
The break occurs on a main steam line between the containment penetration and the main steam isolation valve. The affected steam generator boils dry.
2.
During the rapid boil-off, all activity in the affected steam generator is released to the environment. The secondary side iodine concentration was assumed to be 0.1 microcurie dose equivalent (DE)
I-131 per cc. The steam generator liquid volume was 1690 cubic feet.
3.
Additional activity is released via a primary-secondary leak. This is assumed to be at the maximum allowed by technical specifications, which for Surry is.347 gallons per minute (gpm) to any one steam generator (assumed to be the affected generator), and a total of 1. gpm to all the steam generators.
4.
Pre-accident spike only cese: before the accident, an iodine spike has ocurred which brings the primary coolant activity up to the technical specification limit for 48-hour operation,10 microcuries per cc DE I-131.
5.
Iodine spike caused by accident case: before the accident, the primary coolant activity is at the technical specification limit for long-term operation, 1.0 microcurie per cc DE I-131. With the start of the accident an iodine spike begins,, which releases an additional 7400 Ci/hr, for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, from the core to the coolant.
6.
Because the time that the steam generator tube bundles in the affected steam generator are fully covered is small compared to the total duration of the accident, it is assumed (1) that tne primary-to-secondary leak to the affected steam generator occurs entirely in a dry section, and (2) that all the activity in the leaked coolant is released to the environment.
For the unaffected steam generators, it was assumed that the condenser would be unavailable, and that steam release to the environment would take place.
It was assumed, however, that cnly 1% of the iodine in the primary coolant leaked to the unaffected steam generators would be released to the environment.
7.
Primary-to-secondary leaks become negligible after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and all
' releases to the environment cease.
18.
All the noble gases in the leaked coolant, with a concentration equal to the technical specification limit, are released to the environment.
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