ML19263D309
| ML19263D309 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 03/22/1979 |
| From: | Parker W DUKE POWER CO. |
| To: | Baer R, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7903270373 | |
| Download: ML19263D309 (12) | |
Text
9 DUKE POWER COMPANY POWER DUII. DING 422 Sourn Ciruncir STREET, CIIAHIDTTE, N. C. 28242 wituru o. eAn nen.s n.
March 22, 1979 vier PpEsiorst TEttPMoNCI A A E A 704 Strau Paoovcf row 373-4083 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attention:
Mr. Robert L. Baer, Chief Light Water Reactors Project Branch No. 2 Re: McGuire Nuclear Station Units 1 and 2 Docket Nos. 50-369, 50-370
Dear Mr. Denton:
On January 18, 1979 representatives of Duke Power Company and the NRC Staff met to discuss the proposed augmented inservice inspector program for McGuire Nuclear Station. As a result of that meeting, the Staff indicated that Duke's proposal would be acceptable subject to specific requirements. These require-ments or positions were documented in a February 9, 1979 letter from Roger S.
Boyd to Duke Power Company.
Attached are responses to each of the positions contained in Mr. Boyd's letter.
Very truly yours, h
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William O. Parker, Jr. /h " '[
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7903270313
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th Anniversary
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POSITION 1 RESPONSE The criteria applicable for the design of McGuire Nuclear Station against the dynamic effects of postulated high-energy pipe ruptures is described in detail in the McGuire FSAR.
This criteria evolved in parallel with Regulatory Guide 1.46 and staff positions APCSB 3-1 and MEB 3-1.
However, design and construction of McGuire had already begun before any positions relative to pipe rupture had been finalized.
Extensive engineering effort was made to meet requirements of a backfit nature.
A total of 420 break locations inside containment and 65 break locations outside containment were reviewed (per unit) for the effects of pipe whip and jet impinge-
, ment.
Wherever practical, piping was routed to maintain separation of high energy lines from essential plant components.
Of the 485 break locations, 293 locations require no physical protection since adequate separation is provided. At 186 of the remaining 192 locations, pipe whip restraints and/or jet barriers are designed to protect all identif-ied unacceptable targets.
For six break locations neither separation nor restraint structures are feasible. These six locations are the welds at both ends of the first elbow off Reactor Coolant Loops A, C, and D in the 10" accumulator lines.
Unacceptable jet impingement targets are the 2" steam generator blowdown lines, the 16" feedwater lines and the 6" hot leg injection lines.
To provide protection of the 6" hot leg injection by means of separation is impossible, since the hot leg injection line nozzle at the Reactor Coolant Loop is impinged up-on and it cannot be relocated.
Construction of physical restraints has been exten-sively pursued.
Large loads would have to be transferred to the shield wall re-quiring a number of large anchor bolts.
The shield wall is heavily reinforced with two rows of #11 bar.
Installation of high tension anchor bolts is considered impossible and extreme congestion exists in these areas. Modification of the basic building structure is not considered practical or desirable.
The acceptability of breaks at the six locations is discussed in the following response to Position 2.
However, in order to provide additional protection it is proposed that an acoustic emission leak detection system be installed and addi-tional inservice inspection be performed. The inservice inspection techniques for these locations will include a procedure for the detection of stress corrosion cracking.
(This and other procedures are described in Appendix I to topical SRG-78-01, Rev. 1.)
In addition, surface examinations at each weld will be per-formed.
To aid in the detection of small cracks, all examinations will occur with the system under pressure.
Inspection will be performed at intervals prescribed by Section XI.
The leak detection system will continuously monitor these areas of piping and will alarm in the control room if an acoustic signal characteristic of fluid leakage is detected.
POSITION 2 RESPONSE Part a The 16" feedwater line has been analyzed for the jet impingenent loads resulting from circumferential and longitudinal breaks in the 10" accumulator line.
The method of analysis consisted of (1) a determination of these distributed loads on the piping system, (2) reduction of the distributed loads into point loads at specified node locations and (3) inclusion of these point loads in a piping stress analysis along with pressure, tamperature and dead weight loads using the PISOL3A computer code.
As an example, consider the case of a longitudinal crack in the Loop D accumulator line. A RELAP analysis has shown the blowdown thrust load to be 217 kips.
Assuming a 100 half-angle expansion, the pressure load on the impinged feedwater line is 40.1 psia.
Figures 1 and 2 show the jet in relation to both the accumu-lator and feedwater lines. The load on the feedwater line can be found from F = 2KcPA where:
2 = dynamic load factor for a suddenly applied load K6 = shape factor (Reference ANSI 58.2)
= one-half the drag coefficient for submerged flow P = jet impingement pressure on the feedwater pipe A = cross sectional area of the feedwater pipe normal to the jet flow.
The calculated load of 55.6 kips is applied as shown in Figure 3 for input to the piping stress code.
A maximum stress of 26,005 psi is computed for the case of a longitudinal break in Loop D when including the combined effects of internal pressure, weight, jet impingement and thermal expansion.
The faulted allowable is 36,000 psi.
The following table summarizes the results for all loops and crack types CIRCUMERENTIAL BREAKS LONGITUDINAL BREAKS LOOP A LOOP C LOOP D LOOP A LOOP C LOOP D N0DE STRESS (psi STRESS psi STRESS (psi STRESS psi STRESS psi STRESS (psi A
14147 17324 10521 10976 12528 7931 B
12645 14812 13364 10076 10769 9633 C
13618 14506 15679 11024 11398 11981 D
12294 10076 11177 9915 8735 10551 E
11219 10293 9950 9728 8350 9232 F
9790 11278 14963 10798 10766 13704 G
20159 25876 18756 14658 16527 10677 H
29439 36967*
27703 21586 26005 19533
- This stress exceeds the allowable from EQ. (9f) of the ASME Code,Section III.
% Overage = 36967 x 100 = 2.7%
36000 These stresses are conservative in that the thermal stresses will be relieved during the loading process, as they are not primary loads. Also the allowable stresses in Section III are based on minimum specified values.
Material usually exceeds minimum values by 3% to 5%.
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IMPINGEMENT LCAD ON QUADRANT D FEEDWATER LINE
POSITION 2 RESPONSE Part b Containment mass and energy releases were calculated for a combinat{on cold leg break and feedline break for the McGuire Nuclear Station. A 0.5 ft break in the cold leg and a 2 inch feedline break were assumed to simultaneously occur and be-gin releasing mass and energy to the containment.
The October 75 verric.7 of the WFLASH computer code was used for these calculations. A description of the code and its use is documented in WCAP-8339 and WCAP-8970-P-A.
The procedural guidelines followed to provide this information is very similar to that used for ECCS small LOCA analyses with a slight modification to
'fectively and conservatively model the feedline break.
The small break ECCS analysis doc-umented in Chapter 15 of the McGuire FSAR was used as a base.
The standard 15 node WFLASH model was modified to simulate an additional break at the bottom of the broken loop steam generator secondary side node.
This was done by adding a leak flow path to the existing flow path network of the ECCS model.
The initial conditions assumptions and input data for the mass and energy release calculations are the same as those for the DAP small break ECCS analysis.
This includes:
1.
102% of ESDR Power 2.
Appendix K decay heat values.
3.
Metal heat addition.
4.
Minimum safety injection.
5.
Secondary to primary heat transfer modeling.
6.
Containment at atmospheric pressure.
7.
Initial system pressure at 2280 psia.
8.
Steady state operation before break opening.
A more detailed description of code inputs and assumptions are documented in the above referenced topical reports.
Table 1 summarizes the mass '..i energy releases to the containment as a function of time out to 1775 seconds.
POSITION 2 RESPONSE (Cointued)
For the purpose of this analysis the decay heat generated in the core following the end of the WFLASH run was assumed to generate steam.
In actuality the safety injection water would soon be able to condense all of the steam being generated, and the releases would become subcooled.
Once the releases became subcooled, the ice would melt at a very slow rate.
Clearly the assumption that the decay heat generates steam is conservative.
As previously discussed, at 1775 seconds the WFLASH run was terminated and the mass and energy releases were calculated 'by the LOTIC-1 computer code (Ref. WCAP-8354).
The mass and energy calculation was performed in the same manner as the FSAR post-froth calculation, except that the steam boiloff was increased to account for long term steam releases from the secondary side break.
The LOTIC-1 computer code was utilized to calculate the containment responses to this accident. The input assumptions were the same as those detailed in Section 6.2.1 of the FSAR (containment Integrity Analysis) with the exception that pump flowrates versus time were modified to be consistent with the WFLASH calculation.
The peak pressure was calculated to be 13.8 psig, occurring between 6550 and 6980 seconds.
TABLE 1 Time Break Mass Flow Rate Break Energy Flow Rate (Seconds)
(lbs/second)
(million Btu /second) 0 7762 4.392 12 6300 3.607 18 4802 2.779 24 4529 2.633 36 4500 2,598 66 4331 2.434 72 1345 1.481 78 4131 2.267 90 3629 1.821 96 3429 1.656 102 739 0.7722 108 3110 1.441 114 3057 1.408 120 695 0.720 138 640 0.6544 156 571 0.5721 180 370 0.3298 186 341 0.2949 210 269.9 0.2054 222 268 0,2090 240 290 0.2369 270 327 0.2839.
300 340 0,3025 400 368 0,2974 450 337 0,3100 504 308 0,2793 600 261 0.2298 698 223 0.1885 802 250.5 0,2283 850 250.2 0.2084
TABLE 1 (Continued)
Time Break Mass Flow Rate Break Energy Flow Rate (Seconds)
(lbs/second)
(million Btu /second) 906 243 0.2235 1014 214.1 0.1949 1414 198 0.1844 1614 169.3 0.1520 1714 166.2 0.1481 1774 148.9 0.1157
POSITION 2 RESPONSE Part c The following analysis is an evaluation of the consequences of a break in the ten-inch cold leg accumulator line together with a subsequent break in a six-inch hot leg safety injection line.
This analysis is performed to demonstrate that the consequences of a break in the six-inch hot leg safety injection line possibly resulting from the jet forces generated from certain unrestrained nodes in the ten-inch cold leg accumulator injection line are acceptable.
The transient scenario consist of a break in the ten-inch cold leg accumulator injection line followed by the break in the six-inch hot leg safety injection line. The reactor coolant system initially would blowdown through the ten-inch cold leg accumulator injection line break. Subsequently, when the break in the six-inch hot leg safety injection line occurs, the rate of depressurization of reactor coolant system (RCS) would increase. The increase in blowdown would result in an earlier actuation of the emergency core cooling system (ECCS).
Furthermore, the break in the six-inch hot leg safety injection line would pro-mote higher positive core flow, and therefore greater removal of the core stored energy.
In addition, the hot leg safety injection (SI) line break would act as a vent path for the steam generated in the top of the core, and this vent-ing process would result in less steam binding and improved reflood.
- erefore, there would be an earlier quenching of the core with an early recovery of the core mixture level. The net result of this break scenario would be that the peak clad temperature would be less than the limiting break cases analyzed in the FSAR.
The hot leg safety injection is used during the long-term cooling phase after a loss of coolant accident (LOCA) in order to prevent the buildup of unacceptably high boron concentration in the core (approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after the LOCA).
The safety injection system alignment during this phase consists of the RHR pumps supplying flow to two hot leg safety injection nozzles, the safety injec-tion pumps supplying flow to the other two hot leg safety injection nozzles, and the charging pump supplying flow to the cold leg safety injection nozzles.
Assuming single failure of one train and assuming that the safety injection flow to the broken lines is lost, the minimum effective safety injection flow into the core during hot leg recirculation is approximately 47, lbm/sec through the cold legs and approximately 4_2 lbm/sec through the hot leg.
These flows are in excess of the decay heat boil off at this time of approximately 2_0 lbm/sec, and are sufficient to maintain the reactor boron concentration to acceptable levels and to enable decay heat removal.
It is to be noted that ber.ause of the break in the hot leg safety injection line, forced recirculation of the coolant from the cold legs through the core and out through the hot leg break would prevail during this acci-dent, and therefore, boron precipitation would not be a problem.
Accordingly, ability for long-term decay heat removal is maintained.
It is concluded from this, analysis that a break in a six-inch hot leg safety injection line occurring as a result of a break in a ten-inch cold leg accumu-lator injection line is less severe than the limiting LOCA analysis reported in the FSAR.
POSITION 3 RESPONSE Duke Power Company will provide a description of the final design of the Acoustic Emission Leak Detection System including proposed acceptance standards and plant corrective action by June 15, 1979.
This system will be operational prior to exceeding 1% power.