ML20217D322
| ML20217D322 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/17/1998 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9804240274 | |
| Download: ML20217D322 (16) | |
Text
,
Dave M:r:y Sruthern Nucle:r Vee President Operating Compary Farley Project P0. Box 1295 Birmingham. Alabama 35201 Tel 205.992.5131 SOUTHERN NN April 17,1998 Energy to Servehur%rld*
Docket Nos.:
50-348 10 CFR 50.90 50-364 4
U. S. Nuclear Regulatory Commission A'ITN.: Document Control Desk Washington, DC 20555 j
Joseph M. Farley Nuclear Plant Response to Request for Additional Information Related to Power Uprate Facility Oceratina Licenses and Technical Soccifications Channe Reauest Ladies and Gentlemen:-
By letter dated February 14,1997, Southern Nuclear Operating Company (SNC) proposed to amend the Facility Operatmg Licenses and Technical Specifications for Joseph M. Farley Nuclear Plant (FNP) Unit I and Unit 2 to allow operation at an increased reactor core power level of 2775 megawatts thermal (MWt). NRC letters dated July 1,1997; August 21,1997; and October 14, 1997 requested SNC provide additional information. SNC responded by letters dated August 5, 1997; September 22,1997; and November 19,1997 respectively. SNC letters dated December 17 and 31,1997; January 23,1998; February 12 and 26,1998; March 3,6 and 16,1998; and April 13,1998 responded to NRC questions resulting from conference calls. Attachment I of this letter provides a revised response to Question No. 3 of the Attachment to SNC letter dated April 13, 1998. As requested by the NRC Staff, Attachment 11 describes the methodology and input parameters used to estimate primary-to-secondary leak rate for a locked rotor event.
With regard to Question No. 3, the qualitative assessment of the RCP locked rotor and the control rod ejection events with respect to the Main Steam Line Break (MSLB) radiological analysis and steam generator Alternate Repair Criteria (ARC) has been revised. The assessment continues to consider the transient differential pressure between the primary and secondary side of the steam generators, estimate the primary leakage that potentially could occur, and then calculate offsite doses. 'Ihe revised assessment results continue to support the validity of the conclusions of WCAP-12871, Revision 2, "J. M. Ferley Units 1 and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates,"i.e., the MSLB is the most limiting event for ARC.
\\ f 9804240274 980417 PDR ADOCK 05000348 9 1 p
U. S. Nuclear Regulatory Commission Page 2 As agreed to by the NRC Staff, the transient analysis, evaluations and calculations used to support the revised assessment have not been formalized. This qualitative assessment is based on power uprate conditions and conservative engineering judgment. The uprate radiological calculations previously submitted for Staff review will continue to remain the calculations of record. The revised assessment and methodology used to estimate leak rates are being provided to the Staff for information only.
If you have any questions, please advise.
I Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY f 77L6 Dave Morey J
Sworn to and subscribed before me this/SYy of 1998 00AWo bSW
'L NotaryPublic U
My Commission Erpires:
l Nd/
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/
MGE/ cit:pwrup40. doc 1
Attachment ec:
Mr. L. A. Reyes, Region II Administrator Mr. J. I. Zimmerman, NRR Project Manager j
Mr. T. M. Ross, Plant Sr. Resident Inspector
ATTACHMENTI SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units 1 and 2 SNC REVISED RESPONSE TO NRC QUESTION NO. 3 RESULTING FROM NRC/SNC TELEPHONE CONFERENCE CALLS ON APRIL 9 & 13,1998 (REFER TO ATTACHMENT OF SNC LETTER TO NRC DATED APRIL 13,1998)
i SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units 1 and 2 NRC Ouestion No. 3 (Reference April 9 & I,3.1998 NRC/SNC Conference Call) a) With respect to the 10 CFR Part 100 off-site and on-site dose limits, the Farley power uprate analyses, and the assumptions for steam generator tube ARC, are the RCP locked rotor and control rod ejection events (due to potentially large increases in source terms resulting from fuel cladding or pellet damage) more limiting than the main steam line break event?
b) Based on short-term transient parameters, provide a qualitadve assessment for each of these j
events, which includes the potential for accident-induced steam generator tube leakage (similar to j
that estimated for application of ARC to the Farley steam generators) and the resultant impact on the off-site and on-site radiological doses.
SNC Revised Response to Ouestion No_3 l
a) For the Farley power uprate with ARC, the Farley steam line break event radiological consequences are more limiting thave the radiological consequences of postulated locked rotor and rod ejection events. L licensing basis for this statement is provided by the non-steam line break evaluations presented in WCAP-12871, Revision 2, "J. M. Farley Units I and 2 SG Tube Plugging Criteria for ODSCC at Tube Support Plates," February 1992. WCAP-128?l, Section 11.3, concludes ti.at the increased source terms associated with these events are offset by: reduced primary-to-ncondary differential pressure; decreased flashing and increased mixing in the steam generator; and contined coverage of the steam generator tubes at the tubesheet and tube support
]
plate interfaces. 'Ihe assumption of no long-term tube uncovery is supported by WCAP-13247, j
" Report on the Methodology for Resolution of the Steam Generator Tube Uncovery Issue," March 1992. This licensing basis is supported by the qualitative assessment presented below.
b) An assessment of the primary and== nary system pressure transient data associated with the locked rotor and rod ejectian analyses determined that the locked rotor pressure transients are more challenging than the control rod ejection pressure transients. Based on the pressure and temperature transient data from a locked rotor analysis performed especially to address this question, the primary pressure peaks in about 3 seconds, and it levels off after about 25 seconds.
To estimate primary-to-secondary leak rate in the Farley steam generators, scoping calculations were performed to determine the leak rate ratio for a steam line break versus a locked rrtor L
transient. 'Ihe calculation methodology is similar to that applied to adjust measured leak rates fo, pulled tube specimens. Based on the pressure and temperature transient results discussed above, the leak rate ratio was calculated separately for 0 to 25 seconds and >25 seconds time periods.
l' Ratio 0 to 25 seconds
>25 seconds Leak Rate at SLB Condition 3.7 8.4 l
Leak Rate at Locked Rotor Condition 1
W above leak rate ratios are based on the equation for variation of crack opening area with pnmary! o -aa@y differential pressure as described in EPRI Report NP-7480-L, Volume 1, t
Revision 1, " Steam Generator Tubing Outside Diameter Stress Corrosion Cracking at Tube Support Plates - Database for Alternate Repair Limits, Volume 1: 7/8 Inch Diameter Tubing,"
Appendix B, using leak rate test data for 5 pulled tube specimens from FNP Units I and 2 in the ARC database. It is also noted that since significant TSP movement is not expected during a locked rotor event, packed TSP crevices would prc:lude any significant leak through ODSCC indications as discussed in WCAP-14707, Revision 1,"Model 51 Steam Generator Limited Tube Support Plate Displacement Analysis for Dented or Packed Tube to Tube Support Plate Crevices," January 1997. This WCAP result is supported by the in situ leak test of FNP Unit I pulled tube R2C85 which had a throughwall crack of 0.42 inch, but did not leak at normal operating conditions.
As an illustration of applying the above ratios the following example is presented. The steam line break leak rate calculated for the limiting SG in Farley Unit 2 (including UOA indications with PIs and applying the latest ARC database) is 9.34 gpm (at room temperature). The corresponding leak rate estimates for a locked rotor condition obtained by applying the above leak rate ratios are 2.5 gpm (0 to 25 seconds) and 1.1 gpm (>25 seconds). It is noted the steam line break leak rate value l
used (9.34 gpm) was obtamed assunung that leak rate is independent of bobbin voltage. With the recent NRC clarification on the requirements for a leak rate correlation, a leak rate vs. bobbin voltage correlation can now be applied for 7/8" tubes, which leads to a significant reduction in the estimated leak rate. It can be noted that the locked rotor primary-to-secondary pressure differential after 25 seconds is 944 psi, which is much less than the normal operating pressure differential of about 1450 psi. Since there is only a small amount of plastic crack opening from the 1866 psi i
differential during the first 25 seconds, the leak rate after 25 seconds would be less than the normal operation sheiown limit of 0.1 gpm. % very conservative leak rates in the current analysis are a j
consequence of extrapolating from SLB conditions to the low pressure differential rather than the closer extrapolation from normal operating conditions.
An assessment of the potential impact on the off-site doses based on the leak rate ratios presented above follows. Since the accident induced leakage estimated for the locked rotor accident bounds j
that for the control rod ejection, this leak rate will be used for both assessments. For similar steam i
generator tube conditions (i.e., those which result in the limiting leakage of 24 gpm for a main steam line break), the locked rotor leakage is 1/3.7 of the limiting leak rate (i.e., 6.5 gpm) for the 0 to 25 seconds period and 1/8.4 of the limiting leak rate (i.e., 2.9 gpm) for the >25 seconds period.
l These leakage rates are assumed to exist in all three steam generators, and the long-term leakage is i
assumed to last until the RCS and steam generator pressures equalize. For the control rod ejection event, the duration, as described in the Farley Power Uprate NSSS Licensing Report (WCAP-l 14723), is 2500 seconds. For the locked rotor event, it is assumed that the RCS pressure remams l
constant to the RHR cut-in at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. in addition, for the locked rotor, the estimated fuel failure has been reduced from the power uprate radiological assumption of 20% of the gas gap to the Farley-specific value of 6.3% (This is a conservatively large number for the " rods in DNB" calculated for the Farley uprate locked rotor analysis; e.g., the Farley Unit 2 Cycle 13 value is
<0.135%) 'Ihe rod ejection event source term assumption is the same as used in the power uprate radiological analysis. Comparison of the control room X/Q with iodine protection factor and the off-site X/Q indicates the control room thyroid doses for these accidents with accident-induced leakage is not limiting. The limiting off-site thyroid doses compare to the acceptance limits as follows.
2
Partition Total Acceptance Percent Eysol Factor iaaka=a Duration Limit (REM) ofLimit MSLB (ARC) 1 24 gpm 8 hr 30 100 Locked Rotor 0.01 19.5 gpm 8 hr 30 73 l
(t <25 sec) 8.6 gpm l
(t >25 sec) l Rod Ejection 0.01 19.5 gpm 2500 sec 75 61(0 l
(t <25 sec) j 8.6 gpm l
(t >25 sec) l These qualitative results demonstrate that the accident-induced leakage limit determined for the main steam line break remains limiting.
l l
SNC letter dated April 13,1998, listed a value of 63; the correct value should have been 59 for O'
the rod ejection event " Percent of Limit."
W/dh & wjs & vs - 4/15/98 & SCS/ jaw - 4/16/98 l
3 l
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i i
e ATTACHMENTII l
SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal-Joseph M. Farley Nuclear Plant, Units 1 and 2 LEAK RATE ESTIMATES FOR A LOCKED ROTOR EVENT l
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i 1
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SNC Response to NRC Request For Additional Information Related to Power Uprate Submittal - Joseph M. Farley Nuclear Plant, Units 1 and 2 LEAK RATE ESTIMATES FOR A LOCKED ROTOR EVENT Summary This write-up describes the whndatagy and input parameters used to estunate leak rate for a locked mtor event and summarizes the results. leak rate estimate is presented as a ratio ofleak rate at SLB condition to locked rotor condition. Separate values are shown for two time periods in the tmnsient.
Time Period SLB Leak Rate / Locked Rotor Leak Rate l
0 to 25 seconds 3.7 i
> 25 seconds 8.4 1.0 Application of EPRI NP-7480-L, Volume 1, Revision 2, Appendix B From equation B-11, the ratio of leak rates between two conditions is obtained as:
R = U' = apy Db l
l From equation B-15, the factor accounting for crack opening area and difference in pressure differential is (equivalent to ratio ofleak rates for room temperature leak rates):
t A r, - A ri) a = 10 s
)
l From equation B-21 with no adjustment for volumetric leak rate from hot to room temperature conditions, the temperature correction is:
1 Es ai {
l p = E2a2 % Pi l
1 From equation B-22 with C, = 1.0 and flashing at both leak rate conditions, the factor for the difference in l
flashing conditions (difference in primary pressure and saturation pressure) is:
l ipr 2 - P 2) 1 AP2 1 (Pr,-Pp) bPi l
1
r l
l I
where l
Ap2 " PN - p 2 equals primary to secondary pressure differential for condition 2 pa quals saturation pressure at condition 2.
e Step 1: Adjustment of Measured Leak Rate to Standard Conditions (p, = 15 psi and T = 620 'F) at Same Ap R
pya
=
a 1
E ~ E,
ya=)Pr-Ps p,, = A p, + 15 s
t Step 2: Determination of b in alpha term (example based on two test points) 3 Step 1 previously applied to each of the test points to obtain leak rates at 15 psi secondary pressure and 620 'F. R is the ratio of the leak rates for the two test points.
(an-Ard R = ay = y 10 s
bAP2 - api log (ivy) i This equation for b is equivalent to the last equation in Section B.6.1 with the generality that gamma can 1
include flashing at both conditions.
When more than two data points are available over the range ofinterest, a mean value of b can be obtained by fitting a regression curve to the leak rate versus Ap data and obtaining b from the slope of the regression fit. For simplicity, b values in this study were obtained using the leak rates at the highest two pressure differentials included in the tests for Farley pulled tubes. This is equivalent to extrapolating leak rates from SLB conditions down to lower pressure differentials.
Step 3: Calculation of Leak Rate Ratio Between the Measured Point 1 and the Desired Point 2 l
Apply above equations B-11, B-15 and B21 to obtain the ratio R between the leak rates at the locked rotor conditions (point 2) and SLB leak rates (point 1).
2 l
2.0 Evalnation of Lacked Rotor for > 25 Seconds Into the Transient The primary to secondary pressure differential peaks at about 1866 psi at about 3 seconds into the event and then reaches a quasi-steady value of about 944 psi beyond 25 seconds. For tube leakage, the crack opening at the higher pressure consists of clastic and plastic opening area contributions. The plastic opening would not significantly decrease as the pressure differential decreases, but the clastic opening would decrease This condition can be evaluated in two ways. Conservatively, one can assume that the crack opening area does not change between the 1866 and 944 psi conditions. More realistically, the l
plastic opening can be estunated at 1866 psi using the CRACKFLO code and included in the leak rate l
estimate for 944 psi. Both of these methods are described below.
From equation B-12,
'AP2 g, A_2.k A As P
Conservatively assuming the crack area is totally plastic between 1866 and 944 psi conditions, the areas l
are equal and:
bP2 a=
, j,4j APs The gamma factor between the 1866 and 944 psi conditions can be obtained from the expression given i
above from equation B-22. The product of alpha and gamma yields the leakage ratio between the 1866 I
and 944 conditions. Ganuna is calculated to be 0.935 so that the ratio ofleak rates between 1866 psi and l
944 psi pressure differentials becomes 1.32.
j The CRACKFLO calculations for the 1866 and 944 psi conditions show that the crack opening areas arc I
dominantly clastic for both conditions. The plastic area at 1866 psi is only 16% of the clastic area, and at 944 psi, the plastic area is only 3% of the clastic area. The CRACKFLO data can be used to calculate the area ratio in alpha based on the sum of the clastic + plastic areas at 1866 psi relative to the sum of the clastic area at 944 psi + plastic area at 1866 psi. This ratio is calculated to be 1.73 which compares to 2.19 calculated with both crack openings at their respective pressures. Then, the more accurate estimate l
for alpha between 1866 and 944 psiis:
l l
bP2 a = 1.73
, y,44 iAPs The product of alpha and gamma then becomes 2.28. That is, the leak rate after 25 seconds is 1/2.28 =
0.438 of the leak rate during the peak pressure differential at about 25 seconds.
3
r l
I 1.
.When CRACKFLO calculated leak mes for 1866 psi and 944 psi are adjusted for the higher differential plastic opening described above, CRACKFLO yields a leak rate ratio of about 4 compared to the 2.28 obtained above. This indicates the conservatism in the Appendix B procedure.
3.0
'b' Factor Calculated froan Farley Pulled Tube Specimens The following are the values for the 'b' factor calculated applying the following equations to the leak rate data for Farley pulled tube specimens shown in the tables attached (Table I to 4).
M2 b
=
log (R/y) l Specimen h
R4C73 405 l
R21C22 909 R34C53 424 R2C85 674 - 706 R28C35 868 - 953*
(* Measurement at Ap = 1906 psi excluded)
The largest value for 'b' based on the above data is 953.
4.0 Thermal Hydraulic Conditions for a Locked Rotor Event The primary and secondary pressure transients predicted for a locked rotor event show that the primary system pressure and primary-to-secondary pressure differential peak at about 3 seconds and become ner.rly constant after 25 seconds. The transient is disided into 2 periods.
O to 25 seconds l
l The conditions at the 3 seconds peak were applied for this time period.
i Primary pressure 2702 psia
=
Secondary pressure =
838 psia Primary temp.
584*F
=
l Sat. pressure at primary temp. = 1368 psia l
Primary-to-secondary diff. pr. = 1866 psid l
l
> 25 seconds Quasi-steady conditions predicted after 25 seconds.
Primary pressure 2132 psia
=
l kond=y pressure =
1188 psia Primary temp.
583
- F
=
Sat. pressure at primary temp. = 1360 psia r
Primary-to-secondary diff. pr. = 944 psid 4
l Steam Line Break Conditions Primary pressure 2575 psia
=
haAay pressure =
15 psia Primary temp.
607
- F
=
Sat. pressure at primary temp. = 1625 psia Primary-to-secondary diff. pr. - 2560 psid 5.0 Ratio ofleak Rate at SLB condition to Locked Rotor Condition 9 to 25 seconds a factor is given by, (2560 - is66) a = 10 oss a=
5.3 Neglecting the effect of temperature on material properties, p factor is given by r-p = )El Ps 42.1
=
44.4 0.97
=
t l
7 actor is given by, f
(2575-1625) 2560 7,1 (2702 - 1368) 1 I
1866 0.72 y
=
The above calculation utilizes Farley plant specific L value of 607 'F for a SLB event.
l Therefore, the ratio ofleak rates at SLB to locked rotor condition for 0 to 25 seconds is:
R = U"' = apy 5.3 x 0.97 x 0.72 3.7
=
=
IRia 5
I l
.> 25 seconds l
l Based on the evaluation gm.e4 here in Sectuwt 2.0, leak rate at > 25 seconds is acrima'ad to be 1/2.28 of the estimate for 0 to 25 seconds Amordmgly, the ratio ofleak rate at SLB conditions to leak rate at locked rotor conditions for 25 seconds is (3.7x2.28 = 8.4).
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