L-82-179, Responds to NRC Requesting Addl Info Re Pressurized Thermal Shock
| ML17341B139 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 05/03/1982 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR L-82-179, NUDOCS 8205070203 | |
| Download: ML17341B139 (22) | |
Text
REGULATORY QRMATION DISTRIBUTION SY.
M (RIDS)
ACCESSION NBR:8205070203 DOC CHOATE-'2/05/03 NOTARI'ZED:
NO FACIL:50"251 Turkey-Point Planti 'Unit 4i Florida Power and Light C
AUTH'AME AUTHOR AFFILIATION UKRIGg R ~ E ~
Flor ida Power 8 Light Co ~
RECIP./ATE RECIPIENT AFFILIATION VARGAgS ~ AD Operating Reactors Branch 1
SUBJECT:
Responds to NRC 820316 ltr requesting addi info re pressuri'zed thermal shock.
DOCKET 05000251 DISTRIBUTION CODE;.
A049S SCOP lES 'RECEI VED:LTR,
~
ENCL TITLE: Thermal Shock ~to Reactor Vessel NOTES:
S IZE:.
RECIPIENT ID CODE/NAME ORB ¹1 BC 01 INTERNAL ACRS ABBOTTgE AEOD COM LIANg B MURLEY T NRR DIR NRR HAZEL<<TON NRR KLECKER NRR OREILLYiP NRR TKROMzE NRR/DE OIR NRR/DHFS OIR NRR/OL OIR NRR/DL/DRAB 11, NRR/DSI/RAB DIR 05 RES VAGINSg t RES/DRA
<<COP IES LTTR ENCL 7
7 1
1 1
1 1
1 1
1 1
1 1,
1 1
1 1
1 1
1 1
1 1
1 1
1 1
0 1
1 1
1 1
1 1
1 1
RECIPIENT ID CODE/NAME 12 ACRS IGNEgE COM AUSTIN ELD NRR CLIFFORD NRR GOODHIN g E NRR JOHNSON NRR LOISt L NRR RANDALL NRR VI SS ING r G04 NRR/DHFS DEPY09 NRR/DHFS/<<PTRB NRR/DL/ADSA NRR/OSI DIR NRR/DS I/RSB NRR/OST/GIB RES BASDEKA'S RES/DET RGN2 COPIES LTiTR ENCL 1
1 1
1 0
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
1 1
EXTERNAL: ACRS NRC POR NTIS 10 02 16 1
1 1
LPDR NSIC 03 06 1
1 1
TOTAL NUMBER OF, COPIES REQUIRED:
LTTR 63 ENCL 61
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FLORIDAPOWER 5 LIGHTCOMPANY May 3, 1982 L-82-179 Office of Nuc1 ear Reactor Regul ati on Attention:
Mr.'teven A. Varga, Chief Operating Reactors Branch gl Division of Licensing U.S.
Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Varga:
Re:
Turkey Point Unit 4 Docket No. 50-251 Pressurized Thermal Shock
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sec rz We have reviewed your letter dated March 16, 1982 requesting additional information regarding pressurized thermal shock.
The following are our responses:
1.
Provide the followin information related to fluence determi nation:
(A)
Plant specific information which would allow determination of the pressure vessel fluence.
Such information should contain as built core and pressure vessel dimensions, regional material compos>tion and neutron source for a two-dimensional (R-6) and (R-Z) neutron transport solution, and FPL Response Refer to Attachment A for as built core and reactor pressure vessel dimensions.
Refer to Attachment B for required material composition and neutron source.
(B)
Plant specific values of the pressure vessel fluence and its estimated uncertainty.
FPL Response Refer to FPL letter L-81-462'ated October 21, 1981 page one (1) of the attachment for specific values of fluence.
Refer to WCAP-10019 page 41 for estimate of'ncertainty.
2.
Concerni n 0 erator Action (A)
In your evaluation, the actions described do not provide the operator with clear direction for dealing with conflicting concerns that need to be evaluated when considering the operation of HPI and the charging flow as it relates to vessel integrity and o4 ig20507ti~Gd eeoOPjt Ql pisit. mDDGY, <
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PEOPLE... SERVING PEOPLE
a~
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P, '<f~; i 1
l Re:
Turkey Point Unit 4 Docket No. 50-251 Pressurized Thermal Shock maintaining core cooling.
Provide an evaluation of the need and effectiveness of procedure modifications to clearly identify the concerns in the emergency operating procedures themselves.
This should be done in contrast of depending upon upgrading operator training alone.
FPL Response We have reevaluated our Turkey Point Unit Nos.
3 and 4 Emergency-Operating Procedures using the criteria for procedure review presented in Section 2 of the NRC Staff audit report for the Carolina Power and Light Company's Robinson Unit No.
2 of April 15, 1982.
Because the Turkey Point Emergency Operating Procedures are based on the Westinghouse Owners Group Procedure guidelines we conclude the procedural conclusions in this audit report would apply equally.
Accordingly, we intend to revise our Emergency Operating Procedures to address the NRC Staff concerns regarding pressurized thermal shock by implementing a short term and long term program.
The NRC'concerns with the existing procedures will be resolved in our long term program by the resolution of THI Action Item I.C.
1 with the Westinghouse Owner's Group both from a technical and a
human factors standpoint.
The short term program involves a modification and revision of our existing procedures to address the NRC Staff concerns that the existing Emergency Operating Procedures are weighted toward core cooling concerns and do not properly address vessel integrity.
We specifically plan to reduce the 2000 psig safety injection termination criteria and to provide additional guidance to the operator for stabilizing temperatures in the reactor coolant system following a steamline break or feedwater line break (Loss of Secondary Coolant Accident).
Provide a formal commitment to upgrade operator understanding of Pressurized Thermal Shock to the reactor pressure vessel.
FPL Response Florida Power 8 Light Company commits to implementing an augmented training program focusing on the PTS issue with the emphasis on balancing the operator actions with regard to the safety systems to mitigate the consequences of transients and accidents.
Presently classroom training concurrent with simulator training is being provided to licensed operators and shift technical advi sors.
3.
Concerning In ut Data and Assum tions 3 ~
Provide a description of the models or data used for:
(a)
Heat sources (or sinks),
(b)
Decay heat, (c)
ECC and feedwater temperatures (enthalpies) and flow rates, d)
Primary and secondary relief capacities, e)
Empirical correlation coefficients used for PTS evaluations, (f)
Operator
- Actions, (g)
Initial conditions.
Re:
Turkey Point Unit 4 Docket No. 50-251 Pressurized Thermal Shock 3.2 FPL Response Refer to WCAP 8945, May 1977, transmitted with our letter L-77-292 of September 14, 1977 and our letter L-82-26 of January 21, 1982 for Large LOCA and Large Steam Line Break assumption and conditions.
Refer to WCAP 10019 and our letter L-82-26 of January 21, 1982 for Small LOCA and Small Steam Line Break assumptions and conditions.
Provide a list of all transients or accidents by class (for example:
excessive feedwater, operating transients which result from multiple failures including control system failures and/or operator error, steam line break and small break LOCA) which could lead to inside vessel fluid temperatures of 300' or lower.
Provide any Failure Modes and Effects Analyses (FHEAs) of control systems currently available or reference any such analyses already submitted.
Estimate the frequency of occurrence of these events and provide the basis for the estimates.
Discuss the assumptions made regarding reactor operator actions.
For a given initiating event, potential multiple and consequences failures need to be considered to identify those transients which could 1 ead to a
PTS problem.
3.3 FPL Response This information will be provided directly to the NRC Staff by the Westinghouse Owners Group on May 20, 1982.
,Identify all potential PTS events which have occurred at your facility.
Include a designation of the operator actions and identify potential additional failures (including operator) which could-have resulted in a more severe event.
FPL Response We have reviewed Turkey Point Unit No.
4 records since initial operation using the following criteria for such PTS events:
1.
Cooldown transient with reduced pr essure and temperatures, 2.
Duration of severe cooldown transient greater than 10 minutes, 3.
Change of RCS temperature greater than 100'F, and 4.
Conditions 'existed for repressur ization following the cooldown transient.
Our review has revealed no PTS events that meet the above criteria.
Because the initiating events reviewed involved equipment failures and operator actions that did not lead to PTS events, it is our opinion that neither additional multiple failures nor personnel errors (of omission or commission) should be postulated.
Re:
Turkey Point Unit 4 Docket Ho. 50-251 Pressurized Thermal Shock 4.
Concerning a Review of 0 crating Histor Review your operating history at your plant and identify events which have resulted in exceeding 'the cooldown rate of 100'F/hr.
as well as those events which could have exceeded the cooldown rate limit if not mi tgated by plant controls or operator actions.
FPL Response We have reviewed plant records and have found no events that exceed the cooldown rate of 100'F/hr.
Very truly yours, Robert E. Uhrig Vice President Advanced Systems and Technology REU/JEH/mbd Attachment cc:
J.P. O'Reilly, Region II Harold F. Reis, Esquire
Re:
Turkey Point Unit 4 ~
Docket No. 50-251 Pressurized Thermal Shock ATTACHMENT A We have reviewed our as built drawings and reference documents We conclude that the dimension shown on Figure Nos.
1, 2, and 3 represent the Turkey Point Unit No.
4 Vessel.
Specifically, the eccentricity of the inner radius of the reactor pressure vessel shown on Figure 1, is no greater than 0.05 cm (0.011 in.).
The surveillance capsule slots are less than 1'rom the dimension shown.
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I 1.380 A2 2.959 1.849
- 6. 591
)1 3.4 16 4.826 thermal shield Figure 2
- Surveillance capsule dimensions (cmj
A3 27O' po/
lp 30 OO IO
) 00 gpO THERMAL SHIELD REACTOR YESSEL Figure 3
Arrangement ot Surveillance C~-'sules
ATTACHMENT 8 PLANT SPECIFIC FLUENCE DATA Design basis reactor core power distributions which were used by Westinghouse in performing fast neutron fluence evaluations for the Turkey Point Unit 4 pr essure vessel were supplied to the NRC in January 1982.
That submittal consisted of the following items:
II 1.
The material description of the reactor core presented in terms of volume fractions of solid material within a homogenized fuel zone'.
2.
Individual fuel assembly power density levels within one core octant.
This data was presented relative to a core average of 1.0.
3.
Spatial gradients of power density within each of the peripheral fuel assemblies.
This data was presented relative to an assembly average of 1.0.
4.
A time averaged relative axial distribution of power density applicable to long term fast neutron fluence projections.
The methodology used to define these long term design basis power distributions was outlined in WCAP 10019.
Also described in WCAP 10019 was the methodology by which these power distributions were used in subsequent neutron transport calculations.
Following submittal of WCAP 10019 and the 150 day letters, the NRC requested additional plant specific information which would "allow determination of the current pressure vessel fluence".
It is our understanding that this plant specific data is not intended to be used to perform fast neutron fluence projections for future operation.
Therefore, the following information is
'rovided for use by the NRC in their evaluation of the present condition of the pr essure vessel.
The material composition submitted in the January letter was based on a fuel assembly design consisting of a 17 x 17 array of zirconium clad fuel rods.
In actuality, the Turkey Point Unit 4 reactor employs a fuel design consisting of a 15 x 15 array of zirconium clad fuel rods.
A comparison of the material volume fractions for a homogenized reactor core employing each of these fuel designs is given in Table l.
An examination of Table 1 shows that the compositions of the two fuel assemblies are quite similar and in our opinion the differences will have an insignificant impact on reactor vessel fluence calculations.
Plant specific peripheral assembly power distributions for cycles 1 through 7
are tabulated in Table 2.
These data were extracted from the appropriate core design reports (WCAP's 7447,
- 8506, 8756,
- 8954, 9351,
- 9501, and 9826).
Bias factors wer e applied to the design power distributions consistent with the methodology outlined in WCAP 10019.
Also presented in Table 2 are the seven cycle time averaged power distributions for the peripheral assemblies..
These aver age distributions were obtained by burnup weighting of the individual fuel cycle data sets.
A comparison of the 7 cycle average'data with the design
of Figure 1 shows that the plant specific power distribution will result in a
somewhat lower fluence projection than that which would be calculated using the design basis distribution.
It would appear that a reduction in pressure vessel fluence on the order of 204 might be realized.
However, at this time neutron transport calculations using the plant specific power distributions have not been carried out.
These computations must be completed before any reduction in the current pressure vessel fluence can be certified.
It must also be re-emphasized that the plant specific data are applicable only for establishing the present condition of the pressur e vessel.
They should not be used to project forward in time.
An examination was also made of the variations in the power density gradients for the peripheral fuel assemblies at beginning of life and end of life for both 15 x 15 and 17 x 17 fuel rod arrays.
The conclusion of this study was that these spatial gradients, relative to an assembly average power of 1.0, were quite similar in all cases examined.
Therefor e, the gradient information previously provided in the January submittal should also be used to generate plant specific fluence values for Turkey point Unit 4.
Likewise, the time averaged axial power distribution supplied in the January letter is suitable for the current analysis.
In the January 1982 letter, comparisons of calculated and measured fast neutron flux within reactor vessel surveillance capsuley were also provided.
Since that submittal, an error was identified in the Fe~4 (n,p)
Hn5" measurements.
This error was caused by an improper calibration of the counting equipment used in the radiometric analysis.
Updated comparisons are enclosed in this submittal in Table 3.
The agreement between tne updated measurements and the calculations is excellent.
CONPARISON OF TURKEY POINT UNIT 4 ANO DESIGN BASIS PERIPHERAL POINTER DISTRIBUTIONS
B2 Table 1
Material Composition of Reactor Core Region Material Volume Fraction Design Basis Turkey Point Water UD2 Zirc - 4 Inconel - 718 Stainless Steel
.58864
.29967
.10035
.00281
.00062
.5994
.3062
.0842
Table 2
Turkey Point Unit 4 Peripheral Power Assembly No.
.85
.67 1.08
.96
.67
.98
.71
.88
.79
.65 1.18
.95
.66 1.11
.67
.78 3
.68
.71 1.17 1.11
.76
.97
.79
.98 Cycle 4
.53.
.65 1.22
.97
.43 1.18
.82 1.02
.43
.37 1.17
.68
.65 1.17
.51 1.19
.81
.59 1.11
.98
.44 1.08
.47 1.05
.78
.53 1.21 1.05
~ 44 1.12
.45 1.13
~AY
~
.73
.62 1.15
.98
.57 1.08
.83
.99 Burnup 13900 8700 7400 9000 4250 13000 8850 Note:.
The fuel assembly numbers refer to core positions designated in Figure 4 of the January 1982 submittal.
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B4 Table 3
Comparison of Measured and Calculated Fact Neutron Flux for Surveillance Capsule S
Reaction and Axial Location Fe54 n.p Mn54 Top Middle Bottom Average Radial Location
~cm 191.46 191.46 191.46 Adjusted Saturated Activity d s/ms 5.76 (3)'.59 (3) 5.71 (3)
Fast Neutron Flux (n/cm2 - sec)
Ca sule S
Calculated 1.32 (11) 1.28 (11)
.1.31 (11) 1.30 (11) 1.19 (11)
Top Middle Bottom Average 192.46 192.46 192.46 6.21 (3) 5.69 (3) 6.07 (3) 1.43 (11) 1.30 (11) 1.39 (11) 1.38 (11) 1.19 (11)