ML20041F344
| ML20041F344 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 03/12/1982 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Miraglia F Office of Nuclear Reactor Regulation |
| References | |
| SBN-232, NUDOCS 8203160413 | |
| Download: ML20041F344 (16) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _
suam sum IPUBLIC SEAVICE
- .-g.=w osc Companyof NewHampshre 1671 Worcester Road Framingham, Massachusetts 01701 (617) - 872-8100 March 12,1982 W
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United States Nuclear Regulatory Commission
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C}\\ ' et c.d" Washington, D. C. 20555 T:
Attention:
Mr. Frank J. Miraglia, Chief I.h A'
Licensing Branch #3
/,', ;7 -'
Division of Licensing L *'
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket No s. 50-443 and 50-444 (b) USNRC Letter, dated February 12, 1982, " Request for l
Additional Information," F. J. Miraglia to W. C. Tallman Su bjec t :
Responses to 492 Series RAIs; (Core Performance Branch)
Dear Sir:
i We have enclosed responses to the subject RAIs, which you forwarded in Reference (b).
Ve ry truly yours, YANKEE ATOMIC ELECTRIC COMPANY s
John DeVincentis
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Project Manager JDV : ALL: dad Enclosure f
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-t QUESTION 492.2 Operating experience on two pressurized water reactors, not of Westing-house design, indicate that a significant reduction in the core flow rate can occur over a relatively short period of time as a result of crud deposition on the fuel rods.
In establishing the Technical Speci-fications for the Seabrook Units, we will require provisions to assure that the minimum design flow rates are achieved.
Therefore, provide a description of the flow measurement capability for the Seabrook Units as well as a description of the procedure to measure flow.
RESPONSE TO QUESTION 492.2 There has been no case reported to Westinghouse of signif-icant flow reduction in a relatively short period of time due to buildup of cruo on the fuel rods at any Westinghouse plant. Additionally, there has been no report to Westinghouse of a significant flow reduction in a rela-tively short period of time for any reason (excluding steam generator tube plugging) at any Westinghouse plant.
Therefore, Westinghouse-is of the opinion that this portion of the question is not applicable. To address flow measurement technique and frequencies, please see the attached technical specification and discussion of flow measurement techniques which require a monthly calorimetric flow measurement.
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ATTACHMENT TO QUESTION 492.2 1668Q:1
2-IMPROVED ACCURACY RCS FLCM MEASUREMENT TECHNICUE Sceci~!:ati:n 3.2.3, RCS Flow Rate and 0, in the Standard Technical 5:ecifi-03ticr.s recu: es ha total reac:cr ficw (total ficw through the vessel frcm all iccas) be above scme minimum value and if abcVe tha: minimum value allcws a trade off between-red bow penalty and reactor ficw.
The minimum ficw value is thermal design ficw corrected for ficw measurement uncertainties.
Histcrically the uncertainty has been specified as 3.5%.
Flow measurement uncertainties much less than this can be achieved hcwever by using modern statistical error combination techniques and a calorimetric flow measurement' method.
The accuracy claimed for this technique depends primarily on the measurement procedure employed and on how well the instrument errors are understoon and controlled by plant eersonnel.
The calorimetric flow calculation, the measurements required and the measurement uncertainty analysis are described ir, the folicw-ing paragraohs and tables.
Reactor coolant loop flow is determined from the steam generator thermal Output, corrected for the loop's share of the. net pump heat input, and the enthalpy i
rise (ah) of the coolant.
Total reactor flow is the sum of the individual loop ficws.
Table 1 lists the calorimetric equations and defines the terms.
To establish the overall ' flow measurement' uncertainty, the accuracy and relationship to flow of each instrument used for the calorimetric measurements (see Table 2) must be determined.
Ir. most cases there are several cc ?cnents (transducer, converter, isolator, readout device, etc.) which contribute to the overall uncertainty of the measurement.
Table 3 provides a list of typical components involved in the calorimetric loop flow measurement, a corresponding conservative instrument error allcwance and the effect of tne instrument error allowance on the calculatad ;cwer or flow value.
The overall ice? fl;w measure-ment uncertainty is the statistical ccmbination of the individual uncertainties
~and appears at the bottom of Table 3.
Total reactor ficw measurement uncer-tainty is the statistical comtfination of the individual loop flow uncertainties and also appears at the bottom of' Table 3.
In summary, individual loop ficw is determined by performance of a calorimetric and these values summed to arrived at total reactor ficw.
The measurament 4
uncertainty is determined by statistically ccmbining individual component and iccp uncertainties.
A calorimetric flow measurement mus be cerformed to take credit for this particular measurement unter:ainty.
TABLE 1 REACTOR C00LAMT LCOP FLOW CALCULATION
[033 - Q +([Q)]'/c g
0.12 4 g
=
[h
-h]
g c
Loop ficw (gpm) where:
U
=
Steam generator thermal output (Btu /hr.)
Q
=
SG Primary system net heat losses (Stu/hr.)
Q
=
N' ilumber of loops
=
Reactor coolant pump heat adder (Stu/hr.)
Q
=
p Hot leg enthalpy (Stu/lb.)
h
=
g Cold leg enthalpy (3tu/lb.)
h
=
C CON leg specific voiume (cu. h./Ya.)
'I
=
c 03g. = (h -h)W s
7 p -
4.
Steam enthalpy (Stu/lb.)
where:
h
=
s Feedwater enthalpy (Stu/lb.)
h
=
p Feedwater ficw (ib./hr.)
W
=
p 4
KF # P #'?
Wp a
F
T
=
Feedwater venturi flow coefficient where:
K
=
F, Feedwater venturi correction for thermal expansion
=
Feedwater density (ib./cu. ft.)
P
=
g Feedwater venturi pressure drop (inches H O)
=
2 c
i
TABLE 2 MEASUREMENTS REQUIRED Parameter Instrument Functi on 1.
Feedwatar venturi Barton gauge feedwater fl.ow pressure differential 2
Feedwater temperature RTD feedwater enthalpy and.
density venturi thermal expansion 3.
Steam pressure Transducer steam enthalpy 4.
Reactor coolant T Narrow range RTD RCS hot leg enthalpy hot 5.
Reactor coolant T Narrow range RTD RCS cold leg enthalpy cold RCS s;ecific yciume 6.
Reactor coolant pressure Transdneer-RCS enthalpy and specific volume Other information. required for the calculation is as folicws:
7.
Feedwater venturi coefficient from vendor calibration.
8.
Steam generator bicwdown secured during the measurement.
9.
Primary system heat losses and pump' heat input o'otained from calculations.
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iABLE 3 CALORIMETRIC FLO'cl MEASUREME?iT U:!CERTAI:lTIES Uncerninty Instrument 5 Power or.
Ccmconent Error i Flow Feedwater Flcw Venturi K
-+ 0.5% K
-+ 0.5%
Thermal Expansion coefficient Temperature
+ 2.0*F Material T 5.0%
+ 0.06%
Density Temperature
+ 2.0 F
+ 0.09%
Pressure T 60 psi DP Cell Calibration
+ 0.5%
+ 0.39%
_ 1.0%
[0.78%
T OP Cell Reading Uncertainty Feedwater Enthalpy Temperature 1 2.0 F 1 0.28%
Pressure 1 60 psi Steam Enthalpy Transducer Calibration
+ 18 psi
+ 0.07%
Isolator Calibration
~ T 1.8 psi T 0.07%
' f.0.25%
[0.22%
Moisture Carryover Primary Enthalpy
+ 0.2 F
+ 0.38%
Tg RTD
~+ 0.6*F
[1.13%
Tg R/E Converter Tg Readout
. + 0.1 F
' O.19%
T 1.2 F T 2.27%
Tg Temperature Streaming T 30 psi 7 0.24%
Tg PressureiEffect TC RTO T 0.2 F T 0.31%.
TC R/E Can'verter T 0.6 F T 0.94%
T Readout-7 0.1 F E 0.16%
C T. Pressure Effect
[30 psi 10.06%
C tiet Pump Heat Aadition i 20%
i 0.0855 Total Loop Flow Uncertainty
/fe
+ 2.974%
Total Reactor Flow Uncertainty 4-loop
+ 1.5%
3-loop T 1.75%
2-1aop-E2.1%
t-
.m
. TABLE 3 (continued)
ASSU dPTIOll3 The values on cage I ca Table 3 are based cn scme scecific assum tions
- about the instrumen: and readouts.
1.
Feedwater flow is-octained frcm several" readings ~of Sar:an differential pressure guages' installed on the feedwater venturi.
2.
The measurement is performed soon after a caifbration eliminating consideration of instrument drift.
3.
Credit was taken for the 3 tap scoop RTD bypass loop in reducing.
uncertainties due to streaming.
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SB 1 & 2 FSAR POWER DISTRIBUTION LIMITS 3.'4. 2. 3 RCS FLOW RATE AND R LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant 1
System (RCS) total flow race and R, R shown on Figure 3.2-3 for 4 loop operation.2 shall be maintained within the reg Where:
H 1
1.49 l,1. 0 5 0.2 (1.0 - P)]
R1 2
l-RBP(BU)
THERMAL POWER
~
RATED THERMAL POWER F{ g d.
Measured valaes of F H obtained by using the
=
movable incore de::ec{ ors to obtain a power t
distribution map. The measured values of FN shall be used to' calculate R since Figure 3.2-3 g
includes meashroment uncertainties of 3.5% for flow and 4% foe incore measurement of F{ g, and e.
RBP (BU)
Rod Bow Penalty as: a function of. region average burnup as shown in. Figure 3.2-4, where a region is s.
defined as those assemblies with the same loading date (reloads) or enrichment (first core).
APPLICABILITY: MODE I ACTION:
s With the combination of RCS total. flow rate and R, R e
1 2 outside the region of acceptable operation shown in Figure 3.2-3..
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-,.. _---, _,+.,, _ -,--,,,- _ - -,
.n..,-
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-,-----.-,-,.n---------,-,.c
SB1&2 FSAR POWER DISTRIBUTION LIMITS ACTION:
(Continued) a.
Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
1.
Either restore the combination of RCS. total flow ract and R,t R2 to within the above limits, or 2.
Reduce THERMAL POWER to less. than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint to less than or equal to 55% of RATm THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R, R2 and RCS total flow rate t
are restored to within the above limits, or reduce THERMAL POWER to less chan SL of RATED '2HERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Identify and correct the cause of the out-of-limit condition prior c.
to increasing THERMAL POWER above the-reduced THERMAL POWELlimit requ:. red by ACTION items a.1 and/or b.. above; subsequent-POWER.
OPERATION r ay proceed provided that the combination of R1,. K2 and indicated RCS total flow rate are demonstrated, through incore flux mapping and RCS cotal flow race comparison, to be within the region of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:
1.-
A. nominal 50% of RATED TNFeMAf. POWER,.
2..
A nominst 75% of RATED THERMAL POWER,. and 3..
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.1. 3. L The provisions of Specification 4.0.4-are not applicable.
4.2.1.2.
The combination of' indicated RCk total flow race and-R1, RI shall be determined-to be within-the region of acceptable operation:of Figure.
3'. 2-3 :
o 4
3/4 2-10
S3 1 & 2 FSAR POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 1 1
Prior to operation above 75% of RATED THERHAL POWER af ter each a.
fuel loading, and b.
At least once per 31 Effective Full Power Days.
4.2.3.3 The indicated RCS total flow race shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the most recently obtained values of R1 and R, obtained per 2
Specification 4.2.3.2, are assumed to exist.
4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.
l 4.2.3.5 The RCS total flow race shall be determined by measurement at least l *.
once per 18 months.
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492.3 Regulatory Guides 1.133, Revision 1 and 1.70, Revision 3 require (4.4.6) that FSAR Section 4.4.6 contain a description of the loose parts monitoring system (LPMS) which will be installed at the Seabrook i
Station. The information that should be supplied is:
~l)
A description of the monitoring equipment including location; 2)
A description of how alert levels will be determined, including sources of internal and external noise, diagnostic procedures used to confirm the presence of a loose part, and precautions to ensure acquisition of quality data; 3)-
A description of the operation program, including signature analysis during startup, normal containment environment operation, the seismic design, and system sensitivity; 4)
A detailed discussion of the operator training program for operation of the LPMS, planned operating procedures, and recordkeeping procedures; 5)
A report from the applicant which contains an evaluation of the system for conformance to Regulatory Guide 1.133; and 6)
A ccamitment from the applicant to supply a report describing operation of the system hardware and implementation of the loose part detection prograr-
RESPONSE
Please refer to our response to Acceptance Review RAI which was provided in FSAR Amendment 44.
This was:
It is intended that Seabrook Station will be provided with a standard loose-parts monitoring system similar to those currently in use or planned for other plants. The loose parts monitoring system and loose parts detection program to be implemented will comply with Regulatory Guide 1.133.
The description of the equipment, which has yet to be purchased, will be provided to the NRC by June 1982.
Procedures for the utilization of the loose parts monitoring equipment will be developed three months prior to fuel load.
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492.4 State' your intentions with regard to N-1 loop operation.
(4.4)
RESPONSE
The initial OL being sought by PSNH does not include provisions for three-loop operation.
PSNH intends to submit the necessary i
i safety evaluations to support N-1 loop operations in the future.
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e QUESTION 492.5 The staff has developed interim criteria for evaluating the effects of rod bow on DN8 for application to the Westinghouse standard 17x17 fuel assembly.
The resultant reduction in DNBR due to rod bow is given by:
Burnup DNBR Reduction (MWD /MTU)
(%)
0 0
3500 0
5000 0
10000 2.15 15000 4.64 20000 6.74 25000 8.59 30000 10.27 35000 13.07 40000 19.09 Prior to issuance of the Technical Specifications, the applicant should present to the staff an acceptable method of accomodating the thermal margin reduction given above. Also, insert into the bases of the Tech-nical Specifications any generic or plant specific margins that will be used to offset the DNBR reduction due to rod bowing.
I' RESPONSE TO QUESTION 492.5 The DNB analyses described in the FSAR of the Seabrook 17x17 core were performed such that generic DNBR margins described in the " Revised Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors (Revision 1)",
February 16, 1977, are available for offsetting rod bow penalties. The appropriate rod bow penalty and any operating restriction in the tech-nical specifications, if required, will be addressed prior to the issuance of the Operating License of this core.
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