ML17352B149
| ML17352B149 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 01/27/1995 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17352B147 | List: |
| References | |
| CAW-95-779, NUDOCS 9505120136 | |
| Download: ML17352B149 (68) | |
Text
WestinghoUse Electric Corporation Energy Systems Nuclear and Advanced Technology Division Box 355 Pittsburgh Pennsylvania 15230 0355 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 January 27, 1995 CAW-95-779 Attention: Mr. WilliamT. Russell, Director APPLICATIONFOR WITHHOLDINGPROPRIETARY INFORMATIONFROM PUBLIC DIS LOSURE
Subject:
WCAP-13719, Rev. 1, "Westinghouse Revised Thermal Design Procedure Instruments Uncertainty Methodology For Turkey Point Units 3 &4".
Dear Mr. Russell:
The proprietary information for which withholding is being requested in the above-referenced report is further identified in AffidavitCAW-95-779 signed by the owner of the proprietary information, Westinghouse Electric Corporation.
The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Affidavitby Florida Power and Light Company.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-95-779 and should be addressed to the undersigned.
JJD/bbp Attachment cc:
Kevin Bohrer/NRC(12HS)
N. J. Liparulo, Manager Nuclear Safety Regulatory &Licensing Activities NSRLAo38L/WCAP13119 9505120186 950505 PDR ADQCK'5000250 PDR
CAW-95-779 AFFIDAVIT COMMONWEALTHOF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared Henry A. Sepp, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidaviton behalf of Westinghouse Electric Corporation ("Westinghouse" ) and that the averments of fact set forth in this Affidavitare true and correct to the best of his knowledge, information, and belief:
Sworn to'nd subscribed before me this ~~
dey of
, 1995 Henry A. Sepp, Manager Regulatory and Licensing Initiatives Notary Public Notarhl Seal Rose Marte Payre, hfoe~ PuMo Monmvw Sore, sr eflhmy Covnty MyComics Expves h,tov. 4, f Member, Penr6ytroiaAmeaton of Notarfee ts39c.l JD.1:0ls79$
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4 CAW-95-779 (1)
I am Manager, Regulatory and Licensing Initiatives, in the Nuclear Technology Division, of the Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.
(2)
I am making this Affidavitin conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.
The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence ifit falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
1539C-JJD-2:01279S CAW-95-779 (a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a
competitive economic advantage over other companies.
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c)
Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)
The use of such information by Westinghouse gives Westinghouse a
competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)
It is information which is marketable in many ways.
The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
CAW-95-779 (c)
Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage.
Ifcompetitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
(iv)
The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Westinghouse Revised Thermal Design Procedure Instruments Uncertainty Methodology for Turkey Point Units 3 &4", WCAP-13719, Rev.
1 (Proprietary), January, 1995 for Turkey Point Units 3 &4, being transmitted by the Florida Power and Light (FP&L) letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk, Attention Mr. WilliamT. Russell.
The proprietary information as submitted for use by Florida Power and Light Company for the Turkey Point Site is expected to be applicable in other licensee submittals in response to certain NRC requirements for 1539C-Jl&441Z95
CAW-95-779 justification of plant-specific calculations for the uncertainties associated with Pressurizer Pressure, Reactor Coolant System (RCS) Coolant Average Temperature (T,Q, RCS flow and Reactor Power for use in the Revised Thermal Design Procedure.
This information is part of that which will enable Westinghouse to:
(a)
Provide documentation of the analyses, methods, and acceptability of previous examples wherein the Revised Thermal Design Procedure (RTDP) was used for reaching a conclusion relative to the determination of instrumentation errors for several operating parameters and the acceptability of the reactor protection system setpoints for these parameters.
(b)
Establish the relationships between various channel instrument error components and various channel instrument error allowances inasmuch as they are statistically dependent or independent.
(c)
Establish the instrument uncertainties for the RTDP parameters (pressure,
- pressure, RCS Flow, T, and Reactor Power) in terms of total uncertainty and normal, two-sided standard deviation uncertainty for each parameter.
(d)
Assist the customer to obtain NRC approval.
Further this information has substantial commercial value as follows:
i (a)
Westinghouse plans to sell the use of similar information to its customers for purposes of utilizing the RTDP methodology and meeting NRC requirements for licensing documentation.
(b)
Westinghouse can sell support and defense of the RTDP methodology.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar licensing documentation and licensing defense services 1539C.J JD-S:012795
CAW-95-779 for commercial power reactors without commensurate expenses.
Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for development and licensing of this technology.
Further the deponent sayeth not.
ATTACHMENT 1 DESCRZPTXON OF AMENDMENT REQUEST
L-95-131 Attachment 1
Page 1 of 14 DESCRXPTION OF AMENDMENT RE UEST XNTRODUCTION Florida Power and Light Company (FPL) proposes to change Turkey Point Units 3 and 4 Technical Specification (TS) 2.1.1, Reactor Safety Limit; TS 2.2, Limiting Safety System Settings Reactor Trip System Instrumentation Setpoints; TS 3/4.2.5 Power Distribution Limits DNB Parameters; TS 3/4.3.2 Engineered Safety Features Actuation System Instrumentation and the associated BASES.
The proposed revisions to the Technical Specifications implement (a) the implementation of the NRC approved Westinghouse's Revised Thermal Design Procedure (RTDP) and (b) a revision to the Steam Generator Water Level Low-Low trip setpoint.
The Revised Thermal Design Procedure statistically combines the uncertainties on power, pressure, temperature, and flow, with the Departure from Nucleate Boiling (DNB) correlation uncertainty to calculate the Departure from Nucleate Boiling Ratio (DNBR) design limit.
Assumptions used in generating the new DNB limit also include reduced Reactor Coolant System flow, and an increase in F.
These assumptions are conservative with respect to plant operation and a separate Technical Specification submittal is being considered to implement these changes following completion of the reanalysis of the Large Break Loss of Coolant Accident
[LBLOCA].
As a result of the changes to the input assumptions and methodology used to determine the DNBR limit, new core thermal limits and Oveztemperature
~T and Overpower
~T reactor trip setpoints were calculated.
Additionally, a relaxation in the DNBR monitoring Technical Specification is proposed which more accurately reflects the intent of the specification and capability of monitoring DNB parameters.
The Steam Generator Narrow Range Water Level Process Measurement Accuracy (PMA) term and corresponding protection system setpoints have been recalculated to account for additional uncertainties as addressed later in this submittal'he changes that will be addressed for the proposed amendments are as follows:
Overtem erature and Over ower hT Set pints and Uncertainties The RTDP was employed which generated revisions to the Core Safety Limits (TS Figure 2.1-1) and Overtemperature and Overpower dT setpoints and associated uncertainties.
The revised setpoints provide additional operating margin for the proposed Technical Specification Tables 2.2-1 and 3.3-3.
Steam Generator Process Measurement Accurac PMA The current Model 44F Steam Generator Narrow Range Water Level PMA uncertainty terms and corresponding protection system setpoints have been recalculated to account for additional uncertainties.
PMA uncertainties are based on the type of measurement that is performed but aze not directly related to the accuracy of the
I
L-95-131 Attachment 1
Page 2 of 14 measurement device; however, overall instrument channel accuracy is affected.
Reactor Coolant S stem Loss of Flow Set pints The RTDP was employed which generated additional DNB margin.
RTDP is utilized in determining additional operational margin in the DNB parameters and the Loss of Flow trip setpoint identified in Technical Specifications 3/4.2.5 and 2.2.1.
DISCUSSION In order to increase the margins associated with the DNB limits, the changes to the Technical Specifications shown in the attached aze being proposed.
The proposed revisions involve the following:
a) b)
c) d)
use of upgraded process instrumentation equipment (i.e.,
implementation of Eagle-21, etc'
),
use of the RTDP Hethodology, incorporation of additional PMA uncertainties, oz an editorial cozzection.
By surveillance procedures changes and replacement of instrumentation, it has been possible to pzovide additional operational margin to the limits associated with the measurement and indication of RCS Flow.
The RTDP takes advantage of the conservative use of statistical combination of values for reactor
Changes in hot channel factors and RCS flow cause the DNB coze limits to change.
The RTDP methodology uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure",
as approved by the NRC.
With the change to the core thermal limits, the Overpower aT and Overtemperature hT reactor trip setpoints are changed in the analysis.
The values of the setpoints used in the safety analysis as well as the Reactor Core Thermal Limits revision are reflected in the revised Technical Specification pages.
The Overtemperatuze bT reactor trip function is defined in Table 2.2-1, "Reactor Trip System Instrumentation Trip Setpoints",
of the Turkey Point Units 3 and 4 Technical Specifications.
This reactor trip function provides coze protection to prevent DNB for all combinations of pressure, power, flow, coolant temperature when pressure is within the range defined by the Pressurizer High and Low Pressure trips.
The setpoint is automatically varied with:
(1) coolant tempezature to correct for temperature induced changes in density and heat capacity of water, and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer
- pressure, and (3) axial power distribution.
The Overpower bT reactor trip function is also defined in Table 2.2-1 of Turkey Point Units 3 and 4 Technical Specifications.
This reactor trip function is designed specifically to ensure operation within the fuel centerline temperature design limit.
This is accomplished by
L-95-131 Attachment 1
Page 3 of 14 controlling the gross core thermal powez within a prescribed limit (118 percent of nominal full power)
Overpower bT provides assurance of fuel integrity (i.e.,
no fuel pellet melting and less than 1%
cladding strain) under all credible overpower conditions, limits the required range for Overtemperature hT trip, and provides a backup to the Power Range Neutron Flux High Trip.
The setpoint is automatically varied with:
(1) density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation of piping delays from the core to the loop temperature detectors to ensure that the allowable heat generation rate (kw/ft) is not exceeded.
The limits on the DNB-flow parameter assure that the parameter is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
This limit is consistent with the UFSAR assumptions and has been demonstrated adequate for maintaining the required minimum DNBR above the applicable design limits throughout each analyzed transient.
All of the limits have been recalculated with the use of the NRC approved setpoint methodology, WCAP-12745, "Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4."
The methodology used is the "square zoot of the sum of the squares" which has been utilized in other submittals to the NRC.
This methodology has also been used in WCAP-10395, "Statistical Evaluation of LOCA Heat Source Uncertainty,"
and WCAP-8567, "Improved Thermal Design Procedure (ITDP)."
RTDP uncertainties are calculated using the same methodology employed foz ITDP.
WCAP-8567 is approved by the NRC noting acceptability of statistical techniques for the application requested.
Also, various American National Standards Institute (ANSI), American Nuclear Society (ANS), and Instrument Society of America (ISA) standards approve the use of probabilistic and statistical techniques in determining safety-related setpoints (specifically ANSI/ANS Standard 58.4-1979, "Criteria for Technical Specifications for Nuclear Power Stations,"
and ISA Standard S67.04,
- 1987, "Setpoints foz Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants" ).
The methodology used in this license amendment is essentially the same as that used for South Carolina Electric
& Gas Company Virgil C.
Summer Plant in August, 1982:
approved via NUREG-0717, Supplement No.
4, "Safety Evaluation Report related to the Operation of Virgil C.
Summer Nuclear Station, Unit No. 1," Docket No. 50-395, August 1982.
FPL has included in this submittal WCAP-13719, Rev.
1, and WCAP-13718, Rev.
1, entitled "Westinghouse Revised Thermal Design Procedure Instruments Uncertainty Methodology foz Turkey Point Units 3 and 4."
These WCAPs document the plant specific application of the NRC approved RTDP methodology.
L-95-131 Attachment 1
Page 4 of 14 PROPOSED TECHNZCAL SPECZFZCATZON CHANGES FPL proposes to change the following Technical Specifications in the proposed amendments:
TS Fi ure 2.1-1 "Reactor Core Safet Limits Three Loo s in Limits as a result of using the RTDP Methodology.
Justification; The curzent licensing basis utilizes the Standard Thermal Design Procedure (STDP) which uses the W-3 DNB correlation to generate the Coze Thermal Limits.
STDP assumes that the peaking factors are in their most conservative condition.
Uncertainties on power, pressure, and RCS temperature are accounted for in the accident analysis initial conditions and are not reflected in the Core Thermal Limits.
In
- contrast, the RTDP statistically combines the initial condition uncertainties of power, temperature,
- pressure, Reactor Coolant System (RCS) flow as well as peaking factor uncertainties.
In
- addition, RTDP uses the WRB-1 DNB correlation which yields better DNB results because of its tighter range of applicability (i.e.,
coze flow, temperatures, pressures).
The net result is a large increase in DNB margin (i.e., the difference between the design limit and the safety analysis limit DNBR).
Note that there is an insignificant change in the Reactor Coze Safety Limits because they aze based on the safety analysis DNBR limit and not the design (licensing) limit DNBR.
- Thus, a comparison of the current Reactor Core Safety Limits to the proposed limits does not in any way reflect the actual gain of DNB margin that is realized by switching from STDP to RTDP.
Finally, the Reactor Core Safety Limits form the basis for the reactor protection system Overtempezature aT and Ovezpower hT setpoints.
Therefore, the justification foz changing TS 2.1-1 is to ensure that the correct basis for the Overtemperatuze hT and Overpower dT setpoints is presented, even though the gains for these setpoints are unchanged from the current Technical Specifications.
2.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tri Set pints" Functional Unit 5 Overtem erature aT:
Revise the following values:
a) b)
c) d)
Allowance (TA) from "7.2" to "9.0",
from "4 8ll to II7 8511 S from "2.5" to "3.2" and the associated "0" footnote to read "1.2% for pressurizer pressure" instead of "0.5% foz pressurizer pressure."
Justification:
Use of the RTDP Methodology provides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP
L-95-131 Attachment 1
Page 5 of 14 takes advantage of conservative use of statistical combination of values for RCS flow, temperature and pressure to calculate the DNBR limit.
Changes in hot channel factors and RCS flow cause the DNB core limits to change.
This methodology was approved by the NRC in WCAP-11397-P-A and uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure."
With the change to the core thermal limits, the Overpower and Overtemperatuze bT reactor trip setpoints are changed in the analysis.
The values of the setpoints used in the safety analysis as well as the Reactor Coze Thermal Limits revision are zeflected in the revised Technical Specifications.
The revised nominal trip setpoint, Allowable Value, and associated values have been recalculated to provide margin in preserving the revised Safety Analyses Limits as determined by the RTDP analyses.
The RTDP analyses provide additional margin above the current STDP analyses.
The revision to the 4 footnote is used to reallocate allowances for sensor drifts to accommodate the revised "S" tezm.
3.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tri Set pints" Functional Unit 6
Over ower ~T:
Revise the following values:
a)
Allowance (TA) from "5.3" to "4.3", and b)
Z from "3.1" to "3.4".
Justification:
Use of the RTDP Methodology provides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP takes advantage of the conservative use of statistical combination of values for reactor power, RCS flow, temperature and pressure to calculate the DNBR limit.
This methodology was approved by the NRC in WCAP-11397-P-A and uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure."
With the change to the core thermal limits, the Overpower and Overtemperature bT zeactoz trip setpoints aze changed in the analysis.
The values of the setpoints used in the safety analysis as well as the Reactor Core Limits revision are reflected in the revised Technical Specifications.
The revised nominal trip setpoint, Allowable Value, and associated values have been recalculated to provide margin in preserving the revised Safety Analyses Limits as determined by the RTDP analyses.
The RTDP analyses pzovide additional margins above the current STDP analyses and is allocated as defined in TS Figure 2.1-1.
L-95-131 Attachment 1
Page 6 of 14 4.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tri Set pints>>
Functional Unit 10 Reactor Coolant Flow-Low: Revise the following values:
a) b)
c)
Z fzpm n2 7n to n3 3>>
S from >>0
~ 8>> to >>1. 4' Allowable Value from >>88.7% of loop design flow" to
>>88.8% of loop design flow",
Justification:
Use of the RTDP Methodology pzovides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP takes advantage of conservative use of statistical combination of values for reactor power, RCS flow, temperatuze and pressure to calculate the DNBR limit.
Changes in hot channel factors and RCS flow cause the DNB coze thermal limits to change.
This methodology was approved by the NRC in WCAP-11397-P-A and uses the same methodology as defined in WCAP-8567,
>>Improved Thermal Design Procedure."
With the change to the core thermal limit,s, the Overpower sT and Overtemperature dT reactor trip setpoints are changed in the analysis.
The values of the setpoints used in the safety analysis as well as the Reactor Core Limits revision are reflected in the revised Technical Specifications.
The changes to the Z,
S, and the Allowable Value reflects use of the additional margin created by the RTDP and reallocation of instrument uncertainties to provide flexibilityin plant Instrumentation
& Control operations.
5.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tzi Set pints>>
Functional Unit 11 Steam Generator Water Level Low-Lpw and Functional Unit 12 Steam Feedwater Flow Mismatch Coincident With Steam Generator Water Level-Lpw:
Revise the following values:
a) b)
c) d)
Z from >>2.33>> tp n2 77>>
frpm nj 9n tp
>>2 0>>
Trip Setpoint from n>15%>> to ">10%>>,
and Allowable Value from ">13.2%>> to ">8.9%".
Justification:
Westinghouse identified that the potential exists that insufficient margin may have been included in the PMA term for Steam Generator Water Level instrumentation uncertainty calculations.
This would impact the protection functions which use this parameter, i.e.,
Steam Generator Water Level Low and Low-Low setpoints.
Previously a value of +2.0%
span was used foz this term in setpoint uncertainty calculations for all models of steam generator design.
This value was based on various process parameters which are determined by specific plant operating conditions.
L-95-131 Attachment 1
Page 7 of 14 With the inclusion of the additional PMA terms and their treatment as a bias, the existing trip setpoints for Steam Generator Water Level Low and Low-Low provide margin to protect the existing safety analysis limits.
However, small changes of less than 1.0% span aze required for the associated Z and S
values identified in the Technical Specifications.
The total allowance was also adjusted to account for the current reanalysis.
The Turkey Point Steam Generator Water Level Low-Low protection system setpoints have been recalculated to account for additional PMA uncertainties.
In previous calculations the PMA was assumed to be bounded by a 2.0% span allowance.
New calculations have been performed which explicitly identify the impact of the PMA term on the Channel Statistical Allowance.
The reduction in Steam Generator Water Level Low-Low setpoint has been analyzed in Chapter 14 analyses of the Updated Final Safety Analysis Report (UFSAR) and has confirmed that sufficient margin exists to maintain safe plant operation.
As a result of these calculations, margin has been ze-allocated to improve plant operations by reducing the low-low setpoint and accompanying Allowable Value while still maintaining a margin of safety.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tri Set pints" NOTE 1
OVERTEMPERATURE ~T:
Revise the following values:
a) b)
c) d)
e) g)
K, of "1.25", instead of "1 ~ 095",
K, of "0.016" instead of "0.0107",
Kg of "0 0011" instead of "0.000453" Footnote (1):
to read "q ~ between
-46% and +2%,"
instead of reading "q, ~ between
-14% and +10%,"
Footnote (2):
to read "q, ~ exceeds
-46%," instead of reading "q, ~ exceeds
-14%",
Footnote (3):
to read "q ~ exceeds
+2%", instead of reading "q, ~ exceeds
+10%",
and Footnote (3): the wording "the aT trip setpoint shall be automatically reduced by 2.3%" instead of reading "the aT trip setpoint shall be automatically reduced by 1.5%"
Justification:
Use of the RTDP Methodology, provides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP takes advantage of the conservative use of statistical combination of values for reactor power, RCS flow, temperature and pressure to calculate the DNBR limit.
This methodology was approved by the NRC in WCAP-11397-P-A and uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure."
With the change to the core thermal limits, the Overpower
~T and Overtemperatuze hT reactor trip setpoints are changed in the analysis.
The values of the setpoints used in
L-95-131 Attachment 1
Page 8 of 14 the safety analysis as well as the Reactor Core Limits revision are reflected in the revised Technical Specifications.
The existing setpoints are relaxed to provide operational benefits taking advantage of the additional margin gained by implementing the Revised Thermal Design Procedure methodology.
The increased value of Ky to 1.25 (increased from 1.095) permitted f(aI) wings of -46% and
+2% (versus
-14% and
+10%)
which continue to maintain sufficient margin between the Relaxed Axial Offset Control (RAOC) operating band and the f(aI) penalty deadband.
8.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tzi Set pints" OVERTEMPERATURE aT NOTE 2:
Change the
% of instrument span from "1.5%" to "0.73%"
~
Justification:
Use of the RTDP Methodology provides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP takes advantage of the conservative use of statistical combination of values for reactor power, RCS flow, temperature and pressure to calculate the DNBR limit.
Changes in hot channel factors and RCS flow cause the DNB core limits to change.
This methodology was approved by the NRC in WCAP-11397-P-A and uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure."
With the change to the coze thermal limits, the Overpower
~T and Overtemperatuze hT reactor trip setpoints are changed in the analysis.
The values of the setpoints used in the safety analysis as well as the Reactor Core Limits revision aze reflected in the revised Technical Specifications.
Note 2 has been revised to reflect changes to the Allowable Value as calculated utilizing the methodology defined in WCAP-
- 12745, "Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4."
Changes to the Allowable Value reflects use of additional margin created by the RTDP and reallocation of instrument uncertainties to provide flexibility in plant Instrumentation
& Control operations.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tzi Set pints" NOTE 3:
OVERPOWER aT:
Revise the following values:
a)
K, from "1.09" to "1.10", and b)
Kg from "0 00068" to "0 00232".
Justification:
Use of the RTDP Methodology provides additional operating margin to Turkey Point's Technical Specifications by generating increased DNB margin.
The RTDP takes advantage of the conservative use of statistical combination of values for reactor
- power, RCS flow, temperature and pressure to calculate the DNBR limits Changes in hot channel factors and RCS flow cause the DNB coze limits to change.
This
L-95-131 Attachment 1
Page 9 of 14 methodology was approved by the NRC and uses the same methodology as defined in WCAP-8567, "Improved Thermal Design Procedure."
With the change to the core thermal limits, the Overpower
~T and Overtemperatuze dT reactor trip setpoints are changed in the analysis..
The values of the setpoints used in the safety analysis as well as the Reactor Core Limits revision are reflected in the revised Technical Specifications.
The existing setpoints aze relaxed to provide operational benefits taking advantage of the additional margin gained by implementing the RTDP methodology.
TS Table 2.2-1 "Reactor Tri S stem Instrumentation Tri Set pints" NOTE 4:
Change 0 of instrument span from "1.4%" to lf0 4Q ll Justification:
Note 4 has been revised to reflect changes to the Allowable Value as calculated utilizing the methodology defined in WCAP-12745, "Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3 and 4."
Changes to the Allowable Value reflects use of additional margin created by the RTDP and reallocation of instrument uncertainties to provide flexibilityin plant Instrumentation and Control operations.
10.
TS SURVEIILANCE RE UIREMENT 4.2.5.1:
Revise to read as follows:
4.2,5.1 Reactor Coolant System T,> and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 RCS flow rate shall be monitored for degradation at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Justification:
Technical Specification 3/4.2.5 requires that the flow parameter of 277,900 gpm be verified within the limits once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The value of 277, 900 gpm foz reactor coolant system flow is the precision flow calorimetric limit and was not intended to be utilized as the limit for the daily shift surveillance.
The SURVEILLANCE REQUIREMENTS and BASES of TS 3/4.2.5 are revised to provide clarification of the surveillance.
This surveillance requirement is similar to that previously submitted and appzoved by the NRC foz the Shearon Harris and Vogtle plants Technical Specifications.
TS SURVEILLANCE RE UIREMENT 4.2.5.3:
Revise to read as follows:
4.2.5.4 After each fuel loading, and at least once per 18
- months, the RCS flow rate shall be determined by precision heat balance after exceeding 90%
RATED THERMAL POWER, The measurement instrumentation shall be calibrated within 90 days prior to the perfozmance
L-95-131 Attachment 1
Page 10 of 14 of the calorimetric flow measurement.
The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
Justification:
With incorporation of the RTDP analyses, the RCS Low Flow trip setpoint of TS Table 2.2-1, Item 10, was evaluated to account foz changes in performing the precision flow calorimetric which is used for calibration of the flow transmitter which validates the trip setpoint.
Additions to the surveillance requirement provide clarification to the acceptable methodology of performing the flow rate verification.
12.
TS Table 3.3-3 "En ineered Safet Features Actuation S stem Instrumentation Tri Set pints" Functional Unit 1
4 and 8
T ~
-Low:
Revise the Z value from "2.0" to "2.1".
Justification:
With the incorporation of the RTDP analyses, the T,,--Low setpoint was evaluated to account for changes to the temperature related setpoints.
The TS setpoint remains the
- same, however, the five column Westinghouse setpoint methodology dictates a revision to the Z value.
13.
TS Table 3.3-3 "En ineered Safet Features Actuation S stem Instrumentation Tri Set pints" Functional Unit 6
Steam Generator Water LevelLow:
Revise to reflect the following values:
a) b)
c) d)
Z from n2 33n to
~r2 77n from tel 9>> to n2 0n Trip Setpoint from "15%" to "10%", and Allowable Value from "13%" to "8.9%".
Justification:
Westinghouse identified that the potential exists that insufficient margin may have been included in the PMA term for Steam Generator Water Level instrumentation uncertainty calculations.
This would impact the protection functions which use this parameter, i.e.,
Steam Generator Water Level Low and Low-Low setpoints.
Previously a value of +2.0%
span was used for this term in setpoint uncertainty calculations for all models of steam generator design.
This value was based on various process parameters which aze determined by specific plant operating conditions.
With the inclusion of the additional PMA terms and the treatment of the PMA tezms as a bias, the existing trip setpoints for Steam Generator Water Level Low and Low-Low provide margin to protect the existing safety analysis limits.
- However, small changes of less than 1.0% span are required foz the associated Z
and S values identified in the Technical Specifications.
The Turkey Point Steam Generator Water Level Low and Low-Low protection system setpoints, have been recalculated to account foz additional PMA uncertainties.
The reduction in Steam
L-95-131 Attachment 1
Page 11 of 14 Generator Water Level Low-Low setpoint has been analyzed in Chapter 14 UFSAR analyses and has confirmed that sufficient margin exists to maintain safe plant operation.
14.
TS Table 3.3-3 "En ineered Safet Features Actuation S stem Instrumentation Tri Set pints
" Functional Unit G.b.
Steam Generator Water LevelLow-Low Allowable Value:
Change the signs from "<" to ">".
Justification:
By letter dated April 23, 1991, the NRC issued license amendments 140/135 for Turkey Point Units 3 and 4, respectively, to implement the replacement of the resistance thermal detector (RTD) bypass manifold system with fast-response thezmowell-mounted RTDs.
The Technical Specification pages that were transmitted with the NRC letter of April 23,
- 1991, included the correct sign of ">" for the Steam Generator Water Level-Low-Low trip setpoint
~
By letter dated August 26, 1991, the NRC issued license amendments 146/141 for Turkey Point Units 3 and 4, respectively, to implement the Westinghouse setpoints five-column methodology.
The Technical Specification pages that were tzansmitted with the NRC letter of August 26, 1991, incorrectly transposed the sign for the Steam Generator Water Level Low-Low trip setpoint.
This correction will ensure consistency between Technical Specification Tables 2.2-1 (Functional Unit 11) and 3.3-3 (Functional Unit 6.b.)
ANALYSIS OVERTEMPERATURE bT AND OVERPOWER bT Revised core thermal limits reflected in TS Figure 2.1-1 were generated employing the RTDP methodology.
Overtemperature
~T and Overpower bT setpoints and associated uncertainties were calculated based on the new core thermal limits.
A review of the Turkey Point UFSAR was performed to determine those events sensitive to changes in the Oveztemperature and Overpower dT setpoints.
Each of the events have been analyzed to determine if the various acceptance criteria were met.
Xn all cases, the acceptance criteria were met, and therefore the margin of safety is maintained.
STEAM GENERATOR PROCESS MEASUREMENT ACCURACY Westinghouse identified that the potential exists that insufficient margin may have been included in the Process Measurement Accuracy term for Steam Generator Water Level instrumentation uncertainty calculations.
This would impact the protection functions which use this parameter, i.e.,
Steam Generator Water Level-Low and Low-Low setpoints
~
The Steam Generator Water LevelHigh-High setpoint includes the revised PMA term, as approved in license amendments 163/157 for Turkey Point Units 3 and 4.
L-95-131 Attachment 1
Page 12 of 14 Previously a random value of +2.0% span was used for this term in setpoint unceztainty calculations for all models of steam generator design.
This value was based on various process parameters which are determined by specific plant operating conditions.
More recently, an improved understanding of dP level measurement system errors based on scientific work documented in an Instrument Society of America paper (G. E. Lang and J.
P.
Cunningham, "Delta-P Level Measurement Systems",
"Instrumentation,
- Controls, and Automation in the Power Industry", Vol.34, Proceedings of the Thirty-Fourth Power Instrumentation Symposium, June 1991),
has led to a reinvestigation of the Steam Generator Level PMA terms.
The conclusions aze that two other erzor components should be accounted foz explicitly (i.e.,
reference leg temperature changes from calibration temperature, and downcomer subcooling) and that fluid velocity effects should be considered.
These error components are not considered to be random in
- nature, and therefore, are treated as biases.
With the inclusion of the additional PMA tezms and their treatment as
- biases, the existing trip setpoints for Steam Generator Water Level Low and Low-Low provide ample margin to protect the existing safety analysis limits.
- However, small changes of less than 1
~ 0% span aze required for the associated Z and S values identified in the Technical Specifications.
The total allowance was also ad)usted to account for the current reanalysis.
The applicable terms are defined as follows:
Total Allowance:
Z'otal Allowance (TA) is the difference, in percent instrument
- span, between the Technical Specification trip setpoint and value used in the safety analysis limit foz the trip setpoint.
Z, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.
S:
S or Sensor Error, is either the "as measured" deviation of the sensor from its calibration point oz the value specified in TS 2.2-1, in percent
- span, to the analysis assumptions.
Trip Setpoint:
Allowable Value:
Nominal value at which the trip is set.
Allowable Value is a value chosen to accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated.
Operation with setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
L-95-131 Attachment 1
Page 13 of 14 For the Steam Generatoz Water Level Low and Low-Low trip setpoints the previous PMA value included a random independent term in the overall channel statistical allowance.
Explicit values were calculated for process pressure variations, reference leg temperature variations, with respect to the calibration conditions and included arithmetically as a positive bias or a negative bias as appropriate.
Additionally, the safety analysis was performed at 5% to support the Technical Specification nominal trip setpoint of 10% for Steam Generator Water Level Low-Low.
The basis for determination of acceptability of the existing trip setpoints and allowable values is defined in WCAP-12745, "Westinghouse Setpoint Methodology for Protection Systems Turkey Point Units 3
& 4."
DNB PAR&METER SURVEILLANCE REQUIREMENTS By improving procedures and replacement of instrumentation, FPL can gain additional operational margin to the limits associated with the measurement and indication of RCS Flow.
The RTDP calculated precision flow calorimetric uncertainty remains at 3.5% flow.
The 3,5% flow uncertainty was maintained to provide the plant flexibilityby extending the time requirement from calibration to performance of the precision flow calorimetric'lternate methods of taking data were evaluated to reduce the probability of causing spurious trips when performing calozimetrics.
The limits on the DNB-flow parameter assure that the parameter is maintained within the normal steady-state envelope of operation assumed in the tzansient and accident analyses.
The limit is consistent with the initial UFSAR assumptions and has been demonstrated adequate to maintain the required minimum DNBR above the applicable design limits throughout each analyzed transient.
In all
- cases, the acceptance criteria were met, and therefore the margin of safety is maintained.
Technical Specification (TS) 3/4.2.5, requires that the flow parameter of 277, 900 gpm be verified within the limits once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The value listed of 277,900 gpm is the precision flow calorimetric limit and wasn't intended to be utilized as the limit for the daily shift surveillance.
The SURVEILLANCE REQUIREMENT and BASES of the TS are revised to provide clarification of the surveillance.
The SURVEILLANCE REQUIREMENT is similar to that previously submitted and approved for the Shearon Harris and Vogtle plants'S
~
With the incorporation of the RTDP analyses, the RCS low flow trip setpoint of TS Table 2.2-1, Item 10, was evaluated to account for changes in performing the precision flow calorimetric used foz normalization of the trip setpoint.
The TS trip setpoint remains the
- same, however, the five column Westinghouse setpoint methodology dictates revision to the Z,
S, and Allowable Value values.
L-95-131 Attachment 1
Page 14 of 14 SUMbGLRY Revised core thermal limits reflected in TS Figure 2.1-1 were generated employing the Revised Thermal Design Procedure (RTDP) methodology.
Oveztemperature and Overpower AT reactor trip setpoints and associated uncertainties were calculated based on the new core thermal limits.
A review of the Turkey Point Updated Final Safety Analysis Report (UFSAR) was performed to determine those events sensitive to changes in the Overtemperature and Overpower dT reactor trip setpoints.
Each of the events [i.e.,
Rod Withdrawal at Power, Boron Dilution, and Loss of Load] have been analyzed to determine if the various acceptance criteria were met.
In all cases, the acceptance criteria were met, and therefore the margin of safety is maintained.
With the inclusion of the additional Pzocess Measurement Accuracy terms and their tzeatment as biases, the existing trip setpoints for Steam Generator Water Level Low and Low-Low provide ample margin to protect the existing Safety Analysis Limits.
- However, small changes of less than 1.0% span are required for the associated Z and S values identified in the Technical Specifications.
The revised values for Z
and S are reflected in the revised Technical Specifications (TS) Table 3.3-3, Engineered Safety Features Actuation System Instrumentation Trip Setpoints.
With incorporation of the RTDP analyses, the RCS low flow trip setpoint of TS Table 2.2-1, Item 10, was evaluated to account for changes in performing the precision flow calorimetric used foz normalization of the trip setpoint.
The Technical Specification trip setpoint remains the
- same, however, the five column Westinghouse setpoint methodology dictates revision to the Z,
S, and Allowable Values.
ATTACHMENT 2 DETERMINATION OF NO SIGNXFXCANT HAZARDS CONSXDERATXON
L-95-131 Attachment 2
Page 1 of 9
DETERMINATION OF NO SXGNXFXCANT HAZARDS CONSXDERATXON 1.0 CORE THERMAL LIMITS, OVERTEMPERATURE hT and OVERPOWER aT REACTOR TRIP SETPOINT Description of Pxoposed License Amendments Revised core thermal limits reflected in TS Figure 2.1-1 were generated employing the Revised Thermal Design Pzoceduze (RTDP) methodology.
Oveztempezature and Overpower LIT reactor trip setpoints and associated uncertainties were calculated based on the new core thermal limits' review of the Turkey Point Updated Final Safety Analysis Report (UFSAR) was performed to determine those events sensitive to changes in the Overtemperature and Overpower hT reactor trip setpoints.
Each of the events [i.e.,
Rod Withdrawal at Power, Boron Dilution, and Loss of Load] have been analyzed to determine if the various acceptance criteria were met.
In all cases, the acceptance criteria were met, and therefore the margin of safety is maintained.
Introduction The Nuclear Regulatory Commission has provided Standards foz determining whether a significant hazards consideration exists (10 CFR 50.92 (c)).
A proposed amendment to an operating license for a facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
Each standard is discussed below for the proposed amendments.
Discussion (1)
Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.
The revised Overtemperatuze and Overpower hT reactor trip functions do not involve an increase in the probability or consequences of an accident previously evaluated because operation with these revised values will not cause any design or analysis acceptance criteria to be exceeded.
The structural and functional integrity of all plant systems is unaffected.
The Overtemperature and Overpower hT reactor trip functions aze part of the accident mitigation response and are not initiators foz any transient.
Therefoxe, the probability of occurrence previously evaluated are not affected.
The changes to the Overtemperature and Overpower hT reactor trip functions do not affect the integrity of the fission product barriers utilized for mitigation of radiological dose
L-95-131 Attachment 2
Page 2 of 9
consequences as a result of an accident.
In addition, the off-site mass releases used as input to the dose calculations are unchanged from those previously assumed.
Therefore, the off-site dose pzedictions remain within the acceptance criteria of 10 CFR Part 100 limits for each of the transients affected.
Since it has been concluded that the transient, analyses results are unaffected by the parameter modifications, it is concluded that the probability or consequences of an accident previously evaluated are not increased.
The proposed license amendments do not create the possibility of a new or different M.nd of accident from any accident previously evaluated.
The revised Overtemperatuze and Overpower hT reactor trip functions do not create the possibility of a new or different kind of acci'dent from any accident previously evaluated because the setpoint adjustments do not affect accident initiation sequences.
No new operating configuration is being imposed by the setpoint adjustments that would create a new failure scenario.
In addition, no new failure modes or limiting single failures have been identified.
Therefore, the types of accidents defined in the UFSAR continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.
Therefore, it is concluded that no new or different kind of accidents from those previously evaluated have been created as a result of these revisions.
The proposed license amendments do not involve a significant reduction in a margin of safety.
The changes to the Overtemperature and Overpower bT reactor trip functions do not involve a reduction in the margin of safety because the margin of safety associated with the Overtemperature and Overpower hT reactor trip functions, as verified by the results of the accident
- analyses, are within acceptable limits, All transients impacted by implementation of the RTDP methodology have been analyzed and have met the applicable accident analyses acceptance criteria.
The margin of safety required foz each affected safety analysis is maintained.
This conclusion is not changed by the Overtemperature and Overpower hT setpoint modifications.
The adequacy of the revised Technical Specifications values to maintain the plant in a safe operating condition has been confirmed.
Therefore, the changes to the Overtemperatuze and Overpower dT reactor trip functions do not involve a significant reduction in the margin of safety.
L-95-131 Attachment 2
Page 3 of 9
CONCLUSIONS It has been determined that the proposed changes to the Technical Specifications to revise the Oveztemperature and Overpower hT reactor trip setpoint values identified in TS Table 2.2-1 and the Reactor Core Thermal Safety limits in TS Figure 2.1-1 are acceptable.
These revisions do not involve an increase in the probability oz consequences of an accident previously evaluated; they neithez create the possibility of a new or different kind of accident from any accident previously evaluated, nor involve a significant reduction in a margin of safety.
Therefore, it is concluded that the proposed changes do not involve a significant hazard in accordance with 10 CFR 50.92.
L-95-131 Attachment 2
Page 4 of 9
- 2. 0 STEAM GENERATOR PROCESS MEASUREMENT ACCURACY Description of Proposed License Amendments Westinghouse identified that the potential exists that insufficient margin had been included in the Process Measurement Accuracy (PMA) term foz Steam Genezatoz Water Level instrumentation uncertainty calculations.
This would impact the protection functions which use this parameter, i.e.
Steam Generator Water Level-Low and Low-Low setpoints.
Previously a random value of +2.0% span was used foz the PMA term in setpoint uncertainty calculations foz all models of steam generator design.
This value was based on various process parameters which are determined by specific plant operating conditions.
More recently, an improved understanding of hP level measurement system errors based on scientific work documented in an Instrument Society of America paper (G. E. Lang and J.
P.
Cunningham, "Delta-P Level Measurement Systems",
"Instrumentation,
- Controls, and Automation in the Power Industry", Vol.34, Proceedings of the Thirty-Fourth Power Instrumentation Symposium, June 1991),
has led to a reinvestigation of the Steam Generator Level Pzocess Measurement Accuracy terms.
The conclusions are that two other error components should be accounted for explicitly (i.e., reference leg temperature changes from calibration temperature, and downcomez subcooling) and that. fluid velocity effects should be considered.
These error components aze not considered to be random in nature, and therefore aze treated as biases.
With the inclusion of the additional Process Measurement Accuracy terms and their tzeatment as biases, the existing trip setpoints for Steam Generator Water Level Low and Low-Low provide ample margin to protect the existing Safety Analysis Limits.
However, small changes of less than 1.0% span are required foz the associated Z and S values identified in the Technical Specifications.
The revised values for Z
and S are reflected in the revised Technical Specifications (TS) Table 3.3-3, Engineered Safety Features Actuation System Instrumentation Trip Setpoints.
Introduction The Nuclear Regulatory Commission has provided Standards for determining whether a significant hazards consideration exists (10 CFR 50.92 (c)).
A proposed amendment to an operating license for a facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
Each standard is discussed below for the proposed amendments.
L-95-131 Attachment 2
Page 5 of 9
Discussion 2.
Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.
The revised reactor trip setpoint,s on Steam Generator water level do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Operation with these revised values will not cause any design or analysis acceptance criteria to be exceeded.
The structural and functional integrity of any plant system is unaffected.
The Steam Generator Water Level trip functions are part of the accident mitigation response and are not themselves initiators for any transient.
Therefore, the probability of occurrence previously evaluated is not affected.
The changes to the reactor trip setpoints do not affect the integrity of the fission product barriers utilized for mitigation of zadiological dose consequences as a result of an accident.
The Steam Generator Water Level Low-Low trip setpoint assumed in the safety analyses has been revised and acceptable results were obtained.
The Steam Generator Water Level-Iow setpoint is not credited in the safety analysis.
Consequently, the required margin of safety for each affected safety analysis has been maintained.
In addition, the offsite mass releases used as input to the dose calculations aze unchanged from those previously assumed.
Therefore, the offsite dose predictions remain within the acceptance criteria of 10 CFR Part 100 limits foz each of the transients analyses affected.
Since it has been determined that the transient analysis results are unaffected by these parameter modifications, it is concluded that the consequences of an accident pzeviously evaluated are not increased.
The proposed license amendments do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The revised TA, Z, and S Values do not create the possibility of a
new or different kind of accident from any accident previously evaluated.
The setpoint values do not affect the assumed accident initiation sequences.
No new operating configuration is imposed by the TA, Z, and S value adjustments that would create a
new failure scenario.
In addition, no new failure modes or limiting single failures have been identified for any plant equipment.
Thezefoze, the types of accidents defined in the UFSAR continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.
Therefore, the possibility of a new oz different kind of accident from any accident evaluated is not increased.
L-95-131 Attachment 2
Page 6 of 9
The proposed license amendments do not involve a significant reduction in the margin to safety.
The current Technical Specification trip setpoints and allowable values were changed to maintain the current safety analysis limits.
The Steam Genezator Water Level Low-Low trip setpoint assumed in the safety analyses has been revised and acceptable results were obtained.
The Steam Generator Hater Level-Low setpoint is not credited in the safety analysis.
Consequently, the required margin of safety foz each affected safety analysis has been maintained.
- Thereby, the adequacy of the revised Technical Specification values to maintain the plant in a safe operating condition is also confirmed.
Conclusion Based upon the preceding analysis, it has been determined that the proposed change to the Technical Specifications to modify the Steam Generator Water Level-Low and Low-Low reactor trip associated Trip
- Setpoint, Z,
S and allowable value terms does not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from an accident previously evaluated or involve a significant reduction in a margin of safety.
Therefore, it is concluded that the proposed changes do not involve a significant hazards consideration.
The revised zeactor trip functions on Steam Genezator Water Level do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Operation with these revised values will not cause any design or analysis acceptance criteria to be exceeded.
The structural and functional integrity of any plant system is unaffected.
The Steam Generator Water Level trip functions are part of the accident mitigation response and aze not themselves initiators foz any transient.
Therefore, the probability of occurrence previously evaluated is not affected.
L-95-131 Attachment 2
Page 7 of 9
3.0 DNB PARAMETER SURVEILLANCE REQUIREMENTS Description of Proposed License Amendments By improving procedures and replacement of instrumentation, FPL can gain additional operational margin to the limits associated with the measurement and indication of RCS Flow.
The RTDP calculated precision flow calorimetric uncertainty remains at 3.5% flow.
The 3.5% flow uncertainty was maintained to provide the plant flexibility by extending the time requirement from calibration to pezfozmance of the precision flow calorimetzic.
Alteznate methods of taking data were evaluated to reduce the probability of causing spurious trips when performing calorimetrics.
The limits on the DNB-flow parameter assure that the parameter is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
The limit is consistent with the initial UFSAR assumptions and has been demonstrated adequate to maintain the required minimum DNBR above the applicable design limits throughout each analyzed transient.
In all
- cases, the acceptance criteria were met, and therefore the margin of safety is maintained.
Technical Specification 3/4.2.5, requires that the flow parameter of 277,900 gpm be verified within the limits once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The value listed of 277,900 gpm is the precision flow calorimetric limit and was not intended to be utilized as the limit for the daily shift surveillance.
The SURVEILLANCE REQUIREMENT and BASES of the Technical Specifications are revised to provide clarification of the surveillance.
The SURVEILLANCE REQUIREMENT is similar to that previously submitted and approved for the Shearon Harris and Vogtle plants'echnical Specifications.
With incorporation of the RTDP analyses, the RCS low flow trip setpoint of TS Table 2.2-1, Item 10, was evaluated to account foz changes in performing the precision flow calorimetric used for normalization of the trip setpoint.
The Technical Specification trip setpoint remains the
- same, however, the five column Westinghouse setpoint methodology dictates revision to the Z,
S, and Allowable Values.
Introduction The Nuclear Regulatory Commission has provided Standards for determining whether a significant hazards consideration exists (10 CFR 50.92 (c)).
A proposed amendment to an operating license foz a facility involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not (1) involve a significant inczease in the probability oz consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
Each standard is discussed below for the proposed amendments'
L-95-131 Attachment 2
Page 8 of 9
Discussion Operation of the facility in accordance with the proposed amendments would not involve a significant increase in the probability or consequences of an accident previously evaluated.
With the retention of the previous Safety Analyses Limits for Departure from Nucleate Boiling (DNB)
(T.S. 3/4.2.5) and the existing Reactor Coolant System (RCS) low flow trip Nominal Trip Setpoint (NTS), there is no increase in the probability oz consequences of an accident previously evaluated because there is no change to any design or analysis acceptance criteria.
The structural and functional integrity of any plant system is unaffected.
The proposed license amendments revise the surveillance requirements for DNB parameters and incorporate the RTDP uncertainty analysis into the Westinghouse five column methodology for the RCS Loss of Flow determination of the Allowable Value.
The changes to the reactor trip functions do not affect the integrity of the fission product barriers utilized foz mitigation of radiological dose consequences as a result of an accident.
The margin to safety for the RCS Loss of Flow trip remains protected as the trip setpoints assumed in the safety analyses are not revised.
In addition, the offsite mass releases used as input to the dose calculations are unchanged from those previously assumed.
Therefore, the offsite dose predictions remain within the acceptance criteria of 10 CFR Part 100 limits fox each of the transients affected.
Since it has been determined that the transient results are unaffected by these parameter modifications, it is concluded that the consequences of an accident previously evaluated are not increased.
(2)
Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
The revised 2,
S, and Allowable Value numbers do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Revision of the surveillance requirements merely provides clarification to more accurately reflect the surveillance activity.
The setpoint value does not affect the assumed accident initiation sequences.
No new operating configuration is being imposed by the Z and S value adjustments that would create a new failure scenario.
In addition, no new failure modes oz single failures have been identified for any plant equipment.
Therefore, the types of accidents defined in the UFSAR continue to represent the credible spectrum of events to be analyzed which determine safe plant operation.
Therefore, it is concluded that
L-95-131 Attachment 2
Page 9 of 9
no new or different kind of accidents from those previously evaluated have been created as a result of these revisions.
3.
The proposed license amendments do not involve a significant reduction in the margin to safety.
The RCS Loss of Flow setpoint assumed in the safety analysis zemains unchanged.
Since the safety analysis limit setpoint value is unchanged and no safety analysis is affected, the required margin of safety for each affected safety analysis is maintained.
- Thereby, the adequacy of the zevised Technical Specification values to maintain the plant in a safe operating condition is also confirmed.
Therefore, the change to the RCS Loss of Flow setpoint does not involve a significant reduction in the margin of safety.
Conclusion Based upon the preceding analysis, it has been determined that the proposed changes to the Technical Specifications to modify the RCS Loss of Flow associated Z,
S, and Allowable Value terms along with the DNB surveillance requirements do not involve a significant increase in the probability or consequences of an accident previously evaluated; create the possibility of a new or different kind of accident from an accident previously evaluated; or involve a significant reduction in a margin of safety.
Therefore, it is concluded that the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.
ATTACHMENT 3 PROPOSED TECHNZCAL SPECZFZCATZONS Marked up Technical Specification Pagea; 2-2 2-4 2-5 2-7 2-8 2-9 2-10 3/4 2-16 3/4 3-24 3/4 3-26 3/4 3-27 3/4 3-30 B 3/4 2-8
675 665 2400 PSIA 655 50 PSIA UNACCEPTASLE OPERATlON 625 1
PSIA 575 0.0 t
02 03 OA 0.5 0.7 0.8 0.9
).0 t
1.2 OWER (FRACTtON F NOM(NAL)
FIGURE 2. 1-1 REACTOR CORE SAFETY LIMIT " THREE LOOPS IN OPERATION TURKEY POINT - UNITS 3 8
4 2-2 AMENDMENT NQS 137 AND 13
670 660 650 456 PS 2400 PS UNACCEPTABLE OPERATION 640 630 F 620 00 PS 610 806 PSl 600 590 580
.........AGCEFGLBLf OPERATION 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
1.1 1.2 POWER (FRACTION OF NOMINAL}
Figure 2.1-1 Reactor Core Safety Limit-Three Loops in Operation
T, t ~
1 0
p I
~
~
TABLE 2.2-1 REACTOR TRIP SYSTBl INSTRINENTATIOH TRIP SETPOINTS FUHCTIOHAL UNIT 1.
Hanual Reactor Trip 2.
Power
- Range, Neutron Flux a.
High Setpoint b.
Low Setpoint 3.
Interaediate
- Range, Neutron Flux 4.
Source
- Range, Heutron Flux 5.
OverteIperature hT 6.
Overpower hT 7.
Pressurizer Pressure-Low 8.
Pressurizer Pressure-High 9.
Pressurizer Mater Level-High ALLOWANCE TA N.A 7.5 8.3
- 13. 5 5.5 8.0 Z
S TRIP SETPOINT N.A N.A N.A.
4.56 0.0
<109X of RTP*+
4.56 0.0
<25% of RTP++'.41 0.0
<25K of RTP"*
- 10. 01 3
IZ'.0
<10s cps See Hote 1 2.0 See Note 3 1.12 1.4
>1835 psig 1.12 1.4
<2385 psig 6.8 4.0[A'92K of instruaent span 5.0 20
>90% of loop 8esign flo o
rrow r nge instrument span 10.
Reactor Coolant Flow-Low 4.6 11.
Steam Generator Mater Level Low-Low Ql "Loop design flow = 89,500 gpa
- RE = Rated Theraal Power I ~
2.0X span for hT (RTDs) an for pressurizer pressure'LLOMABLE VALUE N.A.
<112.0X of RTP**
<28.0X of RTP""
<31.(C of RTP"*
<1 4 x 10s cps See Hote 2 See Note 4
>1817 psig
<2403 psig
<92 2X of instrument span gg g of loop Bessgn flow g,g of na ow range instrument span
e e
k 1
II I
~ ~
jf I
t 1
1
~/
N
~
F 1
4
~
r:
l r
I
TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEH INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 12.
Steaw/Feedwater Flow Hisaatch Coincident Mith Steaa Generator Mater Level-Low ALLOWANCE TA
- 20. 0 5.0 1
S 3.67 7.3" 2 0 TRIP SETPOINT Feed Flow <20X range instrument span ALLOWABLE VALUE Feed Flow <23.9X below steam flow 8,f%
of narrow
~ range instrument span 13.
Undervoltage - 4.16 kV Busses A and B
20.0 1.12 0.0
>70X bus voltage
>69X bus voltage 14.
Underfrequency - Trip of Reactor 6.5 Coolant Puap Breaker(s)
Open 15.
Turbine Trip a.
Auto Stop Oil Pressure 2.6 b.
Turbine Stop Valve Closure N.A.
16.
Safety Injection Input froa ESF N.A..
0.03 0.0
>56.1 Hz 1.0 0.0,
>45 psig N.A.
N.A.
Fully Closed***
N.A.
N.A.
N.A.
>55.9 Hz
>42 psig Fully Closed**"
N.A.
17.
Reactor Trip System Interlocks a.
Interaediate Range Neutron Flux, P-6 N.A.
N.A.
N.A.
Nominal 1x10-amp
>6.0x10-'mps
- "*Liaitswitch is set when Turbine Stop Valves are fully closed.
H1.7X span for steaa line flow, 2.9X span for feedwater flow and 2.8X span for steam line pressure.
S L
f H
II F
h
~
~
h9oM I
M Ul C
(
1
)
1 + T3S NOTE 1:
OVERTENPERATURE hT hT 1 + TiS 1 + TzS Mhere:
1 + T>s
=
Lead/Lag 1 + TzS TABLE 2.2-1 Continued TABLE NOTATIONS 2~) [T (
1
) - T'] + K3(P P') Pq ())())
(
1 + T5S) 6T by RTD Instrumentation compensator on measured hT; T>=Os, Tz =Os 1 + T3S hTo Lag compensator on measured 6T; T3 = Os Indicated hT at RATED THERNAL POMER I)ZE Kz 4-1+T S 1 + T5S T4P T5 oF.
Q,ol g The function generated by the lead-lag compensator for T dynamic compensation; Time constants utilized in the lead-lag compensator for T
, T4
- 25s, avg' T5 = 3 SP Average temperature, F;
o Ul 1+ T5S Lag compensator on measured T~vg T6 Os T'
574.2 F (Hominal T at RATED THERNAL POMER);
K3
/PSig; 0
K OO L i Pressurizer
- pressure, psig;
~'
ll
. ~
~
NOTE 1:
(Continued) pl
. TABLE 2. 2-1 Continued TABLE NOTATIONS Continued 2235 psig (Noainal RCS operating pressure);
Laplace transfora operator, s-~.
and fq (hI) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chaibers; with gains to be selected based on measured instrument response during plant startup tests such that:
o (1)
For q - qb between~and fs (al) = 0, where qt and q are percent RATER THERHAL PNER in the top and bottoa halves of the core respectivel nd qt+ qb is total THERNL PNER in percent of RATED THERHAL PNER;
-+6 l~
(2)
For each percent that the aagn$ tude of qt - qb exceeds the hT Trip Setpoint shall be autoaatically reduced by 1.5X of its value at RATED THERNL PNER; and
~ Z y (3)
For each percent that the aagnitude of qt - qb exceeds
, the hT Trip Setpoint shall be autoaatically reduced by of its value at RATED THERNL POWER.
z.s%
NOTE 2:
The channels aaxiiw trip setpoint sha not exceed its coiputed setpoint by sore than of instrlnt span.
q 3 s)]r
'I o
~ l I
I
'N f
V
~
~,
~
~
H r
J 4
4
\\
-TABLE 2.2-1 Continued TABLE NOTATIONS Continued NOTE 3:
OVERPNER hT OT 1+sS 1
<OT (K.
K ~
T" K [T(
T")
fs (OI))
+ ss
+ ss o
s s
1+ xsS (1+ zsS) s (1+ zsS)
Mhere:
hT As defined in Note 1, Ss defined in Note 1, 1+xS 1+ tgS As defined in Note 1, As defined in Note 1, 1
~ l c)
Ks 0.02/'F for increasing average teaperature and 0 for decreasing average teaperature, xS
+
=
The function generated by the rate-lag compensator for T dynaaic compensation, avg Tiae constants utilized in the rate-1ag compensator for T
, tz
> 10 s, 1
avg'+
=
Ss defined in Note 1,
NOTE 3:
(Continued)
Q,oozBZ TABLE 2.2-1 Continued TABLE NOTATIONS Continued TN
'F for T > T" for
< T",
As defined in Note 1, Indicated T at RATED THERMAL PNER (Calibration teayerature for 4T instr~ntation,
< 574.2'F),
hs defined in Note 1, and fg (hI)
=
0 for all hI NOTE 4:
The nel's aaxflw trip setpoint shall not exceed its computed trip setpoint by aore than of instrment span.
O,g
p4 g
\\
4l k'
POWER DISTRIBUTION LIMITS 3/4. 2. 5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits:
a.
b.
Pressurizer Pressure
> 2209 psig",
and c.
>277,900 gpm APPLICABILITY:
MODE 1.
ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5X of RATED 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 />.
SURVEILLANCE RE UIREMENTS
- 4. 2. 5gj The RCS flow rate indicators shall be subjected to a.CHANNEL
~
~
CALIBRATION at least once per 18 months.
4.2.5.
iWS~IE~ (C.) hm.e "Limit not applicable during either a THERMAL POWER ramp in excess of 5X of RATED THERMAL POWER per minute or a
THERMAL POWER step in excess of 10K of RATED THERMAL POWER.
TURKEY POINT - UNITS 3 8 4 3/4 2-16 AMENDMENT NOS.l 37 AND 132
I ~
I 4
g Zl
~FUNCTIONAL UNIT 2.
Coincident with:
Steaa Generator Pressure"-Low or T
Low Contaiint Spray ALLOWANCE TA
- 13. 0 1
S 1.16 2.3 TRIP SETPOINT
>614 psig 4.0
>543'F TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEH ALLOWABLE VALUES
>588 psig
>542.5 F
CalI a.
Autoaatic Actuation Logic and Actuation Relays b.
Containaent Pressure High-High Coincident with:
Contaiteent Pressure High 3.
Containoent Isolation a.
Phase "A" Isolation H.A
- 21. 3
- 13. 3 H.A N.A 2.7 0.0 10.3 0.0 N.A.
<20.0 psig
< 4.0 psig N.A.
<22.6 psig
< 4.5 psig Q
R 5
1)
Nanual Initiation 2)
Autoeatfc Actuation Logic and Actuation Relays 3)
Safety Injection N.A
~
N.A.
see itea 1
, N.A. N.A.
N.A. N.A.
H.A.
N.A.
N.A.
N.A.
See Itea 1 above for all Safety Injection Trip Setpoints and Allowable Values.
b.
Phase 484 Isoldtion 1)
Hanual Initiation N.A.
N.A. N.A.
N.A.
N.A.
II h
TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUNENTATIOH TRIP SETPOIHTS FUNCTIONAL UNIT 4.
Steam Line Isolation (Continued) b.
Automatic Actuation Logic and Actuation Relays c.
Containment Pressure High-High Coincident with:
Containment Pressure High N.A.
H.A.
N.A.
N.A.
21.3 2.7 0.0
<20.0 psig 13.3 10.3 0.0
<4.0 psig TRIP
~NA ALLOMABLE VALUES N.A.
<22.6 psig
<4.5 psig d.
Steam Line FlowHigh 16.7 2.86 3.9
<A function defined as follows:
A hp corresponding to 40% steam flow at 0% load increasing linearly from 20%
load to a value corresponding to 120% steam flow at full load.
<A function defined as follows:
A hp corresponding to 42.6% steam flow at 0% load increasing linearly from 20/
load to a value corresponding to 122.6/ steam flow at full load.
Coincident with Steam Line Pressure Low or T, Low 5.
Feedwater Isolation a.
Automatic Actuation Logic and Actuation Relays b.
Safety In)ection c.
Steam Generator Mater Level High-High 13.0 4.0 N.A.
20.0 see item 1
2.3 Z614 psig Z588 psig 1.0
~543 F
Z542.5 F
N.A.
H.A.
N.A.
~ H.A.
See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
18.27 2.0
<80% of narrow
<81.9% of narrow range instrument range instrument span.
span.
a~
TABLE
~ (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEH FUNCTIONAL UNIT 6.
Auxiliary Feedwater (3) a.
Autoaatic Actuation Logic and Actuation Relays ALLOMANCE TA H.A.
TRIP S
SETPOINT N.A. N..
H.A g,o l0%
ALLOWABLE VALUES N.A. ~ 88%
b.
Steaa Generator Mater LevelLow-Low 5.0 of narrow of narrow range instr'ument range instrument span.
span.
c.
Safety Injection d.
Bus Stripping see itea 1 see itea 7 See Item l. above for all Safety Injection Trip Setpoints and Allowable Values.
See Iteti 7. below for all Bus Stripping Trip Setpoints and Allowable Values.
I e.
Trip of All Hain Feedwater N.A.
Puap Breakers 7.
Loss of Power N.A.
N.A.
N.A.
N.A.
a.
4.16 kV Busses A and B
(Loss of Voltage)
N.A.
N.A.
N.A.
H.A.
N.A.
5
I
TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURES ACTUATION SYSTN C) 4 FUNCTIOHAL UHIT I
~8.
Engineering Safety Features Actuation System Interlocks c
a.
Pressurizer Pressure b.
T Low 9.
Control Rooa Ventilation Isolation a.
Autoaatic Actuation Logic and Actuation Rel~s ALLOWANCE TA N.A.
4.0 N.A.
Z S
N.A. N.A.
~ >.0 N.A. N.A.
TRIP SETPOIHT Nominal 2000 psig Hoainal 543'F N.A.
ALLOWABLE VALUES
<2018 psig
)542.5 F
N.A.
I Ca7 ED b.
Safety Injection see itew 1 See Itea l. above for all Safety Injection Trip Setpoints and Allowable Values.
N.A. N.A.
Particul ate (R-ll)
<6.1 x 10s CPH Caseous (R-12)
See (2)
N.A.
c.
Containeent Radioactivity High (1)
H.A.
N.A.
m d.
Contaireent Isolation N.A. N.A.
N.A.
manual Phase A or Hanual 5
Phase e.
Air Intake Radiation Level N.A. N.A.
< 2 aR/hr I
TABLE NOTATIONS
~(1)
Either the particulate or gaseous channel in the OPERABLE status will satisfy this LCO.
Particulate (R-ll)
<6.8 x 10s CPM t'aseous (R-12)
See (2)
N.A.
< 2.83 aR/hr
P 4 <<I
/
POWER OISTRIBUTION LIMITS BASES 3/4.2.4 UAORANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power dis-tribution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during STARTUP testing and periodically during power operation.
The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x-y plane power tilts.
A limit of 1.02 was selected to provide an allowance for the uncertainty asso-ciated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.
In the event such action action does not correct the tilt, the margin for uncertainty on F~ (Z) is reinstated by reducing the maximum allowed power by 3X for each percent of tilt in excess of l.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors or incore thermocouple map are used to confirm that the normalized symmetric power distribution is consistent with the gUAORANT POWER TILT RATIO.
The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.
The two sets of four symmetric thimbles is a unique set of eight detector locations.
These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-ll, N-8.
3/4. 2. 5 ONB PARAMETERS The limits on the ONB-related parameters assure that each of the param-eters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.
The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated ade-quate to maintain a minimum ONBR above the applicable design limits throughout each analyzed transient.
The indicated T
value of 576.6'F and the indicated pressurizer pressure value of 2209 psig correspond to analytical limits of 578.2'F and 2185 psig respectively, with allowance for measurement uncertainty.
The indicated RCS flow value of 277,900 gpm corresponds to an analytical limit of 268,500 gpm which is assumed to have a 3.5X measurement unceMainty.
The above measurement uncertainty estimates assume that these instrument channel outputs are averaged to minimize the uncertainty.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.
TURKEY POINT - UNITS 3 8
4 B 3/4 2-8 AMENOMENT NOSl37 AND 132
fj Y
I
~
I t
Reactor Coolant System T~ and Pressurizer Pressure shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.2.5.2 RCS flow rate shall be monitored for degradation at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
After each fuel loading, and at least once per 1&
- months, the RCS flow rate shall be determined by precision heat balance after exceeding 90%
RATED THERMAL POWER.
The measurement instrumentation shall be calibrated within 90 days prior to the performance of the calorimetric flow measurement.
The provisions of 4.0.4 are not applicable for performing the precision heat balance flow measurement.
The 18-month periodic measurement of the RCS total flow rate is adequate to ensure that the DNB-related flow assumption is met and to ensure correlation of the flow indication channels with measured flow.
Six month drift effects have been included for feedwater temperature, feedwater flow, steam pressure, and the pressurizer pressure inputs.
The flow measurement is performed within ninety days of completing the cross-caLibration of the hot surveillance on a 12-hour basis will provide sufficient verification that flow degradation has not occurred.
A change in indicated percent flow which is greater that the instrument channel inaccuracies and parallax errors is an appropriate indication on RCS flow degradation.
ATTACHMENT 4 PROPRIETARY WCAP 13719, REV.
1 WESTINGHOUSE REVISED THERMAL DESIGN PROCEDURE INSTRUMENT UNCERTAINTY METHODOLOGY and NON-PROPRIETARY WCAP 13718, REV.
1 WEST1NGHOUSE REVISED THERMAL DESIGN PROCEDURE INSTRUMENT UNCERTAINTY METHODOLOGY
0 I.