ML20196F321
| ML20196F321 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/11/1999 |
| From: | Galenbush J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Collins S NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20137U998 | List: |
| References | |
| CAW-99-1335, NUDOCS 9906290179 | |
| Download: ML20196F321 (25) | |
Text
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O Westinghouse Electric Company Box 355 Pittsburgh Pennsylvania 15230 4 355 May 11,1999 CAW-99-1335 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Mr. Samuel J. Collins a
APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
Subject:
WCAP-14787, " Westinghouse Revised Thermal Design Procedure Instmment Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 and 2 (Fuel Upgrade and Uprate to 1656 Mwt-NSSS Power)", April 1999,(Proprietary)
Dear Mr. Collins:
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-99-1335 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. 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 Affidavit by Wisconsin Electric Company.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-99-1335 and should be addressed to the undersigned.
Very truly yours, 8
A:=
J. S. Galembush, Acting Manager Regulatory and Licensing Engineering Enclosures cc: T. Carter /NRC (SE7) 9906290179 990622 PDR ADOCK 03000266 P
PDR,
9 ^[
/cuv0224S. doc 1
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- q
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with reque.c for beneric and/or plant-specific review and approval.
In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in e
the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means oflower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item ofinformation being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types ofinformation Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).
COPYRIGHT NOTICE l
The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to i
make the number ofcopies of the information contamed in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the l
extent such information has been identified as proprietary by Westinghouse, copyright protection
)
o notwiteading. With respect to the non-proprietary versions of these reports, the NRC is permitted to h
make the number of copies beyond those necessary for its intemal use which are necessary in order to have one copy available for public viewmg in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
1 i
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CAW-99-1335 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss e
COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared John S. Galembush, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (" Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
M 4/
John S. Galembush, Acting Manager Regulatory and Licensing Engineering Sworn to and subscribed before me this /#M day of
,1999 U$$g#E$yd'$$$', 5!h
~2
- "'""M
/
Notary Public g
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g, ',,......... < s.,
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/cm/02255. doc CAW-99-1335 (1)
I am Manager, Regulatory and Licensing Engineering, in the Nuclear Services Division, of the Westinghouse Electric Company LLC (" Westinghouse"), 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 \\ estinghouse.
(2)
I am raaking this Affidavit in 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 4ffidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by Westinghouse Electric Company LLC 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 info.ation 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 informa' tion 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 i
Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it 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:
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of j
/cm/02255. doc
)
CAW-99-1335 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, stmeture, 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 1
(f)
It contains patentable ideas, for which patent protection may be desirable.
l There are sound policy reasons behind the Westinghouse system which include the j
following:
l (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.
(c)
Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
/cm/0:25s. doc CAW-99-1335 (d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
l (e)
Unrestricted disclosure would jeopardize the position of prominence of
' Westinghouse inJhe 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.
l (iv)
The information sought to be protected is not available in public sources or available l
information has not been previously employed in the sarae 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 Instrument Uncertainty Methodology for Wisconsin Electric Power Company Point Beach Units 1 and 2 (Fuel Upgrade and Uprate to 1656 Mwt-NSSS Power)", WCAP-14787 (proprietary), April 1999, for Points Beach Units 1 and 2, being transmitted by the l
Wisconsin Power Company letter and Application for Withholding Proprietary Informatio'n from Public Disclosure, to the Document Control Desk, Attention Mr.
Samuel J. Collins. The proprietary information as submitted for use by Wisconsin Electric Power Company for the Point Beach Units 1 and 2 is expected to be applicable in other licensee submittals in response to certain NRC requirements
/cm/0225s. doc
-5 CAW-99-1335 This information is part of that which will enable Westinghouse to:
)
)
(a)
Provide documentation of the analysis and methods for determining operating parameter uncertainties.
(b)
Calculate information which is used in the thermal analysis of the nuclear fuel.
(c)
Assist the customer in obtaining NRC approval.
e Further this information has substantial commercial value as follows:
(a)
Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.
(b)
Westinghouse can sell support and defense of the technology to its customers in the licensing process, 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 services and licensing defense services 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 developing testing and analytical methods and performing tests.
Further the deponent sayeth not.
/cm/02Ms. doc
r I
(
NPL 99-0369 - Discussion of Technical Specification Changes Page 1 of 16 i
Introduction PBNP plans to refuel and operate, commencing with Unit 2 Refueling Outage 24 (currently scheduled to begin September,2000), with upgraded Westinghouse fuel features. The first
)
reload with the upgraded fuel for Unit I would commence with Unit 1 Refueling Outage 26 (currently scheduled to begin April,2001). The upgraded fuel is 0.422" OD,14x14, VANTAGE
+ fuel with PERFORMANCE + features; hereafter referred to as 422V+.
Technical Specification changes are required to incorporate the 422V+ fuel assemblies into the PBNP coms. A summary of the required Technical Specification changes is provided below and are discussed in detail in this Attachment.
Summary ofRequired Technical Specification Changes:
j Changes to the react _or core safety limit curves to reflect transition cores and full 422V+ cores e
Changes in the OTAT and OPAT reactor trip setting limits e
RCS Tog range change to 558.1 F to 574.0 F from the current 557 F to 573.9 F e
RCS flow measurement uncertainty increase to 2.4% from the current 2.1% and e
corresponding increase in RCS raw measured total flow rate to 182,400 gpm from the current j
181,000 gpm Full power F% peaking factor design limit will increase to 1.77 from the current 1.70 e
Maximum Fn(Z) peaking factor limit will increase to 2.60 from the current 2.50 and the K(Z) e e
envelope will be modified Changes to reflect ZIRLO* material Changes to reflect transition cores and full cores of 422V+ fuel assemblies in the core e
Restrictions on primary system pressure (2250 psia) for cores containing 422V+ fuel e
assemblies Restrictions on storing new fuel in the new fuel vault storage cells Corresponding Basis section changes to reflect the above changes A detailed description of the above proposed changes and the basis for these changes follows.
Deletions from the presently approved Specifications are indicated by revision bars and lineout, and additions are indicated by revision bars and are underlined.
Description of Proposed Chances and Supportine Information:
It is proposed that TS 15.2.0 " Safety Limits and Limiting Safety System Settings" be revised as follows.
TS 15 2.1.1 The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 or Ficure 15.2.1-2 as annlicable for Units I and 2.+-
NPL 99-0369 - Discussion of Technical Specification Changes Page 2 of 16
-TS 15.2.1.1 Bases The family of curves in_ofFigure 15.2.1-land 15.2.1 2are_is applicablefor_to a core with any combination of 14 x 14 OFA and 14 x 14 upgraded OFA fuel assemblies. The family of curves in Ficure 15.2.1-2 is applicable to any combination of 422V+ fuel assemblies. bumed 14 x 14 OFA fuel assemblies. and burned 14 x 14 upgraded OFA fuel assemblies. or a full core of 422V+
fuel assemblies. Theerve+also-apply 4o4he+einsertion+f,rev:cusly-depleted 44+44-standard t
fuel-ussemblies-inte-tm-GFA-eore:
TS Figure 15.2.1-1
- This figure applies to core reloads with any combination of OFA and upgraded OFA fuel assemblies. Unit-24 elk + wing 4MR22-amb-Unit 44ellowing U!R24. Prior-to44RG4rFigure 15.2.12 applies 4o Unit-h TS Figure 15.2.1-2
- This figure applies to core reloads with any combination of 422V+ fuel assemblies. burned OFA and burned uperaded OFA fuel assemblies, or a full core of 422V+ fuel assemblies. Unit 4 prior-te4#R24-Fo!!cwing UIR24, Figure 15.2.1 1 applies 4e Unit-b Thisfigure was replaced by a newfigure developed by Westinghouse based on analyses donefor the 422 V+ fuel.
Basis for Change The addition of the reference of Figure 15.2.1-2 as applicable in TS 15.2.1.1 is necessary because there will be a mix of OFA and upgraded OFA fuel and 422V+ fuel in Unit 2 (Unit 2 will be the first core loaded with 422V+ fuel) after U2R24 while Unit I will still contain a full core of OFA and upgraded OFA fuel until after UlR26. Therefore, including two reactor core safety limit curves in the Technical Specifications will be necessary (Figure 15.2.1-1 and 15.2.1-2) to reflect the possible core configurations of both units.
The footnote under TS 15.2.1.1 was added in July 1997 to reflect the analysis done for the replacement of the Unit 2 steam generators. This footnote was necessary to clarify when the reactor core safety limits curves applied to each unit, and were based on when the Unit 2 steam generators were replaced (The generators were replaced during refueling outage U2R22). In order to maintain one set of reactor core safety limits for both units (Figure 15.2.1-1), the Unit 2 safety analyses performed by Westinghouse with the new steam generators also bounded operation of Unit 1. This footnote is being deleted because it is no longer necessary since the Unit 2 steam generators have been replaced and the refueling outage footnote requirements no longer apply. Eliminating this footnote is administrative in nature.
The proposed changes in the bases of TS 15.2.1.1 are necessary to specify under what conditions the reactor core safety limit curves apply to each unit. The deletion of the sentence that the curves apply to reinsertion of previously depleted 14 x 14 standard fuel assemblies was necessary because the analyses performed for the 422V+ fuel no longer bounds this condition.
NPL 99-0369 - Discussion of Technical Specification Changes Page 3 of 16 l
Figure 15.2.1-1 was previously developed by Westinghouse based on prior analyses performed for OFA fuel and needs to remain to allow Unit 1 operation as discussed in the cover letter of this submittal. The proposed change in the footnote to Figure 15.2.1-1 is necessary based on the above discussion provided for the change in the footnote for TS 15.2.1.1 (replacement of the Unit 2 steam generators). The footnote will be reworded to reflect that the figure applies only to cores containing OFA fuel assemblies. Changing the footnote to specify that the curves apply only to full cores of OFA fuel is necessary to distinguish the applicability of this figure from Figure 15.2.1-2.
9 Figure 15.2.1-2 was replaced with new curves that were developed by Westinghouse based on analyses performed for the 422V+ fuel. The analyses were performed (and the corresponding Figure 15.2.1-2 curves developed) to bound core conditions containing any combination of burned OFA, burned upgraded OFA, and 422V+ fuel assemblies or a full core of 422V+ fuel assemblies. The proposed changes clarify the applicability of these curves to the fuel assemblies contained in the cores. The proposed change in the footnote for Figure 15.2.1-2 is necessary to distinguish the applicability of this figure from Figure 15.2.1-1, and eliminates the applicability contingency to the replacement of the Unit 2 steam generators as discussed above.
It is proposed that TS 15.2.3 " Limiting Safety System Settings, Protective Instrumentation" be revised asfollows.
TS 15.2.3.1.B(2)
High pressurizer pressure *- - $2385 psig for operation at 2250 psia primary system pressure
- $2210 psig for operation at 2000 psia primary system pressure and cores not containinc 422V+ fuel assemblies
- These values app!y :o Uni: 2 falkming U2R22 and ic Unit-1 fc!!cwing UlR24. Priorao-UlR24the4dghpressurker-pressurereaetor4 rip-setpoint-for Uni: ! :s g2385-psig:
TS 15.2.3.1.B(3)
Low pressurizer pressure *
- 21905 psig for operation at 2250 psia primary system pressure
- 21800 psig for operation at 2000 psia primary system pressure and cores not containinc 422V+ fuel assemblies
+ These.alues-apply 4o Unit 2 following-U2R&nd4o4Jnit-1-following U!R24-Priordo-UIR24the4ew prewurizer pressure +eaetorarip-setpoint-for Uni: I a 21790 psig.
TS 15.2.3.1.B(4)
Overtemperature AT (1 + r3S) s ATo (K:- K2(T(1 + r4S)-T')(1 + r21 + r'S ) +
1 1
S where (values are applicable to operation at both 2000 psia and 2250 psia unless otherwise indicated)
NPL 99-0369 - Discussion of Technical Specification Changes Page 4 of 16 indicated AT at rated power, F ATo
=
average temperature, F T
=
T s
569.0 F (for cores containing 422V+ fuel assemblies)
T' s
572.9 f*(for cores not containing 422V+ fuel assemblies) pressurizer pressure, psig P
=
2235 psig (for 2250 psia operation only)
P'
=
1985 psig (for 2000 psia operation and cores not containing 422V+ fuel P'
=
assemblies only)**-
L 5
1.16 (for 2250 psia operation and cores containing 422V+ fuel assemblies)
Ki 5
1.19 (for 2250 psia operation and cores not containine 422V+ fuel assemblies only)
Ki s
1.14 (for 2000 psia operation and cores not containing 422V+ fuel assemblies only)**-
K2 5
0.0149 (for 2250 psia operation and cores containing 422V+ fuel assemblies) 0.025 (for 2250 psia operation and cores not containing 422V+ fuel assemblies K2
=
only) 0.022 (ffg_2000 psia operation and cores not containing 422V+ fuel assemblies K2
=
only P 0.00072 (for 2250 psia operation and cores containine 422V+ fuel assemblies)
K3
=
0.0013 (for 2250 psia operation and cores not containing 422V+ fuel assemblies K3
=
only) 0.001 (for 2000 psia operation and cores not containing 422V+ fuel assemblies K3
=
only)**-
25 see Ti
=
3 see T2
=
2 sec for Rosemont or equivalent RTD T3
=
0 see for Sostman or equivalent RTD
=
2 sec for Rosemont or equivalent RTD T4
=
0 sec for Sostman or equivalent RTD
=
and f(AI)is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests, where qi and qn are the percent power in the top and bottom halves of the core respectively, and q + qd is total core power in percent of rated power, such that:
(a) for gi - go within -17, +5 percent, f(AI) = 0 for cores not containing 422V+ fuel assemblies: for a, - qqwithin -12. +5 percent. f(AI) = 0 for cores containing 422V+ fuel assemblics.
(b) for each percent that the magnitude of gi-qs exceeds +5 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power for cores not containine 422V+ fuel assemblics and reduced by an equivalent of 2.12 percent of rated power for cores containine 422V+ fuel assemblies.
NPL 99-0369 - Discussion of Technical Specification Changes Page 5 of 16 (c) for cores not containing 422V+ fuel assemblies. for each percent that the magnitude of gi
- qu exceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power: for cores containing 422V+ fuel assemblies. for each nercent that the magnitude of asexceeds -12 percent. the AT trin setpoint shall be automatically reduced by an eauivalent of 2.0 percent of rated power.
These-valuewpply4o4hii+-24eHowing&2REend4oUnit-1-folk 5 wing-UI R24-PrioHo 2
UlR24rthe+ aloes-are: T' f 573.9 F, P' = 2235 psigrKcf-1-3GrK2+0.020GrandK3-0.000791.
TS 15.2.3.1.B(5)
Oveq)ower 1
T3S 1
1 AT (1 + T3S) 5 AT.[K4-K5(T S + 1)(1 + T,S)T-K.[T(1 + T S)- T'))
5 4
where (values are applicable to operation at both 2000 psia and 2250 psia) indicated AT at rated power, F ATo
=
T average temperature, F
=
T 5
569.0 F (for cores containine 422V+ fuel assemblies)
T' s
572.9 F*-(for cores not containinc 422V+ fuel assemblies)
L 5
1.10 of rated power (for cores containinc 422V+ fuel assemblies)
K4 5
1.09 of rated power *-(for cores not containine 422V+ fuel assemblies)
K 0.0262 for increasing T 5
=
0.0 for decreasing T
=
5 0.00103 for T 2 T' (for cores containing 422V+ fuel assemblies)
K 0.00123 for T 2 T' (for cores not containing 422V+ fuel assemblies) 6
=
0.0 for T < T'
=
T5
=
10 see 2 see for Rosemont or equivalent RTD T3
=
0 see for Sostman or equivalent RTD 2 see for Rosemont or equivalent RTD T4
=
0 see for Sostman or equivalent RTD (6)
Undervoltage - 23120V (7)
Indicated reactor coolant flow per loop 290 percent of normal indicated loop flow (8)
Reactor coolant pump motor breaker open (a)
Low frequency set point 255.0 HZ (b)
Low voltage set point 23120V Theenluewpply-tuUnit-24ellowing&2R42-und4oUnit4-following441R24-PrioHo UlR24 4he value:, for4Jnit ! are: T' C 573.9 F and-KA-h089-of-ratedp+ wen 7
l
NPL 99-0369 - Discussion of Technical Specification Changes Page 6 of 16 TS 15.2.3.1 (other reactor trips)
LOther reactor tr ps:
i
' (1)
High pressurizer water level - $95% of span
-(2) lew-low steam generator water level -
220% of narrow range instrument span I
25% of nancv range instrument span (Uni: 1)$
' This se::ing !!nnt-applies :c Unit I until $e narrow range !cwcr ap4s+ hanged ic $e
+
!cwer positica cons:::ent with Uni: 2.
TS 15.2.3 Bases The overpower AT reactor trip prevents power density anywhere in the core from exceeding 408% 118% of design power density, and includes corrections for change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The specified setpoints meet this requirement and include allowance for instrument errors.(2)
The overpower, overtemperature and pressurizer pressure system setpoints for OFA and uncraded OFA fuel include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints for 422V+ fuel do not include the effect of reduced system
. pressure coeration: therefore. cores containing 422V+ fuel must be ooerated at 2250 psia. The setpoints will not exceed the core safety limits as shown in Figums 15.2.1-1 (for OFA and upgraded OFA fuel only cores) and 15.2.1-2 (for cores containing 422V+ fuel).
References
- FSAR 14.1.1
- FSAR 14.0, Fage 14-5
- FSAR 14.1.2
- FSAR 14.1.89
- FSAR 14.2.6
- FSAR 7.2,7.3
- FSAR 14.1.10 and 14.1.11 Basis for Change The addition of the restrictions in TS 15.2.3.1.B(2) and (3) to disallow core operation at 2000 psia with the 422V+ fuel was necessary because the analyses done to support the 422V+ fuel bound core operation at 2250 psia primary system pressure with any combination of 422V+ and burned OFA fuel, but does not bound core operation at 2000 psia primary system pressure with a core containing 422V+ fuel. The proposed restrictions were added to reflect this condition.
The change to delete the footnotes in TS 15.2.3.lB(2), B(3), B(4), and B(5) is unrelated to this amendment request. The footnotes were added in July 1997 to reflect the setpoint analysis done a
NPL 99-0369 l - Discussion of Technical Specification Changes Page 7 of 16 for the replacement of the Unit 2 steam generators and was related to the timing when the setpoint changes would be made. These footnotes are being deleted because they are no longer necessary and the refueling outage footnote requirements no longer apply. Eliminating these footnotes is administrative in nature.
I TS 15.2.3.1.B(4) gives the Overtemperature Delta T (OTAT) reac:ar trip setpoint function and parameter values. TS 15.2.3.1.B(5) gives the Overpower Delta T (OPAT) reactor trip setpoint function and parameter values. The old OTAT/OPAT trip setpoints and parameter values need to remain as discussed in the cover letter of this submittal to allow operation of Unit i for a period of time with OFA fuel. The proposed revisions identified above to the inputs of the OTAT/OPAT setpoint functions are necessary as a result of the analyses performed by Westinghouse for the 422V+ fuel assuming the most conservative core thermal limits, and increase the operating margins associated with these trip functions. These core thermal limits bound transition cores of burned OFA, burned upgraded OFA,422V+, and full cores of 422V+.
The changes associated with the OPAT/OTAT setpoints have been confirmed as being acceptable in the FSAR Chapter 14 accident analyses by showing that the DNB design basis is met.
The addition of the contingencies in the OTAT and OPAT variables on the type of fuel assemblies contained in the cores is necessary as discussed in the cover letter of this TSCR. The analyses performed for the 422V+ fuel and the different loading schedules dictate that the cores operate at different operating pressures for a period of time. The proposed contingencies (on fuel contained in the cores and operating pressures) were added to reflect these conditions.
The proposed changes in the OTAT and OPAT reactor trip setting limits provide adequate protection over the full range of expected reactor coolant system operation. Included in is the revised FSAR Figure 14.0-1 which shows these reactor protection functions graphically in relation to the core safety limits.
The OTAT reactor trip is one of the reactor protection functions utilized in the analysis of a uncontrolled rod withdrawal at power as described in FSAR {l4.1.2 and analysis of the loss of load accident as described in FS AR sl4.1.9.
The OPAT reactor trip function is not utilized as the reactor protection initiator in any of the FSAR Chapter 14 accident analyses for PBNP. As stated previously, FSAR Figure 14.0-1 shows the Overpower reactor protection limit in relation to the core safety !!mits.
The analyses to support the proposed changes were performed in accordance with NRC approved methodologies and the results indicate that all design basis acceptance criteria continue to be met.
Therefore, the proposed changes provide adequate reactor protection over the required ranges that are applicable for these functions.
l The revised OPAT and OTAT setpoints require that the reference average temperatures (T') for cores containing 422V+ fuel are maintained les:- than or equal to 569.0 F for any operating i
r NPL 99-0369 - Discussion oi Technical Specification Changes Page 8 of 16 temperatum in the Tavg window. This requirement is essential to ensure that the actual plant conditions required to generate an OPAT and OTAT trip signal are bounded by the assumptions made in the safety analyses. The analysis allows T' to remain as a fixed value (less than or equal to 569.0 F), resulting in a constant reference temperature for a range of indicated T-avg. For instance, this would allow for an end-of-cycle Tavg coastdown without having to constantly re-scale the reference temperature in the OPAT and OTAT trip setpoints.
The revised OPAT and OTAT setpoints require the f(AI) function changes as described in the Technical Specifications. These requirements for the 422V+ fuel are more rest;ictive than the current Technical Specifications. This change was required to avoid potential stress violations in the Westinghouse Fuel Rod Design criterion. Recent PBNP core designs have shown that the clad stress criterion has approached the limit requirements. Since the cycle length and power histories of the fuel rod are principle factors in analyzing the stress criterion, the only other factor that significantly impacts the stress criterion is the transient limits which are developed from the negative wing of the f(AI) function. To gain margin in the fuel rod suess criterion, the negative wing of the f(AI) function has been shifted inwards (i.e., more restrictive from an operational standpoint). This change has no impact on any other analysis other than fuel rod design and operations.
A more detailed discussion of the above changes is provided in Attachment 2.
The addition of "C." to TS 15.2.3.1 (other reactor trips) is administrative in nature and is necessary to put a TS numerical identifier on this section. The letter "C" was used as an identifier of this section on the original plant Technical Specifications, but was inadvertently deleted over the history of Amendments. The addition of"C."is necessary to clearly indicate the numericalidentifier of the Technical Specification section as 15.2.3.1.C. The change to delete the "5% of narrow range instrument span (Unit 1)*" and its corresponding footnote in TS 15.2.3.lC(2) is unrelated to this amendment request. This allowance and its corresponding footnote was made to distinguish between the different instrument span on the Unit I and Unit 2 steam generators. This allowance and its corresponding footnote no longer applies because the level instrumentation taps on the Unit I steam generators have been modified to be consistent with Unit 2's.
The proposed change in the Basis of TS 15.2.3 to change the 108% hmit on design power density to 118% is consistent with the analysis performed for the 422V+ fuel. The OPAT reactor trip, in conjunction with the OTAT, is designed to ensure operation within the fuel temperature design basis and is accomplished in the safety analysis by controlling the core thermal power within 118% of nominal power (the OPAT trip function ensures this). This 118% limit is a parameter that is input into the development of the OTAT and OPAT reactor trip setpoints. The 118%
value is supported by the reanalysis and is based on the most conservative core thermal limits.
NPL 99-0369 - Discussion of Technical Specification Changes Page 9 of 16 The proposed changes in the Basis of TS 15.2.3 to insen "for 14x14 OFA and 14x14 upgraded OFA fuel"is necessary to clarify that only the analyses for OFA fuel includes the effect of reduced system pressure operation (i.e. 2000 psia). The next sentence was added to specify that the analyses done to support the 422V+ fuel did not include the effects of reduced system operation and that cores containing 422V+ fuel must therefore be operated at 2250 psia. The qualification of Figures 15.2.1-1 and 15.2.1-2 in the next sentence was added to distinguish the applicability of the figures based on fuel contained in the reactor cores.
The changes to the FSAR references in the TS 15.2.3 Bases are unrelated to this amendment request smd are necessary to reflect the appropriate FSAR sections. These changes are administrative in nature.
It is proposed that TS 15.3.1 " Limiting Conditionsfor Operation" be revised asfollows.
TS 15.3.1.G Operational Limitations The following DNB related parameters shall be maintained within the limits shown during rated power operation:
- 1. L shall be maintained 2558.1 F and $574.0 F for cores containing 422V+ fuel assemblies._
T,yg shall be maintained 2557 F and $573.9 F for cores not containing 422V+ fuel assemblies.
- 2. Reactor Coolant System (RCS) pressurizer pressure shall be maintained:
22205 psig during operation at 2250 psia, or 21955 psig during operation at 2000 psia for cores not containing 422V+ fuel assemblies.
j
- 3. Reactor Coolant System raw measured Total Flow Rate shall be maintaiaed 2182.400 gpm for cores containing 422V+ fuel assemblies, or 2181,800 gpm for cores not containine 422V+ fuel assemblies.
TS 15.3.1 Basis:
The reactor coolant system total flow rate of 182.400 gpm for cores containing 422V+ fuel assemblies is based on an assumed measurement uncertainty of 2.4 oercent over thermal design flow (178.000 gpmL The reactor coolant system total flow rate of 181,800 gpm for cores not containing 422V+ fuel assemblies is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow (178,000 gpm).
Basis for Chance The full power average RCS temperature operating range (T,v, ) in TS 15.3.1.G.1 will change slightly as a result of the analyses performed for the 422V+ fuel from T,vg 2 557 F and s 573.9 F to T.v,2 558.1 F and s 574 F. The RCS average temperature operating range provides
NPL 99-0369 - Discussion of Technical Specification Changes Page 10 of 16 operational flexibility for reactor operation. Analyses and evaluations have been performed that show all the applicable acceptance criteria continue to be met with the proposed changes to the T.,g temperature range. As discussed in the cover letter of this TSCR, the old T yg temperature range will remain for cores not containing 422V+ fuel assemblies to allow for operation of Unit 1 for a period of time to support the loading schedules of the 422V+ fuel.
The addition of the restriction in TS 15.3.1.G to disallow core operation at 2000 psia with the 422V+ fuel was necessary because the analyses done to support the 422V+ fuel bound core operation at 2250 psia primary system pressure with any combination of 422V+ and burned OFA fuel, but does not bound core operation at 2000 psia primary system pressure willi a core containing 422V+ fuel. The proposed restriction was added to reflect this condition.
As a result of the analyses for the 422V+ fuel, the minimum reactor coolant system raw measured total flow rate in TS 15.3.1.G.3 will increase slightly from 2181,800 to 2182,400gpm.
This is due to the fact that the analyses was performed with an RCS flow uncertainty increase from 2.1 to 2.4 percent. RCS flow is monitored by the performance of a calorimetric flow measurement at the beginning of each cycle. As discussed in Attachment 5 (RTDP), the calculated uncertainty for the calorimetric measurement of RCS flow is 11.9% flow with a
+0.26% bias. The calculated value for uncertainty is significantly less than the 2.4% uncertainty assumed in the safety analyses performed for the 422V+ fuel. The 2.4% uncertainty is a conservatively high value which provides additional margin to the analyzed RCS flow and requires a corresponding increase in the Reactor Coolant System raw measured flow or minimum l
measured flow to 182,400 gpm in the Technical Specifications. The minimum measured flow of 182,400 gpm in the Technical Specifications will ensure that the actual RCS flow is larger than assumed in the accident analyses, thereby providing the necessary initial margin to DNB. The old reactor coolant system raw measured flow rate will remain for cores not containing 422V+ fuel assemblies to allow for operation of Unit I for a period of time to support the loading schedules of the 422V+ fuel. These proposed changes (assumed measurement uncertainty increase and increased flow rate) are also reflected in the Basis section for TS 15.3.1.
It is proposed that TS Table 15.3.5-1 " Engineered Safety Features Initiation instrument Setting Limits" be revised asfollows.
TS Table 15.3.5-1 l
No.
Functional Limit Channel Setting Limit 7
Low-low Steam Generator Auxiliary Feedwater 220% of narrow range instrument Water Level Initiation kWof-narrow-range 4nstrument (Unit 4r*
l
-e-Thiwetting4imiHiprdiewo-Unit 44mtiMe-narrow-range 4eweraatsehanged4e4he4 ewer position-consistent-with-Unit 4 l
NPL 99-0369 - Discussion of Technical Specification Changes Page 11 of 16 1
Basis for Change The proposed change to TS Table 15.3.5-1 Item 7 and its corresponding footnote is unrelated to this amendment request and is administrative in nature. This allowance and its corresponding footnote was made to distinguish between the different instrument span on the Unit I and Unit 2 steam generators. This allowance and its conesponding footnote no longer apply because the level instmmentation tape on the Unit I steam generators have been modified to be consistent with Unit 2's. This change is administrative in nature.
It is proposed that TS 15.3.10 " Control Rod and Power Distribution Limits" be revised as follows.
TS 15.3.10.E.1.a The hot channel factors defined in the basis shall meet the following limits:
For OFA and Uncraded OFA Fuel For 422V+ Fuel Fo(Z) S(2.50)/P x K(Z)
EdZ) Si2.60)l? x K(Z) for P > 0.5 Fo(Z) 55.00 x K(Z)
EclZ) S5.20 x K(Z) for P s; 0.5 F su < l.70 x [1 + 0.3 (1-P)]
Eh < l.77 x I1 + 0.3 (1-P)1 Where P is the fraction of full power at which the core is operating, K(Z) is the function in Figure 15.3.10-3 or Figure 15.3.10-3a. as applicable. and Z is the core height location of Fq.
TS Figure 15.3.10-3 title Hot Channel Factor Normalized Operating Envelope for OFA and Upgraded OFA Fuel New TS Figure 15.3.10-3a Hot Channel Factor Normalized Operating Envelope for 422V+ Fuel TS 15.3.10 Basis As a result of the increased peaking factors allowed by the new 422V+ fuel. a new column was N
l added to TS 15.3.10.E.1.a. The full power F Ag peakin faClor desien limit (radiai neaking l
factor) for 422V+ fuel will increase to 1.77 from the 1.70 value for the OFA fuel. The maximum i
Eg(Z) neakine factor limit (total peakine factor) for 422V+ fuel will increase to 2.60 from the 2.50 value for the OFA fuel. The OFA fuel desien will retain the current F"w and Fo(Z) peakine l
factors of 1.70 and 2.50. respectively. In addition. the K(Z) envelope for the new 422V+ fuel was modified and a new TS figure 15.3.10-3a was develoned and inserted in the Technical Specifications. The K(Z) envelope in TS Figure 15.3.10-3 remains for the OFA fuel, i
An-The upper bound envelope of450_fg (defined in 15.3.10.E) times the normalized pealang factor axial dependence of Figure 15.3.10-3 for OFA and Upgraded OFA Fuel and Figure l
15.3.10-3a for 422V+ Fuel (consistent with the Technical Specifications on power distribut.on I
control as given in Section 15.3.10) was used in the large and small break LOCA analyses. The l
t
NPL 99-0369 - Discussion of Technical Specification Changes Page 12 of 16 envelope was determined based on allowable power density distributions at full power restricted to axial flux difference (AI) values consistent with those in Specification 15.3.10.E.2.
The results of the analyses based on this upper bound envelope indicate a peak clad temperature ofless than the 2200 F limit. When an Fq measurement is taken, both experimental error and manufacturing tolerance must be taken into account. Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three N
percent is the appropriate allowance for manufacturing tolerance. In the design limit of F AH, there is eight percent allowance for uncertainties which means that normal operation of the core N H ; 1.70/1.08 for OFA and Upgraded OFA fuel and 1.77/1.08 is expected to result in a design F 5
for 422V+ fuel.
Basis fgChange To allow for the increased peaking factors allowed by the new 422.
fuel analyses, a column N
was added to TS 15.3.10.E.1.a. The full power F AH Peaking factor design limit (radial peaking factor) for 422V+ fuel will increase to 1.77 from the 1.70 value for the OFA fuel. The maximum Fn(Z) peaking factor limit (total peaking factor) for 422V+ fuel will increase to 2.60 from the 2.50 value for the OFA fuel and the K(Z) envelope will be modified. These increases will permit more flexibility in developing fuel management schemes for longer fuel cycles, and improvement N
of fuel economy and neutron utilization. The OFA fuel design will retain the current F agg Fq(Z) peaking factors of 1.70 and 2.50, respectively.
The implementation of increased radial and total peaking factor limits will have minor impacts on the core power distributions and peaking factors experienced at PBNP. The increased radial peaking factor limit allows the concept of low leakage fuel management to be extended by placing additional burned 422V+ fuel on the periphery of the core. The reduction in power in the l
peripheral assemblies is offset by increased power in the remaining assemblics. This increased radial peaking is accommodated by increasing the radial and total peaking factor limits.
Figure 3-3 of Attachment 2 shows a comparison of the radial peaking factors between the core models used in the analysis. A comparison of the total peaking factor versus cycle length for each of the core models used in the analysis is provided in Figure 3-4 of Attachment 2. Other
)
changes to the core power distributions and peaking factors are the result of the normal cycle-to-cycle variations in core loading patterns. The normal methods of feed enrichment variation and insertion of fresh burnable absorbers are employed to control peaking factors to ensure compliance with the peaking factor Technical Specifications. A more detailed discussion of the increased peaking factors is included in Attachment 2.
As a result of the increased total peaking factor (Fn(Z)) for the 422V+ fuel, the K(Z) envelope for the 422V+ fuel was modified. A new TS figure 15.3.10-3a was developed and inserted in the Technical Specifications as a result of the analysis performed to increase Fn(Z) for the 422V+
fuel. The K(Z) envelope in TS Figure 15.3.10-3 will remain for the OFA fuel, with the title changed to clarify that the figure is for OFA and Upgraded OFA fuel only. The addition of"or
.I
NPL 99-0369 - Discussic n of Technical Specification Changes Page 13 of 16 Figure 15.3.10-3a, as applicable" was added to the sentence under the hot channel limits in TS 15.3.10.E.1.a to distinguish the applicability between the two K(Z) function figures.
The corresponding basis to TS 15.3.10 has been slightly revised to add discussion on the new peaking factors and the modified K(Z) envelope for the 422V+ fuel, clarify the description of the peaking factor equations, distinguish the applicability of TS Figures 15.3.10-3 and 15.3.10-3a to the fuel type, and to add discussion of the 422V+ F% design limit allowance for uncertainties.
3 It is proposed that TS 15.5 " Design Features" be revised asfollows.
TS 15.5.3.A 1.
General The uranium fuel is in the form of slightly" enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 or ZIRLO tubing to form fuel rods. The reactor core is made up of 121 fuel assemblies. Each fuel assembly nominally contains 179 fuel rodsm, Where safety limits are not violated, limited substitutions of fuel rods by filler rods consisting of Zircaloy 4. ZIRLO
,or stainless steel, or by vacancies, may be made to replace damaged fuel rods ifjustified by cycle specific reload analysis.
2.
Core A reactor core is a core loading pattern containing any combination of 14x14 OFA and 14x14 upgraded OFA. or any combination of 422V+ and burned 14x14 OFA or burned 14x14 upgraded OFA fuel assemblies. The core may also ecatain pre..ausly depk^ted 44*44-standard fuel assemblics. The use of previa: sly depleted 14x M stamlard these fuel assemblies will be justified by a cycle specific reload analysis.
References (1) FSAR Section 34-3 3.2 (2) Deleted (3) Deleted (4) FSAR Section 343 3.2 (5) Deleted (6) FSAR Table 4.1-9 Basis for Chance The proposed change to TS 15.5.3.A.1 to add the "ZIRLO* " advanced zirconium alloy material is necessary because the 422V+ fuel assemblies consist of mid-grids, fuel cladding, guide thimbles, and instrumentation tubing manufactured from ZIRLO alloy to obtain ac'tkional operational benefit from the alloy's improved corrosion resistance and dimensional stability underirradiation. ZIRLO cladding and ZIRLO* fabricated components are known as l
l
1 NPL 99-0369 - Discussion of Technical Specification Changes Page 14 of 16 VANTAGE + features (422V+ fuel is VANTAGE + fuel with PERFCRMANCE + features) which have been submitted to the NRC (WCAP-12610-P-A) and received generic NRC approval for lead rod burnups up to 60,000 MWD /MTU. PBNP's highest lead rod burnups have historically been well below this limit. PBNP core designs are currently performed by Westinghouse, and internal controls are in place at Westinghouse to limit lead rod burnups to less than 60,000 MWD /MTU for core designs at PBNP. In addition, PBNP evaluates and concurs with cycle design burnups during design initialization meetings with Westinghouse using the loading pattern risk assessments completed by Westinghouse.
A more detailed discussion on the "ZIRLO
" alloy and the 422V+ fuel design is provided in.
The proposed change to TS 15.5.3.A.2 to add "or any combination of 422V+ and burned 14x14 OFA or burned 14x14 upgraded OFA"is necessary for implementation of the 422V+ fuel and reflect that the reactor cores could contain this type of fuel. The analyses performed for the 422V+ fuel assumed that the transition cores contained at least once burned OFA or Upgraded OFA fuel; therefore, it was necessary to specify burned OFA or bumed Upgraded OFA fuel when describing cores containing 422V+ fuel.
The deletion of the sentence that the cores may contain previously depleted 14x14 standard fuel assemblies was necessary because the analyses performed for the 422V+ fuel no longer bounds this condition. The sentence on cycle specific reload analyses was clarified so that it is clear that these analyses are done on a cycle specific basis that encompasses the fuel assembly designs that are to be used in each core reload.
The changes to the references in the TS 15.2.3 Bases are unrelated to this amendment request and are necessary to reflect the appropriate FSAR sections. These changes are administrative in nature.
It is proposed that TS 15.5.4 " Fuel Storage" be revised asfollows.
TS 15.5.4.2 The new and spent fuel storage racks are designed so that it is impossible to store assemblies in other than the prescribed storage locations. The fuel is stored vertically in an array with sufficient center-to-center distance between assemblies to assure Km<0.95 with the storage pool filled with unborated water and w h the fuel loading in the assemblies limited to 5.0 w/o U-235, a
with or without axial blanket twvs.t Each assembly with a fuel loading greater than 4.6 w/o U-235 must contain Integral F x1 Bemble Absorber (IFBA) rods in accordance with Figure 15.5.4-1 for the spent fuel nooi. M.re*referenee4nfinhe-muhiplicathm4aetorrK.-lew4han+r equal 4o-h4M64r hieh4nelude:; a 1%6K-renelivhybias. Fresh fuel assemblics with the w
maximum enrichment of up to 5.0 w/o U235 and a minimum of 321.25X IFBA rods can utilize all available new fuel vault storace celis An inspection area shall allow rotation of fuel m
assemblies for visual inspection, but shall not be used for storage.
NPL 99-0369 - Discussion of Technical Specification Changes Page 15 of 16 TS 15.14.3 The spent fuel storage pool shall be filled with borated water at a concentration of at least 2100 M00 ppm boron whenever there are spent fuel assemblies in the storage pool.
TS 15.5.4.4 Eweptfor-the tv.a s!crage beations adjacenHo4he<lesignated slo: fw4he+pt'nt fuel s!crtige-rtic4 neutron-absorbing-material +utveiliance specimea4rradiatiendspent fuel assembly storage locations immediately adjacent to the spent fuel pool perimeter or divider walls shall not be occupied by fuel assemblies which have been suberitical for less than one year.
Basis for Chance The addition of "for the spent fuel pool" after Figure 15.5.4-1 in TS 15.5.4.2 was added to clarify that this figure applies to storing fuel in the spent fuel pool only. The deletion of the sentence "or have a reference infinite multiplication factor, L, less than or equal to 1.49364, which includes a 1% AK reactivity bias" is unrelated to this amendment request and is being shown as deleted for information only to avoid confusion. The change to delete this sentence is being made separately under a different Technical Specification Change Request. Westinghouse has issued Nuclear Safety Advisory Letter (NSAL)99-003, dated February 26,1999, which communicated information that Westinghouse had recently determined that use of the reference L method could lead to IFBA requirements that are non-conservative when compared to the IFBA enrichment curve methodology. Westinghouse is discontinuing use of the reference L methodology and deleting this sentence in the PBNP TS is, therefore, necessary. All fuel stored in the spent fuel pool at PBNP continues to meet the existing storage requirements without the use of the reference L method. A more detailed discussion of this change is provided in the TSCR deleting this sentence.
The change to TS 15.5.4.2 to add restrictions on fuel stored in the new fuel vault is being made as a result of new analyses performed by Westinghouse. A PBNP condition report documented that the center-to-center distance of cell locations CC-12 and DD-12 in the new fuel vault was found to be approximately 19.875 inches. The existing criticality analysis for the new fuel vault assumed that the center-to-center distance was 20 inches. This assumption was not conservative with respect to the as-found distance identified in the condition report, and resulted in restrictions on the use of the new fuel vault storage cells. Therefore, a new analysis has been performed by Westinghouse for OFA, Upgraded OFA and 422V+ fuel assemblies assuming a conservative center-to-center distance of 19 inches. This analysis supports unrestricted location loading of OFA, Upgraded OFA and 422V+ assemblies and requires assemblies with a maximum enrichment of up to 5.0 w/o U235 to contain a minimum of 321.25X IFBA rods. Therefore, fresh OFA, Upgraded OFA and 422V+ assemblies with the maximum enrichment up to 5.0 w/o U235 and a minimum of 321.25x IFBA rods can utilize all available new fuel vault storage cells.
The proposed TS change is being made to be consistent with this analysis.
l l
3
i NPL 99 0369 - Discussion of Technical Specification Changes Page 16 of 16 The change in TS 15.5.4.3 to increase the required boron concentration in the spent fuel pool to at least 2100 ppm from the old value of j 800 ppm is conservatively being made to be consistent with the minimum boron concentration for refueling. TS 15.3.8.5 currently states "During reactor vessel head removal and while loading and unloading fuel from the reactor, a minimum boron concentration of 2100 ppm shall be maintained in the primary coolant system." The proposed change matches the minimum boron concentration requirement used in the reactor cavity and refueling canal during refueling operations. TS 15.3.8.5 was recently changed from requiring 1800 ppm to requiring 2100 ppm as a result of the PBNP cores implementing extended fuel cycles of 18 months. Without a corresponding change to the minimum spent fuel pool boron concentration, during refueling operations a slow dilution of the reactor coolant boron concentration with spent fuel pool water could occur. The proposed TS change is being made to prevent this situation from occurring and to have consistent minimum boron concentration reajrements during refueling operations.
The change in TS 15.5.4.4 to remove the exception for fuel placement near the spent fuel storage rack neutron absorbing material surveillance specimen is being proposed per the following discussion. Gamma heating from spent fuel assemblies induce thermal stresses in the spent fuel pool wall structure. Earlier analyses of high density spent fuel storage racks showed that with peripheral fuel assemblies having experienced one year of decay, thermal stresses in the pool walls would not be significant. The most limiting condition was the placement of two freshly discharged assemblies (three days decay) adjacent to the poison material (Boraflex) surveillance samples. It was originally required to place two freshly discharged fuel assemblies adjacent to the poison samples to maximize their dose rates consistent with the poison material surveillance program. The Technical Specifications were changed to allow placing fresh assemblies in these two locations for sample irradiation, and this change was approved by NRC letter dated 10/5/83.
The requirement to perform the sample surveillance ended when NRC Ictter dated 2/2/90 approved " Blackness Testing"in lieu of the original Boraflex sample testing. However, the poison samples continued to be irradiated for informational purposes.
The TS 15.5.4.4 restrictions on fuel placement at the pool perimeter remain valid for 422V+ fuel, l
. because the decay heat rates for 422V+ fuel is similar to standard or OFA and Upgraded OFA fuel. Divider wall gamma heating calculations were originally based on standard fuel, and the OFA and Upgraded OFA fuel evaluations did not consider that any changes were warranted.
Since the 422V+ fuel has a larger diameter than OFA and Upgraded OFA fuel for self attenuation of gamma radiation, divider wall gamma heating should not be significantly greater for the average 422V+ fuel than for previous fuel designs. However, since the Boraflex surveillance program is no longer required, the practice of storing recently discharged fuel at these two locations will be discontinued. Accordingly, the Technical Specifications were revised to remove the exception statement,"Except for the two storage locations adjacent to the designated slot for the spent fuel storage rack neutron absorbing material surveillance specimen irradiation" in TS 15.5.4.4. This proposed change would prevent unnecessary thermal stressing of the divider wall, and avoid the need for a new divider wall gamma heating calculation.
NPL 99-0369 - Safety Evaluation -Safety Evaluation for implementation of the 422V+ fuel assemblies at Point Beach Nuclear Plant Units 1 and 2 1
Included in this attachment is a summary of the safety evaluations and analyses that were performed by Westinghouse to confirm the acceptable use of the 422V+ fuel assembly design at Point Beach Nuclear Power Plant Units I and 2.
1 I
._