ML20147E630
| ML20147E630 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/10/1997 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20147E628 | List: |
| References | |
| NUDOCS 9703180064 | |
| Download: ML20147E630 (12) | |
Text
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j ATTACHMENT A-1 Beaver Valley Power Station, Unit No. 1
)
Proposed Technical Specification Change No. 238 j
The following is a list of the affected pages:
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9703180064 970310
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'DPR-66 5.0 DESIGN FEATURES 5.1 SITE LOCATION The Beaver Valley Power Station Unit No. 1 is located in Shippingport Borough, Beaver County, Pennsylvania, on the south bank of the Ohio River.
The site is approximately 1 mile southeast of Midland, Pennsylvania, 5 miles east of East Liverpool, Ohio, and approximately 25 miles northwest of Pittsburgh, Pennsylvania.
The exclusion area boundary has a minimum radius of 2000 feet from the center of containment.
5.2 REACTOR CORE 4
5.2.1 FUEL ASSEMBLIES The reactor shall contain 157 of fircaloy {1:d fuel assemblies.
Each. assembly shall consist of a matrix fuel rods with an initial l
composition of natural or slightly enriched uranium dioxide (UO ) as 2
fuel material.
Limited substitutions of zirconium alloy or stainless stee:
Piller rods for fuel
- rods, in accordance with approved applisations of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been 1
analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design i
bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
5.2.2 CONTROL ROD ASSEMBLIES The reactor core shall contain 48 full length and no part length control rod assemblies.
The full length control rod assemblies shall contain a nominal 142 inches of absorber material.
The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.
All control rods shall be clad with stainless steel tubing.
1 5.3 FUEL STORAGE 5.3.1 CRITICALITY 5.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum U-235 enrichment as set forth in Specification 3.9.14; b.
K,gg 5 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in UFSAR Section 9.12; BEAVER VALLEY - UNIT 1 5-1 Amendment No.
Otoys& LAI &m, y )
O
- DPR-66 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)
Anderson to K.
Kniel (Chief of Core Performance Branch, 4.
T.
Mf
Attachment:
Operation and Safety NRC) January 31, 1980 Analysis Aspects of an Improved Load Follow Package.
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control 5.
NUREG-0800, Standard Review Plan, U.
S.
Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981.
Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.
2, July 1981.
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical
- limits, core thermal-hydraulic
- limits, ECCS
- limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS
- REPORT, including any mid-cycle revisions or supplements thereto, shall be provided on issuance, for each reload cycle, to the NRC Document control Desk.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.
S.
Nuclear
~
Regulatory Commission, Document Control
- Desk, within the time period specified for each report.
These reports shall be submitted l
covering the activities identified below pursuant to the requirements of the applicable reference specification:
a.
ECCS Actuation, Specifications 3.5.2 and 3.5.3.
b.
Inoperable seismic Monitoring Instrumentation, Specification 3.3.3.3.
c.
Inoperable Meteorological Monitoring Instrumentation, Specificatior. 3.3.3.4.
d.
Seismic event analysis, Specification 4.3.3.3.2.
e.
Sealed source leakage in excess of limits, Specification 4.7.9.1.3.
f.
Miscellaneous reporting requirements specified in the Action Statements for Appendix C of the ODCM.
g.
DELETED (o. WCM-lM/O - P-k "VANT~i466 + fuel Assedly)Aafs+enen &n Rep + f," S Neskgom A,p,'e)aq. A%dolyy aglod %pi) its.rik hilowQ Syst.kcJie'n : t.a.n, Nst Amendment No. @
BENEdVA
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6-20 O*) Meed idedsf)
ATTACHMENT A-2 Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 114 The following is a list of the affected page:
Affected Page: 6-20 4
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NPF-73 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) 2.
WCAP-10266-P-A Rev.
2
/
WCAP-11524-NP-A Rev.
2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code,"
- Kabadi, J.
N.,
et al., March 1987; including Addendum 1-A
" Power Shape Sensitivity Studies" 12/87 and Addendum 2-A
" BASH Methodology Improvements and Reliability Enhancements" 5/88.
Methodology applied for the following Specification:
3.2.2, Heat Flux Hot Channel Factor-F (Z).
q 3.
" POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT,"
September 1974 (Westinghouse j
Proprietary).
Methodology applied for the following Specification; 3.2.1, Axial Flux Difference-Constant Axial Offset Control.
4.
T.
M. Anderson to K. Kniel (Chief of Core Performance Branch, i
NRC)
January 31, 1980
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control.
5.
NUREG-0800, Standard Review
- Plan, U.S.
Nuclear Regulatory Commission, Section
'4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev.
2, July 1981. Methodology applied for the following Specification:
3.2.1, Axial Flux Difference-Constant Axial Offset Control.
operating limits shall be determined so that all applicable The core limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic
- limits, ECCS
- limits, nuclear limits such as shutdown margin, and transient and -accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS
- REPORT, including any mid-cycle revisions or supplements
- thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk.
SPECIAL REPORTS.
6.9.2 Special reports shall be submitted to the U.S.
Nuclear Regulatory Commission, Document Control Desk within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable refe'ence specification:
r a.
ECCS Actuation, Specifications 3.5.2 and 3.5.3.
b.
Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.
G. WCAP-/MIC-#- A
- VANTA6E+ kl 4sse,dl Ae4emee Cort Report >" httil 1997(Wes;hyouse Pro heidle Go,nl Acht -4(s)y spat.c;cho;,f+Et ety), kJ appHet he A 4thW
- 7.2.x i M
BEAVER VALLEY - UNIT 2 6-20 Amendnent No. 44-YNy6$tt)$$$0te$Enh
a ATTACHMENT B Beaver Valley Power Station, Unit No. 1 and No. 2 Proposed Technical Specification Change Nos. 238 and 114 ZIRLO CLAD FUEL A.
DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would modify the technical specifications based on our previous request (letter dated September 9,
1996) currently under NRC review concerning modification of the technical specification (TS) design feature section to reflect the Improved Standard Technical Specifications (ISTS) of NUREG-1431.
This change would revise Unit 1 Design Feature 5.2.1, Fuel Assemblies, to allow use of ZIRLO as an alternate zirconium based fuel rod material and remove the word clad since it has been eliminated from the text of the ISTS.
Limited substitution of fuel rods by ZIRLO filler rods is currently permitted.
In addition, reference to the NRC approved ZIRLO topical report is added to Specification 6.9.1.12 for both units.
ZIRLO was approved for use at Unit 2
by Amendment No.
82 dated September 13, 1996.
B.
BACKGROUND Design Feature 5.2.1 requires fuel rods to be constructed with zircaloy.
Zirconium alloy or stainless steel filler rods may be substituted in place of fuel rods in accordance with approved applications of fuel rod configurations.
The proposed amendment modifies Design Feature 5.2.1 to also allow fuel rods to be constructed with ZIRLO.
C.
JUSTIFICATION The change to ZIRLO clad fuel is consistent with 10 CFR 50.44 and 10 CFR 50.46.
The change is also consistent with NRC approved topical
- report, WCAP-13060,
" Westinghouse Fuel Assembly Reconstitution Evaluation Methodology," which meets the intent of Supplement 1
" Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications."
" Standard Technical Specifications for Westinghouse Plants," specifically includes ZIRLO as an acceptable material when supported by a
plant specific review.
Changing to ZIRLO is one phase of a transition to higher burnup fuel.
Future core designs may feature longer cycles, higher capacity
- factors, and ultimately, higher discharge burnups.
Using higher discharge burnup fuel in the reactor core design reduces the number of fuel assemblies required per reload.
This will save money by paying less for fuel fabrication and by using less spent fuel storage space.
In order to support the required fuel enrichment and burnups, advanced alloys of zirconium must be used to maintain fuel integrity.
B-1
ATTACHMENT B, continund Proposed Tschnical Specification Change Nos. 238 and 114 Page 2 D.
SAFETY ANALYSIS l
In Federal Register Volume 57, Number 169, dated August 31, 1992, the NRC published amended regulations to reduce regulatory burden on nuclear licensees.
The NRC revised the acceptance criteria in 10 CFR 50.44 and 10 CFR 50.46 relating to evaluations of emergency core cooling systems and combustible gas control applicable to Zircaloy fuel rods to include ' ZIRLO fuel rods.
ZIRLO is a preferred material since it provides a significant
)
improvement in corrosion margin and fuel integrity.
The NRC noted that the revision to include ZIRLO as an acceptable zirconium based material along with Zircaloy will reduce the licensee burden but will not reduce the protection of the public health or safety.
The change eliminates the need to obtain 4
exemptions in order to use fuel material not presently addressed in the regulations.
An analysis of the safety implications is provided in a NRC letter to Westinghouse dated July 1, 1991, titled " Acceptance for Referencing of Topical Report WCAP-12610, ' Vantage + Fuel Assembly Reference Core Report' (TAC NO 77258)."
This safety evaluation and a later one titled " Acceptance for Referencing of Licensing Topical Reports WCAP-12610, Appendices F,
"LOCA NOTRUMP Evaluation Model:
ZIRLO Modifications,"
and G,
"LOCA Plant Specific Accident Evaluations" approved the use of the Vantage +
fuel design, i.e.,
ZIRLO clad fuel, described in WCAP-12610 and found it acceptable. The WCAP-12610 report supports the following conclusions:
1.
The mechanical design bases and limits for the ZIRLO clad fuel assembly design are the same as those for the previously licensed Zircaloy-4 clad fuel assembly design, except those for clad corrosion.
2.
The neutronic evaluations have shown that ZIRLO clad fuel nuclear design bases are satisfied and that key safety parameter limits are applicable.
The nuclear design models and methods accurately describe the behavior of ZIRLO clad fuel.
3 3.
The thermal and hydraulic design basis for the ZIRLO clad fuel is unchanged.
4.
The methods and computer codes used in the analysis of the non-loss of coolant accident licensing basis events are valid for ZIRLO clad fuel, and all licensing basis criteria will be met.
5.
The large break and small break loss of coolant accident (LOCA) evaluation models have been modified to reflect the behavior of the ZIRLO clad material during a LOCA.
It is concluded that the revised evaluation model satisfies the intent of 10 CFR 50.46 and Appendix K of 10 CFR 50.
There B-2
j ATTACHMENT B, continusd l
Proposed Technical Specification Change Nos.'238 and 114
,Page 3 4
i is no significant impact on typical large break and small break LOCA analyses results for the ZIRLO model revisions.
In addition, bounding -LOCA rod heatup cases were evaluated and j
all acceptance criteria were
- met, including
.those in i
Other LOCA-related accident analyses (Long Term Core Cooling, Hot Leg Switch Over, and hydraulic forces on the reactor vessel components) have been evaluated and are not affected by the implementation. of ZIRLO clad fuel.
Adequate margin.to
,the peak clad temperature limit of 2200 F is i.
' maintained.
i The effect of ZIRLO on the UFSAR Section 14.2.7 locked rotor transient is minimal.
Sensitivity analyses performed by Westinghouse demonstrate that the impact on peak cladding temperature and metal-to-water reaction on the locked rotor transient results is insignificant.
4 The rod control cluster assembly (RCCA) ejection event (UFSAR Section 14.2.6) was analyzed at hot full power and hot zero power.
'The analysis demonstrated that any consequential damage to the core or the reactor coolant system would not prevent long i
i term core - cooling and that offsite dose would remain within the guidelines of 10 CFR 100.
WCAP-12610 includes results of sensitivity analyses performed by Westinghouse that ' demonstrate that the impact of ZIRLO on RCCA ejection event analyses results in an insignificant change in both the fraction of fuel melted at the hot spot as well as the peak fuel stored energy.
ZIRLO is an alloy of zirconium; therefore, the use of ZIRLO filler rods is addressed in Design Feature 5.2.1 as zirconium alloy.
WCAP-13060 delineates the methodology used to evaluate applicable design criteria associated with reconstituted fuel assemblies that have. solid filler rods replacing uranium filled fuel rods.
Evaluations and analyses of fuel assembly-l reconstitution will be performed on a cycle specific basis whenever reconstituted fuel assemblies are used in the reactor 4
. core.
The WCAP included proposed technical specification changes based on the conclusions in the WCAP and on the guidelines of GL 90-02.
Fuel configuration, size, enrichment and cladding material shall i
.be limited to those designs that have been analyzed with 2 -
applicable NRC-approved codes and. methods, and shown by test or cycle specific reload analyses to comply with all fuel safety design bases.
The use of ZIRLO fuel cladding or filler rods wil]
be justified by a-cycle specific reload analysis, in accordance 4
with NRC approved applications of fuel rod configurations.
The
]
justification of the core analysis methods must address the effect on core-wide analyses of permissible core configurations l
with the reconstituted fuel.
I, B-3
ATTACHMENT B, continund Proposed Technical Specification Change Nos. 238 and 114 Page 4 i
i The proposed change modifies Design Feature 5.2.1, Fuel Assemblies, to allow the use of the enhanced corrosion resistant fuel rod material ZIRLO.
Adding the WCAP-12610 reference to Administrative Control 6.9.1.12, Core Operating Limits Report, l
for both units ensures the analytical methods used to determine the core operating limits are consistent with those previously reviewed and approved by the NRC.
Changes to applicable sections of the UFSAR will be incorporated in the next UFSAR update following issuance of the amendment to reflect the change to i
ZIRLO clad fuel.
This change is consistent with the ISTS, the amended regulations, the NRC approved WCAP-12610 and is similar to changes incorporated by other plants (i.e.,
Byron and Braidwood, Amendments dated December 19, 1995, and Beaver Valley Unit 2 dated September 20, 1996).
In addition, the required i
accident analyses described above have been evaluated and found i
acceptable; therefore, the proposed change is safe and provides 4
l an effective alternative to zircaloy.
I Schedule Recuirements Unit 1 is planning to load fuel with ZIRLO cladding during the twelfth refueling
- outage, currently scheduled to begin on September 5, 1997.
Therefore, it is respectfully requested that the NRC Staff review and approve this license amendment request i
no later than September 5,
1997, so that the amendment is in i
~
place prior to loading the new ZIRLO clad fuel.
E.
NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below:
The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a
facility licensed under paragraph 50.21(b) or - paragraph 50.22 or for a
testing facility involves no significant hazards consideration, if operation of the. f acility 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.
The following evaluation is provided for the no significant hazards consideration standards.
B-4
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ATTACHMENT B, continund Proposed Technical Specification Change Nos. 238 ane 4,
t 3
Page 5 1.
Does the change involve a
significant increase in the probability or consequences of an accident previously evaluated?
i.
The methodologies used in the accident analyses h' ave been modified to reflect the requirements provided in WCAP-12610, VANTAGE + Fuel Assembly Reference Core Report.
Reference to this NRC approved ZIRLO topical report has been added to l
Specification 6.9.1.12, for both units to ensure the 1
analytical methods used to determine the core operating limits are consistent with those previously approved by the l
4 NRC.
The proposed changes.do not change or alter the design i.
assumptions for the systems or components used to mitigate the consequences of an accident.
Use of ZIRLO fuel rod material does not adversely affect fuel performance or impact nuclear design methodology.
Therefore, accident analysis results are not impacted.
i The operating limits will not be changed and the analysis
[
methods to demonstrate operation within the' limits will remain in accordance with NRC approved methodologies.
Other than the changes to the fuel assemblies, there are no physical changes to the plant associated with this technical specification change.
A safety analysis will continue to be performed for each cycle to demonstrate compliance with all fuel safety design bases.
VANTAGE SH fuel assemblies with ZIRLO fuel rods meet the same fuel-assembly and fuel rod design bases as other VANTAGE 5H fuel assemblies.
In addition, the 10 CFR 50.46 criteria are applied to the ZIRLO fuel rods.
The use of these fuel assemblies will not result in a change to the reload design and safety analysis limits.
Since the original design criteria are met, the ZIRLO fuel rods will not be an initiator for any new accident.
The fuel rod material is similar in chemical composition and has similar physical and mechanical properties as Zircaloy-4.
- Thus, the fuel rod integrity is maintained and the structural integrity of the fuel assembly is not affected.
ZIRLO improves corrosion performance and dimensional stability.
No concerns have been identified with respect to the use of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel rods.
The dose predictions in the safety analyses are not sensitive to the fuel rod material used; therefore, the radiological consequences of accidents previously evaluated in the safety analysis remain valid.
A reload analysis is completed for each cycle, in accordance with NRC approved methodologies.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
B-5
t ATTACHMENT B, continued Proposed Technical Specification Change Nos. 238 and 114 l
Page 6 l
2.
Does the change create the possibility of a new or different l
kind of accident from any accident previously evaluated?
VANTAGE SH fuel assemblies with ZIRLO fuel rods satisfy the same design bases as those used for other VANTAGE SH fuel
' assemblies.
All design and performance criteria continue to i
be met and no new failure mechanisms have been identified.
The ZIRLO fuel rod material offers improved corrosion resistance and structural integrity.
i The proposed changes do not affect the design or operation of any' system or component in the plant.
The safety functions j
of the related structures, systems, or components are not changed in. any
- manner, nor is the reliability of any 3
structure, system, or component reduced.
The changes do not
{
affect the manner by which the facility is operated and do not change any facility design feature, structure, or system.
i No new or different type of equipment will be installed.
Since there is no change to the facility or-operating j
procedures, and the safety functions and reliability of f
structures,
- systems, or components are not affected, the l.
proposed changes do not create the possibility of a new or different kind of accident from any accident previously F
evaluated.
i 3.
Does the change involve a significant reduction in a margin of safety?
The use of Zircaloy-4, ZIRLO, or stainless steel filler rods in fuel assemblies will not involve a significant reduction in the margin of safety because analyses using NRC approved j
methodology will be performed for each configuration to I
demonstrate continued operation within the limits that assure acceptable. plant response to accidents and transients.
These analyses will be performed using NRC approved methods that i
have been approved for application to the fuel configuration.
Use of ZIRLO as fuel rod material does not change the VANTAGE 5H reload design and safety analysis limits.
The use of
^
these fuel assemblies will take into consideration the normal core operating conditions allowed in the technical i
specifications.
For each reload core, the fuel assemblies i
will be evaluated using NRC approved reload design methods, including consideration of the core physics analysis peaking i
factors and core average linear heat rate effects.
i Based on the above, it is concluded that the proposed license amendment request does not result in a significant reduction i
in' margin with respect to plant safety as defined in the UFSAR or any plant technical specification BASES.
i l
B-6
w ATTACHMENT B, continund Proposed Technical Specification Change Nos. 238 and 114 Page 7 F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfies the no significant hazards consideration standards of 10 CFR 50.92(c)
- and, accordingly, a
no significant hazards consideration finding is justified.
G.
UFSAR CHANGES The UFSAR will be revised to include ZIRLO fuel rod material.
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