ML20207S421
| ML20207S421 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 03/13/1987 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19292G953 | List: |
| References | |
| NUDOCS 8703190423 | |
| Download: ML20207S421 (16) | |
Text
'
ATTACHMENT I TO JPN-86-63 Rev. 1 PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING RELOAD 7/ CYCLE 8 (JPTS-86-023) l i.
l NEW YORK POWER AUTHORITY l
JAMES A.
FITZPATRICK NUCLEAR POWER PLANT l
Docket No. 50-333 l
DPR-59 l
8703190423 870313 PDR ADOCK 05000333 P
PDR I
l
JAFNPP LIST OF FIGURES
!~
Figure
' Title Page 3.1-1 Manual Flow Control 47a l 3.1-2 OperatingLimitMCPRversus2I 47b 4.1-1 Graphic Aid in the Selection of an Adequate Interval Between Tests 48 4.2-1
' Test Interval vs. Probability of System Unavailability.
87 3.4-1~
Sodium Pentaborate Solution of System Volume-Concentration Requirements 110 3.4-2
-Saturation Temperature of Sodium Pentaborate Solution 111 3.5-1 Thermal Power and Core Flow Limits of Specifications l
3.5.J.1, 3.5.J.2 and 3.5.J.3 134 3.5-6 (Deleted) 135d 3.5 (Deleted) 135e 3.5-8 (Deleted) 135f
(
3.5-9 (Deleted) 135g 3.5-10 MAPLHGR Versus Planar Average Exposure Reloads 4 & 5, P8DRB299 135h 3.5-11 MAPLHGR Versus Planar Average Exposure Reloads 6 & 7, BP8DRB299, QUAD +
1351 3.5-12 MAPLHCR Versus Planar Average Exposure Reload 7 BD319A 135j 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4.6-1
' Chloride Stress Corrosion Test Results at 500*F 164 6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260 Amendment No. M, N, 48, $(, ll, N, 86, jW vii
JAFMPP 1
3.5 (cont'd) 4.5 (cont'd) i condition, that pump shall be considered inoper-2.
Following any period where the LPCI subsys-able for purposes satisfying Specifications tems or core spray subsystems have not been l
3.5.A, 3.5.C. and 3.5.E.
required to be
- operable, the discharge j
piping of the inoperable system shall be t
H.
Average Planar Linear Heat Generation Rate vented from the high point prior to the
]
(APLHGR) return of the system to service, i
)
During power operation, the APLHGR for each type 3.
Whenever the HPCI, RCIC, or Core Spray System 1
of fuel as a function of axial location and is lined up to take suction from the conden-average planar exposure shall be within limits sate storage tank, the discharge piping of based on applicable APLHGR limit values which the HPCI, RCIC, and Core Spray shall be have been approved for the respective fuel and vented from the high point of the system, lattice types.
When hand calculations are and water flow observed on a monthly basis.
required, the APLHGR for each type of fuel as a function of average planar exposure shall not 4.
The level switches located on the Core Spray exceed the limiting value for the most limiting and RHR System discharge piping high points lattice (excluding natural uranium) shown in which monitor these lines to insure they are Figures 3.5-10 through 3.5-12 during two full shall be functionally tested each month.
recirculation loop operation.
During single loop operation, the APLHGR for each fuel type shall H.
Averate Planar Linear Heat Generation Rate not exceed the above values multiplied by 0.84 (APLHGR)
(see Bases 3.5.K.
Reference 1).
If at anytime during reactor power operation greater than 25%
The APLHGR for each type of fuel as a function of of rated power it is determined that the limiting average planar exposure shall be determined daily value for APLHGR is being exceeded, action shall during reactor operation at2.25% rated thermal then be initiated within 15 minutes to restore power.
operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed limits within two (2)
- hours, an orderly reactor pcwer reduction shall be l
commenced immediately.
The reactor power shall l
be reduced to less than 25% of rated power within the next four hours, or until the APLHGR is returned to within the prescribed limits.
l Amendment No.
123 i
JAFNPP 3.5 BASES (cont'd) requirements for the emergency diesel generators.
are within the 10 CFR 50 Appendix K limit.
The limiting values for APLHGR are given in Figures I
G.
Maintenance of Filled Discharge Pipe 3.5-10 through 3.5-12.
Approved limiting values of APLHGR as a function of fuel type are given in If the discharge piping of the core spray, LPCI, NEDO-21662-2 (as amended) for Reload 5 and 6 RCIC, and HPCI are not filled, a water hammer can fuel.
Approved limiting values of APLHCR as a develop in this piping when the ' pump (s) are function of fuel and lattice types are given in started.
To minimize damage to the discharge NEDC-31317P for Reload 7 fuel.
These values are piping and to ensure added margin in the opera-multiplied by 0.84 during Single Loop Operation.
tion of these systems, this technical specifica-The derivation of this multiplier can be found in tion requires the discharge lines to be filled Bases 3.5.K Reference 1.
whenever the system is required to be operable.
If a discharge pipe is not filled, the pumps the I.
Linear Heat Generation Rate (LHGR) supply that line must be assumed to be inoperable for technical specification purposes.
However.
This specification assures that the linear heat if a water hammer were to occur, the system would generation rate in any rod is less than the still perform its design function.
design linear heat generation.
H.
Average Planar Linear Heat Generation Rate The LHGR shall be checked daily during reactor (APLHGR) operation atl 25% rated thermal power to deter-mine if fuel burnup, or control rod movement, has This specification assures that the peak cladding caused changes in power distribution.
For LHGR temperature following the postulated design basis to be a limiting value below 25% rated thermal loss-of-coolant accident will not exceed the limit power, the ratio of local LHGR to. average LHGR specified in 10 CFR 50 Appendix K.
would have to be greater than 10 which is pre-cluded by a considerable margin when employing The peak cladding temperature following a
any permissible control rod pattern.
postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly.
Since expected local variations in power distri-bution within a fuel assembly affect the calcu-lated peak clad temperature by less than i 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures Amendment No. [, k,
,[
130
9 JAFNPP~
Figure 3.5-9 i
(This page is intentionally blank.)
t Amendment No. f4, M 135g
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JAFNPP 1
Figure 3.5-11 j
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Reload 6 BP8DRB299
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Reload 7 QUAD +
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12-eg mw u) 25 5a d"
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a 5
10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
Versus Planar Average Exposure For single-loop operation, these MAPLHGR References NEDO-21662-2 values are multiplied by 0.84.
(As amended December 1984)
Amendment No. JMr7%,
1351
i JAFNPP.
i Figure 3.5-12
. Maximum Averace Planar Linear Heat Generation Rate (MAPLHGR)
Versus Averace Planar Exoosure i
Reload 7 BD319A-l 13 -
Reference:
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Ed b5 11 -
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10 15 20 25 30 35 40 45 50 i
Planar Average Exposure (GWd/St)
For single-loop operation, these This curve represents the limiting MAPLHGR values are multiplied by 0.84.
exposure dependent MAPLHGR values.
i Amendment No.
135j
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8 ATTACHMENT II TO JPN 63' Rev. 1
..f SAFETY E'7ALUATION FOR PROPOSED TECHNICAL fiPEb5FI{J6 TION CHANGES REOARDING RELOAD 7/ CYCLE 8 (JPTS-8 6-02 3 )
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,1 NEW YORK POWER AUTHORITY i
- JAMES A. FITZPATRICX NUCLEAR POWER PLANT-Docket No. 50-333 DPR-59 s
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DESCRIPTION OF THE PROPOSED CHANGES On page vii, the entry for Figure 3.5-9 has been deleted and an e'ntry for a new Figure 3.9-12 is added.
Also on page vii, changes are made to the entries for Figures 3.1-2 and 3.5-1 to add text inadvertantly missing from the Authority amendment request. submittal, approved and issued by the NRC a's Amendment 98.
These changes involve inserting the Greek letter "'g ", and the reference to specification 3.5.J.3.
On page 6, Section 1.0.U.2 is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB fuel loaded as Reload 7, and that the LHGR limit remains 13.4 KW/ft for the remainder of the m
fuel.
i On page 9, Section 2.1.A.1.c, the definition for MFLPD is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB i
Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On page loa, Section 2.1.A.1.d, the definition for MFLPD is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On page 13, the BASES for Section 1.1.A is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On pages 30f through 31a, Section 3.1.B has been revised and reorganized.
Minor changes are made to the text to correct typographical errors and for clarification.
On page 31, Section 3.1.B.1, the "MCPR Operating Limit for Incremental Cycle Core Average Exposure" table has been revised to reflect the transient analyses performed for the Reload 7/ Cycle 8 core (Reference 3).
On page 31a, Section 4.1.E.3, the values ofj9 andor are revised from 0.723 and 0.054 to 0.706 and 0.016 repectively in
- accordance with the accepted procedure as documented in Reference 5.
On page 35, Bases 3.1.B, the reference for the Cycle 8 Loss-Of-Coolant-Accident analyses is added.
On page 43, Notes for Table 3.1-1, Item 12, the definition for MFLPD is changed to reflect the 14.4 Kw/ft LHGR limit for the GE8X8EE Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
Also in this note, the definition
.of "S"
is changed from " percent of initial" to " Percent of rated the2nal power.
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'. On page 47b, the " Operating Limit MCPR Versus 77 (Defined in l
Section 3.1.B.2) for All Fuel Types" graph is revised to reflect the transient analyses performed for the Reload 7/ Cycle 8 core (Reference 3).
On page 123, Section 3.5.H is revised to reflect the removal of Figure 3.5-9 and the addition of Figure 3.5-12.
Other changes to this Specification are made in accordance with Reference 9.
L On page 124, Section 3.5.I is changed to reflect the 14.4 KW/ft LHGR limit for the GE8X8EB Reload 7 fuel, and that the LHGR limit remains 13.4 KW/ft for the remainder of the fuel.
On page 130, Basis 3.5.H is revised to reflect the removal of Figure 3.5-9 and the addition of Figure 3.5-12.
The other changes to this paragraph are made for clarity.
,,Page'135g has been deleted.
Page 1351 is changedsto show that Figure 3.5-11 is applicable to the Quad + fuel design.
Figure 3.5-12 is added,on a new page 135j.-
On page 245, Section 5.2 is changed to reflect the current fuel designs used in the FitzPatrick core.
II.
PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed changes is to support plant start-up and operation after the Reload 7/ Cycle 8 refueling outage.
During this outage, 188 fuel bundles are to be removed from the reactor core and replaced with new fuel.
As part of a program to qualify Westinghouse Nuclear Energy Systems as a vendor of nuclear fuel for FitzPatrick, four (4) demonstration lead test assemblies of their QUAD + design are included in Reload 7.
The changes to the Technical Specifications involve deleting specifications associated with the discharged fuel and with Cycle 7 specific analyses.
These are replaced with ones appropriateforthenewfuelandbasedonCycle8 specific i
l analyses.
l' The changes to page vii are purely administrative.
The " List
\\
of Figures" is updated to reflect changes made as part of this proposed change.
Other changes correct existing administrative errors on this page.
The changes to pages 6, 9,
10a, 13, 43, and 124 reflect a new LHGR limit"of 14.4 KW/ft for the fuel type GE8X8EB added as Reload 7 and asfaccepted and documented in Reference 5.
J '
i I
t The reorganization and other administrative changes to pages 30f through_31a have been made to facilitate use of Section 3.1.B.
The changes to the "MCPR Operating Limit for Incremental Cycle Core Average Exposure" table in Section 3.1 on page 31 and the " Operating Limit MCPR Verses 17 (Defined in Section 3.1.B.2)
For All Fuel Types" graph on page 47b reflect cycle specific transient analyses performed by General Electric for the Reload 7/ Cycle 8 core (Reference 3).
The results of these analyses are included in this application as Attachment III.
Changes to the values of 6 and a cn1 page 31a are made to reflect the use of the GEMINI transient methods for calculating the transient A CPR.
This General Electric methodology has been approved by the NRC and is documented in Reference 5.
General Electric used this methodology in their Cyclo 8 specific transient analyses for FitzPatrick. (Reference 3).
The changes to page 35 add the reference to the Cycle 8 specific Loss-of-Coolant-Accident analysis.
The changes to pages 123 and 130 relect the deletion of Figure 3.5-9 and the addition of Figure 3.5-12.
This reflects the removal of fuel type P8DRB2h4H and the addition of fuel type BD391A as part of this reload.
The other changes to Specification 3.5.H are made in accordance with Reference 9.
The differences between Reference 9 and the proposed changes are made to adapt the generic Specification to FitzPatrick specific analyses and Technical Specifications.
The GE8X8EB fuel added as Reload 7 contains several 1r.ttice types of varying gadolinium content.
To determine the proper MAPLHGR value for a particular axial location in a fuel bundle, the MAPLHGR tables in the LOCA analysis (Reference 4) will be programmed into the plant process computer and backup computer system.
When hand calculations are necessary, the most limiting enriched uranium lattice MAPLHGR value is applicable.
The exposure dependent limiting MAPHIGR values are shown in Figure 3.5-12.
MAPLHGR limits for the QUAD + demonstration assemblies will be those applicable to the Reload 6 BP8DRB299 bundles following the practice approved by the NRC in Reference 7 in which the licensee applied the MAPLHGR limits for the General Electric assembly to the matched QUAD + assemblies.
Section 5.2 on page 245 describes the physical design of fuel bundles in the FitzPatrick core.
This section is revised to reflect the addition of fuel types GE8X8EB and QUAD + inserted as 1
Reload 7.
Fuel type 8X8R is no longer used in the FitzPatrick core and is deleted from this section.
Fuel type BP8X8R, added as Reload 6, was inadvertantly omitted from this section in the Technical Specification amendment submittal for Reload 6/ Cycle 7.
III. IMPACT OF THE PROPOSED CHANGES The impact of the proposed changes would be to allow startup and operation of FitzPatrick following the upcoming Reload 7/ Cycle 8 refueling outage.
This outage is currently scheduled to begin in January 1987.
1 Of the 188 new fuel bundles to be inserted into the FitzPatrick core for Reload 7, four (4) are supplied by Westinghouse Nuclear Energy Systems and are designated as fuel type QUAD +.
These bundles are demonstration lead test assemblies to be used as part of the program to qualify Westinghouse as a vendor of nuclear fuel for FitzPatrick.
A description of the nuclear, thermal, and mechanical characteristics of this bundle i
type has previously been transmitted to the NRC for review in i
Reference 6.
The QUAD + fuel bundles were designed to be compatible with 4
the General Electric fuel used in the FitzPatrick core.
The nuclear and hydraulic properties of the QUAD + bundles have been i
i designed to be as nearly identical to the BP8DRB299 bundles inserted as Reload 6 as practical.
For the purpose of the General Electric Cycle 8 specific analyses, fuel bundles of type BP8DRB299 were assumed to reside in the core locations designated for the QUAD + bundles.
These bundles are very similar to the QUAD + bundles approved for use in the Brown's Ferry Unit 2 core (Reference 7).
The core loading constraints applied by Westinghouse and the TVA for the Brown's Ferry QUAD + lead test assemblies also apply at FitzPatrick.
These six constraints are given in Reference 6.
The calculations done to support the conclusion that the QUAD +
assemblies will operate with a margin of at least 20% in bundle power to the lead assembly during cycle 8 are contained in Reference 8.
The application of the six constraints and the calculations demonstrating the required 20% power margin support the conclusion as given in the NRC safety evaluation report that the QUAD + assemblies will have no adverse safety impact on Cycle 8 operation at FitzPatrick.
The remaining 184 bundles of Reload 7 are of fuel bundle type i
GE8X8EB and are designated BD319A.
This fuel type incorporates the design features described in Reference 5 for GE8X8EB fuel.
Analyses on the Cycle 8 core at FitzPatrick have resulted in small changes in the MCPR values over the Cycle 7 values.
Use of these new MCPR values, as documented in this application as Attachment III, will assure no adverse safety impact in the reloaded core.
- - - - - - - - - - - - - - - - - - - - - - - - ~ ~ ~ ~ - ' ~ ~ ~ ' -
a.
A;new. loss of coolant accident (LOCA) analysis has.been performed for FitzPatrick in_ order to support operation of the-GE8X8EB fuel to its design power levels (Attachment IV).
The-analysis has been. performed using codes and methods approved by-the NRC.~
The-full-break spectrum-has been. analyzed with appropriate failure criteria applied and nominal input values assumed.
Peak clad temperatures (PCT) based on Appendix K criteria were calculated at discrete points in the break spectrum, as provided in the approved application methodology. 'These. peak clad temperatures show considerable margin _to PCT limits:and demonstrate no adverse safety impact associated with the use of the.GE8X8EB MAPLHGR limits as given in this application..
The maximum linear heat' generation rate (MLHGR) limit of 14.4.KW/ft for GE8X8EB fuel as applied for-in this application is the MLHGR-considered in the NRC review of this fuel type. ' Generic NRC acceptance of this' fuel design is documented as Amendment 10-to Reference 5.
The eo and 0" parameters used for scram speed surveillance j
for MCPR limits.are changed to be consistent with the requirements of the GEMINI transient-methods approved as Amendment ll to Reference 5.
IV.
EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatrick plant in accordance with the porposed' Amendment would not involve a significant hazards consideration as statedJin 10 CFR 50.92 since it would not:
~1.
involve a significant increase in the probability or consequences of an accident previously evaluated.
Approved methodologies and codes have been used~to perform all analyses concerning-the General Electric fuel to be loaded'at this refueling (Reference 5).
The fuel design has been reviewed and approved for use at FitzPatrick under the constraints and methodologies detailed in Reference 5.
There are no unique aspects of this fuel or its application which have not undergone prior NRC review and' approval.
p i
i
9 s
The Westinghouse fuel to be introduced has been designed and will be operated within the constraints detailed in a previous NRC safety evaluation report approving this type of demonstration program generically for BWRs 3 through 6 (Reference 7).
The four QUAD + assemblies were designed to be compatible with currently used bundles and possess, as closely as possible, the operational characteristics of a previously used and accepted 4
General Electric bundle.
Constraints are applied to the use of the QUAD + bundles to assure that none of them will become a lead assembly during operation, nor a limiting bundle under transient conditions.
Therefore, the probability or consequences of the accidents previously evaluated and described in the FSAR have not been increased.
2.
create the possibility of a new or different kind of accident from any accident previously evaluated.
Refueling the FitzPatrick reactor is a periodic evolution. performed in accordance with appropriate procedures and controlled by the Technical specifications.
The fuel bundles inserted as Reload 7 are not sufficiently different from previously used bundles as to create the possibility of a new or different type of accident.
The assemblies have been fully reviewed and approved for use in power reactors by i
the NRC (References 5 and 7).
3.
involve a significant reduction in a margin of safety.
The analyses performed in support of this reload assure 4
maintenance of existing margins of safety.
These analyses have resulted in core wide (MCPR) and bundle specific (MLHGR and MAPLHGR) limits for General Electric fuel which, when applied to the reloaded core, assure operation within the design criteria previously approved y
in Reference 5.
Through core positioning constraints and inherent design features, use of' appropriate existing bundle specific limits and core-wide MCPR limits for the QUAD +
s assemblies is conservative as concluded in Reference 7.
In the April 6, 1983 FEDERAL REGISTER (48FR14870), the NRC published examples of license amendments that are not likely to involve significant hazards considerations.
Example number (iii) of that list is applicable to this proposed change and states in part:
"For a nuclear power reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are involved."
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V.
IMPrRMENTATION OF THE PROPOSED CHANGE
-l Implementation of these changes, as proposed, will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.
VI.
CONCLUSION The change, as proposed, does not constitute an unreviewed safety question as defines in 10 CFR 50.59, that is it:
will not change the probability nor the consequences of i
a.
j an accident or malfunction of equipment important to safety as previously evaluated in the safety Analysis Report; b.
will not increase the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; will not reduce the margin of safety as defined in the c.
l basis for any technical specification; d.
does not constitute an unreviewed safety question; and 1
involves no significant hazards consideration, as e.
defined in 10 CFR 50.92.
VII.
REFERENCES 1.
James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR) as updated.
2.
James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER).
3.
General Electric Report, " Supplemental Reload Licensing Submittal for James A. FitzPatrick Nuclear Power Plant Reload 7," 23A4825, November, 1986.
(Included as Attachment III).
4.
General Electric Report, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR - LOCA Loss-of-Coolant Accident Analysis,"
NEDC-31317P, October 1986.
(Included as Attachment IV).
5.
General Electric Licensing Topical Report, "GESTAR II General Electric Standard Application for Reactor Fuel," NEDE 240ll-P-A-8, May 1986.
.,0
't 6 ~.
-Westinghouse letter, E.P. Rahe to R.M.
Bernero, dated December 10, 1986, (NS-NRC-86-3187), submitting Westinghouse report WCAP-11159 for NRC review.
7.
NRC letter, M. Grotenhuis to S.A. White-(TVA), dated August 19,-1986, providing the Safety Evaluation supporting Amendment No. 125 to TVA's Bwowns Ferry Unit 2.
8.
-Westinghouse letter,'J.P. Ducruet to G.L. Rorke, dated December 15, 1986, providing a report entitled, "FitzPatrick QUAD + Demonstration Assembly Power Margin and Monitoring."
9.
General Electric letter, J.S. Charnley to M.W. Hodges (NRC),
dated March 4, 1987, (MFN-021-087/JSC-027-087), " Recommended MAPLHGR Technical Specifications for multiple lattice fuel designs."
10.
General Electric letter, J.S. Charnley to G.C. Lainus-(NRC),
dated October 21, 1986, " Proposed Amendment 18 to GE Licensing Topical Report NEDE-240 11-P-A (GE8X8NB Fuel)".
___ __ _ _ _