ML20065T102
| ML20065T102 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/17/1990 |
| From: | Beck G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20065T104 | List: |
| References | |
| NUDOCS 9012270151 | |
| Download: ML20065T102 (8) | |
Text
_ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _
l 10 CFR 50.90 PHILADELPHIA ELECTRIC COMPANY 4
NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.
WAYNE. PA 19087 5691 (sie sao sooo December 17, 1990 Docket Nos.
50-277 50-278 License Nos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission ATTN Document Control Desk Washington, D. C.
20555
SUBJECT:
Peach Bottom Atomic Power Station, Units 2 and 3 Technical Speci'ications Change Request
Dear Sir:
Philadelphia El"ctric Company hereby submits Technical Specifications Change Reguest No. 90-11, in accordance with 10 CFR 50.90, requesting an amendment to the Technical Specifications (Appendix A) of the Peach Bottom Facility Operating Licenses.
These changes are necessary to account for r.cw fac1 type being used in Cycle 9 operation of Units 2 and 3.
Cycle 9 of Unit 2 is scheduled to begin first on March 19, 1991.
Miscellaneous administrative cha igea are also propotad. to this letter describes the proposed changes, and provides justification for the changes.
The fuel related changes were selected in accordance with NRC-approved methods. contains the revised Technical Specifications pages.
If you have any questions, please do not hesitate to contact Mr. Frank Iear of my staff at (215) 640-6786.
Very truly yours, p
Gk 6
L G.
J.
Beck Manager-Licensing Section Nuclear Engineering & Services
Enclosure:
Affidavit Attachments 1, 2
/
cc:
T.
T. Martin, Administrator, Region I, USNRC 1
J. J. Lyash, USNRC Senior Resident Inspector
/
T. M. Gerusky, Commonwealth of Pennsylvania t(
9012270151 901217 ADOCK0500g{7 PDR 9j ans
' C,OMMONWEALTH OF PEN!JSYLVANI A i
ss.
COUNTY OF CHESTER D. R. Helwig, being first duly sworn, deposes and says:
That he !.s Vice President of Philadelphia Electric Company; the Applicant herein; that he has read the attached request (number 90-11) for changes to Peach Bottom Pacility Operating Licenses DPR-44 and DPR-56, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
(
Vice Pr ent Subscribed and sworn to before me this /71ay of O,144 4 1990.
NA4
- )D.M.,
Notary Public UOTAR'AL OEAL CATHER'NE A. MENDE2, H?cy Puthe trot,ffnn Tru, Cheer Count /
Mv Co*Qygn,Egs & set 41993
\\)
ATTACHMENT 1 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 TECHNICAL SPECIPICATIONS CHANGE REQUEST No. 90-11
" Minimum Critical Power Ratio Safety Limits" f
l l
Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 INTRODUCTION Cycle 9 operation of Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3 necessitates revision of the Technical Specifications (TS) Minimum Critical Power Ratio (MCPR) Safety Limits since the cores will be reloaded with a new fuel type, GE8X8NB (commonly referred to as GE9B fuel).
Unit 2 Cycle 9 is scheduled to begin on March 19, 1991 and Unit 3 Cycle 9 is scheduled to begin on November 19, 1991.
PECo hereby requests that, once approved, these changes be " effective upon start-up in Cycle 9" for each Unit.
DESCRIPTION OF CHANGES Technical Changes:
The current Unit 2 TS MCPR Safety Limits are 1.07 for two-recirculation loop operation and 1.08 for single recirculation loop operation (page 9 of TS).
The current Unit 3 TS MCPR Safety Limits are 1.04 for two-recirculation loop operation and 1.05 for single recirculation loop operation (page 9 of TS).
However, use of GE9B fuel in Unit 2 and Unit 3 during Cycle 9 requires MCPR Safety Limits not less than 1.06 for two-loop operation and 1.07 for single loop operation.
Since the Cycle 9 cores of both units will be a reload of GE9B fuel, revision of the MCPR Safety Limits to 1.06 for two-loop operation and 1.07 for single loop operation is requested for both Units.
These changes are in accordance with Revision 9 of
" General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A-9, September 1988 (CESTAR), which was approved by the NRC in the letter f rom Ashok C. Thadani (NRC) to J.
S.
Charnley (GE) dated May 12, 1988.
GESTAR specifies a MCPR Safety Limit of 1.06 for D-Lattice reactors in two-loop operation.
Units 2 and 3 are D-Lattice reactors.
The Limits for two-loop operation are determined by using NRC-approved " General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application," NEDO-10958-A, January 1977.
The Limit is increased by 0.01 for single loop operation as described in " Peach Bottom Atomic Power Station, Units 2 and 3 Single-Loop Operation", NEDO-24229-1, May 1980, which was submitted to the NRC on January 9, 1981 to support license amendment-for single recirculation loop operation at PBAPS (amendments suosequently approved, No. 78 for Unit 2 and No. 77 for Unit 3).
The reload fuel for Cycle 9 operation of both Units will be GE9B, with the exception of twelve or less qualification fuel bundles (OPBs) in Unit 2. ___
Docket Nos. 50-277-50-278 License Nos. DPR-44 DPR-56 However, these OFBs will be loaded-in non-limiting locations such that the OPBs will not have a significant impact on the core-wide MCPR Scfety Lin ts.
This was the subject of PEco'c Aovember 21, 1990 letter to..he NRC.
Administrative Changes:
On June 15, 1990 the NRC issued a Safety Evaluation Report approving PECo Report No. PECo-FMS-0006, " Methods for Performing BWR Reload Safety Evaluations."
PECo requests that this report be referenced on the following pages of the Unit 2 and Unit 3 TS: 17, 24, 140a, 140b, 140c and 256a.
PECo proposes to add parentheses around the abbreviation "MCPR" on page 9 of both Units' TS (Specification No. 1.1.A), and to change "NEDO-24011-P-A" to "NEDE-240ll-P-A' on page 140c of both Units' TS (Reference No. 7).
These-changes correct typographical errors.
PECo proposes to add to the list of references on page
'24 of both Units' TS "NEDE-240ll-P-A" (GESTAR), which is currently " spelled out" in the text on page 17 of both Units' TS.
On Page 17 the document can now be referred to by its reference unmber (on page 24 as revised).
This change is in the interest of convenience and consistency.
INFORMATION SUPPORTING A FINDING OF NO SIGNIFICANT HAZAlD_S_
Technical Changes:
The MCPR Safety Limits are set such that no fuel damage-Eis calculated to occur if the limit is not violated.
Since-the parameters which result in fuel damage are not directly observable during reactor' operation, the thermal hydraulic
-conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.
Although it is recognized that.a departure from E
nucleate boiling would not necessarily result in damage to BWR Icel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.
L However, the uncertainties in monitoring the core operating state L
and in the procedure used to calculate the critical. power result-l in an uncertainty in the value of critical power.
Therefore, the MCPR Safety Limit is defined as the critical pcwer ratio for which more than 99.9% of the fuel rods in the core are expected to a"old boiling transition during the most severe moderate L
fregaency transient event, considering the power distribution L
wittin'the core and all uncertainties.
l l l l
o
Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 As discca' d previously, the proposed MCPR Safety Limits have been establi
.d in accordance with NRC-approved methods.
In addition, concervative MCPR operating limits will also be established using NRC-approved methods in accordance with T" 6.9.1.e(1) and (2) and will be published in the Core Operating Limits Report (COLR) for Cycle 9.
The COLR will be submitted to the NRC upon issuance in accordance with TS 6.9.1.e(4).
The accidents previously evaluated which are potentially impacted by this change are the limiting Anticipated Operational Occurrences (AOOs) specifically analyzed for each operating cycle.
These APGs are Rod Withorawal Error, Loss of 100 F 0
Feedwater lleeting, Generator Load Rejection Without Bypass, Feedwater Controller Fallure, Fuel Loading Error, and Rotated Bundle r:r r or.
These events are described in the United States s upplc.ne nt to GESTAR.
PECo proposes that the changes to the MCPR Safety Limits do not involve significant hazards considerations for the following reasons.
1)
The proposed changes do not involve a signiL2 cant increcse in the_ probability or consequences of an, accident previously evaluated.
Because the MCPR Safety l
Limits are operational thresholds analytically selected using proven methods, they cannot, therc selves, initiate l
an accident.
The probability of occur. 'r e of l
transients is determined by the frequency of operator er rors anc equipment failures, not by the adcquacy of tie MCPR Safety Limits selected.
Because the proposed
'sCPR Safety Limits have been selected such that no fuel damage is calculated to occur during the most severe moderate frequency transient events, they will ensure that the consequences of these events are not increased.
The response of the plant to transients will be within the bounds of the discuscion in Chapter 14 and Appendix G of the Updated Final Safety Analysis Repoct since the proposed MCPR Safety Limits will accomplish &hs same objectives as the previous limits.
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DocketLNon 277
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50-278 q
W' Y
License Nos..DPR-44 DPR-56 11)' The-proposed changes do not create the possibility of a new or-different kind of accident from any accident previously evaluated because the proposed MCPR Safety-Limits have been seTected such that the: design basis is satisfied.
The MCPR' Safety Limits are operational
-threshholds analytically selected using proven methods;
.therefore, they cannot, themselves, initiate an
- accident.
An improperly selected limit could result in fuel; damage, which is a'conse'aence of previously i
evaluated 1 accidents.
Thus, no=new or different. type of-J accident could be created by revising the limits..
111) The proposed = changes do not-involve a significant reduction-in.a margin of safety because the proposed MCPR Safety Limits have been selected such that the design' basi's isisatisfied and nuch that the-conservatisms described in the Bases for the' Fuel i
- Cladding Integrity Safety Limit TS are maintained.
.Thus,~ margins of-safety _with the proposed MCPRJSafety
-Limits are the'same.as with the pavious limits. -
a
-fAdministrative Changes:
.o The NRC has provided guidance concerning _the
?
of the standards forLdeterminingJwhether license amen. application e.
dments i
involve;no.significantahazards considerations by providing
_.examplesh(51LFederalLRegister 7751)..An example of=a change:that
'involvesino1significant hazards..considerationsEls::"a purely administrativeLchange to technicalespecifications: for' example,_a change _to/achieverconsistency throughout1the-technical-
. specifict';1ons, ~ correction Tof-an error,- or a chanta ~1n nomenclature"..
The : proposed.-administrative changes; clearly conform :tolthisLNRC example, and PECo; proI :ses that 4 these L,
.adminintrat'ivefchanges-do not involve ~ sigh'cicant-hazards-
~
considerations'for the.following reasone.
1)
- Th'e' proposed changes-do not-involve:a.significant
-increase-inothe-probability or consequences of an-
. accident previously-evaluated because:they do not: affect operation,; equipment, or anv safety-related activity.
Thus,ithese; administrative changes cannot7 affect =the
- probabilityuor consequences of any accident.
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Licenso Nos. DPR-44 DPR-56' i
11)
The__ proposed changes do not create the-possibility of a 3
new or different kind of accident from any accident--
previously evaluated because-these changes are purely j
administrative and-do not affect the plant.
Therefore, these. changes cannot create the possibility of any accident.
Tili) The' proposed-changes do not involve a significant reduction in a margin-of safety because the chat;ges do
-not-affect any safety related activity.or equipment.
Thesel changes are' purely administrative in. nature'and-increase the probability that the: Technical.
Specifications are correctly' interpreted by> adding-appropriate references and correcting errors.
- Thus, these changes cannot reduce any-margin of safetu l
. ENVIRONMENTAL IMPACT-An: environmental-assessment-is not required for'the
'changestreguestedoby thisl Application becauseEthe requested 7 changes conform to th=e-criteria for "actionsieligible for categorical exclusion"Las-specified inJ10LCFR'51.22(c),(9). :The 5
1-requestedichangesDhave been:shown by this ApplicationLnot to adverselyiaffect the systems andiequipment-that prevent the 4
uncontr'olled release-of" radioactive! material to the environment.
t The Application Involves.no'significant hazards considerations as demonstrated-Tin the preceding. sections. =The Application involves no;significantDchange in the types or:significant Increase-in-the-
-~ amounts ~.of.any effluents that may be: released-offsite, and there
?willibe no1significantLincrease in-. individual:or. cumulative
- occupational radiation exposure.-
[,:_
- CONCLUSION The-Plant' Operations Review Committee and'the Nuclear ReviewSBoard:have reviewed these proposed changes _to the
-Technical ~ Specificationsoand determined that they donot-involve D
- an Unreviewed Safety-Question and.willunot'. endanger the health b
andisafety of theEpubl-ic.-
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