ML20236Q498
| ML20236Q498 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/09/1987 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20236Q500 | List: |
| References | |
| NLR-N87207, NUDOCS 8711200093 | |
| Download: ML20236Q498 (13) | |
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.4 Pubhc Service Electric and Gas Compar,y Corbin A. McNeill, Jr.
Public Service Elect:ic and Gas Company P.O. Box 236, H ancocks Bridge, NJ 08038 609 339-4800 Sanior Vice President -
Nuclear November 9, 1987 NLR-N87207 United States. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with the requirements of 10CFR50.90, Public Service Electric and Gas Company (PSE&G) hereby transmits a request for amendment of Facility Operating License NPF-57 for Hope Creek Generating Station (HCGS).
In accordance with the requirements of 10CFR170.21, a check in the amount of $150.00 is enclosed.
In accordance with the requirements of 10CFR50.91(b)(1), a copy of this request has been sent to the State of New Jersey as indicated below.
This amendment request revises Technical Specifications 3/4.9.2 and 3/4.10.7, as well as associated bases and definitions, regarding Source Range Monitors ( S RM s ) and Refueling Operations.
Specifically, the changes identified in Attachment 3 define the terminology Spiral Reload and Spiral Unload and eliminate the SRM operability requirement for channel count rate when sixteen or fewer fuel bundles are in the core.
Attachments 1 and 2 contain further discussion and justification for these proposed revisions.
These changes are necessary prior to commencing the first refueling outage; hence this amendment request should be issued by January 18, 1988 and effective on February 1, 1988.
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Document Control Desk 2
11-9-87 l
This submittal includes one (1) signed original, including affidavit, and thirty-seven (37) copies pursuant to 10CFR50.4(b)(2)(ii).
Should you have any questions on the subject transmittal, please do not hesitate to contact us.
Sincerely,
& s/Lh'a y C lh^d Enclosure (check)
Affidavit Attachments (3)
C Mr.
G. W. Rivenbark USNRC Licensing Project Manager Mr.
R. W. Borchardt USNRC Senior Resident Inspector Mr.
W. T.
Russell, Administrator USNRC Region I Mr.
D. M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628 1
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Reft.LCR 87-18.
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, STATE OFiNEW' JERSEY
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COUNTY OF SALEM
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Steven E. Miltenberger,>being duly sworn according to law deposes and says:
I am Vice President of.Public Service Electric and Gas
. Company,'and as such, I find the matters set forth in our letter concerning Facility Operating License dated November 9, 1987 --,
NPF-57 f'or Hope Creek Generating Station,. are true to the best of Lmy' knowledge, information and belief.
. x; Subscribed and Sworn to before me this 9fM day of ~//p8 den /44 1987 Yh Notary Public of New Jersey E!LEEN M. 0CHS NOTARY PUBLIC OF NEW JERSEY My Commission expires on My Commission Expires July 16,1932 i
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ATTACHMENT 1 I'
PROPOSED CHANGE IN THE TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO.
50-354 LCR 87-18 l
I I.
DESCRIPTION OF THE CHANGE Revise the Technical Specification Index, Section 1. 0
( Definitions), Specifications 3/4.9.2 ( Ref ueli ng Operations
- I ns t r ume n t a t i on) and 3/4.10.7 ( Special Test Exception -
1 Instrumentation - Initial Core Loadi ng), and Bases Sections 3/4.3.7.6 ( Source Range Monitors), 3/4.9.2 ( Ref ueling Operations - I ns t r ume nt a t i on) and 3/4.10.7 ( Speci al Instrumentation - Initial Core Loading) as shown in Attachment 3.
These changes will define the terminology Spiral Reload and Spiral Unload and eli mi na t e the Source Range Monitor ( SRM) operability requirement for channel count rate during refueling operations when sixteen or fewer fuel bundles are in the core.
II.
REASON FOR THE CHANGE Hope Creek Generating Station ( HCGS) is currently required to demonstrate that at least two of the four SRM channels are OPERABLE during OPERATIONAL CONDITION 5 ( Ref ueli ng) by in part ve ri f yi ng that the channel count rate is at least
- 0. 7 counts per second ( cps) provided the signal-to-noise ratio is greater than or equal to 2, otherwise the count rate must be at least 3 cps ( Tec hni c al Specification 4.9.2.c).
This verification must be performed prior to control rod withdrawl, prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS, and at least once per twenty-four hours.
As indicated in the Bases Section, this requirement ensures that redundant capability is available to detect changes in the r e a c t i vi t y condition of the core (B 3 / 4. 9. 2), which during low power and startup conditions is only available from the SRMs (B 3 / 4. 3. 7. 6)
Should conditions require that HCGS completely unload the core, or if a situation exists which requires core unloading to be temporarily halted when only a small number of bundles remain in the reactor pressure vessel ( RPV)
(i.e.
sixteen or f e we r), then SRM OPERABILITY must be demonstrated as described above prior to reloading the RPV or continuing the core unload.
While this requirement does not in and of itself present a problem, due to either the lack of bundles in the core or the low number of bundles and their relative positions in the RPV, a portable source may have to be inserted into the core region to generate enough neutrons Page i
E such'that a count rate of 3 cps ( or 0. 7 cps, as appropriate) can be achieved.
This action would produce unnecessary delays in CORE ALTERATIONS which would-negatively impact i
outage scheduling.
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Therefore, Public Service Electric and Gas Company- ( PSE&G) proposes.to revise the referenced Technical Specifications removing the requirement to demonstrate SRM OPERABILITY whenever sixteen or. fewer fuel bundles are in the core, thus avoiding the necessity of obtaining, inserting and r emovi ng a portable sample during refueling operations.
III. JUSTIFICATION FOR THE CHANGE The ability to begin ref ueling following a complete core unload, or to resume refueling after a temporary halt when i
I only a small number of-bundles are in the RPV (i.e.
sixteen or f ewe r), without' requiri ng t he minimum SRM channel count rate to be ' verified in demonstrating the OPERABILITY of the I
SRMs has already been addressed in Technical Specification Special Test Exception 3/4.10.7.
However,- this exception only applied during the initial core loading within the Startup Test Program in order to allow sufficient source-to-detector' coupling ~such that the mi nimum count rate can be achieved on an.SRM.
This exception was granted because of the significant margin to criticality while loading the initial sixteen bundles ( Tec hnical Specification Bases Section 3 / 4.10. 7).
The proposed change would delete this exception 'in f avor of an actual change to Technical Specification Section 3/4.9.2 which extends this same reasoning to any sixteen fuel bundles which meet:the criteria identified by General Electric Company ( GE) in the letter from A.D, Vaughn ( GE) to E. S.
Rosenfeld ( PSE&G) dated July 31, 1987 (see Attachment 2).
As discussed in the Paragraph IV below, this change is justified since inadvertent criticalit y with the four fuel bundles around an SRH is not possible and in fact a significant margin to criticality is maintained with sixteen or fewer fuel bundles based upon the fuel reactivity calculations performed by GE and the si gni fi c a nt hazards l
consideration evaluation completed by PSE&G.
J V.
SIGNIFICANT HAZARDS CONSIDERATION EVALUATION I
The proposed changes to the HCGS Technical Specifications-i (1)
Do not involve a significant increase in the probability or consequences of an accident previously i
analyzed.
The f oll owi ng discussion addresses ( i) the sequence of events associated with the proposed change as well as the GE criticality analysis due to the loading
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' sequence,.and ( ii) an evaluation of the proposed change L
against the accident analysis in the Final Safety Analysis Report ( FS AR)- Section.15.4.7 - Misplaced Bundle Accident.
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(i)
The analysis. performed in j justification of Technical l
Specification Special Test Exception 3/4.10,7 l
recognized only new, unexposed fuel bundles as the initial sixteen fuel bundles loaded into the RPV.
The proposed changes involve'using exposed fuel bundles in the two-by-two array surrounding an SRM and the possibility of loading two i
of these four bundles into the positions surrounding the SRM when the final' core cycle configuration requires the eventual removal of these bundles.
This latter issue involves the loading of two of the four bundles simply to L
provi de a sufficient count rate and then prior to completing l-the reload, re movi ng and replacing the two bundles with l'
bundles scheduled to occupy the subject locations during the planned cycle.
With the proposed changes to the Technical Specifications, i
i after a complete core unload, HCGS would begin the refueling l
l operation without inserting a portable neutron source and
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verifying an SRM channel count rate of 3 cps.
Instead, two exposed bundles would be loaded into the core around each SRM in the positions which they would occupy for the subsequent cycle.
If an SRM count rate of 3 cps is observed, then a spiral loading would proceed from the SRM instrument outward ( see the discussion of Spiral Unload and Spiral Reload' in Subparagraph ( 2) bel ow).
However, if sufficient counts are not observed, additional high exposure bundles may be inserted to complete the two-by-two array around each SRM in order to achieve 3 c ps.
These additional fuel bundles would not normally have been loaded into these locations and would solely be inserted in order to assist in satifying the SRM count rate surveillance requirement.
At this point if a minimum of 3 cps was not observed, refueling operations would be halted until the SRM instrumentation is checked.
If the required count rate is observed, then the Spiral Reload would proceed from the SRM instrument outward eventually encountering the fuel bundles loaded around the other three SRMs.
The core configuration at this time would be different from the scheduled configuration for the next cycle in two manners.
First, the core would only be partially loaded (i.e.
up to 16 bundles) and second, the second pair of bundles loaded around each SRM to obtain the minimum count rate may be different from the bundles scheduled to occupy those locations.
As long as the cold re ac ti vi ti e s (zero voids) of the high exposure fuel bundles temporarily loaded around the SRMs are individually less than the cold re ac ti vi ti e s of the respective bundles scheduled for the subject locations, the cold shutdown margin calculation performed for the scheduled core loading bounds the L
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-partially reloaded core (see Attachment 2).
Hence, this L
feriteria is required when temporarily loading the latter two bundles.around an SRM in order to satisfy the SRM channel L
. count rate operability requirements.
This requirement is L
currently being satisfied and will ' continue to be met through the use of station administrative procedures.
L As. discussed in Attachment 2, GE has performed fuel reactivity; calculation to determine k-effective when four
.GE bundles, restricted'to the cold reactivity criteria identified above and an uncontrolled lattice k-infinity of j
less t ha n 1. 31, are arranged in a two-by-two array u
surrounding an. SRM wi t h a. mi nimum of 12-inches between'them and,any surrounding bundles.
The analysis, based on the'GE lattice physics 'models previously reviewed and approved by the NRC in GESTAR, indicate that for the conditions specified, k-effective will be less than 0.95 which bounds the highest enriched lattice designs allowed under GESTAR.
As a result, the need for SRM count rates when sixteen or fewer' bundles are in the RPV is unnecessary since an inadvertent criticality is not possible.
The margin to criticality of 5% is sufficient to assure that the proposed changes are at least as conservative as-the basis for Technical Specification 3/4.10.7.
(ii) PS AR Section 15. 4. 7 discusses the accident analysis associated with a misplaced bundle.
Since the proposed j
change would. permit the deliberate misloading of fuel bundles in order to create sufficient neutron flux to satify the SRM channel count rate operability requirement, a potential exists in.which a bundle may be inadvertently left out of position for the subsequent cycle.
This possibility is extremely remote in that two bundles would have to be misplaced.
In other words, the first bundle is that bundle deliberately misloaded for the SRM operability demonstration, whereas the second bundle is that bundle which is scheduled to occupy the position occupied by the first misloaded bundle but the original misloading is overlooked and thus a second misloading occurs.
Additionally, following any core reload, a core verification
- is performed to assure that the bundles have been located properly.
This verification step would have to fail.in picki ng-up the two above-identified loading errors.
These three errors, as well as the accident sequence discussed in FSAR Table 15.4-5, form the basis for the accident analysis.
Further information regarding the accident analyeis is contained in the response to Question 491.2.
PSE&G has re-evaluated the accident scenario in light of the propsced changes and has concluded that the results presented in FSAR Table 15.4-7 are still applicable, bounding the proposed changes.
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Therefore, the proposed changes do not increase the l
probability or consequences.of an accident previously.
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Do not create the possibility of a new or different ki nd ' of accident from any accident previously
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evaluated.
.The proposed changes do not change the physical
~ design or operation of the Source Range Monitors nor require any-hardware' modifications or core redesign.
Rather, the proposed changes only require procedural revisions which indicate when the SRHs are to be declared operational and how the refueling. sequence will be handled.
FSAR Section 7.7.1.1.2.2.a.4 discusses rod block signals generated when the reactor mode switch is in the "Startup" or " Ref uel" position and Section 7.7.1.1.2.2.b discusses rod block. bypasses including those permitted for the SRM channels.
The information discussed in these sections remains accurate in light of the proposed changes and hence the functioning of the rod block circuitry remains as-is.
FSAR Section 14,2.12.3.6 discussed the startup tests completed in conj unction with the SRMs.
However, this testing was completed after fuel loading and not prior-to or during partial core loading; hence, the proposed changes do not require any additional testing nor invalidate any
' testing completed during the Startup Test Program.
In fact, the use of Technical Specification 3/4.10.7 precluded the need to demonstrate SRM operability prior'to installing the
.first sixteen bundles, the same condition sought with this proposal.
Two new definitions are identified in the proposed change Spiral Reload and Spiral Unload.
Spiral Reload is a core loading methodology employed to refuel the core after a complete core off-load.
During a Spiral Reload, the fuel is loaded into i ndi vi d ual control cells (i.e.
four bundles surrounding a control-blade) in a spiral fashion centered on an SRM and moving outward.
Before initiating a Spiral Reload, up to four bundles may be loaded in the four bundle locations immediately surrounding each of the four SRMs to L
obtain the necessary SRM channel count rates.
S i mi l a rl y, a
S pi ral Unload is a core unloading methodology wherein the fuel from the outermost control cells is removed first and is employed when completely unloading the core.
Unloading continues in a spiral fashion by removing fuel from the outermost periphery to the interior of the core, symmetric about.the SRMs, except for the four bundles around each of L
the four SRMs.
Again, four fuel bundles may be left around each of the four SRMs to maintain the necessary SRM channel l
count rate.
However, when sixteen or fewer fuel bundles are
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in the core, the requirement to maintain the SRMs OPERABLE L
based upon the channel count rate is unnecessary.
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This type of; loading sequence simply descri bes the pattern to load or unload fuel and in no way changes the i.
- requirements for the incore nuclear instrumentation, i. e.
L ESRMs, Intermediate Range. Monitors ( I RMs) -and Local Power Range Monitors ( LPRMs).
The fact.the SRMs need not be l-
. operable with sixteen or fewer f uel bundles in the core, simply reflects the GE criticality analysis which has no bearing on the loading sequence other than that discussed above.
Therefore, the Technical Specifications are being revised to reflect the loading methodology simply to provide a better definition of the loading sequence thereby maintaining consistency with the proposed change.
Therefore, the proposed changes do not create the possibility of a new or different ki nd of accident from any accident previously evaluated.
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-( 3)
Do not involve a significant reduction in a margin of safety.
The proposed changes represent a condition associated with the SRMs which has been previously reviewed and approved by the NRC during the preparation of the Technical Specification Special Test Exceptions.
Although the current exception identified in Technical Specification Section 3/4.10.7 is applicable to the i ni ti al core only, for that situation,.the margin of safety for the SRMs is not reduced when the Limiting Conditions for Operation ( LCO) are
' satisfied.
One of these conditions is that no more than 16 bundles can be in the core at any one time.
This specification and the GE analysis discussed above provide j
the basis for utilizing this same LCO for any subsequent core cycles provided the limitations stated in Subparagraph (1) above are s a t i s i fi e d.-
Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
V.
CONCLUSION Finally, PSE&G concludes that this amendment request conforms to Example ( i v) for Amendments That Are Not Li kel y To Involve Significant Hazards Considerations ( published in Federal Register Volume 51, Number 44, dated March 6, 1986) which in part states that relief will be granted
...upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation was not yet demonstrated
( ass umi ng)... t he operating restriction and the criteria to be applied to a request for relief have been established in a prior review."
PSE&G proposes that the Special Test Exception cited was provided only for the initial core loading since an analysis Page 6
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had not yet been performed which j ustified that for 16'or l,
fewer exposed fuel bundles,'SRM channel count rate operability was not required.
With the completion of the GE analysis cited, the fact the Special Test Exception existed for the Startup Test Program, and the discussions provided
' in the above Paragraphs, PSE&G concludes that the proposed changes and hence this Amendment request do not involve a Significant Hazards Consideration.
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ATTACHMENT 2 Letter ( ADV87229). from A D, Vaughn ( Gene ral Electric Company) to Elliott S.
Rosenfeld ( Public Service Electric and Gas Company) dated July 31, 1987
Subject:
A c hi e vi ng Minimum SRM Count Rate With Exposed Fuel.
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GENER AL$ ELECTRIC u
-.useos -
GENEAAL ELECTec COMPANY
- 175 CURTNER AVENUE o SAN JOSE, CAUFCNNIA 95195 July l31', 1987-cc: w attachment
'i ADV87229' R. J. Gennone J. M. Haun R. A. Schmidt Mr. Elliott S. Rosenfeld; Manager Nuclear Fuel.
Public Service Electric and Gas Co.
Hope Creek Generating Station i
P.O.~ Box 236 f
Hancocks Bridge, NJ- 08038
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SUBJECT:
Achieving Minimum SRM Count Rate with Exposed Fuel l
This responds to your request.for information concerning the loading of fuel j
bundles around source range monitors (SRMs) during refueling at Hope Creek in order.to achieve the minimum count rate of three counts per second (CPS).
General Electric has performed fuel reactivity (k-effective) calculations ~
which bound the highest' enriched lattice designs allowed under the license of.
1 the General. Electric Manufacturing Plant and these calculations demonstrate
.that any four GE fuel' bundles with uncontrolled lattice k-infinity <l.31, arranged in a square (2X2) array surrounding an SRM, and with a minimum of 12 inches.between them and any surrounding bundles, will have a~ k-effective less lthan 0.95.
This analysis was based on the GE lattice physics models
-previously reviewed and-approved by the NRC. Thus any loading combination of 1
four GE Fuel bundles around each SRM in the Hope creek plant, before obtaining the minimum cps count rate, should not constitute a safety concern.
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It is our understanding that, after a complete core unload, PSE&G would plan to begin refueling by loading two exposed fuel bundles around each SRM and then attempting'to observe at least 3 cps. These bundles'would be those scheduled to occupy those locations in the next cycle.
If an acceptable count rate.is then observed. spiral loading of the core would proceed from the SRM instrument outward.
If insufficient counts were observed, additional high exposure fuel bundles would be inserted to complete the 2X2 array around each SRM to attempt to reach a minimum of 3 cps. These additional fuel bundles would normally not have been loaded in these locations and would be so inserted solely to assist in the SRM instrumentation check.
If at this point the minimum of 3 cps were not obtained, refueling would be halted until the SRM instrumentation was checked.
If the required count rate is observed, then spiral loading would proceed from the SRM instrument outward eventually encountering the fuel bundles loaded around the other SRMs. The
. core configuration present at this time would be different from the scheduled full core loading in two respects.
First, the core would only be partially
GENERAL $ ELECTRIC Core Loading for Minimum July 31, 1987 SRM Countrate
. loaded, and, second, the second pair of bundles loaded around each SRM may be different from the bundles scheduled to occupy those locations. As long as the cold reactivities (zero voids) of the high exposure fuel bundles temporar-ily loaded around the SRMs are individually less than the cold reactivities of the respective bundles scheduled for these locations, the cold shutdown margin calculation performed for the scheduled core loading bounds the partially reloaded core. The selection of the four bundles to be initially loaded around each SRM is, thus, not arbitrary, but the aforementioned criterion is not expected to hamper refueling operations.
If you have further questions on this subject, please do not hesitate to contact us.
Very truly yours, dbM k
A. D. Vau Fuel Project Manager Hope Creek Plant 408/925-1618 Approved:
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' W. A.' Golub, Manager Core Nuclear Design Approved:
MQ p C. Stirn, Manager Systems Integration Engineering l
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