ML20211M253

Sanitized Version of TS Page Changes Re SLMCPR

ML20211M253
Person / Time
Site:	Hope Creek
Issue date:	07/16/1997
From:	Public Service Enterprise Group
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ML20046D868	List: ML20211M230 ML20211M240 ML20211M245 ML20211M253 ML20211M257 ML20211M264
References
NUDOCS 9710140107
Download: ML20211M253 (10)

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ATTACRMENT 3 LR-N97433 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET No. 50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SIJ4CPR) CHANGES TECHNICAL SPECIFIC . ION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:

Technical Specification Page 2.1.2 2-1 Bases 2.0 B 2-1 3.4.1.1 3/4 4-1 9710140107 971003 PDR P

ADOCK 05000354 pg

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2.0 SAFETY t.!MIT$ ANO t.!MITING SAFETY SYSTEM SETTINGS 2.1 SAFETY t.!MITS TwERMAL DOWER, Low pressure or Low Flow l

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- 2.1.1 THERMAL power small not exceed 25% of RATED THERMAL power with the 4

reactor ressel steam done presure less tnan 785 psig or core flow less than j 10% of rated flow.

APPLICA8tLITY_: OPERATIONAL CONDITIONS 1 and 2.

, ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam does pressure less than 785 psig or core flow less than 10% of rated flow.

be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the reouirements of Spe,ification 6.7.1.

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\.\% \* \ 0 THERMAL power. Hiah Pressure and Hich Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less han -h4. th ith single l

two recirculation loop operation and shall not be less than I

recirculation loop operation, in both cases with the reactor vessel steam does pressure greater than 785 psig and core flow greater than 105 of rated flow.

APPLICA8tLITY: OPERATIONAL CON 0!TIONS 1 and 2.

ACTION: .go

]

With MCPR less than th two recirculation loop operation or less than

>-1 8 with single recirculation loop operation and in both cases with the reactor,

' vessel steam does pressure greater than 785 psig and core flow greater than 10%

\ .\ 'l of rated flow, be in at least HOT SHLITDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRE 55URE 2.1. 3 The reactor coolant system pressure, as esasured in the reactor vessel.

l steam does, shall not exceed 1325 psig.

APPLP Q : OPG ATIONAL CON 0!TIONS 1, 2, 3 and 4.

I M:

A With the reacter coetant system pressure, as seasured in the reactor vessel i

• steam does, ateve 1325 psig, be in at least HOT SHilTD0tm with reactor coolant systas pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and coesly with j

the requirements of Specification 6.7.1.

21 Amendment No. 15 i HOPE CREEK i

s I 2.1 SAFETY LIMITS

! BASES l

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary systes piping 3

are the principal barrists to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these j

barriers during normal plant operations and anticipated transients. The fuel l cladding integrity Safety Limit is set such that nofuel fueldamage dasageisisnot calculated

Because directly M

! to occur if the limit is not violated. roach is used to establish a Safety Limit such that l g.t o observable, a step-back the MCPR is not less than,;i app # for two recirculation loop operation and

! MCPR greater than 4 M ffor two re- yo j m for single rectreulat.1on circulation loop operation loopan opera ion.or single recirculation loo $ operation i

represents a conservative marg n relative to the conditions required to maintain

! fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom f m perforations or cracking. Although soon corrosion or use related cracking may occur during the life of the cladding, fission product migrationFuel from this source cladding is incre-perforations, i

mentally cumulative and continuously measurable.

however, can result from thermal stresses which occur from reactor operation l

j significantly above design conditions and the Limiting Safety Systen Settings.

While fission product afgration from cladding perforation is [ustngasperforations measurable the thermally caused cladd' as thatafrom signal use related threshold beyondcracking,ill which st greater thermal stresses may cause gross Therefore, the fuel cladding rather than incremental cladding deterioration. ,

Safety Limit is oefined with a margin to the conditions which would produceThe onset of transition boiling, MCPR of 1.0.

i j

ficant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER. Low Pressure or Low Flow l

The use of the applicable NRC-approved critical power correlation is not l

valid for all critical power calculations performed at reduced pressures belowTh

' 785 psig or core flows less than 105 of rated flow. This is done by estab-integrity Safety Limit is established by other means.

lishing a limiting condition on core THERMAL POWER with the following basis.

Since the pressure drop in the bypass region is essentially all elevation head, the caro pressure drop at low power and flows will always be greater than 4.5 psi. 1.nalyses show that with a bundle flow of 28 x los 1bs/hr, bundleThus, pressure drop is nearly independent of bundle power and has a value of 3.5 psi.

the bundle flow with a 4.5 psi driving head will be grea that the fuel assembly critical power at this flow is approximately 3.35 Wt.

With the design peaking factors, this corresponds to a THERMAL than 505 of RATED THERMAL POWER.

THERMAL POWER for reactor pressure below 785 psig is conservative.

Amendment No. 42 B 2-1 HOPE CREEK

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_ _ . _ ___ _ _ _ _ _ _ _ _ _ _ .__._ _ _ _ _ _ _ _ _ _ . _ . _ . . ~ . _ _ . _ _ _ _ _ _ _ . . - _ _ _ _ _ _

3/4.4 P u m R COOLANT SYSTEM i

3/4.4.1 nacIncutATIon SYSTEM i

REcipCUIATIM_LQQPS LIMITING CONDITION FOR OPERATION l

Two reactor coolant systes re:irculation loops shall be in operation 3.4.1.1 I

with

a. Total core flow greater than or equal to 45% of rated core flow, or

b. THERMAL POWER less than or equal to the limit specified in Figure 4

3.4.1.1-1.

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i APPLI MBILITYs OPERATIONAL CONDITIONS 1 and 2 .

l M3 --

in operationt

a. With one reactor coolant system recirculation loop not f

1. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> a)

Place the recirculation flow control system in the Local Manual code, and and Reduce THERMAL POWER to n 70% of RATED THERMAL POWER, b) safety Limit c) Increase the MINIMUM tRITICAL POWER RATIO (MCPR)

L, 0.^1 tr 1 #per specification 2.1.2, and d) Reduce the Nazir.us Average Planar Linear Heat Generation Rate (MAPLEGR) limit to a value of 0.34 times the two recirculation loop limit per specification 3.2.1, and I e) DELETED.

f) Limit the speed of the operating recirculation pump to less

, than or equal to 90% of rated pump speed, and g) Perfocia surveillance requirement 4.4.1.1.2 if THERMAL POWER is '

s 30% of RATED TRERNAL POWER or the recirculation loop flow in l the operating loop is s 50% of rated loop flow.

2. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the Average Power Range Monitor (APRM)

! scram Trip setpoints and Allowable values to those applicable for single recirculation loop operation per specifications 2.2.1 and 3.2.2; otherwise, with the Trip setpoints and

. Allowable values associated with one trip system not reduced to

thoes applicable for single recirculation loop operation, place the,affected trip system La the tripped condition and within

the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, reduce the Trip setpoints and Allowable 5

Values of the af fected channels to those applicable for single recirculation loop operation per specifications 2.2.1 and 3.2.2.

3. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reduce the APRM Control Rod Block Trip 4 .

1 See special Test Exception 3.10.4.

Amendment No. 63 l 3/4 4-1 HOPE CREEK

ATTACHMENT 4 LR-N97433 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE-HPF-57 DOCKET No. 50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SIJ4CPR) CHANGES AFFIDAVIT AND BASIS FOR WITHHOLDING INFORMATION CONTAINED IN LCR H97-05 FROM PUBLIC DISCLOSURE

Att~hm:nt Affidavit

1. Ralph J Reds, being duly sworn, depose and state as follows:

(1) I am Manager, Fuels and Facility Licensing, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the attachments to the letter numbered LR N97433, LCR H97 05, Public Service Electric & Gas Company to The United States Nuclear Regulatory Commission, Request for Change to Technical Specifications (Supplement),

Safety Limit hiinimum Cntical Power Ratio (SLSICPR), Hope Creek Generating Station, Factlity Operating License NPF 57, Docket No. 50354.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act

("FOIA"), 5 USC Sec. 552(bX4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(aX4) and 2.790(aX4) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential"(Exemption 4). W material for which exemption from disclosure is here sought is all " confidential commercial information," and some portions also qualify under the narrower definition of " trade secret,"

within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, critical Mau Enernv Prole-t v. Nuclear Regulatnry Comminion. 975F2d871 (DC Cir.1992),

and Public Citiren Health Rewarch Groun v. FDA. 704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which 'it into the definition of proprietary informat ion are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product;

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals asrcts of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

e. Infonnation which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

Page1 i

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, Att-hrrrnt (5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in con 0dence by GE, and is in fact so held. Its initial designation as proprietary informatien, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

l (8) The information identified in paragraph (2) is classified as proprietary because it would provide

! other parties, iactuding competitors, with information related to detailed results of analytical l

models, methods and processes, including computer codes, which GE has developed, requested NRC approval of, and applied to perform evaluations of the BWR. The development of the

, evaluation process along with the interpretation and application of the analytical results is l derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The fuel design and analpical methodology are part of GE's comt ehensive BWR safety and technology base, and their commercial value extends beyond the original development cost. 'Ihe value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and :noney by GE.

The precise value of the expertise to devise an evaluation process and apply the correct malytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

Page 2 l

. Attm"hment The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitise advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

State of North Carolina )

County of New Hanover )

Ralph J. Reda, being duly sworn, deposes and says:

That he has read the foregoing amdavit and the matters stated therein are true and correct to the best of his knowledge, information, and belicf.

l Executed at Wilmington, North Carolina, this 7 day of July,1997, l gotiip ues, M ,]

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f ,?, .* .

/ NOTARY $,j p' . Reda lM q i 3 ii, PUBUC

~ * ~ C eneAljp! Electric Company

$$ 0 '~

~,,,, h'{5-Subscribed and sworn before me this 7 day of July,1997.

r My commission expires on )J ,,./p_, f 7 Notary Public, State of North Carolina Page 3

ATTACKMENT 5 LR-N97433 HOPE CREEK GENERATING STATION FACIL:TY OPERATING LICENSE NPF-57 DOCKET No. 50-354 SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR) CHANGES ADDITIONAL INFORMATION CONCERNING THE HOPE CREEK CYCLE 7 SLMCPR ANALYSES Comparison of the pin by pin R-factor distribution in the Hope Creek Cycle 7 bundles versus that of the generic GE9B SLMCPR analysis:

The limiting case for the Cycle 7 SLMCPR (BOC) was reviewed. It was determined that bundles which contributed of the total pins undergoing boiling transition had (out of the 60 in the bundle) which were within -factor of the most limiting pin. This compares to the bundle in the ceneric GE9B analysis which has 1lllllggofthelimitingpininthebundle. By this measure it was concluded that the generic GE9B bundle had a flatter pin by pin R-factor distribution than the bundles which dominated transition boiling in the Hope Creek Cycle 7 SLMCPR analysis. The R-factor distribution controls the pin by pin MCPR distribution at any given setpoint. Thus, a bundle with a large number of pins near the limiting R-factor will have a large number of pins operating near the same MCPR (so if the lead pin is near the SLMCPR there will be a large number of other pins that are also close to the SLMCPR),

Sensitivity results demonstrating how the greater percentage of uncontrolled bundles for Hope Creek Cycle 7 can overcome the greater flatness of the generic GE9B pin by pin R-factor distribution and cause the SLMCPR to increase.

There are no specific studies for Hope Creek which quantify the relationship between SLMCPR and core flatness as described in terms of the number of uncontrolled bundles

@ or flatness as described by the number of pins close to the limiting R-factor. The Cycle 7 SLMCPR information transmittal quantified the flatness of the cle 7 core by determining the number of bundles that were THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION

- NOT FOR PUBLIC DISCLOSURE -

Page 1 of 2 i

Documsnt Control Desk LR-N97433 l Attachmsnt 5 LCR H97-05 l

} and contrasted those totals with the 4 generic GE9B core. It should be noted that the number of bundles

, has much more influence over the

' determination of the SLMCPR than the number of bundles M M Since the Cycle 7 core had more bundles closer to 4

the core MCPR this was taken as an ob]ective measure that the l Cycle 7 core was flatter. This is a trend which would cause the l Cycle 7 core to have a higher SLMCPR than the generic GE9B core j (all other things being equal). However, a ccmparison of the

{ Cycle 7 pin by pin R-factors wl-th the generic GE9B pin by pin R-j factors suggested that the generic bundle was flatter. This R-i factor comparison alone would indicate that the Cycle 7 S LMC PR I would be smaller than the generic GE9B core. Since the i calculated Cycle 7 SLMCPR was actually 0.01 more limiting than i the generic SLMCPR, it was concluded that the core MCPR

) distribution was the dominant factor and the reason primary l reason for the increase in the SLMCPR over the genetic result.

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) THE ATTACHMENTS TO THIS LETTER CONTAIN PROPRIETARY INFORMATION i

- NOT POR PUBLIC DISCLOSURE -

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