ML18143A416

The Commission Has Issued the Enclosed Amendment No. 10 to Provisional Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant

ML18143A416
Person / Time
Site:	Ginna
Issue date:	03/30/1976
From:	Purple R Office of Nuclear Reactor Regulation
To:	White L Rochester Gas & Electric Corp
References
Download: ML18143A416 (29)

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Docket No. 50-244

'DISTRIBUTION'arch 30, 1976 Local PDR ORB-1 Reading KRGoller/TJCarter SMSheppard TVtltambach

Attorney, OELD Rochester Gas and Eloctric Corporation OI6E (3)

ATTIC:

Mr. Leon D. Nhite, Jr.

BJones (4)'

Vice President BScharf (15)

Electric and Steam Production JMcGough 89 Hast Avenue JSaltzman Rochester, New York 14604 ACRS (16)

CMiles, OPA Gentlemen:

TBAberaathy, DTIE JRBuchanan, NSIC The Commission has issued the enclosed Amendment No. 10 to Provisional Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant.

The amendment consists of changes to the Technical Specifications and is in response to your requests of March 27 and September 22, 1975.

The amendment changes the Technical Specifications to revise the core thermal limits curvo based on the application of the approved, updated Nestinghouse model for fuel densification and clad flattening, a revision in the overpower dT and overtemporature hT set points, a deletion of the allowance for relaxing end-of-lifo control rod insertion limits, and revised shutdown margin requirements based on a reanalysis of the main steam line break accident.

Copies of the Safety Evaluation and the Federal Register Notice are also enclosed.

Sinceroly,-

Original signeg gy Robert A. Purple, Chief Operating Reactors Branch 81 Division of Operating Reactors

Enclosures:

1.

Amendmont No. 10 2.

Safoty Evaluation 3.

Federal Register Notice cc w/cncls:

See next page x27433 SURNAMS~

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UNITED STATES UCI+lR REGULATORY COMMISSION V(ASHINGTOt4, D. g. 20655

'ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO; 50-244 R. E.

GINNA NUCLEAR POWER PLANT

'AMENDMENT'TO PROVISIONAL OPERATING'LICENSE Amendment No. 10 License No.

DPR-18 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The applications for amendment by Rochester Gas and Electric Corporation (the licensee) dated March 27 and September 22,

1975, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized.

by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's regula-tions; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of

. the public; and E.

An environmental statement or negative declaration need not be prepared in connection with the issuance of this amendment.

2.

Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment.

3.

This license amendment is effective as of the. date of issuance.

FOR THE NUCLEAR REGULATORY COMMlSSION

Attachment:

Changes to the Technical Specifications Karl R. Goller, Assistant Director for Operating Reactors Division of Operating Reactors Date of Issuance:

March 30, 1976

ATTACHhKNT TO LICENSE AhKNDhKNT NO. 10 PROVISIONAL OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Revise Appendix A as follows:

Remove Pa es Insert Revised Pa es 2.1-3 2.3-1 2 ~ 3 2 2 ~ 3-3

2. 3-6 Figure 2.1-1 3.1-1

3. 10-1

3. 10-2 3.10-6

3. 10-7 3.10-9

3. 10-10 Figure 3.10-1 Figure 3.10-2

2. 1-3 2.3-1 2 ~ 3 2

2 ~ 3 3 2.3-6 Figure 2.1-1

3. 1-1 3.10-1 3.10-2 3.10-6 3.10-7

3. 10-9

3. 10-10 Figure 3.10-1 Figure 3.10-'2

tl The curves are based FN

~

on the followingnuc)ear hot channel factors:

2-72, FN aH 1;66 Hot.channel factors are defined as:

Nuclear heat flux eeakin~ factor. This is the ratio of the maximum Lineal power in the core to the

'verag lineal power in the core.

Vhere:

(FN FN x FN

)

R z

'N z

Nucl e ar axial veakina f ac tor.

This is the ratio oi the maximum lineal power in the maximum power rod to he average lineal power in the maximum power rod.

Nuclear enthaln rise creaking factor This is equaL to the ra io of the maximum channel power to the average in the core.

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2. 1-3 Ainendment No.

10 VZR 3 01976

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-2:3 Limiting Safet S ster

Settings, Protective inst"umentation I*bi '

Applies to trip settings for instruments monitoring reactor power, reactor coolant pressure, temper-

ature, and Qow; and pressurizer level.

~ot'

'o provide for automatic protective action in the event that the principal process variables approach a safety limit.

2. 3. 1 Protective instrumentation for reactor trip settings shall be as follows:

2. 3. 1. 1 Startuo Protection High fluv, power range (low set point) -

~25% of rated power.

2. 3, l. 2 Core Protection a.

High flue, power range (high set point) -

<109% of rated powei.

b.

High pressurizer pressure

~ 2385 psig.

c.

Low pressurizer. pressure

> 1865 ps>g.

2. 3-1 Amendment No.

10 MAR 3 01976

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Overtemperature hT 1 + v S 2

c T

[K> +,

(P P->) - K>(T-T~) (~) -T(ET))

1+ ~2S where:

I, o T

~ indicated LPi it. rated power, F

T

~ average'emperature, F

TI 573 55F

~ pressurizer

pressure, psig P

~ 2235 psig KI

~ 1.12

'K2 0.0007356 Kg

~ 0 01577 25 sec T2 5 sec and f (~I) is a function of the indicated diEference between top and bottom detectors oz the power-range nuc'ear ion cha-...be with gains to be selected based on =easured instrument respo-..s during plant startup tests where q

and q

are the perce..t power in the op and bottom halves of the core respective ",

and qt + q

's the total core power in percent oz ".area po"er, such that:

(i) foz q - q

~ithin -18, +10 percent, f (Il:)

0 (ii) for each percent hat the magn'".ude of q qb exceeds t

+]0 percent, the dT trip set poiwt snail be automat'ca reduced by an equivalent oE 2

7 percent of ratec power (iii) for each percent that the magnitude oz qt qb exceeds 18 percent, the IT set point shall be automatica'ly reduced by an equivalent oz 2, percent of. rated

power,

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10

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Overpover hT

(*-* ).-

- E( )I h o

. tg S+1 where:

-LENT

~ indicated% at rated pover, 'F T

~ average zemperature, F

T

~ indicated Lavg at nominal conditions at rated pover, 'F

= 1.083 Kg

~ 0.0 for T<T~

~ 0. 001 for T>Ti K6

~ 0.0262 for increasing T

< 0.0 for decreasing T

~ 10 sec

'I f(hl) ~ as defined in 2.3.1.2.d."

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2. 3-3 Amendment No.

10 MAR 3 0 1976

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s transient is slow with respect to piping transi delays froze the core

~ 0 to the temperature detectors (about 4 seconds),

and (Z) 'pressure ee

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is within the range bet'ween the high and low pressure reactor trips.

VFith norznal axial power distribution, the reactor trip limit, with allowance for errors,

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is always below the core safety limitas shown on Figure 2. 1-'l. Ifaxial peaks are greater than design, as

'ndicated by difference beSveen top and bottom power range nuclear detectors, the reactor trip limitis automatically reduced.

I The overpower 4T reactor trip prevents power density anywhere in the-core from exceeding a va1ue at which fuel pellet centerline melting would in NCAP-.8058, E

Unit 1, Cycle occur as described in Section 7.2.3 of the FSAR and t

"Fuel Densification, R. E. Ginna Nuclear Power Plant 3" and'ncludes corrections for axial power distribution, change in density and heat capacity of water and temperature, and dynamic compensation for piping delays from the core to the loop temperature. detectors.

The specified set points meets the requirement

~ I and include allowance for instrument error The overpower and overtemperature protection setpoinM include consideration of the effects of fuel densification on core safety limits.

The low flow reacto" trip pro ects the core against Dht'2 in the event

~

of a sudden loss of power to one or both reactor coolant pumps.

The set point specified is consistent with the value used in the accident analysis.~

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The underfrequency reactor trip protects against a de-crease in flow caused by loiv electrical frequency.

The specified set point assures, a reactor trip signal before the low flow trip point is reached..

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2. 3-6 Amendment No-10 MAR 4 -Q 3976..

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Core DNB Safety Limits Two Loop Operation, 100+go Flow Amendment No.

10 MAR 3.0 1976

I G Ct'~:L>ITIO S r OB, OPERATION 3.1 c

Coolant S) s tcm 0 ~

A lic"bilit::

Applies to the operatin~ status of the Reactor Coolant System.

'oi'o specify those conditions of thc Reactor Cool" nt System rvhich must, be mct to assure safe reactor operation.

S ccificat ion:

3. 1. 1

~O>er':l!nba! C::s:.nnrr.'..".

~ 3. 1. 1. 1 Coolant Pu:.:n s a.

AL least onc rcac'or coolant puiv.p or the residual hc: t removal 'ystcm shall, bc in operation chen a rcduct,.'on is mac.e i > the boron.conccnt.ration of the reactor cool-t t..

Vfhcn the react,or is crit.icaL and abov'c l~~~ tncrival po::c at least onc rc:tctor coolant pump shall be in operation.

C ~

(i)

Reactor power shall not be maintained above 130 h5A'8.5~~) unless both reactor coolant pumps are in operation.

If either reactor coolant pump ceases operating, immediate power reduction shall be initiated under administrative control. If the shutdown margin meets the one loop requirements of Figure 3.10-2, then the power shall be reduced to les's than 130

!~fiA'. If the one loop shutdown margin of Figure 3.10-2 is not met, the plant'shall be taken to tht: hot shutdown condition and the one loop shutdown margin shall be met.

3.1-1

-r Amendment No.

10 MAR 3 01976

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3. 10 Control Rod =nd Power Distribution Limits Applies to the operation of the control rods and power distribution limits.

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'o ensure (1) core subcriticality after a reactor trip,'2) limited potential reactiv'ity insertions from a nypo-thetical control rod ejection, and (3) an acceptable core power distribution during power operation.

5 ecification

3. 10. 1 Control Rod Ins er tion Limits

3. 10.1.1 When the reactor is subcritical prior to startup, the hot shutdown margin shall be at least that shown in Figure

3. 10-2.

The shutdown margin as used here is defined as the amount by which the reactor core would be subcrit'.cal at hot shn".down conditions (e47 oF) if aii cortroi ".ods were tripped,. assuming that the highest worth co..trol rod re-mained fullywithdrawn, and assuming no changes in +enon, boron, or part-length rod position.

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3. 10-1 e

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10 MAR 3 01976 s

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3. 10. 1. 2

'xe reactor is critical e t

physics tests

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and control rod exercise's, the shutdown control rods shall be fullywithdrawn.

3 10. 1; 3 When the "reactor is critical, except for physics tests and control rod exercises, each group of control rods shall be no furth'er inserted than the Limits shown by the lines on Figure 3.10-L.

Furthermore, the control rod banks are moved sequentiaLly with a.100 t'+S)step overlap between successive banks.

3.10.1.4 During physics tests and control rod exercises indicated in Table 4. 1-2, the insertion limi s need not be observed but the Figure 3. 10-2 must be observed.

3.10.1.5 The part Length control rods villnot oe inse ted except for physics tests or :or axi 1 offset calibration per-for'r;ed at 75/

QG ver or less."

3. 10. 2 Power Distribution L'... -;ts and Misa'.ironed Control Bod

3. 10. 2. 1 The movable detector system shall be used to measure power distribution after each fuel reloading prior to operation of the plant at 50~la of rated power to ensure that design limits are not exceeded.

If the core is operating above 75~o power with one excore nuclear channel out of service, then the quadrant to

3. 10-2 Amendment No; 10 MAR 3 01976

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inoperable rod has a potential reactivity in'se'rt'.on upon ejection greater than 0. 365% Ak/k.

The control bank insertion limits shown in Figure 3. 10-1 shall be used until the potential reactivity insertion of the inoperable rod has been confirmed to be less than

0. 365% E k/k at greater control 'bank insertion.

Basis:

The reactivity "control concept is that reactivity changes accompanying changes in reactor power are compensated by control rod motion.

Re-activity changes associated with xenon, samarium, fuel depletion, apd large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated by changes in the soluble boron con-centration.

During power operation, the shutdown groups are fu11y withdrawn and control of reactor power is by the control groups.

A reactor trip occurring during power operation willput the reactor in o the hot shutdown cond'tion.

The control rod insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the hignest worth control rod remains fullywithdrawn with sufficient margins to meet the assumptions used in the accident analysis.

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In addition, they provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical

3. 10-6 I

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V Amendment No.

10 MAR.3 0 1976

~ 0 rod ejection,. and pro'vide for acceptable nuclear peaking factors.

The lines shown on Figure 3.10-1 meet the shutdown requirement.

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The maximum shutdown margin requirement'ccurs at end-of-cycle life and is based on the value used in analysis of the hypotheti-al steam break accident.

Early in cycle life less shutdown margin is

required, and Figure 3.10-2 shows the shutdown margin equivalent to that which is required at the end-of-life with respect to an h

uncontrolled cooldown.

All other accident analyses are based on 1'eactivity shutdown margin.

Part length rod insertion has been eliminated for'this c'ycle to eliminate potential adverse power shapes and to preclude rapid local power changes caused by part. length rod travel through the core.

The various control rod banks (shutdown banks, control banks A, B, C. D, and part-length rods) are each to be moved as a bank; that is, with all rods in the bank within one step (5/8 inch) of the bank position.

Position indication is provided by two methods:

a digital count of actuation pulses which shows the demand position of the banks and a linear posit-'on indicator (LVDT) which indicates the actual rod position.

The 15 inch permissible m'salignment provides an (2) 3.10-7 Amendment No 10 MAR 3 0 1976

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,=-ace abc.-.".~~

a."A a ply:" ~'.".u".d" a is p- " -= "

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Thc specified rod drop time is consistent wi.h safety analyses that have bccn performed.

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An inopcrab)c rod imposes additional dcma>>ds on thc operator.

Thc permissible nun>ber of i>>opcrablc control rods is limitcc. to one except duri>>" physics tcsHng, in order to limit t)ic magniludc of thc operating burden, but such a failure v:6uld not-prc; cnt dropping of thc operable rod" upon x cac (or trip.-

Thc reactivity worth )imit for an i>>operable co>>trol rod is co>>sis-tent xvith thc va)uc found tolerable in the analysis of the hypothec t'ical rod ejection accident.

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Thc initi;1-core physics tcstin~ shoivcd thc maximum worth to bc 1css than 0. 365~o when the controlling Group D divas morc than 60~o withdrawn, whereas larger worths werc possible with thc controlling bank fully i>>scrted.

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3. 10-9 Amendment No. 10-

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0 MAR.3 O>976

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~ 0 Referer.ces:

{1) Technical Suppiemcn Accomoany;ng, App".cation to Increase Popover - Section 14

'2)

FSA"C, Section 7. 3 (3) FSAR, Section

14. 2. 6

{4) Techn cal Supplei-..ent - Append.'.:: A, P".. 120 ci f ~

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FIGURE 3. 10-1 CONTROL ROO INSERTION LIt(ITS VERSUS CORE POMER FOR OOL TIIROUGH EOL

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UNITED'STATES NUCLEAR'REGULATORY COhMISSION

'OCKET 'NO.

SO-'244 ROCHESTER 'AS AND

" ELECTRIC "CORPORATION

'OTICE'OF ISSUANCE'OF AhfENDhKNT TO PROVISIONAL

OPERATING LICENSE Notice is hereby given that the U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No.

10, to Provisional Operating License No.

DPR-18 issued to Rochester Gas and Electric Corporation (the licensee) which revised the Technical Specifications for operation of the R. E. Ginna Nuclear Power Plant located in l(ayne County, New York.

The amendment is effective as of its date of issuance.

The amendment changes the Technical Specifications to revise the core thermal limits curve based. on the application of the approved, updated Nestinghouse model for fuel densification and clad flattening, a revision in the overpower hT and overtemperature hT set points, a

deletion of the allowance for relaxing end-of-life control rod insertion limits, and revised shutdown margin requirements based on a reanalysis of the main steam line break accident.

The applications for the amendment comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations.

The Commission,has made

'ppropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendme'nt.

Notice of Proposed Issuance of Amendment to Provisional

Operating License in connection with this action was published in.the

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FEDERAL REGISTER on April 10, 1975 (40 F.R. 16249).

No request for a hearing or petition for leave to intervene was filed following notice of the proposed action.

The Commission has determined that the issuance of this amendment will not result in any significant environmental impact and that pursuant to 10 CFR 551.5(d)(4) an environmental statement, negative declaration, or environmental impact appraisal need not be prepared in connection with issuance of this amendment.

For further details with respect to this action, se'e (1) the applications for amendment dated March 27, and September 22, 1975, (2) Amendment No.

10 to License No. DPR-18, and (3) the Commission's related Safety Evaluation.

All of these. items are available for public inspection at the Commission's Public Document

Room, 1717 H Street, N. W., Washington, D. C.

20555 and at the Lyons Public Library, 67 Canal Street,

Lyons, New York 14489 and at the Rochester Public Library, 115 South Avenue, Rochester, New York 14627.

A copy of items (2) and (3) may be obtained upon request addressed to the U., S.'uclear Regulatory Commission, Washington, D. C.

20555, Attention:

Director, Division of Operating Reactors.

Dated. at Bethesda, Naxyland, tQis 30th day of March 1976.

FOR THE NUCLEAR REGULATORY COMMISSION Robert A, Purple, ief Operating Reactors Branch ;il Division of Operating Reactors

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UHITED STATES V%PAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE'OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT'NO. 10'TO PROVISIONAL OPERATING'LICENSE'NO. DPR-18 ROCHESTER GAS AND ELECTRIC'CORPORATION R.'E.

GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Intr'oduction By application dated March 27, 1975, Rochester Gas and Electric Corporation (RG5E) proposed changes to the Technical Specifications for the R. E.

Ginna Nuclear Power Plant.

RG5E had reanalyzed the Ginna core using the NRC approved, updated Westinghouse model for fuel densification and clad flatteni'ng OICAP 8218 and 4'CAP 8377 proprietary versions and WCAP 8219 and WCAP 8381 non-proprietary versions).

As a result of this reanalysis and the replacement of fuel that had low internal prepressurization with fuel that has higher internal prepressurization, an increase in the operating 'pressure of the primary coolantback to the original value of 2250 psia was proposed and a slight change in the core thermal safety limits was proposed.

Associated with these

changes, RG5E proposed the corresponding changes to the overpower hT and ovextemperature dT set-points.

In addition, the application proposed increasing the low pressurizer pressure trip set point, correcting the temperature of the core indicated in Specification 3.10.1.1 for the hot zero power condition, and deleting the allowance for deeper control rod insertion limits for end of core life operation.

By application dated September 22, 1975 RG5E proposed changes to the Technical Specifications that would'ncrease the shutdown margin/

requirements based on a reanalysis of the main steam line break accident, RG5E has already imposed the maximum shutdown margin requixements resulting from this reanalysis pending NRC appxoval of these proposed

changes, Evaluation The proposed core safety limit curves and the associated overpower and overt'emperature bT setpoints are based on calculations using the methods described in the NRC-approved, updated t<estinghouse models for fuel densification and clad flattening.

These limits and setpoints relate avexage core coolant tempexature and coolant pressure to the allowable

reactor power level based on the margin to departure from nucle'ate boiling (DNB).

As a result of eliminating clad flattening provisions, since no clad flattening is predicted for the'ore's fuel and as a result of the application of the approved fuel densification model, these new limits and setpoints (including operation at 2250 psia) result in no reduction of margin to DNB.

We, therefore, find them acceptable.

The proposed increase in the minimum low pressurizer pressure trip set point from 1715 psig to 1865 psig will produce a 'reactor trip before safety injection occurs.

The safety injection set point will remain at 1715 psig.

This increase in the low pressurizer pressure reactor trip set-point increases the margin of safety over what was previously'ound acceptable and is therefore a'cceptable.

The elimination of the allowance for deeper control rod insertions near the end of coie life also increases the margin of safety over what was previously found acceptable and is therefore acceptable.

As reported in a letter from L. D. White, Jr. to James P. O'Reilly, Directorate of Regulatory Operations dated November 7,.1974, it was determined that the Safety Injection System delivery curve previously employed for steam line break analysis was inconsistent with the

updated, slower delivery curve used for Loss df Coolant Accident analyses and that, until further analysis

>>as performed, a shutdown margin at 2.45~op should be maintained.

RG5E imposed such a restriction.

A reevaluation was completed assuming an end-of-cycle shutdown margin of 2.45'~p and the revised delivery curve.

It was confirmed that sub-criticality is maintained for the credible break and a DNBR of >1.3 is maintained for the hypothetical break cases.

A revised steam line break analysis was submitted by letter dated September 22, 1975 which included proposed Technical Specification changes increasing the shutdown margin requirements and specifying different shutdown margin requirements dependent upon whether both primary coolant loops were operating or only one.

In'he event of a reactor coolant pump failure while operating below 50~~'f full power, immediate power reduction wi.ll be initiated.

(Above 50~~ power, the reactor trips automatically).

If the one-loop shutdown margin is met, po>>er will be reduced to below 130 i~h~t (8.5'f full power). If the one-loop margin is not met, the plant will be taken to hot shutdown and the one-loop margin established before returning to a power below 130 hhvt.

The revised analyses included a spectrum of steam line break accidents both inside and outside containment, during various modes of operation and with or without offsite power.

The accident which resulted in the most

severe consequences was determined and evaluated. using a suitable mathematical

model, and the parameters used as input to this model were suitably conservative.

The results of the analysis of the spectrum of steam line break accidents showed that no loss of core cooling'apability resulted.

The minimum departure from nucleate boiling ratio (DiVBR) experienced by any fuel rod was shown to be greater'than 1.3 in alI cases.

Based upon our review of this revised main steam" line break analysis, we conclude that, with the incoxporation of the proposed increased shutdown margin requirements, there is no increase in the consequences of this accident and there is no decrease in the margin of safety.

We, therefore, find the proposed changes acceptable.

We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendment involves an action which's insignificant from the standpoint of environmental impact and, pursuant to 10 GFR 551.5(d)(4), that an environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Conclusion We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated:

March 30, 1976