ML20134C967

Proposed Tech Specs 2.0 Re Safety Limits

ML20134C967
Person / Time
Site:	Pilgrim
Issue date:	01/24/1997
From:	BOSTON EDISON CO.
To:
Shared Package
ML20007F895	List: BECO-LTR-97-002, Application for Amend to License DPR-35,requesting Rev to Safety Limit Minimum Critical Power Ratio.Encl Attachment 1 Re Description & Evaluation of Proposed TS Change to Min Critical Power Ratio Safety Limit Withheld ML20134C967
References
NUDOCS 9702040206
Download: ML20134C967 (19)

Text

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2.0 SAFETY LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow <

10% of rated core flow:

THERMAL POWER shall be 5 25% of RATED THERMAL POWER.

2.1.2 With the reactor steam dome pressure 2 785 psig and core flow 2 10% of rated core flow:

MINIMUM CRITICAL POWER RATIO shall be 2 2.1.3 Whenever the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the normal active fuel zone.

2.1.4 Reactor steam dome pressure shall be s 1325 psig at any time when irradiated fuel is present in the reactor vessel.

2.2 Safety Limit violation With any Safety Limit not met the following actions shall be met:

2.2.1 Within one hour notify the NRC Operations Center in accordance with 10CFR50.72.

2.2.2 Within two hours:

A. Restore compliance with all Safety Limits, and B. Insert all insertable control rods.

2.2.3 The Station Director and Senior Vice President - Nuclear and the Nuclear Safety Review and Audit Committee (NSRAC) shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.2.4 A Licensee Event Report shall be prepared pursuant to 10CFR50.73. The Licensee Event Report shall be submitted to the Commission, the Operations Review Committee (ORC), the NSRAC and the Station Director and Senior Vice President - Nuclear within 30 days of the violation.

2.2.5 Critical operation of the unit shall not be resumed unti authorized by the Commission.

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Amendment No. 15;-27 1-42r-721-133;-146 2-1

,.o'3 BASES:

2.0 FAFETY LIMITS INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish a Safety Limit such that the Minimum Critical Power Ratio (MCPR) is not less than the limit specified in Specification 2.1.2. MCPR greater than the specified limit represents a conservative margin 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 from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling (i.e., MCPR of 1.0). These conditions represent a significant departure from the condition intended by

][ v s e; _dr_ design for planned operation. The MCPR fuel cladding integrity

" Safety Limit assures that during normal operation and during j4g anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling.

FUEL CLADDING CE critical power correlations are applicable for all critical INTEGRITY (2.1.1) power calculations at pressures at or above 785 psig or core flows at or above 10% of rated flow. For operation at low pressures and low flows another basis is used as follows:

(Cont)

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Amendment No. 157 72 1-105 -129 5-133 5 2 B2-1

BASES:

2.0

• SAFETY LIMITS (Cont)

FUEL CIADDING Since the pressure drop in the bypass region is essentially all INTEGRITY (2.1.1) elevation head, the core pressure drop at low power and flows will (Cont) always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 108 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATIAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critica'.

power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

(2.)

The Safety Limit MCPR is dete ned us n t e General Electric Thermal Analysis Basis, CETAB , which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.

The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. These conditions are:

Pressure: 800 to 1300 psia Mass Flux: 0.1 to 1.5 Mlb/hr-ft2 Inlet Subcooling: 0 to 70 Btu /lb Axial Profile: 1.5 chopped cosine 1.7 inlet peaked 1.7 outlet peaked R-Factor 0.95 to 1.20 Rod Array 9X9 CE 11 array MINIMUM CRITICAL The fuel cladding integrity Safety Limit is set such that no fuel POWER RATIO damage is calculated to occur if the limit is not violated. Since i (2.1.2) the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and 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 nucleate boiling would not result in damage to BWR fuel rods, the critical power at (Cont)

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• Amendment No. 15-42-72,-105,-129,-133,TR 3 1 B2-2

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BASES:

2.0 SAFETY LIMITS (Cont)

MINIMUM CRITICAL which boiling transition is calculated to occur has been adopted POWER RATIO as a convenient limit. However, the uncertainties in monitoring (2.1.2) (Cont) the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit MCPR is. determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1. Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis.

The statistical analv 'a used to determine the MCPP. safety limit is based on a model of the BUR core which simulates the process computer function. The reactor core selected for these analyses was a large 764 assembly, 251 inch reload core. Results from the rge reload core analysis apply for all operating reactors for all reload cycles, including equilibrium cycles. Random Monte Carlo selections of all operating parameters based on the uncertainty ranges of manufacturing tolerances, uncertainties in measurement of core operating parameters, calculational uncertainties, and statistical uncertainty associated with the critical power correlations are imposed upon the analytical l representation of the core and the resulting bundle critical powe I ratios. Details of this statistical analysis are presented in Reference

, REACTOR WATER With fuel in the reactor vessel during periods when the reactor is LEVEL (Shutdown shutdown, consideration must be given to water level requirements Condition) due to the effect of decay heat. If reactor water level should (2.1.3) drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be cooled sufficiently should the water level be reduced to two thirds the core height.

Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be cont!nurusly monitored.

(Cont) 1Mvision 4 474 Amendment No. 151-423 -72 2-133 B2-3

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BASES:

2.0 SAFETY LIMITS (Cont)

I I I I I REACTOR STEAM The Safety Limit for the reactor steam dome pressure has been DOME PRESSURE selected such that it is at a pressure below which it can be show-(2.1.4) that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and ]

Pressure Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%,

i 1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 1

psig, as measured by the reactor steam dome pressure indicator, is i equivalent to 1375 psig at the lowest elevation of the reactor l l

coolant system. The reactor coolant system is designed to the  !

USAS Nuclear Power Piping Code, Section B31.1.0 for the reactor recirculation piping, which permits a maximum presso.4 transient of 120% of design pressures of 1148 psig at 562 *F for suction piping and 1241 psig at 562

• F for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure j allowed by the applicable codes.

REFERENCES 1. " General Electric Standard Application for Reactor Fuel," l NEDE-240ll-P A (Applicable Amendment specified in the CORE OPERATING LIMITS REPORT).

2. General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Applicatica, General Electric Co. BWR Systems Department, January 1977, NEDE-10958 PA and NEDO-10958-A.

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Amendment No -15;-1331 146 B2-4 l

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S7Mpn O TS I?ME i Reactor Cera SLs B 2.1.1 BASES l

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! BACKGROGC Operation abcVe the boundary of the nucleate boiling regime

! (continued) could result in excessive cladding temperature because of

! the onset of transition boilinfi and the resultant sharp i reduction in heat transfer coe"ficient. Inside the steam l .

film, high cladding temperatures are reached, and a cladding j - water (zirconium water) reaction may take place. This

% chemical reaction results in oxidation of the fuel cladding 3M to a structurally weaker form. This weaker fom may lose its integrity, resulting in an uncontrolled release of

" A, activity to the reactor coolant.

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l PLICABLE The fuel cladding must not sustain damage as asult of 1

S Y ANALYSES normal operation and A00s. The reactor cor Ls are -

established to preclude violation of the f 1 design criterion that an MCPR limit is to be es 4blished, such that

at least 99.9% of the fuel rods in the ore would not be j expected to experience the onset of ansition boiling.

l The Reactor Protection System set ints (LC0 3.3.1.1, eactor Protection System (RPS) nstrumentation"),in j c ination with the other LCO , are designed to prevent any i ant pated combination of tr nsient conditions for Reactor i Coolan System water level, ressure, and THERMAL POWER level th would result i reaching the MCPR limit.

3 2.1.1.la Fuel ladd a Intearity IGeneral Electric Company I (GE) FGm1Y GE critical powe cor lations are applicable for all s critical power alculat ns at pressures 2 785 psig and core flows 210% o rated flow. For operation at low pressures or low flow , another basis s used, as follows:

Sin the pressure drop in e by> ass region is e entially all elevation he tie core pressure p at low power and flows wi always be

> 4.5 psi. Analyses (Bef. 2) sh that with a bundle flow of 28 x 10' lb/hr, bu pressure drop is nearly independent of bundle er and has a value of 3.5 psi. Thus, the bun flow with a 4 5 psi driving head will be

> 28 x 10' lb/hr. Full scale ATLAS test da taken at pressures from 14.7 psia to 800 psia (conti ed)

BWR/4 STS B 2.0-2 Rev 1, 04/07/95

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• NOTES FOR TABLE 3.2.C-1

1. ,

With the number of operable ch:nnels:

a. One less than required by the minimum operable channels per trip function requirement, restore an inoperable channel to operable status within 7 days or place an inoperable channel in the tripped condition within the next hour.

b. Two or more less than required by the minimum operable channels per trip function requirement, place at least one inoperable channel in the tripped condition within one hour. ,

2. a. With one RBM Channelinoperable:

(1) restore the inoperable RBM channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; otherwiv. place one tod block monitor channel in the tripped condition within -

, the next hour, and; (2) prior to control rod withdrawal, perform an instrument function test of the l operable RBM channel.  ;

b. With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within one hour.

3. If the number of operable channels is less than required by the minimum operable channels per trip function requirement, place the inoperable channel in the tripped condition within one hour.

, 4. SRM operability requirements during core alterations are given in Technical Specification l 3.10.

j 5. RBM operability is required in the run mode in the presence of a limiting rod pattom with

reactor power greater than the RBM low power setpoint (LPSP). A limiting rod pattem l exists when

MCPR < or reactor power ? 90%

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MCPR < [0)or reactor power < 90%

The allowasble value for the LPSP is < 29% of rated core thermal power.

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6. When the reactor mode switch is in the Refuel position, the reactor vessel head is removed, and control rods are inserted in all cure cells containing one or more fuel assemblies, these Rod Block functions are not required.  ;

7. With one or more Reactor Mode Switch - Shutdown Position channels inoperable, suspend l i

control rod withdrawal and initiate action to fully insert all insertable control rods in core  ;

cells containing one or more fuel assemblies immediately. -

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-RevisionA Amendment No. 15,27,42,SS,77,110,138Ih. 3/4.2-22 l

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ATTACHMENT 4 GE AFFIDAVIT l

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Att: chm:nt

. i Affidavit j 1

l I, Ralph J. Reda, being duly sworn, depose and state as follows:

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

(2) The information sought to be withheld is contained in the letter, E. T. Boulette (Boston Edison ,

Co.) to the U. S. Nuclear Regulatory Commission Document Control Desk, Boston Edison letter j number 97-002, Docket No. 50-293, License No. DPR-35. ,

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

("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4) and 2.790(a)(4) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential"(Exemption 4). The 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 Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992),

and Public Citizen Health Research Groun v. FDA. 704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information 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 l license from General Electric constitutes a competitive economic advantage over other l compames; j l 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, prod tion capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric; j e. Information which discloses patentable subject matter for which it may be desirable to l 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.

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l (5) The information sought to be withheld is being submitted to NRC in con 6dence. The information is of a sort customarily held in con 6dence by GE, and is in fact so held. Its initial designation as proprietary information, aad the subsequent steps taken to prevent its-

! unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be l withheld has, to the best of my knowledge and belief, consistotly been held in con &dence by j GE, no public disclosure has been made, and it is not available in public sources. All disclosures 3 t

to third parties including any required transmittals to NRC, have been made, or must be made, l pursuant to regulatory provisions or proprietary agreements which provide for maintenance of l_ the information in confidence.

(6) Initial approval of proprietary treatment of a document is made by the manager of the l'

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.

(8) The information identi6ed in paragraph (2) is classined as proprietary because it would provide other parties, including competitors, with information related to detailed results of analytical  !

models, methods and processes, including computer codes, which GE has developed, requested i 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 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 f GE's competitive position and foreclose or reduce the availability of proGt-making  ;

opportunities. The fuel design and analytical methodology are part of GE's comprehensive

• BWR safety and technology base, and their commercial value extends beyond the original  !

development cost. The value of the technology base goes beyond the extensive physical  !

database and analytical methodology and includes development of the expertise to determine .j 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 money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is~dif6 cult 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.

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, l 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 competitive advantage to seek an ,

I adequate return on its large investment in developing these very valuable analytical tools.

t State of North Carolina ) l SS'-

County of New Hanover )

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

l That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of -  !

l his . knowledge, information, and belief.

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l Executed at Wilmington, North Carolina, this 17 day of January,1997. ,

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Ralph eda -

General Electric Company l

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Subscribed and sworn before me this 17 day of January,1997.

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$$ . Olj a Q E My commission expires on

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,,y Notary Public, State'of North Carolina ,,j4....

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