ML20210T660

Application for Amends to Licenses DPR-53 & DPR-69, Reflecting Proposed Tech Spec Changes Including Rev to Surveillance Requirement 4.4.10.1.2, Augmented Inservice Insp Program for Main Steam.... Fee Paid

ML20210T660
Person / Time
Site:	Calvert Cliffs
Issue date:	10/01/1986
From:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
To:	Thadani A Office of Nuclear Reactor Regulation
Shared Package
ML20210T667	List: ML20210T660 ML20210T679
References
NUDOCS 8610090128
Download: ML20210T660 (8)

Text

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BALTIMORE GAS AND ELECTRIC CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 JOSEPH A.TIERNAN Vict ParslDENT NUCLEAR ENERGY October 1,1986 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 ATTENTION: Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Amendment

REFERENCE:

(a) BG&E Letter dated March 21, 1977, from A. E. Lundvall, Jr.

(BG&E), to D. L. Ziemann (NRC)

Gentlemen:

The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53 and DPR-69 for Calvert Cliffs Unit Nos.1 & 2, respectively, with the submittal of the proposed changes to the Technical Specifications.

CHANGE NO.1 (BG&E FCR 86-166)

Change page 3/4 4-28.of the Unit 1, and page 3/4 4-29 of the Unit 2 Technical Specifications as shown on the marked up copies attached to this transmittal.

DISCUSSION April 1,1987, starts the second 10-year Inservice Inspection (ISI) interval for Calvert Cliffs Unit Nos.1 & 2. In accordance with 10 CFR 50.55a(g)(4Xii), we must revise our inspection requirements to comply with the latest edition and addenda of ASME Code Section XI that is referenced in 10 CFR 50.55a(b) twelve months prior to the start of the next inspection interval. The code in effect during this timeframe, and currently the latest edition of the code referenced in 10 CFR 50, is the 1983 Edition (with Addenda through the summer 1983 Addenda).

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Mr. Ashok C. Thadani October 1,1936 Page 2 10 CFR 50.55a(gX5)(li) requires the licensee to apply to the NRC for amendment of the Technical Specifications to conform to the revised ISI program six months prior to the start of the new inspection interval. Accordingly, a review was performed to identify areas where the revised ISI program conflicts with the existing Technical Specifications. One major area of difference concerns snubber testings. The 1983 edition of the code now addresses snubber visual and functional testing. We plan to continue snubber testing in accordance with the Technical Specifications for the immediate future, while industry-wide studies of code snubber testing (and their acceptability to the NRC) are evaluated.

The only change to Technical Specifications necessary prior to adopting the 1983 edition of the code is an administrative change to Surveillance Requirement 4.4.10.1.2, /

" Augmented ISI Program for Main Steam and Main Feedwater Piping." This change simply updates the reference in the Surveillance Requirement to address the 1983 code.

Additionally, since we are beginning our second 10-year inspection program, we propose to delete two references to inspections performed during the first 10-year interval. The first of these changes simply deletes paragraph 4.4.10.1.2.a, regarding a 100% volumetric examination of each weld prior to exceeding 18 months of operation. The second deletes reference to the first 10-year interval in paragraph 4.4.10.1.2.b, and rewords the Surveillance so it is applicable to any succeeding 10-year interval.

DETERMINATION OF SIGNIFICANT HAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or l

This change does not involve a significant increase in the probability or consequences of an accident previously evaluated. The 1983 edition of the ASME code has been previously approved for industry use by the NRC.

(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or No changes to plant equipment configuration are involved with this l change. Therefore, this change does not create the possibility of a l new or different type of accident.

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Mr. Ashok C. Thadani October 1,1986 Page 3 (iii) involve a significant reduction in a margin of safety.

Updating our ISI program to the current ASME code does not involve a reduction in a margin of safety. The 1983 code maintains strict limitations on inspection and testing, and provides updated industry criteria for Section XI programs.

CHANGE NO. 2 (BG&E FCR 86-175)

Change page 3/4 3-20 of the Unit 1 & 2 Technical Specifications as shown on the marked-up copy is attached to this transmittal.

DISCUSSION We propose to change the required response time for the containment purge isolation valves from five to seven seconds. Since these valves are no longer opened during reactor operation, five seconds is overly restrictive.

The Engineered Safety Features Actuation System (ESFAS) response time for the containment purge isolation valves was reduced to five seconds in September 1980 in accordance with NRC guidance to minimize the release of activity following a postulated Loss of Coolant Accident (LOCA). At the time, these valves could be opened to purge the containment during reactor operation. In December 1981, in response to the continued NRC concerns related to closure of these valves in a post-LOCA environment, administrative controls were put in place to assure the valves would be closed during MODES 1 through 4.

Since the containment purge isolation valves can no longer be opened during operating MODES in which a LOCA could occur, the basis for the automatic closure of these valves is a fuel handling incident in the containment. Increasing the required ESFAS response time for the containment purge isolation valves from five to seven seconds would not significantly change our ability to isolate the containment in response to a fuel handling incident. A seven second response time is bounded by the conservative assumption regarding containment isolation made in our evaluation of the incident in Reference (a):

Based on the normal ventilation and cooling system which will be in operation during refueling, the activity will be well mixed inside the containment.

In the most probable case, the radiation monitors would detect the release in a few seconds and shut down the purge system automatically. However, for this analysis it was assumed that the activity was released, r..ixed by the ventilation system throughout 1/2 of the containment volume and purged at the resulting concentration for one minute. The detectors then alarmed and shutdown the exhaust (containment purge) system.

Mr. Ashok C. Thadani October 1,1986 Page 4 This evaluation concluded that the fuel handling incident in the containment is bounded by the fuel handling incident in the Auxiliary Building, as discussed in Section 14.18 of the Updated Final Safety Analysis Report. Both incidents result in doses at the site boundary which are significantly less than 10 CFR 100 limits.

DETERMINATION OF SIGNIFICANT HAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or Automatic closure of the containment purge isolation valves is required to mitigate the consequences of a fuel handling incident in the containment. An ESFAS response time of seven seconds for these valves is bounded by the assumption regarding containment isolation used in our previous evaluation of this incident. Therefore, the consequences of this incident would not be significantly increased.

(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or This proposed change does not involve a change to system design or method of operation. The surveillance test procedure change involved would not create the possibility of a new or different accident.

(iii) involve a significant reduction in a margin of safety.

Since this proposed change is bounded by the assumptions made in our previous evaluation of a fuel handling incident in the containment, this change would not involve a significant reduction in a margin of safety.

CHANGE NO. 3 (FCR 86-3000)

Replace Unit 1 & 2 Technical Specification pages 3/41-17 through 3/41-19 and B 3/41-4 with the attached marked-up pages.

DISCUSSION A change is proposed to Technical Specification 3/4.1.3, Movable Control Assemblies Full Length CEA Position and its bases to permit Calvert Cliffs Units 1 & 2 to continue operations at steady state power for some period of time following a dropped Control Element Assembly (CEA). This period of time will be used by the Operators to realign

Mr. Ashok C. Thadani October 1,1986 Page5 the dropped CEA. If the CEA cannot be realigned within the allowed time period, appropriate guidelines will be observed to assure that the increase in peaking factors caused by xenon redistribution effects will not compromise thermal margin. The allowed pe 'od of time will depend on the pre-drop value of the integrated radial peaking factor (F ) measured at the plant during normal power distribution surveillances.

A second, purely administrative change is proposed to Technical Specification 3/4.1.3 eliminating the term " full length" when referring to CEAs. Such a change is appropriate since all part length CEAs have been removed from the reactor and there are no plans to reinstall them.

When a CEA drops into the core, it causes a substantial decrease in power in the vicinity of the dropped rod. The drop also adds negative reactivity causing a prompt drop in reactor power varying from 4% to 20% depending on the worth of the dropped rod. Since the heat extraction by the secondary plant remains essentially constant, the primary side temperature also drops. These decreases in power and moderator temperature in turn, affect the core power level via the feedback effects af ter the length of time necessary for the fuel (fuel time constant) and coolant (loop cycle time) to respond. Since the doppler and moderator temperator coefficients are both negative, the decreasing fuel and coolant temperatures add positive reactivity. Consequently, af ter several minutes core power returns to its initial value, due to the positive reactivity inserted, and core temperature levels out at a value lower than its initial value.

i The presence of the dropped CEA results in a distorted core power distribution and corresponding increased peaking factors which results in degradation in the margin to the l DNBR Specified Acceptable Fuel Design Limit (SAFDL). The event is protected by l building into the DNBR LCO (Technical Specification 3.2.5) enough margin to accommodate the initial effects of the worst CEA drop at any time during the core cycle.

For a dropped CEA, the required overpower margin that is factored into the DNBR LCO l does not include the radial peaking effects of xenon redistribution following the dropped l CEA. Xenon redistribution effects start to become significant approximately five to ten minutes af ter the CEA is dropped. Therefore, the current action statement for Technical Specification 3/4.1.3 require immediate operator action in reducing power to less than or equal to 70% power within one hour after the drop. This reduction in power

level is intended to offset the potential increase in radial peaking factor following a l dropped CEA.

The proposed changes to Technical Specification 3/4.1.3 allow the Operator some time to realign the CEA prior to reducing powe.g levels. The allowed time to realign the CEA varies as a function of the pre-drop F value (see Figure 3.1-3). This time limit it 8

permitted for the following reasons:

.. E Mr. Ashok C. Thadani October 1,1986 Page 6 1.

The DNBR margin calculations are based which support on a steady-state radial the power peaking distribuy=n factor LCO for of F r 1.65.

2. When the actual Ffis less than 1.65, additional margin exists.

3. This additional margin can be credited to offset the increase in Ff that can occur following the CEA drop due to xenon redistribution.

The requirement to reduce power level af r the time limit of Figure 3.1-3 has been reached offsets the continuing increase in F that occurs due to xenon redistribution. A reduction in power leve to 75% of the value prior to the CEA drop is sufficient to offset the worst increase in F due to xenon redistribution. A power reduction is not required when operating below 0% because there exists sufficient conservatism in the DNBR power distribution LCOs to completely offset the xenon redistribution effect.

A detailed analysis of the effects of a dropped CEA was performed using the ROCS code to determine the worst change in radial peaking factor as a function of time due to xenon redistribution following a CEA drop. This analysis covered the possible combinations of core power versus CEA insertion allowed by the Power Dependent Insert Limit (Technical Specification 3.1.3.6). The CEA drop simulations also bound beginning and end-of-cycle conditions.

A bounding limit curve was generated for the variation in FT with time following a CEA drop. Sufficient conservatism was added to this curve to Eound future cycles. Figure 3.1-3 was g erated from this generic curve by trading o the margin loss due to the increase in F with the margin available when the pre-drop F is less than 1.65.

A power reduction is required af ter the time limi(permitted by Figure 3.1-3 is exceeded in order to offset the continuing increase in F; with time. The ROCS calculations showed that a power reduction to 75% the value prior to the CEA drop would completely offset the worst increase in F with time. This limit includes sufficient conservatism to bound future cycles.

For Calvert Cliffs Units 1 & 2, the transients which are limiting for setting the DNBR LCO are Loss of Flow, CEA Withdrawal, and CEA Drop. Each transient is limiting over different regions of the allowed operating space. A reanalysis of the DNBR LCO was performed using only the overpower margin data set by CEA drop in order to quantify the available conservatism in the DNBR LCO relative to CEA drop requirements. _ This analysis showed that below 50% power there is sufficient conservatism to completely offset the xenon redistribution effects. Therefore, a reduction in power is not necessary for CEA drops occurring below 50% power.

Mr. Ashok C. Thadani October 1,1986 Page 7 DETERMINATION OF SIGNIFICANT HAZARDS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:

(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or The proposed changes to Technical Specification 3/4.1.3 do not result in an increase in the probability or consequences of an accident previously evaluated because no changes in the assumptions and analytical inputs to the transient analyses were required.

(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or The proposed changes to Technical Specification 3/4.1.3 do not create the possibility of a new or different kind of accident from any accident previously evaluated because it does not involve a change to the configuration of the plant, equipment design, or equipment used.

(iii) involve a significant reduction in a margin of safety.

The proposed changes to Technical Specification 3/4.1.3 will not result in a reduction in the margin of safety because the action statements that specify the Operator actions following a CEA drop will assure that sufficient margin exists to offset any adverse effects caused by xenon redistribution.

SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.

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Mr. Ashok C. Thadani October 1, 1986 Page 8 FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. 1082433 in the amount of $150.00 to the NRC to cover the application fee for this request.

Very tjuly ours,

-i. . Aws

+ J. A. T ernan STATE OF MARYLAND :

TO WIT:

CITY OF BALTIMORE :

Arthur E. Lundvall, Jr., being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth;.that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation.

WITNESS my Han'd and Notarial Scal: u,,jd , .

Motary Public Y

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My Coinmission Expires: 7-/-f6 ~

[ //. /ff6 Date' JAT/BEH/MTF/JFW/gla Attachments cc: D. A. Brune, Esquire J. E. Silberg, Esquire S. A. McNeil, NRC T. Foley, NRC T. Magette, DNR

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