ML17221A337

Exemption from 10CFR50,App J Requirement Re Testing of Air Locks Opened During Periods When Containment Integrity Not Required by Plant Tech Specs

ML17221A337
Person / Time
Site:	Saint Lucie
Issue date:	08/19/1987
From:	Varga S Office of Nuclear Reactor Regulation
To:	FLORIDA POWER & LIGHT CO.
Shared Package
ML17221A338	List: ML17221A337
References
NUDOCS 8708250048
Download: ML17221A337 (17)

Text

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of FLORIDA POWER AND LIGHT COMPANY (St. Lucie Plant, Unit No. I)

Docket No. 50-335 EXEMPTION Florida Power and Light Company (the licensee) is the holder of Facility Operating License No.

DPR-67 that authorizes the operation of the St. Lucie Plant, Unit No.

1 (the facility) at a steady-state power level not in excess of 2700 megawatts thermal.

The facility is a pressurized water reactor (PWR) located at the licensee's site in St. Lucie County, Florida.

The license

provides, among other things, that the facility is subject to all rules, regulations and orders of the Cemission now or hereafter fn effect.

10 CFR 50.54(o) states that primary reactor containments shall be subject to the requirements set forth in Appendix J to this part.

Appendix J to 10 CFR Part 50, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,"

sets forth the detailed requirements for containment leakage testing.

These test requirements provide for preoperational and periodic verification by tests of the leak-tight integrity of the primary reactor containment and systems and components which penetrate containment of water-cooled power reactors, and establish the acceptance criteria for such tests.

Section III of Appendix J addresses the specific leakage testing requirements.

8708250048 870819 PDR ADOCK 05000335 P

An exemption request was submitted by the licensee by letter dated January 9, 1987; IV.

The licensee's proposed leak.testing of containment air locks is in compliance with the requirements of Appendix J to 10 CFR Part 50, with one exception.

The licensee has requested an exemption from paragraph III.D.2(b)(ii) of Appendix J, which states:

Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such periods at not less than P

Mhenever the plant is in cold shutdown, containment integrity is not required.

however, if an air lock is opened during cold shutdown, paragraph III.D.2(b)(ii) requires that an overall air lock leakage test at not less than P

be conducted after it is closed.

The existing air lock doors are so designed a

that a full pressure, i.e.,

P (39.6 psig), test can only be performed after strong backs (structural bracing) have been installed on the inner door.

Strong backs are needed since the pressure exerted on the inner door during the test is in a direction opposite to that of the accident pressure direction.,

The strong backs are extremely difficult to install ahd the outer door must be opened to remove the strong backs.

Installing strong backs, performing the

test, and removing the strong backs, is a cumbersome process requiring at least 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> during which access through the air lock is prohibited.

Alternatively, the licensee proposes to test the seals of the inner and outer doors by pressurizing the area between the seals and verifying an accept-able leakage rate prior to returning to a plant operating condition requiring containment integrity.

The licensee contends this proposal will provide adequate assurance of air lock integrity without imposing undue delays on return to power operation.

If the periodic 6-month test of paragraph III.D.2(b)(i).and the test required by paragraph III.D.2(b)(iii)are current, there should be no reason to expect an air lock to leak excessively just because it has been opened during cold shutdown or refueling.

Moreover, if maintenance has been performed on the air lock since the last successful test pursuant to paragraph III.D.2(b)(i), an overall air lock test will be performed by the licensee instead of the seal test 'described

above, Accordingly, the staff concludes that the licensee's proposed

approach, consisting of testing the seals of the inner and outer doors by pressurizing the area between the seals and verifying an acceptable leakage rate prior to returning to a plant operating condition requiring containment integrity, is acceptable.

Based upon this alternative.testing, an exemption from the requirements paragraph III.D.2(b)(ii) of Appendix J to 10 CFR Part 50 is proper.

V.

Accordingly, the Commission has determined that, pursuant to 10 CFR 50.12(a),

this exemption as described in,Section IV is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the common defense and security.

The Commission further determines that special circumstances as provided in 10 CFR 50.12(a)(2) are present justifying the exemption.

In a letter dated January 9, 1987, the licensee provided information to the "special circumstances" finding required by revised 10 CFR.50.12(a)

(See 50 FR 50764).

The licensee stated that thc application of the requirements of 10 CFR 50, Appendix J, Paragraph III.D.2(b)(ii) is not necessary to serve the underlying purpose of these regulations.

Thc basis for this statement is the fact that the alternatives presented limit the postulated accident doses to within the 10 CFR Part 100 guidelines.

Therefore, the special circumstances of Section 50.12(a)(2)(ii) apply to these specific exemption requests.

Therefore, thc Cereission hereby grants the exemption request identified in Section IV above.

Pursuant to 10 CFR 51.32 thc Commission has determined that the granting of the Exemption will not result in any significant impact on the environment (52 FR 10273).

This exemption is effective upon issuance.

Dated at Bethesda, Maryland, this 19th day of August, 1987.

FOR THE NUCLEAR GULATORY COMMISSION

arga, ircc Division of Reactor Pro ts-I/II Office of Nuclear Reactor egulation

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LICENSE AUTHOIIITY.FILE COP/

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 March 11, 1988 00 KT REMOVE POS+dd2-Pnz2?'.

'F/

~ DP8-Ca7 Docket Nn. 50-335 Mr.

W.

F.

Conway Actinq Group Vice President Nuclear Energy Florida Power 8 Light Company P.

0.

Box 14000 Juno

Reach, Florida 33408

SUBJECT:

ST.

LUCIE UNIT 1 -

ISSIIANCE OF AMENDMENT RE:

SPFNT FUEL POOL EXPANSION (TAC NO. 65589)

The Commission has issued the enclosed Amendment No. 91 to Facility Operating License No.

DPR-67 for the St. Lucie Plant, Unit No.

J..

This amendment consists of changes to the Technical Specifications in response to your application dated. June 12,

1987, as supplemented by letters dated September 8, 1987, October 20, 1987 (three letters),

December 21,

1987, December 22,

1987, December 23, 1987 (three letters),

and January 29, 1.988.

I This amendment allows the expansion of the.'spent fuel pool storage capacity from

'he current 728 fuel assemblies to the proposed 1706 fuel assemblies.

The expansion is to be achieved by removing the existing racks and installing new, higher density ones.

The request for the amendment was individually noticed ir the Federal Re ister on August 31, 1987 (52 Fl? 32852), followed by a biweekly notice on eeeotem er.3, 1987

<52 FR 38513).

A request for a public hearing was filed on September 30, 1987 by Mr. Campbell Rich.

Ry undated letter, Mr. Rich subsequently filed 16 contentions.

The 16 contentions are addressed in the enclosed Safety Evaluation.

s The Safety Evaluation also includes a Final Determination of No Significant Hazards Consideration.

Under NRC regulations, the Commission may issue and make an amendment immediately effective, notwithstanding a request for a hearing, in advance of'o1ding the hearing where, as here, it has been determined that the amendment involves no significant hazards consideration.

Such issuance is also consistent with Section 132 of the Nuclear Waste Policy Act of'982; which requires the Commission to encourage and expedite the effective use of available storage at civilian reactor sites.

The Environmental Assessment related to this action was transmitted to you on February 29, 1988.

The Notice of Issuance of Environmental Assessment and Finding of No Significant Impact was published in the Federal

~Re later on March 4, 1988 (53 FR 7065),

0

s Mr. ll. F.

Conway A c'py of the Safety Evaluation is enclosed.

A copy of Notice o~ Issuance and Final Determination of No Significant Hazards Consideration is also enclosed The Notice of Issuance will also be included in the Commission s bi-weekly Federal

~Re ister notice.

Sincerely Enclosures l.

Amendment No.

91 to DPR-67 2.

Safety Evaluation 3.

Notice of Issuance cc w/enclosures See next page DISTRIBUTION NRC 8 Local PDRs PDII-2 R/F S.

Varga G. Lainas D. Miller E. Tourigny OGC-h'F D.

Hagan J. Partlow T. Barnhart (4)

Manda Jones E. Butcher ACRS(10)

GPA/PA ARl1/LFMB Gray File Original signed by E.

G. Tourigny, Project Manager Project Directorate II-?

Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation 2

er O / i/Ss gny:bd keA F 03/y/88 03//I/88 OGC-MF N~/<<<~j 03///88

Mr.

W. F.

Conway F~orida Power IC Light Companv St. Lucie Plant cc Nr. Jack Shreve Office of the Pub'ic Counsel Room 4, Holland Building Tal'ahassee, Florida 32304 Resident Inspector c/o U.S.

NRC 7585 S.

Hwy AlA Jensen

Beach, Florida 34957 State Planninq 8 Development Clearinqhouse Office nf Planning 4 Budget Executive Office of the Governor The Capitol Building Tallahassee, Florida 32301 Harold F. Reis, Esq.

Newman 8 Holtzinger 1615 L Street, N.W.

Washington, PC 20036 John T. Butler, Esq.

Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Administrator Department of Environmental Peaulation Power Plant Siting Section State of Flnrida 26CO Blair Stone Road Tallahassee, Florida 32301 Nr. Weldon B. Lewis, County Administrator St. Lucie Cnunty 2300 Virginia Avenue, Room 104 Fort Pierce, Florida 33450 Yr. Charles B. Brinkman, Manager Washington - Nuclear Operations Combustion Engineering, Inc.

7910 Wnodmnnt Avenue

Bethesda, Maryland

>0814 Jacob Daniel Nash Office of Pad'.ation Control Department o< Health and Rehabilitative Services 13l7 Winewood Blvd.

Tallahassee, Florida 32399-0700 Reginnal Admiristrator, Region II ll.S. Nuclear Regulatorv Comnissinn Executive Pirectnr for Operations 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323

~gR REGS (4

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FLOP IDA POWER 5 LIGHT COMPAtIY DOCKET N0. 50-335 ST.

LUCIE PLANT UNIT Ã.

1 AHFNDNENT TO FACILITY OPFPPTING LICENSF Amendment No.

91 License No.

DPR-67 1.

The Nuclear Pegulatory COIImission

(.he COIImission} has found that:

A.

The application for amendment by Florida Power 8 Light Company, (the licensee) dated June 1P.,

l,9o87, as supplemented by 'Ietters dated September 8, lqP7, October 20, 1987 (three letters),

December 21,

1987, December 22,

1987, December 23, 1987 (three lItters),

and January 29, 1988, complies with the standards and reouirements 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; 8.

The facility will operate in conformity with the application, the provis'nns 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 endanaerino the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comoission's regulations; D.

The issuance of this amendmen

. will not be inimical to the common defense and security nr to the health and safetv of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the ComIission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, Facility Operating License No.

DPP-67 is amended by changes to the Technical Speci+ications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.(P) to read as fol'ows.

(2)

Technical S ecifications The Technical Specifications contained in Appendices A

and 8, as revised throuoh Amendment No.

9>

, are herebv incoroorated in the license.

The licensee shall operate the fac'.lity in accordance with the Technical Specifications.

3.

This license amendment is effective as o+ the date of its issuance.

FOR THE NUCLEAR REGl!LATORY CONYISSION erb~rt N. Berkow, Director Prn ect Directorate II-P, Division of Reactor Proiects-I/II Office of Nuclear Reactor Reoulatinn

Attachment:

Charges to the Technical Specifications Date of'ssuance:

March 11, 1988

ATTACHMENT TO LICFNSE AMFNDYENT NO.

TO FACILITY OPERATING LICFNSF. NO.

DPR-67 DOCKET NO. 50-335 Replace the followinq pages of the Appendix "A" Technical Specifications with the enclnsed paqes.

The revised page's ar~ identified bv amendment number and contain vertical 1':nes indicating the area of change.

The corresponding overleaf pages are also provided to maintain document com-pleteness.

~ll.

P 8 3/4 9-3 5-6 Insert Pa es 8 3/4 9-3 5

5-6 5-6a 5-6b

I REFUELING OPERATIONS BASES 3/4.9.12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assump-tions of the accident analyses.

3/4.9.13 SPENT FUEL CASK CRANE The maximum load which may be handled by the spent fuel cask crane is limited to a loaded single element cask which is equivalent to approxi-mately 25 tons.

This restriction is provided to ensure the structural integrity of the spent fuel pool in the event of a dropped cask accident.

Structural damage caused by dropping a load in exceed of a loaded single olement cask could cause leakage from the spent fuel pool in excess of the maximum makeup capability.

3/4.9.14 DECAY TIME -

STORAGE POOL The minimum requirements for decay of the irradiated fuel assemblies in the entire spent fuel storage pool prior to movement of the spent fuel cask into the fuel cask compartment insur e that sufficient time has elapsed to allow radioactive decay of the fission products.

The decay time of 1180 hours0.0137 days <br />0.328 hours <br />0.00195 weeks <br />4.4899e-4 months <br /> is based upon one-third of a core placed in the spent fuel pool each year during refueling until the pool is'filled.

The decay time of 1490 hours0.0172 days <br />0.414 hours <br />0.00246 weeks <br />5.66945e-4 months <br /> is based upon one-third of a core being placed in the spent fuel pool each year during refueling following which an entire core is placed in the pool to fill it.

The cask drop analysis assumes that all of the irradiated fuel in the filled pool (7 2/3 cores) is ruptured and follows Regulatory Guide 1.25 methodology, except that a Radial Peaking Factor of 1.0 is applied to all irradiated assemblies.

ST.

LUGIE - UNIT 1

8 3/4 9-3 Amendment No. gg, fg, 91

t~ )

DESIGN FEATURES CONTROL ELEMENT ASSEMBLI ES 5.3.2 The reactor core shall contain 73 full length and no part length control element assemblies.

The control element assemblies shall be designed and maintained in accordance with the original design provisions contained in Section

4. 2.3.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4. 1 The reactor coolant system is designed and shall be maintained:

a.

b.

C.

In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, For a pressure of 2485 psig, and For a temperature of 650'F, except for the pressurizer which is 700'F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 11,100

+ 180 cubic feet at a nominal T

of 567'F.

5.5 EMERGENCY CORE COOLING SYSTEMS

5. 5. 1 The emergency core cooling systems are designed and shall be main-tained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5. 6 FUEL STORAGE CRITICALITY 5.6. l.a The spent fuel storage racks are designed and shall be maintained with:

l.

A k ff equivalent to less than or equal to 0.95 when flooded witk unborated water, which includes a conservative allowance of 0.0065 hk for uncertainties.

ST.

LUCIE - UNIT 1

5-5 Amendment No. gg,g7, 7$,

91

DESIGN FEATURES CRITICALITY (Conti nued 2.

A nominal 10.12 inches center to center distance between fuel assemblies in Region 1 of the storage racks and a nominal 8.86 inches center to center distance between fuel assemblies in Region 2 of the storage racks.

3.

A boron concentration greater than or equal to 1720 ppm.

4.

Neutron absorber (boraflex) installed between spent fuel assemblies in the storage racks in Region 1

and Region 2.

b.

Region 1 of the spent fuel storage racks can be used to store fuel which has a U-235 enrichment less than or equal to 4.5 weight percent.

Region 2

can be used to store fuel which has achieved sufficient burnup such that storage in Region 1 is not required.

The initial enrichment vs.

burnup requirements of Figure 5.6-1 shall be met prior to storage of fuel assemblies in Region 2.

Freshly discharged fuel assemblies may be moved temporarily into Region 2 for purposes of fuel assembly inspection and/or repair, provided 'that the configuration is maintained in a checkerboard pattern (i.e., fuel assemblies and empty locations aligned diagonally).

Following such inspection/repair activities, all such fuel assemblies shall be removed from Region 2 and the requirements of Figure 5.6-1 shall be met for fuel storage.

c.

The new fuel storage racks are designed for dry storage of unirradiated fuel assemblies having a U-235 enrichment less than or equal to 4.0 weight percent, while maintaining a

k ff of less than or equal to 0.98 under the most reactive condition.

DRAINAGE 5.6.2 The fuel pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 56 feet.

CAPACITY 5.6.3 The spent fuel pool is designed and shall be maintained with a storage capacity limited to no more than 1706 fuel assemblies.

5. 7 SE ISNIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as seismic Class I

in Section 3.2.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 3.7 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirement.

ST.

LUCIE - UNIT 1

5-6 Amendment No. 77,22.28~88.

75~

91

0 DESIGN FEATURES

5. 8 METEOROLOGICAL TOllER LOCATION 5.8.1:The meteorological tower location shall be as shown on Fi'gure 5.1-1.

5.9 CO! IPONENT CYCLE OR TRANSIENT LIMITS 5.9.1 The components identified in Table 5.9-1 are designed and sha11 be maintained within the cyclic or transient limits of Table 5.9-1.

ST.

LUGIE - UNIT 1

5-6a Amendment No.

91

QV 4

25 z

20 x

4 15 ACCEPTABLE BURNUP DOMAIN UNACCEPTABLE BURNUP DOMAIN 10 1.5 2.0 2.5 3,0 3.5 INITIALENRICHMENT, WT % U-235 4.0 4.5 F 16URE

5. 6-i 1K1T IAL EKR1CHNEKT VS TURNUP REOU1REllEKTS FOR STORA6E OF FUEL ASSENbLlES lN kE610K L ST. LUCIE PLANT UNIT 1 5-6b ggendnent t(o.

01