ML18026A925

Safety Evaluation Supporting Amends 88 & 61 to Licenses DPR-52 & DPR-68,respectively

ML18026A925
Person / Time
Site:	Browns Ferry
Issue date:	03/06/1984
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8404020184
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gl4 9N W~*~4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 88 TO FACILITY OPERATING LICENSE NO.

DPR-52 AND AMENDMENT NO.

61 TO FACILITY OPERATING LICENSE NO.

DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 DOCKET NOS. 50-260 AND 50-296

1. 0 INTRODUCTION By letter dated June 2, 1983 (TVA BFNP TS 188)

(Reference 1), the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License Nos.

DPR-52 and DPR-68 for Browns Ferry Nuclear Plant, Units 2 and 3.

The proposed amendments would revise the Technical Specifications to permit the licensee to increase the coolant flow through the reactor core during coastdown operations (reactor coastdown conditions occur at the end of a reactor cycle, prior to fuel reloading, at which time reactor power has to be reduced due to fuel burnout).

Increasing the reactor coolant flow would reduce the amount of power reduction required of Browns Ferry Units 2 and 3

during the end-of-cycle coastdown operations.

A similar amendment has been previously approved for Unit 1.

(Amendment No. 88'to Facility Operating License No.

DPR-33 issued March 14, 1983.)

In support of this application, the licensee submitted safety evaluations prepared by the General Electric Company (GE),

NEDO-22245 and NEDO-22149, "Safety Review of Browns Ferry Nuclear Plant Units 2 and 3 at Core Flow Conditions Above Rated Flow During Cycle 5," (Reference 2).

The proposed changes to the Technical Specifications are to permit Browns Ferry Units 2 and 3 to operate with core flows up to 105% of rated flow for the rest of the fuel cycle.

The increased core flow would permit the unit to generate about 3% more power than would otherwise be attainable during the coastdown mode of operation.

These amendments do not authorize Browns Ferry Units 2 and 3 to exceed the thermal power limit authorized by License Nos.

DPR-52 and DPR-68.

ii 8404020184 840306

'.O EVALUATION PDR ADQCK 05000260 P

PDR 2.1 Thermal and H draulic Desi n

The objective of the review is to confirm that the thermal-hydraulic design of the core has been accomplished using acceptable

methods, and provides an acceptable margin of safetv conditions which could lead to fuel damage during normal and anticipated operational transients, and is not susceptible to thermal-hydraulic instabilit

The review includes the following areas:

(1) safety limit minimum critical power (MCPR), (2) operating limit MCPR, (3) thermal-hydraulic stability, and (4) changes to Figures 3.5.K-1 and 3.5.2 of the Technical Specifications.

The licensee has submitted analysis reports for each unit for Cycle 5

operation at core flow conditions above rated flow (Reference 2).

The reports rely on a generic document (Reference 3), which has been reviewed and approved (Reference

4) by the staff.

Discussion of the review concerning the thermal-hydraul'ic design for Cycle 5 operation at core flow conditions above rated flow follows:

Safet Limit MCPR The safety limit MCPR has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients.

As stated in Reference 3, the safety limit MCPR is 1.07.

The same safety limit MCPR of 1.07 is used for the Browns Ferry Units 2 and 3 Cycle 5

operation.

0 eratin Limit MCPR The most limiting events have been analyzed by the. licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (bCPR).

The hCPR values given in Table 2-1 of Reference 2 are plant specific values calculated by using the ODYN methods.

The calculated LCPRs are adjusted. to reflect either Option A or Option B LCPRs by employing the conversion method described in Reference 6.

The MCPR values are determined by adding the adjusted ACPRs to the safety limit MCPR.

Table 6. 1 of Reference 2

presents both the cycle MCPR values for the non-pressurization and pressurization events.

The maximum cycle MCPR values (Options A and B) in Table 6. 1 are specified as the operating limit MCPRs and incorporated into the Technical Specifications.

Since the approved method in Reference 3 was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated transients, we conclude that these limits are acceptable.

Thermal-H draulic Stability The results of the thermal-hydraulic analysis (Reference

2) show that the maximum reactor core stability decay ratio while operating with increased core flow during Cycle 5 is bounded by the Reload-4 licensing submittal(s).

These were approved for Browns Ferry Unit 2

,by Amendment No.

85 to Facility Operating License No.

DPR-52 issued March 11, 1983 and for Browns Ferry Unit 3 by Amendment No.

51 to Facility Operating License No.

DPR-68 issued March 29, 1982.

O.

Therefore, we conclude that the thermal-hydraulic stability results are acceptable for increased core flow operation during Cycle 5.

Chan es to Fi ures 3.5.K-1 and 3.5.2 of the Technical S ecifications Figures 3.5.K-1 of the Technical Specifications has been modified to include the operating limit MCPR for Cycle 5 extended flow operation.

Using Option A, for Unit 2 the operating limit MCPRs shall be 1.35 for P8X8R fuel, and 1.32 for SXS and SXSR fuel; for Unit 3 the operating limit MCPRs shall be 1.33 for PSX8R fuel, and 1.32 for SX8 and 8X8R fuel types.

Using Option B, for Unit 2 the operating limit MCPRs shall be 1.26, 1.26 and 1.25 for PSXSR, SX8 SX8R fuel types, respectively; for Unit 3 the operating limit MCPRs shall be 1.26 for P8XSP.,

SX8 and SXSR fuel types.

Figure 3.5.2 has been changed to include a note to reflect that the K factor is equal" to 1.0 for core flows greater than or equal to rated core flow.

The staff has reviewed the Technical Specification changes requested by the licensee.

We find that for the determination of the

OLMCPR, credit is assumed for operation of the highwater level (L8) trip and turbine bypass system.

In this regard, we have concluded that this subject should be treated"as a generic issue, and we plan to handle it in accordance with our internal procedures for dealing with such issues.

We have also determined, based on preliminary analysis, that the risk of operating without Technical Specifications concerning surveillance of the highwater level turbine trip or turbine bypass systems until the generic issue is resolved is small.

Accordingly, we find that the results of analyses are consistent with the proposed OLMCPRs and safety limit MCPR and conclude that the proposed OLMCPRs are acceptable for operation during the remainder of Cycle 5.

Fuel Bundle Liftoff GE reevaluated the bundle liftoffmargin for 105 percent core flow.

The method used was described in a letter from R. Gridley (GE) to D.

Eisenhut (NRC) dated July ll, 1977.

The new analysis yielded a bundle liftoffmargin of 132 lbs., which is 15 lbs. less than the old analysis using 100 percent core flow.

We conclude that this is a

small variation and an adeauate liftoffmargin is maintained for the increased core flow during Cycle 5 operation.

2.2 ~12 The rod block monitor is programmed to block rod withdrawal when its output is 106 percent of full power (0.66 W + 40).

If the program were not changed, at 105 percent flow the block would occur at 109.3 percent of full power.

This would result in a change in CPR of 0.31 for SX8 fuel - an unacceptably high value.

Accordingly the RBM

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upscale flow biased setpoint is clipped at 106 percent rated power.

The change in CPR would then be 0.19 for this event for the SX8 fuel.

This is an acceptable procedure and result.

Table 3.2.C of the Technical Specification has been modified to show this change.

The rod drop accident is a low flow startup event that is not affected by the change in flow except for end of cycle where the initial conditions are slightly altered.

However, end of cycle conditions are not limiting for this event and the previous analysis is still valid.

2.3 Summar of Evaluation We find thermal-hydraulic methods have been used which have 'been approved generically by Reference 4 and that the results of analyses support the proposed limit MCPRs, which avoid violation of the safety limit MCPR for design transients.

We, therefore, conclude that the core flow increase beyond the rated flow will not adversely affect the capability to operate Browns Ferry Nuclear Plant, Units 2 and 3 safely during Cycle 5 extended flow operation and that the proposed changes to Figures 3.5.K-1 and 3.5.2 of the Technical Specifications discussed above are acceptable.

Based on the discussion in Section 2.2 above we conclude that clipping the Rod Block Monitor at 106 percent of rated power will permit the plant to be operated within the limits shown on Figure 3.5.K-1.

The proposed Technical Specifications (Table 3.2.C) have been changed to require this clipping.

We find this acceptable.

In summary, we conclude that operation during the remainder of Cycle 5

for Units 2 and 3 with extended flow will not endanger the health and safety of the public.

3. 0 ENVIRONMENTAL CONSIDERATIONS We have determined that these amendments do 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 determina-

tion, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental

impact, and pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.

4.0 CONCLUSION

S We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will

i5

not be endangered by operation in the proposed

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

Principal Contributors:

W. Brooks, S.

Sun and S. L.

Wu Dated:

March 6, 1984

REFERENCES 1.

Letter from L. Mills (TVA) and attachments to H. Denton (NRC) dated June 2, 1983.

2.

NEDO-22245 and NED0-22149, "Safety Review of Browns Ferry Nuclear Plant Units 2 (dated October 1982) and 3 ('ated June 1983) at Core Flow Conditions above Rated Flow During Cycle 5."

3.

NED0-24011-A-4, "General Electric Boiling Water Reactor Generic Reload Fuel Applications," January 1982.

4.

Letter from D.

G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978.

I 5.

V1003J01A40 and Y1003J01A19, Supplemental Reload Licensing'ubmittal for Browns Ferry Nuclear Plant Units 2 (dated November 1981) and 3

(dated March 1981),

Reload No.

4 (Cycle 5).

6.

Letter from R. Buchholz (GE) to P.

Check (NRC), Response to NRC Request for Information on. ODYN Computer Model,.September 5,

1980.

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~.PGULn I C'RY CGi/Ital ASS< ~I)

January 25, 1984 Docket Nos. 50-259/

60 96 Mr. Hugh G. Parris Manager of Power Tennessee Valley Authority 500 Chestnut Street, Tower IE Chattanooga, Tennessee 37401

Dear Mr. Parris:

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On December 27, 1983, we issued Amendment Nos. 94, 87 and 60 to Facility Operating License Nos.

OPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Unit Nos. 1, 2 and 3.

There were errors on several

pages, due to changes to the Technical Specifications approved subseouent to your submittal of Uarch 25, 1983.

A complete set of oages changed by the above Amendments is enclosed and should be substituted for the pages sent you on December 27, 1983.

We apologize for any inconvenience this may have caused you.

Sincerely,

Enclosures:

As Stated Rs ard J.

lark, Project Manager Operating Reactors Branch 82 Division of Licensing cc w/enclosures:

See next page

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Mr. Hugh G. Parris Tennessee Valley Authority Browns. Ferry Nuclear Plant, Units 1, 2 and 3

cc H. S. Sanger, Jr., Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E 11B 330 Knoxville, Tennessee 37902 Mr. Ron Rogers Tennessee Valley Authority 400 Chestnut Street, Tower II Chattanooga, Tennessee 37401 Mr. Charles R. Christopher

Chairman, Limestone County Commission Post Office Box 188

Athens, Alabama 35611 U. S. Environmental Protection Agency Region IV Office Regional Radiation Representative 345 Courtlana Street, N.

W.

Atlanta, Georgia 30308 Resident Inspector U; S. Nuclear Regulatory Commission Route 2, Box 311

Athens, Alabama 35611 Mr.. Oonald L. Williams, Jr.

Tennessee Valley Authority 400 West Summit Hill Orive, W10885 Knoxville, Tennessee 37902 Ira L. Myers, M. O.

State Health Officer State Oepartment of Public Health State Office Building Montgomery, Alabama 36130 Mr. H. N. Culver 249A HBO 400 Commerce Avenue Tennes'see Valley Authority Knoxville, Tennessee 37902 George Jones Tennessee Valley Authority Post Office Box 2000

Decatur, Alabama 35602 Mr. Oliver Havens U. S. Nuclear Regulatory Commission Reactor Training Center Osborne Office Center, Suite 200 Chattanooga, Tennessee 37411 James P. O'Reilly Regional Administrator Region II Office U. S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100

Atlanta, Georgi a 30303

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ATTACHMENT TO LICENSE AMENDMENT NO.

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FACILITY OPERATING LICENSE NO. DPR-33 DOCKET'O. 50-259 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically nUmbered pages.

180

'181 219 2.

The marginal lines on these pages denote the area being changed.

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ATTACHMENT TO LICENSE AMENDMEHT NO.

87'ACILITY OPERATING LICENSE NO. OPR-52 OOCKET NO. 50-260 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages.

180 181 219 2.

The marginal lines on these pages denote the area being changed.

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ATTACHMENT TO LICENSE AMENDMENT NO. 60 FACILITY OPERATING LICENSE NO. DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

1.

Pemove the following pages and replace with identically numbered pages.

191 192 224 2.

The marginal lines on. these pages 'denote the area being changed.

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I LICENSE AUTHORITY FILE C(Q UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 December 27.,

1983.

Docket Nos.

50-259 260 296'r.

Hugh G. Parris Manager of Power Tennessee Val-ley:Authority 500A Chestnut Street,.

Tower II Chattanooga, Tennessee-37401

Dear Mr'.,

P.arris:

Rhks~ S4QL o moT@.Roy Pa~I a

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~ QPg.-sw The Commission.has issued the enclosed Amendment Nos. 94',. 87 and 60 to Facility Operating License Nos.. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Unit Nos.

1, 2'nd 3.

These arne'ndments are in response Co your appl'ication dated March 25, 1983 (TVA BFNP TS 186) as modified wi.th the concurrence of your staff.

The amendments change the Technical Specifications to add more stringent requ'irements to Section 3.6.C. on allowable, primary coolant leakage into the drywell.

A copy of the Safety Eval'uation is also enclosed.

The notice of'ssuance will be incl'uded in the Commission's next monthly Federal Register notice.,

Si ly,

Enclosures:

1'.

Amendment No.

94. to License No.

DPR-33 2.

Amendment No. 87 to License No.

DPR-52 3.

Amendment No.

60 to License No.

DPR-68 4.

Safety Evaluation cc w/enclosures:

See next page Cl k,, Projdct anager perating Reactors Branch P2 ivision of Licensing

llew

Mr. Hugh G. Parris Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2 and 3

CC:

H. S. Sanger,, Jr., Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E 11B 330 Knoxville, Tennessee

37902, Mr. Ron Rogers Tennessee Valley Authority 400 Chestnut Street, Tower II Chattanooga, Tennessee 37401 Mr. Charles R. Christopher

Chairman, Limestone County Commission Post Office Box 188

Athens, Alabama 35611 U. S. Environmental Protection Agency Region IV Office Regional Radiation Representative 345 Courtland Street, N.

W.

Atlanta, Georgia 30308 Resident Inspector U. S. Nuclear Regulatory Commission Route 2, Box 311

Athens, Alabama 35611 Mr. Oonald L. Williams, Jr.

Tennessee Valley Authority 400 West. Summit Hil7 Orive,. W10885 Knoxvikle, Tennessee 37902 Ira. L., Myers,. M. O..

State Health Officer State Oepartment of Public Health State Office Building Montgomery,

Alabama, 36130 Mr. H. N. Culver 249A HBO 400 Commerce Avenue Tennessee Valley Authority Knoxville, Tennessee 37902 George Jones Tennessee Valley Authority Post. Office Box 2000 Oecatur,,

Alabama 35602 Mr. Oliver Havens U., S. Nuclear Regulatory Commission Reactor Training Center Osborne Office Center, Suite 200 Chattanooga, Tennessee 37411 James P. O'Reilly Regional Administrator Region II Office U., S. Nuclear Regulatory Commission 1Q1 Marietta Street, Suite 3100 Atlanta, Georgia 30303

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROMNS FERRY NUCLEAR PLANT', UNIT 1 AMENDMENT, TO FACILITY OPERATING LICENSE Amendment No. 94 License No.

DPR-33 1.

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

A'.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 25, 1983, complies 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 application, 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 conducte'd in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security nr to the health and safety of the public; and E.

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

2.

Accordinaly, the license is amended by chanaes to the Technical Specifications as indicated in the attachment to this license amendment and.paragraph 2.C(2) of Facility Operatina License No.

DPR-33 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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This license amendment is effective a's of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes. to the Technical Speci'fications

. >-- Domenic B. Vassallo, Chief Operating Reactors Branch 82 Division of Licensing Date of Issuance:

December 27, 1983

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ATTACHMENT TO LICENSE AMENDMENT NO.

94 FACILITY OPERATING LICENSE NO.

DPR-33 DOCKET NO. 50-259 Revise Appendix A as follows:

1.

Remove the following pages and replace with-identically numbered pages.

180 181 219 2.

The marginal lines on these pages denote the area being changed.

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HDUNDAAY tt ~ 6 PRTHAB Y SY~ivee BOUHDXAY C.

Coolant Leakage C.

Coolant Leakage k

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t '- >. Any'ttte irradiated fuel is in the ream"ar: vessel and reactor coolant tetnpe acure is. above; 2 12etP',

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coolant. leakage inta the-primary containment I=am unidenti "ied scu=ces shaH nct, exceed 5

grimm Xn addi the total.

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coolant system leakaqe inta the ttrimary containment shall not exceed 25 gpm.,

b. Anytime, the reactor, is in RUN made, reactor coolanr.

leakage into the p&~

concafnment irom, unident~~~ied sources shall not increase by more: than, 2 gpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in a&4h the reactor

's in the ROM mode except as dezMed in 3.&.C.'l.c b~t. Drywell leakage shall be measured and recorded every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

20 Reac cr caolant s gstem leakacte shalJ.

be checked by the suma and ai samalincr system and record at Least once ger.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

With the a'amplinq svst m inocerable, q ah samales shall be obtained and analyted at leas once; eve y 2'n hcursm.

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c. During the i=-st 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 'z the, iEH mde-follotting start<~,

an increase in reactor coolant leakage &to the p.~ry containmenr, oi 02 gpm dm acceatabie as lang as the reauirements oi 3.&.C.l.a are met.

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I:M~TIiTG COhhDITIONS FOR OPZBATION SURVEILMC-REQUZR~MTTS 3.&.C Coolant Leakage 4.6.C Coolant Leakage 2.

3e Boch the sump and ai sampling systems, shaLL be operable during reactor power operation.

From and airer the date that one of these systems is made or-found to be inoperable for any reason, reactor paver operation is permissible only duW~g the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The air sampling system may be removed E~ service Eor a.

period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Eor cali-

bration, Eunction testing, and maintenance without providing a temporary monitor.

1'f the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the reactor sha11 be shut-down in the Cold Condicion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Relief Valves

1. ApprtocLmately one-half of all relief valves shall be bench-.

checked or replaced with a beach-checked valve each.

operat~~g cycle.

All 13 valves v~ have been checked.

or teolac d coon the comole-t=on. oi every second cyc'e.

Z. Once during each operat'ag cyc'e, each raliei valve shaLL be manua13.y opened uncs the~couples and acoust'c monitors downstream of the valve indicate steam is Eloving E~ the valve.

3. The integri~ of tNe rel'eil safety valve beLLovs shaLL be cont~~uously monitored.

wo Je Relief'alves

'~hen aura than one relief valves are known. to be fa&ed, an orderly shutdown shall.be initiat-ed and the reactor depressu~ "ed to Less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Jee Porte Wherever the reootor te ie the startup or run modes, alL jet pumps shall'be operabLe.

Zf it is determined that a jec pump is inoperable, or U tvo or more jet pump Elov inst~t failures occur and cannot, be cor ected within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an orderly shutdovn sha11 be inf.t~=

ated and the reactor shal1 be shutdown in the Cold Condition wiJoin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

oe e 4.

At, lease one reliei valve shall be disassembled and ~pected each operating cycle.

Jet Pumos

1. 'whenever. there is tec'rculat'on E'ov with the reactor in the startup or tun modes virh both tec'ulation pumps tunnince jet puma operability shall be checked dailv by ver'fv'ng that the following conditions do not occur simultaneous lv:

a.

.he tvo recirculation loons have a Elov ~moalance oi

,"; or =ore when

=he

~u.=os are operated at the same speed Amendment No. 9g, 94 1S1

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detected reasonably Ln a matte.

of fee hours ut'Ll= ng the availabL Leakage detection

scnemes, ancl if the.ar gm cannot be. determined

.in a z'easanabLy snort'ime che unit should 0e shut dawn to al'aw.uc.her Cnvesti~ticn and corrective act'cn.

The 2 gpm'Xmf.c for cooLant. Le~>a~c raze 'nc enon over cxy 2~ hou per'oci is. a Limit specified by che 1RC (Reference 2.)

This LMc.

applies only dung che 3LPil mode to avoid being penaI'red for the expected coolant leakage inc ease du 'ng pressur~ "ation.

The total Lea}cage-race con iats af aLL, Leakage, ident'fied and un'dent'f'ed, which flcws to the d.-twell f'aac-d~in and equipment drain sumps.

The caoacity of the dryweLL f'cor sump puma is 50 gym anrt

".e caoacity af tne dryMeLL equ'~en mg pump is also 50 ppm Removal, af Z5 gpm fram either af these sumac can be accomplished with considerable ~a 1..

Nuclear System Leakage Rate ~its (BEBOP PS'ubsec"ion 4.10) 2.

Saf ety Evaluation Repor (SM) an W Bullets~~

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3. 6. D/4. 6. D Rel'8 Valves To meec the sniecy basis, chirccen relief valves have been i<<stall@el o<<chv <<nit with a tota1 i'apace ty nf'4.IX of n<<clear boiler rated sceam flow ac. a reference pressure oi (J.'105 + 12) pain.

Tho analysis ol th<<. worse.. <<vvrprussuri transient. (3-siwond closure of a11. main stcam Linc Lsolatfon vniv<<s), <<<<1;louring cho direct scram (valve position scram),

results, Ln a maximum vessc1. yrussur<<which, C! a <<outran flux scram.

Ls assumed considering L2 valves operable, resuLcs i<<adequacc margin to chc cudu aLLowable overpc'cssura 1~it of 1375 paly,.

To meet opvrucio<<al design, che analysis of che pi.anc Molacion,transient (generacor load re)ecc wich bypass valve iailure co open) shows that. 12 oi the 13 rel'ei valves limic peak system pressure co a value. which is weil below the allowed vessel overpressurc oi 1375 psig.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-260 BROMNS FERRY NUCLEAR PLANT UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE 0

Amendment No. 87 License No.

DPR-52 1.

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

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated March 25, 1983, complies 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 application, 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 regulations; 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.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No.

DPR-52 is hereby amended to read as follows:

(2)

Technical S ecifications The. Technical Specifications congained in Appendices A and B,

as revised through Amendment No. 87, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications I

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Domenic B. Vassallo, Chief Operating Reactors Branch k'2 Division of Licensing Date of Issuance:

December-27, 1983'

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ATTACHMENT TO-LICENSE At1ENDMENT N0.87 FACILITY OPERATING LICENSE NO.

DPR-52 DOCKET'O. 50-260 Revise Appendix A'. as follows:

1.,

Remove the following pages and repl'ace with identically numbered pages.

180 181.

219 2.

The marginal lines on these pages denote the area being changed.

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'Nj>> INC CQNO i ONS Poa OPERA>> TON SNV~+WHV'rran: 72Ci iS

1. 6 pzmaa Y s.~

scDNDan Y 5.

PRIMARY SY=T~ BOUNDARY, C

Coolant Leakage

l. a. any ime irradiated fuel is in the etc or vessel and reac or coolant.

ttmpe ature is above 2t2~%',,

eamr.

coolant leakage into the yrmarv containment cm uniden -'ed sou=ces sha3~ n~ exceed'.

@pe, Zn adMtion~

the nota1 rea~r coolant system leakage into the pr mary containmenc shall not exceed 25 gpm.

C, Cool on t Leakage Reac or coolant s fs em Leakage shall he checked hy the suma and al

. samolinc system'nd cco<<

M>>i't Least once ger.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Aith the sir samaiinq.

system inoperahle, qrah samoles snail he oh oined and analy" d at Leas once every 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />s~:

b. Anytime the reactor is in KH,mde, reactor coolant leakage

~mto the pr~

contafnrnmt E~ unident~ ied sources shIllnot inc Ease'y more than. 2 gpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per<<od, in which the reac"o.

is in che iHJN mode except as dezined in 3.6.C.L.c belo~.

Dzywe3.l leakage shall be measured and recorded every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Ou"ag the.: -st 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

-he i'ode following star un, an increase

~w "tactor coolant leakage

~+to the pr~zy conta~ent, oz

)2 gran is acceotaola as long as the reaui-ements af 3.6.C'.L.a are met.

Amendment No. gg, 87 I nn

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0 IMZ DIG CORDI-ZQNS FOR OPERA XCH SQ VTEZLLZsNC= RZQUZ~~FZS 3.6.C Coolant Leakacce 4.6.C Coolant E.eakaae 3

3oth the s~

and ai

. sampling systems shall be operab3.e during zeactoz power opezation.

From and aiter the date that, one oi these systems M made or found.

co be inoperable, for. any. reason, reactor power operat~wn-is permissible only during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />..

T~ he a'- sampling system may be-removed E~ service Eoz a period, oi 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Eor caL'-

bration, Eunc ion tasting, and.

aaintenance without pzav<<A<<mg a temporary monitor.

if the candit~4 in. l oz 2 above cannoc be met, an ozderLy shutdown shall be init~mted and the.reactor shaLL', be shut-down in the Cold. Condi~~

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Rel<<ei Va3.ves Appzo~tely one-ha3

~. or a

zecC.ef valves shall be bench checked az replaced wf.th a bench~hecked valve each operating cycLe.

AL1 L3 valves wi~'c have been checked oz replaced upon. the comple-tion oi every second cycle.

2. Once dur<<wg each operating cyc'e.

each zel'ei valve shaLL be aanua3'y opened unt<<M the~couples and acoustic monitors downstream oi the valve indicate steam

<<m flowing fram the va3.ve.

3. The ~mtegz ty oi the relief/

saiecy valve bellows shaLL. be cant<<~uausly monitared.

De l.

Relief Valves When care than. one relief, valves aze. known to be failed, an order3y shutdcfnx shalL be in't~mt>>

ed and the zeac ar depzessur<<ed

.ta Less. ~

L05 psig wi"bin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Jee Peeee L. tiheeerer rhe reeeeer re re rhe starzup or ~ mades, all jet pumps shalL be operable.

Zf it is determined that. a jet pump is inoperable, oz

<<D two or sore jet pump Elow Lnstnment failures occur and cannot be car.ec ed within l2 hours, an order3.y shutdown shaL3. be.m aced and the reac az shaLL be shutdown Xn the Col'd Condit"on within 24 haurs.

we

4. Ac least one ze" ei valve sha3.'e disassembled and <<~peered each operating cyc'e.

Jet Pumps I.. Wenevez there is zec'zculat'an E'aw with the reactor in the star up oz rm modes with both zecirculat'on pumps running.

jet pump opezabf'ty sha3.'e checked dai'y by verify"'ag "hat

he following canc't'ons do co" oc"ur sim3.taneously:

a.

The two rec'zc'.at-cn

'oops have a."aw <<maLance or

.'5Z or =are cchen

.".e ~uncs aze operated at

=he same speed.

LSL Amendment No. 35, 46, 68, 85, Ri 8@mucM~

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detected rea onably

'.n a catter of f'ev hours. ut'l'= n-he availabl.

leakage detect'cn

schemes, and, if-the or ig1n cannot be.deters'ned fh ~

geasonably shor me "he un't should be shut Covn to allow.ur he>

lnveszigat'on and correct've action.

.he 2 gpm Mr. ior coolant le~>a~c raze inc=anno ovcz my Zc hour peri.od is

a. Limi.c spec~""ied by zhe ~C (Rezezence Z.)

Th~ l'-'"

apolies only d~g, the %Pi mode to avoid, being. penal'-ed. =or the expected coolanr. Leakage

'=c=ease du='ng pzessu~ "ation.

The total leakage rate consists ot ~~

leakage,

'.dent'fied and mf.cent'f'ed, vhich ~ovs to Ne C. pell Poor Min and cpu'~ent dram swaps.

.he caoacity of the C~l'oor sump cubo 's 50 gym abc

=Me cacnchty of the CryMeU. equ'~ 'c=p puuO 's also 50 @pe.

Removal. of 25 ~."~ ei her of'hese cunns can be accccO3.ished vith considerable m II~EHC-Hucleaz System Leakage Race ~4'ts (SHARP FSM Subsec=-'on 4.10) 2.

Safacy Eva1uaa~

Ra@are (SM) oa ZE Sel~ac~~

82-03

3. 6.D/4. 6.D Relief Valves To neet the safety basis, th'rteen re" ef valves have been insmlled on the unit with a total capacity of 84.J~+ of nuclear boiler rated steam flow.

The analysis of the worst overpressure transient

(>-second closure of all main s earn Line isolation valves) neglecting the direct sc~

(valve positicn scram) results in a ~um vessel pressure wh'ch, if a neut. on flux scram is a sumed considering

$ 2 valves operable, results in

-de"uate na."gin to the code M~o able overpressure " 't of 1375 psig ~

To neet operational design, the analysis of the plant isolation trmsient (generator load r feet with. byp ss valve failure to open) shows that

$ 2 of the 13 relief valves l' peak system pressure to a value which is well below the allowed vessel overpressu."e of 1375 psig.

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment,No.

60 License No. DPR-68 1..

The Nuclear Regulatory Commission (the Commission) has ound tha+:

A..

The application for amendment by Tennessee Valley Authority (the licensee) dated March 25, 1983, complies 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 application, 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 sich activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and pecurity or to the health and safety of the public; and E.

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

2.

Accordingly, the. license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No.

DPR-68 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and 8,.

as revised through Amendment No. 60, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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0 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications

~-'Domenic B. Yassallo, Chief Operating Reactors Branch ¹2 Division of Licensing Date of Issuance:

December 27, 1983

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ATTACHMENT TO LICENSE AMENDt1ENT NO.. 60 FACILITY OPERATING LICENSE NO.

DPR-68 DOCKET NO. 50-296 Revise Appendix A as follows:

1.

Remove the following paces and replace with identically numbered pages.

191 192 224 2.

The marginal lines on these pages denote the area being changed.

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t.LHt a I'Nc COHOI TQHS FOB QPXSA'T[QN suavxz~Mc-nz~c 7 LENT 5

PRIMARY SYS,

.I aOUNOARY 4 ~ 6 PRZ!WARY SY~K.'~

eOONOA RY C

cool. ant Leakage c'calant Leakage oAny t~e. irradiat.ed fue3. is in the reac=or vessel.

and reactor coolant temperature is-above 212~P, reactar caalan leakage '-. a the pr~4ary containment am, unidentified, sau ces shalL. not exceed 5

gpm-In additian, the total reactor car@ant system 3.eakage inta the primary containment.

shaLL not exceed 25

'gpm o 2 0 Reactor caol.ant system leakage shalL be checked by sump and az samal~g system and ecarded at least, once pe 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

MXth the air samal'ng system inaperacle, grah samales shall be obtained and analyted at 3.east once every

'24'ours.

b. Anytime the reactor is in MPiT mode, reactor coolant leakage; into the p&~ry containment i am, unident~

ed sources,shall not increase by more than

2. cpm averaged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pezfad in which the zeac or is in the RUN cade except as defamed in 3.6.C.l.c below.

Oryxe11 leakage.

sha11 hh rgeasured and recorded eyery-8 hours.

c.

Du Mg the-B=st 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the RUiV made iollavtng

staztup, an inc ease in reactor coolant leakage into the primary containment oi

>2

!cpm is acceptab3.e as

'ons as the reauirements ai 3. 6.C. l. a are met.

Amendment No. gP, 60

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LXMTMG CONDX XONS FOR OPZEATZON SURVEXLLANC= R:"QU ~~~S

3. 6 PRMJLRY SY~

SOUNDARY 4.6 PRW~Y SY~c BOUNDARY 2.

3ath the sump and ai= sampL'ag sampling systems shall be operable during reactor power operat"on.

ram and after the date rhat one af these systems is made ar found to be

~moperable for any reasan, reactor power operat"on

~m permissibLe only duz4mg the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The air samp~4g system may be removed from service ior a.

period o.f 4.hours for calibra-tion, functional testing, and mainr.enance without providing a temporary monitor.

3.

Lf the condition in 1 or 2 above cannot be met, an orderly shutdown shalL be initiated and the. reactor shall be shutdown the Cold Candit'an within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

Relief Valves L. Approximately one-haLf oi all relief valves sha11.

be bench~ecked or raolaced with a beach-checked

vaLve, each operating cycle.

Ail L3 valves wi'1 have been checked or reolaced.

upon the complec'an oi every second cycle..

Z. Once during each aperat'ng

cycle, each rellei valve shall be manually opened uncs thermocouples and.

acaust" c mon'tors down-.

stream of,the valve mdicate steam

~m flawing i~

the'alve.

D:

Relief Valves When more than ane relief valve Ls known to be failed, an orderly shutdown shall be initiated and the reactor depressuriaed to less than L05 psig'ithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

At least one teliei valve shalL be disassembled and Lnspected each operatina cycle.

Amendmen~

No.

O'J, 60 102

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3.6/4 6

SASZS limit specified for unidentified leakage, the probabi3.ity is small that imoerfections or cracks associated with such leakage would g.ow rapidly.

However, the establishment of allowab3.o unidentified leakage greater than that given in 3 6.C on. the basis of the data presently available would be premature because of uncertaint'es associated with the data.

For leakage of the order of 5~,

as spec'~ied in 3.6.C, the exper~tal and analytical data suggest a reasonable mar in of safety that such leakage magnitude would not result from a c=ack approaching the critical si=e for rapid propagat'on.

Leakage less. than the magnitude spec'fied can be detected reasonably in a matter oi few hours utQi=w~ g the av ~'ab3.e leakage detect'on

schemes, and if the origin,cannot be determined

~m a reasonably short t~e the unit should, be shut'down to allow further investigation and corrective action.

The 2 gpm 3.~t for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a lent specified by the NRC (Reference 2).

This 1&.t applies only during the HUN mode to avoid being penali d for the expected coolant leakage inc ease dur~mg pressuri"at'on.

The total leakage rate consists of all leakage,

'dentif 'ed and unidentified, which flows to the drywell floor drain and equipment drain sumps.

The capacity of the drywaU. f1oor sump pump 's 50 gpm and the capacj.~

of the drywell equipment sump'pump is also 50 gpm.

Removal of 25 gpm f"om either of these sumps can. be accompMhed with considerable mar~~w.

RE.=~HCZS L.

lfuclear System Leakage Hate Limits (3FAP FSAK Subsection 4.10)

Safety Evaluation Report (SWr) on TE Bulletin 82-03 Amendment No. PJ, 60 224 y I-a<EY

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UNiTED STATES AIUCLEAR R EG ULATORY COMMiSSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 94'O FACILITY OPERATING LICENSE NO.

DPR-33 AMENDMENT NO. 87 TO FACILITY OPERATING LICENSE NO. DPR-52 AMENDMENT NO.

60 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY'UTHORITY

. BROWNS FERRY. NUCLEAR PLANT, UNIT NOS; 1, 2 AND 3 DOCKET NOS. 50'-259, 50-260 AND 50-296 1.0 I'ntroduction By letter dated March 25, 1983 (TVA BFNP TS 186) the Tennessee Valley Authority (the licensee or TVA) requested amendments. to Facility Operating License Nos.. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Unit Nos. 1, 2 and 3.,

The application by TVA was in response to a reauest by the NRC staff on March 11',, 1983 to provide revised Technical Specifications for Browns Ferry Unit 2 with more stringent requirements on unidentified 1'eakage in the drywell.

The requested changes were the same as those in the BWR Standard Technical Specifications.

During-the refueling and modification outage of Unit 2,. which extended from July 30, 1982 to March 18, 1983, TVA found indications of cracks in two of the ten sweep-o-1'et to manifold welds in the recirculation system.

TVA proposed to opera e in Cycle 5 with these two indications.

We performed an independent materials and fracture mechanics. evaluation and concluded that operation throughout the next cycle with these indications was acceptable but that certain additional. compensatory measures were warranted, such as more stringent requirements, on unidentified leakage.

2.0 Discussion The staff requested TVA to submit a change to the Technical Specifications limiting the rate of increase for unidentified drywell leakage to 2 gallons per minute (gpm) in a 24-hour period'.

This is the same requirement

~hat is in the BWR Standard Technical Specifications (NUREG-0123, Rev. 3).

The requested change was submitted by TVA's letter of March 25, 1983.

On March 4, 1983, IE Bulletin 83-02 was, issued requiring augmented inservice inspection of recirculation and residual

%eat removal system piping for BWRs shutting down for refueling after February 1983.

Inspections performed in accordance with this Bulletin rev'ealed indications of pipe cracks in most facilities.

As a result, the staff concluded that even more enhanced surveillance of possible leakage was warranted.

Specifically, the staff proposed that the frequency for checking the leakage rate be increased once per day to once per shift and that the allowable period for plant operation without the leakage monitoring sysiems

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in operation be reduced from 7 to 3 days.

TVA subseavently imposed these limits. administratively on all three Brnwns Ferry units. Similar limits on unidentified leakage have been incorporated in the Technical Specifications for numerous BWRs during the past year.

The NRC staff proposed these additional surveillance requirements, to TVA as a supplement to the triarch 25, 1983 submittal.

The.additional changes were accepted by the TVA staff, and,, as noted above, were administratively imposed voluntarilv by TVA.

Thus,. the changes to the Technical Specifications encompassed by these amendments are already in effect.

However, since Browns Ferry Unit 1 will be returning to power in C'vcle 6 in the near. future with nine unrepaired

welds, the staff has determined that the chanqes should be incorporated in the Technical Specifications by amendments.

3.0 Environmental'onsiderations We have determined that these amendments do not authorize a change in effluent types or total'mounts. nor an increase in power level and will not result in any significant environmental impact.

Having made this determina-tion, we have further concluded that these amendments involve an action which is insignificant from the standpoint of environmental

impact, and pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these"amendments.

4.0 Conclusion We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safetv of'he 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 these amendments will not be inimical to the common defense and security or to the health and safety of'he public.

Principal Contributors:

W. Hazelton, W., Koo and R. Clark Oated:

Oecember 27, 1983.'

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