ML20113B185

Application for Amends to Licenses DPR-33,DPR-52 & DPR-68, Revising SLMCPR & Bases Description of RHR Removal Suppl Fuel Pool Cooling Mode

ML20113B185
Person / Time
Site:	Browns Ferry
Issue date:	06/21/1996
From:	Salas P TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20113B188	List: ML20113B185 ML20113B195 ML20113B197
References
TVA-BFN-TS-377, NUDOCS 9606260236
Download: ML20113B185 (13)

Text

,

l HM  ;

Tennessee Valley Authority Post Office Box 2000. Decatur, Alabama 35609 June 21, 1996 TVA-BFN-TS-377 10 CFR 50.4 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document' Control Desk Washington, D.C. 20555

{

l Gentlemen: '

In the Matter of ) Docket Nos. 50-259 i Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 -

TECHNICAL SPECIVICATION (TS) 377 - CHANGE IN SAFETY LIMIT MINIMUM CRITICAL POWER RATIO (SLMCPR) AND REVISION TO BASES DESCRIPTION OF RESIDUAL HEAT REMOVAL SUPPLEMENTAL FUEL POOL i COOLING MODE  !

In accordance with the provisions of 10 CFR 50.4 and 50.90, TVA is submitting a request for an amendment (TS-377) to licenses DPR-33, DPR-52 and DPR-68 to change the TSs for Units 1, 2, and 3. The proposed change revises the SLMCPR.

As discussed in the General Electric 10 CFR Part 21 notification that was submitted to NRC on May 24, 1996, it was determined that the generic calculated SLMCPR may be non-conservative when applied to some actual core and fuel designs. A Unit 2 Cycle 9 specific calculation has shown that the SLMCPR:is non-conservative with respect to Technical Specification 1.1.A.1. A Unit 3 Cycle 7 specific calculation has shown that the current SLMCPR is valid. Therefore, the Unit 2 calculation has been used as the basis to provide a new bounding SLMCPR for all three BFN units until the long-term resolution of this issue is implemented. In addition, the description of the Residual Heat Removal supplemental fuel pool cooling mode in Bases Section 3.10.C is being revised in order to resolve a discrepancy identified by the NRC Project Manager during a review of spent fuel cooling and core offload practices.

. Prior to startup for Unit 2 Cycle 9, TVA revised the administrative Operating Limit Minimum Critical Power Ratio 4

based on preliminary information from General Electric to bound any potential non-conservatisms in the generic SLMCPR 9606260236 960621 a3 /

PDR P

ADOCK 05000259 PDR hjb j I

l U.S. Nuclear Regulatory Commission Page 2 June 21, 1996 calculation. This operating limit penalty ensured that the SLMCPR would not have been exceeded for any analyzed transient during the current operating cycle.

TVA has determined that there are no significant hazards considerations associated with the proposed change and that the change is exempt from environmental review pursuant to the provisions of 10 CFR 51.22 (c) (9) . The BFN Plant Operations Review Committee and the BFN Huclear Safety Review Board have reviewed this proposed change and determined that-operation of BFN Units 1, 2, and 3 in accordance with the l

, proposed change;will not endanger the health and safety of l l the public. Additionally,'in accordance with i

10 CFR 50.91(b) (1), TVA is sending a copy of this letter and enclosures tc. the Alabama State Department of ' Public Health.

l l Enclosure 1 to this letter provides the description and i evaluation of the proposed change. This includes TVA's i

determination that the proposed change does not involve a significant hazards consideration, and is exempt from i environmental review. Enclosure 2 contains copies of the

appropriate TS pages.from Unit" 1, 2, and 3 marked-up to show the proposed change. Enclosure 3 forwards the revised TS-

! pages for Units 1, 2, and 3 which incorporate the proposed

! change. Enclosure 4 contains the revised Unit 2' Cycle 9 Core j operating Limits Report, which is required by Technical i Specification 6.9.1.7.d to be submitted to NRC. Enclosure 5 4

contains copieslof the Units 1, 2 and 3 Bases Section 3.10.C i pages, which have been marked-up to denote that the Residual j Heat Removal system is a means of providing additional fuel ,

pool decay heat' removal, but its use is not required. i Enclosure 6 forwards the revised Bases pages. l-l l TVA requests that the revised TS be made offective for l l Units 1 and 2 within 30 days of NMC approval. The Unit 3  ;

! Cycle 7 specific calculation has shown that the current

! SLMCPR is valid. Therefore, TVA requests the TS be made f

i

)

i l

1 I

U.S. Nuclear Regulatory Commission Page 3 June 21, 1996 effective for Unit 3 during the cycle 7 outage, which is currently scheduled to begin on February 21, 1997. If you have any questions about this change, please contact me at (205) 729-2636.

Sincy ezy, ,

Pedro Salas Manager of Site Licensing Enclosures cc: see page 4 Subscribed and sworn to before me on this Olot day of 3 M t. 1996.

boa Notary Public R.RmL My Commission Expires N '

i l

U.S. Nuclear Regulatory Commission Page 4 June 21, 1996 Enclosures

, cc (Enclosures):

Mr. Johnny Black, Chairman Limestone County Commission 310 West Washington? Street '

Athens, Alabama 35611 Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission l Region II l 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, Box 637 Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager l

U.S. Nuclear Regulatory Commission  :

One White Flint, North 11555 Rockville Pike ,

Rockville,' Maryland 20852 l,

Dr. Donald E. Williamson l State Health Officer Alabama State Department of Public Health 434 Monroe Street i Montgomery, Alabama 36130-3017

1 ENCLOSURE 1 TENNES8EE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) l l

UNITS 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE T8-377 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE I. DESCRIPTION OF THE PROPOSED CHA3GE TVA is proposing a change in the Safety Limit Minimum Critical Power Ratio (SLMCPR). The specific changes are i described below: l

1. Units 1, 2 and 3 TS pages 1.1/2.1-1, Safety Limit 1.1.A.1, currently reads:

"When the reactor pressure is greater than 800 psia, l the existence of a minimum critical power I ratio (MCPR) less than 1.07 shall constitute violation of the fuel cladding integrity safety i limit." l l

The proposed safety limit reads:

"When the reactor pressure is greater than 800 psia,

the existence of a minimum critical power j

ratio (MCPR) less than 1.10 shall constitute violation of the fuel cladding integrity safety limit."

l

2. Units 1, 2 and 3 TS pages 1.1/2.1-8, Section 1.1, Bases: Fuel Cladding Ir.tegrity Safety Limit, currently j reads:

! "Because fuel damage is not directly observable, the i Fuel Cladding Safety Limit is defined with margin to j the conditions which would produce onset transition j boiling (MCPR of 1.0). This establishes a safety j Limit such that the minimum critical power ration'(MCPR) is no less than 1.07. MCPR > 1.07

represents a conservative margin relative to the j conditions required to maintain fuel cladding

. integrity."

t The revised Bases state:

"Because fuel damage is not directly observable, the Fuel Cladding Safety Limit is defined with margin to the conditions which would produce onset transition boiling (MCPR of 1.0). Maintaining the MCPR greater than the Safety Limit MCPR represents a conservative margin relative to the conditions required to maintain fuel cladding integrity."

._ .m_ _ _ . . _ . . . _ _ _ . . . . _ . _ _ . _ . - . _ _ _ _ _ _ . _ _ _ . . _ _ _ _ _

i

3. Units 1, 2 and 3 TS pages 1.1/2.1-8, Section 1.1, i Bases: Fuel Cladding Integrity Safety Limit, currently  ;

{ reads: 1 1

} "The Safety Limit (MCPR of 1.07) has sufficient i

conservatism to assure that in the event of an

, abnormal operational transient initiated from a

! normal' operating condition (MCPR > limits specified ,

j in Specification 3.5.K) more than 99.9 percent of l i the fu'el rods in the core are expected to avoid l boiling transition. The margin between MCPR of 1.0 (onset'of transition boiling ) and the safety limit i 1.07 is derived from a detailed statistical analysis )

4 considering all of the uncertainties in monitoring 1

the core operating state including uncertainty in  !

i the boiling transition correlation as described in

Reference 1."

i l The revised Bases state: '

"The Safety Limit has sufficient conservatism to 1 assure that in the event of an abnormal operational 1

transient initiated from a normal operating j j condition (MCPR > limits specified in' l

1 Specification 3.5.K) more than 99.9 percent of the  !

j fuel rods in the core are expected to t avoid boiling i transition. The margin between MCPR of 1.0 (onset j of transition boiling ) and the Safety Limit MCPR is derived from a detailed statistical analysis '

considering all of the uncertainties in monitoring the core operating state including uncertainty in

, the boiling transition correlation as described in i Reference 1."

4. Units 1, 2 and 3 TS pages 1.1/2.1-9, Section 1.1, i Bases: Fuel Cladding Integrity Safety Limi'c, currently l reads: ,

t

! "Because the boiling transition correlation is. based  ;

on a large quantity of full scale data there is a very h'gh i confidence that operation of a fuel assembly at the condition of MCPR = 1.07 would not produce boiling transition."  ;

The revised Bases state:

"Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR equal to the Safety' Limit MCPR would not produce boiling transition."

El-2

i

, 5. Units 1, 2 and 3 TS pages 1.1/2.1-13, Section 2.1.A.1, Bases: Neutron Flux Scram - APRM Flow-Biased High Flux Scram Trip Setting (RUN Mode), currently reads:

" Analyses of the limiting transients show that no scram ' adjustment is required to assure MCPR > 1.07 l when the transient is initiated from MCPR limits specified in specification 3.5.k."

i The revised Bases state:

i " Analyses of the limiting transients show that no scram adjustment is required to assure MCPR is

. greate'r than the Safety Limit MCPR when the ,

transient is initiated from MCPR limits specified in '

4 Specification 3.5.k."

l l

6. Units 1, 2 and 3 TS pages 1.1/2.1-14, Section 2.1.A.4, Bases: Neutron Flux Scram - IRM Flux Scram Trip l

Setting, currently reads *

)

! "The results of this analysis show that the reactor l j is scrammed and peak power limited to one percent of I i rated power, thus maintaining MCPR above 1.07."

The revised Bases state:

l "The results of this analysis show that the reactor j is scrammed and peak power limited to' one percent of

~

rated power, thus maintaining MCPR above the Safety Limit MCPR." j l

7. Units 1, 2 and 3 TS pages 1.1/2.1-15, Section 2.1.C, i Bases: Neutron Flux Scram - Reactor Water Low Level i Scram and Isolation (Except Main Steam Lines), currently j reads:

1 "The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than 1.07 in all cases, and system pressure does not reach the safety valve settings."

The revised Bases state; "The results reported in FSAR Subsection 14.5 show that scram and isolation of all process lines (except main steam) at this level adequately protects the fuel and the pressure barrier, because MCPR is greater than the Safety Limit MCPR in all cases, and system pressure does not reach the safety valve settings."

El-3

_ . .. . >_ _ _ . _ . . _ _ _ . _ _ _ - _ ~ . _ . - - . . _ _ _ . _ _ _ . _ _ _ . . - _ _

4 s

8. Units 1, 2 and 3 TS pages 3.3/4.3-17, Section C, Bases

Scram Insertion Times, currently reads:

! "The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent

~

fuel damage; i.e., to prevent the MCPR from becoming less than 1.07. The limiting power I

transients are given in Reference 1. Analysis of i these transients shows that the negative reactivity l

rates resulting from the scram with the average 1 l respon'se of all drives as given in the above I

! specifications provide the required protection and '

] MCPR remains greater than 1.07."

The revised Bases state:

"The control rod system is designed to bring the l

reactor.subcritical at a rate fast enough to prevent ,

fuel damage; i.e., to prevent the MCPR from l l becoming less than the Safety Limit MCPR. The l l limiting power transients are given in Reference 1.

Analysis of these transients.shows that the negative reactivity rates resulting from the scram with the j-

~

average response of all drives as given in the above specifications provide the required protection and MCPR er' mains greater than the Safety Limit MCPR."

l i  ;

l II. REASON FOR THE PROPOSED CHANGE d

As discussed in the General Electric 10 CFR Part 21 l

notification that was submitted to NRC on May 24, 1996, it was determined that the generic calculated Safety Limit i Minimum Critidal Power Ratio (SLMCPR) may be i non-conservative when applied to some actual core and fuel l designs. A Unit 2 Cycle 9 specific calculation has shown j

that the SLMCPR is 1.09, which is non-conservative with respect to the 1.07 value specified in Technical Specification'1.1.A.1. A Unit 3 Cycle 7 specific

calculation has shown that the current SLMCPR of 1.07 is i valid. Therefore, the Unit 2 calculation has been used as i the basis to provide a new bounding SLMCPR of 1.10 for all three BFN units until the long-term resolution of this issue is implemented.

1 4

4 e

El-4 4

5 t

1 III. SAFETY ANALYSIS i

l The objective of the thermal and hydraulic design of the

core is to achieve power operation of the fuel over the life  ;

of the core without sustaini'ng fuel damage. The thermal i

! hydraulic design of the core provides the following j characteristics: 1

a. The ability to achieve rated core power output 3 throughout the design lifetime of the fuel without i

! sustaining fuel damage. '

I

! b. The flexibility to adjust core power output over the l .. range of' plant load and load maneuvering requirements 4

without sustaining fuel damage.

The thermal hydraulic design of the core also establishes limits for use in setting devices of the nuclear safety l- systems so thst no fuel damage occurs as a result of i abnormal operational transients. Fuel damage is defined for i design purposes as perforation of the cladding, which i permits release of fission products. In addition, the j thermal hydradlic design of the core establishes a thermal i hydraulic safety limit for use in evaluating the safety j margin relating the consequences of fuel barrier failure to j public safety.

j There are three different types of boiling heat transfer to water forced convection systems: nucleate boiling, j transition boiling, and film boiling. Nucleate boiling, at

! lower heat transfer rate, is an extremely efficient mode of heat transfer,' allowing large quantities of heat to be

{ transferred with a very small temperature rise at the heated

! wall. As heat transfer rate is increased the boiling heat

transfer surface alternates between film and nucleate

! boiling, leading to fluctuations in heated wall 4 temperatures. The point of departure from the nucleate l boiling region into the transition boiling region is called a the boiling transition. Transition boiling begins at the i critical power, and is characterized by fluctuations in j cladding surface temperature. Film boiling occurs at the

! highest heat transfer rates, it begins as transition boiling i comes to an end. ,

The thermal hydraulic design objectives discussed above are achieved, in part, by maintaining nucleate boiling and i avoiding a transition to film boiling. The figure of merit utilized for core monitoring with respect to this objective

is the critical power ratio (CPR). The CPR is the ratio of 1 the critical p,ower (bundle power at which the onset of j boiling transition occurs) to the operating bundle power. A j value of 1.0 for the CPR indicates that fuel rods in the I

bundle are at'the onset of transition from nucleate to film boiling while a value greater than 1.0 indicates margin to l.

the onset of transition boiling.

I El-5 i

The boiling transition does not necessarily correspond to the fuel damage threshold, especially in the high steam-quality range. Boiling transition is identified as the heat transfer rate below which clad overbeating does not occur. Damage would not actually occur until well into the film boiling regime. For example, during inpile tests, Zircaloy-clad uranium dioxide fuel was purposely operated at heat transfer' rates well into film boiling for a total time exceeding 5 minutes, then operated at typical boiling water reactor conditions.for 10 days. Post-irradiation examination showed evidence of overheating but no cladding failure. To ensure good performance and long life of the cladding, conservative limits have been established to.

ensure that normal operations remain well below the transition boiling regime.

Thus, if nucleated boiling around the fuel cladding continues for an extended period, fuel cladding cracking could result due to a loss of cooling. Therefore, a minimum value of the critical power ratio (MCPR) is specified to maintain adequate margin to the onset of transition boiling.

The MCPR thermal limits are derived from the following design basis requirement:

Transients caused by a single operator error or equipment malfunction shall be limited so.that, considering. uncertainties in monitoring the core operating state, more than 99.9% of the fuel rods are expected to avoid boili'ng transition.

In summary, this design basis requirements is applied as shown below:

MCPR = 1.0 Fuel rods in the bundle are at the onset of transition from nucleate to film boiling.

Safety Limit Established to account for MCPR (SLMCPR) uncertainties in test data and core operating state.

Operating Established to account for margin Limit MCPR between normal steady state 1 (OLMCPR) operations and the worst case core fue1 transient.

In summary, the core is sized with sufficient coolant flow to assure that the MCPR is maintained greater than the operating limit at rated conditions. The fuel is designed so that fuel cladding cracking is not expected to occur during normal' steady-state' full power operation or abnormal transients. To ensure.that an adequate MCPR margin is maintained, a ' design requirement was established based on a

-statistical analysis that required 99.9% of the fuel rods

' avoid boiling' transition during anticipated operational occurrences. This requirement is satisfied by establishing El-G

l a safety limit MCPR based on a statistical analysis of the core near the limiting MCPR conditions and then setting an l OLMCPR for normal steady state operations. l The MCPR and the other core operating limits have been calculated for BFN Unit 2 using NRC methods approved in General Electric Licensing Topical Report NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel.

Recent calculations indicated that the generic analysis did l not always yi41d the most conservative result. Using cycle-specific procedures, a new safety limit MCPR was determined to be 1.09 for Browns Ferry: Unit 2. In order to provide for future changes, the proposed amendment increases the safety limit MCPR to'1.10 for all three units. l IV. WO SIGNIFICANT HAEARDS CONSIDERATION DETERMINATION j e  ;

TVA has concluded that operation of Browns Ferry Nuclear

Plant (BFN) Units 1, 2, and

3 in accordance with the l proposed change to the technical specifications does.not involve a significant hazards consideration. TVA's l conclusion is' based on its evaluation, in accordance with 10

CFR 50. 91(a) (1) , of the three standards set forth in 10 CFR l 50.92(c). '

A. The proposed amendment does not involve a sianificant increase ':Ln the nrobability or consecuences of an accident'oreviousiv evaluated.

The proposed change in the Safety Limit Minimum Critical Power Ratio (SLMCPR) does not increase the frequency of the precursors to design basis events or l operational transients analyzed in the Browns Ferry Final Safety Analysis Report. Therefore, the probability of an accident previously evaluated is not significantly increased.

The proposed change in the SLMCPR ensures that 99.9 percent of the fuel rods in the core are expected to avoid boiling transition during the most limiting anticipated operational occurrence, which is the design and licensing basis for the analysis of accidents and transients described in the Browns Ferry Updated Final Safety Analysis Report (UFSAR). It does not change the nuclear safety characteristics of any safety system or containment system. Therefore, the consequences of an accident,' operator error, or malfunction of equipment important to safety previously evaluated in the UFSAR has not een increased.

El-7 i

B. The cronosed amendment does not create the nossibility of a new~or different kind of accident from any accident nreviousiv evaluated.

The proposed change to the Technical Specification requirements for the safety limit minimum critical power ratio does not involve a modification to plant equipment. No new failure modes are introduced. There is no effect on the function of any plant system and no new system interactions are introduced by this change.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

C. The cronosed amendment does not involve a sianificant reduction in a marcin of safety.

The proposed change will ensure that during any anticipated operational transient, at least'99.9% of the fuel' rods would be expected to avoid boiling transition which is consistent with the licensing basis. Since the margin to safety is being increased with this change, the proposed amendment does not involve a reduction in a margin of safety.

V. ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a change in the types of, or increase in, the amounts of an? effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22 (c) (9) . Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of

, the proposed change is not required.

I i

i i

I El-8

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _