ML20140B929
| ML20140B929 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 06/02/1997 |
| From: | Abney T TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20140B934 | List: |
| References | |
| TVA-BFN-TS-387, NUDOCS 9706090009 | |
| Download: ML20140B929 (33) | |
Text
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Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 June 2, 1997 TVA-BFN-TS-387 10 CFR 50.4 10 CFR 50.90 10 CFR 50.91 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Wa.shington, D.C. 20555 Gentlemen:
In the Matter of-
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Docket Nos. 50-259 Tennessee Valley Authority
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50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 -
TECHNICAL SPECIFICATION (TS) 387 - SINGLE RECIRCULATION LOOP OPERATION (SLO)
In accordance with the provisions of 10 CFR 50.4 and 50.90, TVA is submitting a request for an amenament (TS-387) to licenses DPR-33, DPR-52, and DPR-60 to allow continued plant operation with a single reactor recirculation loop in service.
The current TS provide for only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of
/j operation in single-loop mode.
7 SLO is a recognized operating practice for Boiling Water Reactor (BWR) operation and many BWRs have TS which allow hgggg indefinite operation with a single recirculation pump in service.
The ability to onerate in single-loop mode provides operational flexibility ar.
allows continued power operation in the event of the loss of a recirculation loop due to component malfunction.
The majority of active components in the recirculation system are readily accessible and can be repaired with the reactor in service.
The NRC has previously determined SLO is generically acceptable as set forth in Generic Letter 86-09, Technical Resolution of Generic Issue B-59-(N-1) Loop Operation in BWRs and PWRs.
Also, SLO is recognized as a standard mode of operation in the BWR/4 Improved Standard TS (ISTS).
9706090009 970602 PDR ADOCK 05000259 P
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l U.S. Nuclear Regulatory Commission Page 2 June 2, 1997 l to this letter provides the description and safety analysis of the proposed change, and the significant hazards and environmental impact considerations.
General Electric analysis report, NEDO-24236, " Browns Ferry Nuclear Plants, Units 1, 2,
and 3, Single-Loop Operation, May 1981" is provided as Attachment 1 to Enclosure 1 and is the primary basis document for the proposed TS changes.
Attachnent 2 of is a copy of the cycle-specific reload analysis for the current Unit 3 operating cycle and is provided as an example of how SLO is addressed in individual core relcad analyses.
Enclosures 2 and 4 contain mark-up copies of the appropriate pages from the current Units 1, 2, and 3 custom TS,.and from j
the proposed ISTS change (TS-362) showing the revised TS.
TS-362 is BFN's conversion package to ISTS and was submitted to NRC on September 6, 1996.
Enclosures 3 and 5 forward the word processed custem TS and ISTS pages which incorporate the proposed changes. is a summary of the commitments made in this submittal.
TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the TS change qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).
The BFN Plant Operations Review Committee and the Nuclear Safety Review Board have reviewed J
this proposed change and determined that operation of BFN 1
Units 1, 2,
and 3 in accordance with the proposed change will not endanger the health and safety of the public.
Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health.
l A high priority is requested for the review of this TS to avoid reactor shutdowns in the event of a recirculation
.nystem problem.
If the TS is approved in custom format prior to the approval of the ISTS package, TS-362, then the SLO l-changes in this TS need to be incorporated into TS-362.
Otherwise, we request the TC changes ae issued with TS-362.
TVA requests that the revised TS be made effective within 30 days of NRC approval for all three BFN units.
Folloaing the approval of the SLO TS changes, the cycle-specific Core 1
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l U.S. Nuclear Regulatory Commission Page 3 June 2, 1997 Operating Limits Reports (COLRs) will be revised to provide i
adjustment factors for certain core thermal limits and instrument setpoints as stipulated in the proposed change.
For Units 2 and 3, the cycle-specific reload analyses for the current core loads which support the SLO COLR changes have been completed.
Since Unit 1 is currently in an indefinite nonoperational status, the cycle-specific reload analysis with SLO has not yet been completed.
This cycle-specific analysis will be performed prior to the restart of Unit 1.
If you have any questions about this change, please contact me at (205) 729-2636.
Sincerely,
/
T.
E. Abney Manager of nsing and Indus.ry Affaire Enclosures cc: see page 1' Subscribed and sworn to before me on is 2nd June
- 1997, day of dd& Y.
d # 1 r>< '
N6tary Public My Commission Expires N l
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i U.S. Nuclear Regulatory Commission Page 4 June 2, 1997 J
Enclosures
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cc (Enclosures)
Chairman Limestone County Commission j
310 West Washington Street Athens, Alabama 35611 1
Mr. Mark S. Lesser, Branch Chief U.S.
Nuclear Regulatory Commission Region II 1
61 Forsyth Street, S.W.
Suite 23T85 Atlanta, Georgia 30303 l
NRC Resident Inspector Brownc Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager 1
U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 Dr. Donald E. Williamson State Health Officer Alabama State Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-3017 i
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1 ENCLOSURE 1 l-l TENNESSEE VALLEY AUTHORITY l
BROWNS FERRY NUCLEAR PLANT (BFN) j UNITS 1, 2, and 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-387 DESCRIPTION AND EVALUATION OF THE PROPOSED CHANGE l
INDEX l
I.
DESCRIPTION OF THE PROPOSED CHANGE El-2 i
II.
REASON FOR THE PROPOSED CHANGE
...................El-13 1
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III. SAFETY ANALYSIS
..................................El-14 IV.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION
.....El-23 l
V.
ENVIRONMENTAL IMPACT CONSIDERATION
...............El-26 l
l VI.
REFERENCES
.......................................El-27 VII. ATTACHMENTS NEDO-24236, Browns Ferry Nuclear Plants, Units 1, 2, and 3, Single-Loop Operation l
l Supplemental Reload Licensing Report for Browns Ferry l
Nuclear Plant Unit 3, Reload 7 Cycle 8 I
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l I.
DESCRIPTION OF THE PROPOSED TS CHANGE TVA is requesting changes to Units 1, 2, and 3 TS to include provisions for continued reactor operation in i
i single recirculation loop operation (SLO) mode under certain specified conditions.
The current TS provide for J
only 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation in single-loop mode.
Provided below is a description of each requested change to the existing custom TS (CTS) and to Improved Standard 3
Technical Specifications (ISTS).
The changes in ISTS
)
format are based on NUREG-1433, Revision 1, " Standard i
Techt.ical Specifications for General Electric Boiling Water g
Reactors (BWR/4)," which includes provisions for SLO.
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. Enclosures 2 and 4 provide mark-up copies of the custom TS and ISTS pages, respectively., which show the proposed changes for SLO.
The mark-up pages in ISTS format are from BFN's proposed conversion package from custom TS to ISTS which was submitted to NRC as TS-362 on September 6, 1996.
Erclosures 3 and 5 are word processed copies of the TS changes in custom and ISTS format.
I PROPOSED CHANGES TO EXISTING CUSTOM TS 2
o Described below are the proposed changes to the current custom TS and the associated Bases changes to incorporate 4
provisions for SLO.
Page numbers are presented as Unit 1; Unit 2; Unit 3.
If the page numbers are the same for all three units, then a single page number is listed.
1.
CTS Page 1.1/2.1-1 In the SAFETY LIMIT specification 1.1.A.1, a phrase is inserted to differentiate between the Safety Limit i
Minimum Critical Power Ratio (SLMCPR) for two recirculation loop operation and SLO.
The new text is in bolded italics.
For Unit 1 the change is as follows-When the reactor pressure is greater tuan 800 psia, i
the existence of a minimum critical power ratio (MCPR) i less than 1.07 for two-recirculation-loop operation (or 1.09 for single-loop operation) shall constitute violation of the fuel cladding integrity safety limit.
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4 El-2
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t For Units 2 and 3 the change is as follows:
When the reactor pressure is greater than 800 psia, the existence of a minimum critical power ratio (MCPR)
Less than 1.10 for two-recirculation-loop operation (or 1.12 for single-loop operation) shall constitute violation of the fuel cladding integrity safety limit.
2.
CTS Page 1.1/2.1-2 LIMITING SAFETY SYSTEM SETTING 2.1.A.1.a, Neutron Flux Trip Settings, is modified to include a recirculation drive flow correction factor for SLO.
Tne modified equation is:
Ss(0.58(W-aW) + 62%)
Also, a definition is added for the flow correction factor as follows:
AW = Difference between two-loop and single-loop effective recirculation drive flow at the same core flow.
AW = 0 for two-loop operation, j
3.
CTS Page 1.1/2.1-8 l
The second paragraph in Bases 1.1 for fuel cladding integrity is revised to acknowledge the change in SLMCPR applicable to SLO.
Changes are in italics, j
For Unit 1 the change is as follows:
The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational... boiling (MCPR of 1.0).
This establishes a Safety Limit such that the minimum critical power ratio (MCPR) is no less than 1.07 (for two-recirculation-loop operation).
MCPR > 1.09 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
This Safety Limit MCPR is increased for single-recirculation-loop operation to account for instrument uncertainties as discussed in Reference 3 and as documented in the COLR.
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El-3
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'For Units 2 and 3 the change is as follows:
The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational... 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.
This Safety Limit MCPR is increased for single-recirculation-loop operation to account for instrument uncertainties as discussed in Reference 3 and as documented in the COLR.
4.
CTS Page 1.1/2.1-10 The BFN NEDO on SLO is added to the list of references as Reference 3.
3.
Browns Ferry Nuclear Plants, Units 1, 2,
and 3, Single-Loop Operation, NEDO-24236, May 1981.
5.
CTS Pages 3.1/4.1-7; 3.1/4.1-7; 3.1/4.1-6 Note 24 to Table 3.1.A, REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENTS, applying to the Average Power Range Monitor (APRM) scram function is supplemented to reference the need to apply a recirculation drive flow correction factor when operating in SLO mode.
A 6-hour allowance for making the required setpoint adjustments is added.
Also, the change corrects the existing reference to Figure 2.1-1 to Figure 2.1-2.
Figure 2.1-1 was deleted in a previous TS change.
After the change, note 24 will read as follows.
24.
The Average Power Range Monitor scram function is varied (Reference Figure 2.1-2 and 2.1.A.1.a) as a function of recirculation loop flow (W).
The trip setting of this function must be maintained in accordance with 2.1.A.
For single-recirculation-loop operation, a flow correction factor (hW) is also applied as specified in 2.1.A.1.a.
Following entry into single-loop operation, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to make the setpoint adjustment.
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6.
CTS Pages 3.2/4.2-26; 3.2/4.2-26; 3.2/4.2-25 Note 2 for Table 3.2.C, INSTRUMENTATION THAT INITIATES ROD BLOCKS, relating to the APRM Upscale Flow Bias rod block setpoint is revised to reference the need to apply a recirculation drive flow correction factor when operating in SLO mode.
Note that the actual rod block setpoints are documented in the CORE OPERATING LIMITS REPORT (COLR).
A 6-hour allowance is provided for making the setpoint adjustments.
The revised Note 2 will read as:
2.
The trip level settings shall be as specified in the CORE OPERATING LIMITS REPORT (COLR).
Fbr j
single-recirculation-loop operation, a flov i
correction factor (bH) is also applied to the settings as specified in the COLR.
Following entry into single-loop operation, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed to make the setpoint adjustment.
7.
CTS Pages 3.2/4.2-27; 3.2/4.2-27; 3.2/4.2-26 Note 13 for Table 3.2.C relating to the Rod Block Monitor (RBM) Upscale Flow Bias rod block setpoint is revised to reference the need to apply.a drive flow correction factor when operating in SLO mode.
Note that.the actual rod block setpoints are documented in the COLR.
A 24-hour allowance for making the required setpoint adjustments is added.
The revised Note 13 will read as:
13.
The trip level settings and clipped value for this setting shall be as specified in the CORE OPERATING LIMITS REPORT (COLR).
Fbr single-recirculation-loop operation, a flow correction factor (bH) is also applied to the settings as specified in the COLR.
Following entry into single-loop operation, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed to make the setpoint adjustment.
8.
CTS Pages 3.5/4.5-33; 3.5/4.5-31; 3.5/4.5-34 In the Bases section for Limiting Condition for Operation (LCO) 3.5.I, Average Planar Linear Heat Generation Rate (APLHGR), a discussion is added I
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El-5 i
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l indicating that when in SLO, a multiplier is applied l
to the APLHGR core thermal limit as specified in the l
COLR.
The new sentence is:
In single-recirculation-loop mode, an APLHGR l
multiplier is applied by the process computer to the l
APLHGR core thermal limit.
The multiplier is documented in the COLR.
9.
CTS Page 3.6/4.6-12 1
1 l
LCO 3.6.F.1 on Recirculation Pump Operation is replaced as shown below.
Changes are in italics.
This change makes continued SLO contingent on implementing the referenced instrumentation (drive flow correction factor) and core thermal limits adjustments (APLHGR multiplier and MCPR adder).
The existing 24-hour LCO time limit in current TS is maintained.
3 Operation in single-recirculation-loop operation (SLO) made is permitted provided the SLO requirements specified in 2.1.A.1.a, Table 3.1.A, and Table 3.2.C are met, and provided core thermal limits are adjusted as specified in the COLR for SLO.
Otherwise, the plant shall be placed in a HOT SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the loop is sooner returned to service.
10.
CTS Page 3.6/4.6-32 In the Bases for 3.6.F/4.6.F, Recirculation Pump Operation, the following sentence is added as the first sentence to reference the new SLO requirements in LCO 3.6.F.1.
Operation in single-recirculation-loop mode is permitted provided adjustments are made to the flow-biased scram and rod block setpoints, and core thermal limits are adjusted as specified in the COLR.
11.
CTS Page 6.0-26a and 6.0-26b; 6.0-26a; 6.0-26a A footnote is added to TS 6.9.1.7.a, CORE OPERATING LIMITS REPORT, indicating core thermal limits and i
instrument setpoint adjustments associated with SLO l
will be documented in the cycle-specific COLR.
The l
l El-6 l
l new footnote is as follows and applies to TS 6.9.1.7.a subitems (1), (3), (4), and (5).
The CORE OPERATING LIMITS REPORT shall also i
document core limits and instrument setpoint adjustments associated with single-recirculation-loop operation.
l l
Proposed Changes To ISTS Conversion Package (TS-362)
Below are the proposed changes in ISTS format to l
incorporate provisions for SLO.
The conversion package to ISTS was submitted to NRC as TS-362 on September 6, 1996.
)
l The requested SLO changes are based on NUREG-1433, Revision
]
1,
" Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/4)" which already contains TS provisions for operating in SLO mode.
Page numbers are l
the same for all three units.
l 12.
ISTS Page 2.0-1, SAFETY LIMIT (SL), 2.1.1.2 l
In Reactor Core SL 2.1.1.2, the following text is l
added to differentiate between the SLMCPR Limit for two recirculation loop operation and SLO.
The new text is in bolded italics.
1 2.1.1.2 With the reactor steam dome pressure 2 785 psig and core flow 2 10% rated flow:
l MCPR shall be 2 1.10 for two recirculation loop operation or 2 1.12 for single loop operation.
13.
ISTS Page 3.3-6, Table 3.3.1.1-1, Reactor Protection System Instrumentation For FUNCTION 2.b, APRM Flow Biased Simalated Thermal l
Power - High, a new footnote (b) is superscripted to l
the APRM flow biased scram line ALLOWABLE VALUE to provide a recirculation drive flow correction factor for SLO.
In the equation, RTP is rated thermal power.
l The modified scram setpoint value is in the new footnote which reads as follows.
(b)
[0.58 W + 62% - 0.58 AW] RTP when reset for I
single loop operation per LCO 3.4.1,
" Recirculation Loops Operating."
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14.
ISTS Page 3.4.1, LCO 3.4.1, Recirculation Lcops Operating LCO 3.4.1 is revised to provide for single-loop operation.
This change makes continued SLO contingent on implementing the referenced instrumentation setpoint changes (flow correction factor) and core thermal limits adjustnents (APLHGR multiplier and MCPR adder).
Also, a requirement to operate outside the exclusion regions of Figure 3.4.1-1, THERMAL POWER VERSUS CORE FLOW STABILITY REGIONS, is included in the same manner as that for two-loop operation.
The following text is added immediately following the existing LCC statement on two-loop operation.
9R One recirculation loop may be in operation with core j
flow as a function of THERMAL POWER outside Regions I and II and the Operation Not Permitted Region of o
Figure 3.4.1-1 and provided the following limits are applied when the associated LCO is applicable:
i a.
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; b.
" MINIMUM CRITICAL POWER RATIO (MCPR),"
single loop operation limits specified in the COLR; c.
" Reactor Protection System (RPS)
Instrumentation," Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power - High), Allowable Value of Table 3.3.1.1-1 is reset for single loop operation; d.
" Control Rod Block Instrumentation,"
Function 1.a (Rod Block Monitor Upscale (Flow l
Biased)), Allowable Value of Table 3.3.2.1-1 is reset for single loop operation.
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15.
ISTS Page 3.4-1, ACTION C l
l ACTION C is revised to match BWR/4 ISTS for plants licensed to operate in single loop as adapted to the l
BFN stability monitoring TS.
The proposed COMPLETION J
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i TIMES are the same as those in BWR/4 ISTS for the equivalent REQUIRED ACTIONS.
Current CONDITION C l
CONDITION REQUIRED ACTION COMPLETION TIME l
l C.
One recirculation loop C.1 Restore two 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> not in operation.
recirculation l
loops to operation.
i Revised CONDITION _C CONDITION REQUIRED ACTION COMPLETION TIME C.
Requirements of the C.1 Satisfy the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LCO not met for requirements reasons other than of the LCO.
A or B.
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16.
ISTS Page B 3.2-2, BASES Section 3.2.1, APLHGR In the APPLICABLE SAFETY ANALYSES BASES for Section 3. 2.1, APLHGR, a new paragraph is added which describes the APLHGR multiplier which is applied when in SLO.
The text is the same as the BWR/4 ISTS except that the COLR is referenced as the source of the APLHGR correction factor and the references are BFN specific.
The new paragraph reads:
For single recirculation loop operation, an APLHGR multiplier is applied to the APLHGR limit (Ref. 5 and Bef. 7).
The multiplier is documented in the COLR.
This multiplier is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe heat-up during a LOCA.
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17.
.STS Page B 3.2-2, BASES Section 3.2.1, APLHGR In the LCO BASES for Section 3.2.1, a discussion concerning the APLHGR multiplier for SLO is added as l
follows.
The wording is similar to BWR/4 ISTS except that BFN does not yet utilize power-and j
flow-dependent APLHGR multipliers associated with the APRM and RBM TS (ARTS) improvement options.
New text is shown in italics.
j The APLHGR limits specified in the COLR are the result i
of the fuel design, DBA, and transient analyses.
With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1,
" Recirculation Loops Operating," the linit is
}
determined by multiplying the exposure dependent limit by an APLHGR corroction factor (Ref. 5 and Ref. 7).
18.
ISTS Page B 3.2-3, BASES Section 3.2.1, APLHGR I
In the list of references, the BFN SLO NEDO is added as new Reference 7 as shown next.
7.
NEDO-24236, " Browns Ferry Nuclear Plant Units 1, 2,
and 3, Single-Loop Operation," May 1981.
19.
ISTS Page B 3.2-4, BASES Section 3.2.2, MCPR l
In the first sentence in the APPLICABLE SAFETY ANALYSES BASES, a new ref. B is added to reference the BFN SLO NEDO.
20.
ISTS Page B 3.2-8, BASES Section 3.2.2, MCPR In the list of references, the BFN SLO NEDO is added as new Reference 8 as shown next.
8.
NEDO-24236, " Browns Ferry Nuclear Plant Units 1, i
2, and 3, Single-Loop Operation," May 1981.
a 21.
ISTS Page B 3.4-3, BASES Section 3.4.1, Recirculation l
Loops Operating In the APPLICABLE SAFETY ANALYSES BASES, two new paragraphs are added describing the SLO Loss of Coolant Accident (LOCA) and transient analyses.
Refer to the mark-up pages in Enclosure 4 for placement of
{
the paragraphs which are repeated below.
The new i
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l paragraphs use the same wording as BWR/4 ISTS except that BFN specific references are inserted and a discussion related to the need to modify the RBM flow-biased setpoints is included.
Plant specific LOCA analyses have P-un performed assuming only one operating recirc' ation loop.
These analyses have demonstrated that, 1-the event of a LOCA caused by a pipe break in the, aperating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Refs. 7 and 8).
l The transient analyses of Chapter 14 of the FSAR have I
also been performed for single recirculation loop operation (Ref. 7) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients l
analyzed provided the MCPR requirements are modified.
During single recirculation loop operation, i
modification to the Reactor Protection System (RPS) l average power range monitor (APRM) instrument and RBM setpoints is also required to account for the different relationships between recirculation drive l
flow and reactor core flow.
setpoints for single loop operation are specified in the COLR.
The APRM Flow Biased Simulated Thermal Power-High setpoint is in LCO 3.3.1.1,
" Reactor Protection System (RPS) Instrumentation" and the RBM Flow Biased Upscale setpoint is in the COLR as referenced by LCO 3.3.2.1,
" Control Rod Block Instrumentation."
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22.
ISTS Page B 3.4-4, BASES Section 3.4.1, Recirculation Loops Operating l
A new sentence is inserted into the LCO 3.4.1 BASES l
describing the requirements for SLO specified in the i
revised LCO.
Refer to the mark-up copies for l
placement of the sentence which is repeated below.
l Wording is the same as BWR/4 ISTS except for adding i
BFN specific references and the addition of a i
reference to resetting the RBM rod block line.
With only one recirculation loop in operation, i
modifications to the required APLHGR Limits (LCO 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE l
El-11 r
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( APLHGR)"), MCPR limits (LCO 3.2.2,
" MINIMUM CRITICAL POWER RATIO (MCPR)"), APRM Flow Biased Simulated Thermal Power-High Setpoint (LCO 3.3.1.1), and RBM Flow Biased Upscale Setpoint (LCO 3.3.2.1) may be applied to allow continued operation consistent with the assumptions of References 7 and 8.
24.
ISTS Page B 3.4-5, BASES Section 3.4.1, Recirculation Loops Operating In the BASES discussion for ACTION C.1, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is substituted for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (two places) to be consistent with the proposed change to the subject ACTION.
The current 12-hour requirement derives from the existing restrictions on operating in SLO.
Also, the following text is inserted as the second paragraph in this BASES section.
With these changes, the BASES will read the same as BWR/4 ISTS for SLO licensed plants.
Alternatively, if the single loop requirements of the LCO are applied to the operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.
25.
ISTS Page B 3.4-8, BASES Section 3.4.1, Recirculation Loops Operating Two new references 7 and 8 are added as follows.
7.
NEDO-24236, " Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.
l 8.
NEDC-32484P, " Browns Ferry Nuclear Plant Units 1, 2,
and 3, SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 1, February 1996.
26.
ISTS Page B 3.4-11, BASES Section 3.4.2, Jet Pumps In the BASES for the jet pump SURVEILLANCE REQUIREMENT 3.4.2.1, the following sentence is inserted into the i
first paragraph to match BWR/4 ISTS wording.
Refer to the mark-ups for placement.
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Similarly, initial entry into extended single loop operation may also require establishment of these relationships.
II.
REASON FOR THE PROPOSED CHANGE The reactor coolant recirculation system provides forced coolant flow through the reactor core and, in combination with control rods, provides a means to control and change reactor power over a broad range.
The recirculation system consists of two recirculation pump loops and drive units, each with a separate variable speed motor generator (MG) set, recirculation pump, and piping loop.
The individual recirculation pumps are located in the drywell and provide drive flow to two separate jet pump headers inside the reactor vessel which, in turn, provide core recirculation flow.
For a detailed discussion of the system components and operating characteristics, refer to the description in Chapter 4.3 in the Updated Final Safety Analysis Report (UFSAR).
During normal power operation both recirculation pumps are operated at matched speeds to provide forced recirculation flow.
Recirculation pump speed and flow can be changed using the variable speed recirculation system MG set and, thus, be used to change core power.
Power generation with a single recirculation loop in service is also a recognized mode of operation for Boiling Water Reactors (BWRs) and many BWRs have TS which allow for SLO.
The most obvious benefit of SLO is the ability to continue power operation in the event of the loss of a recirculation loop due to component malfunction.
The majority of active components in the recirculation system are located in the i
reactor building and are readily accessible during power operation.
This includes the MG set drive motcrs, fluid l
couplers, generators, and associated oil coolers.
Also accessible are the recirculation system controllers, logic l
relaying, and system electrical panels and breakers.
l Typically, most of these components can be repaired with the reactor in service with no impact on power operation other than the unavailability of the affected loop itself, f
While the reactor recirculation system is a very reliable system, the temporary unavailability of a recirculation loop due to components problems is occasionally l
experienced.
Current TS 3.6.F.1 provides only a 24-hour El-13 l
l l
l
l LCO time allowance in SLO mode.
As might be expected, it is not alwa}J possible to diagnose and effect repairs to the system within these time frames.
BFN most recently had a recirculation system TS required shutdown in September 1996 due to a bus bar failure on a recirculation set drive generator.
In that case, the repair ended-up being relatively simple, but could not be diagnosed, planned, and implemented within the current LCO time frame.
In this submittal, TVA is providing the reactor analyses which form the basis for changing the TS to allow indefinite SLO provided certain core limits and instrument setpoints are adjusted.
These TS changes would allow continued operation in the event of the trip or loss of a single recirculation loop.
As noted above, it is expected that, in most cases, a malfunctioning loop could be l
repaired with the reactor in service, and normal operations resumed.
Thus, for these circumstances, approval of this TS would have a positive economic impact in allowing continued power generation in SLO.
Similarly, the avoidance of shutdowns due to recirculation system failures
)
would eliminate risk factors associated with unnecessary reactor shutdowns and restarts.
III.
SAFETY ANALYSIS With two recirculation loops in service, the reactor recirculation system, in combination with the control rods, l
is designed to support the full range of power operations.
Power operation with a single recirculation loop in service is also a viable mode of operation although with a single loop in service, full core flow and, hence, full core power can not be achieved.
Maximum power output in single-loop is about 70% of that attainable with two loops in service.
Reactor control and operation in single-loop is very l
similar to that in two-loop recirculation mode.
The primary difference is that as the drive flow on the operating pump is increased, part of the total flow from the active jet pump loop will backflow through the inactive l
jet pumps.
This effect reduces the net achievable core l
flow and limits the power level that can be achieved i
l compared to two-loop operation.
I i
l Many BWRs (Hatch, Peach Bottom, Cooper, Duane Arnold, Grand j
Gulf, and others) have previously provided justification for SLO and have obtained TS changes allowing operation in 5
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i l
l l
single-loop.
Also, BEN Unit 1 has previously operated in single-loop mode for several weeks in 1978 (References 1 1
and 2).
Thus, there is ample industry experience demonstrating the viability of operating in single-loop and precedent in analyzing the safety considerations of SLO.
The NRC has previously determined SLO is generically acceptable as set forth in Generic Letter 86-09, Technical Resolution of Generic Issue B-59-(N-1) Loop Operation in BWRs and PWRs (Reference 3).
SLO is also recognized as a standard mode of operation in NUREG-1433, BWR/4 Rev.1 ISTS.
The primary analysis of the safety considerations in support of BFN's proposed SLO TS is presented in NEDO-24236, " Browns Ferry Nuclear Plants, Units 1, 2,
and 1
3, Single-Loop Operation," May 1981, which is attached for reference.
NEDO-24236 was prepared by General Electric (GE) for TVA for the express purpose of evaluating the effects of SLO on the plant transient and accident i
analyses.
In NEDO-24236, an evaluation was performed related to the l
effects of operating in single-loop mode on the affected transient and accident analyses in Chapter 14 of the UFSAR.
In some cases, as discussed below, improved analysis l
methods have subsequently been implemented since the original issue of NEDO-24236.
Nonetheless, the basic i
)
conclusions of NEDO-24236 remain valid in that the SLO l
analyses are bounded by the corresponding two-loop full l
power transients.
i
(
For Units 2 and 3, SLO has been included as an operating i
flexibility option in the cycle-specific core reload l
analyses which are performed for BFN by GE each fuel cycle.
These analyses are performed in accordance with the latest l
approved version of GE Licensing Topical Report NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel."
Since Unit 1 is in an extended nonoperational status, a cycle-specific analysis with the l
SLO option has not yet been prepared.
Prior to the return of Unit 1 to service, this SLO analysis will be performed.
Experience indicates that the Unit i results will be very similar to the current Unit 2 and 3 analyses.
Safety Limit Minimum Critical Power Ratio (SLMCPR) l The SLMCPR is established such that no fuel damage is 3
calculated to occur if the limit is not violated during i
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reactor transients.
The limit is calculated using a statistical model which includes considerations for uncertainties.
For SLO as discussed in NEDO-24236, the SLMCPR increases by.01 due to increased uncertainties in the total core flow and Traversing In-Core Probe (TIP) readings compared to two-loop operation.
Except for these readings, the uncertainties used in the statistical analysis for SLMCPR are not dependent on whether core flow is provided by one or two recirculation pumps.
The SLO'SLMCPR adjustment factor has been recalculated as documented in latest cycle-specific reload analysis for Units 2 and 3.
A copy of the current Unit 3 cycle-specific core reload analysis is provided as an example in and shows a.01 SLMCPR adder for SLO in Section 11, Cycle MCPR values.
The SUMCPR adjustment for the current Unit 2 core is.02 which is the value included in the proposed TS for all three BFN units to avoid the need to request additional TS changes for future core reloads.
Discussions with GE indicate that a.02 value should be bounding for future considerations.
Regarding the proposed TS changes, in CTS format, change 1 revises the SLMCPR specification 1.1.A.1 to add.02 to the value of SLMCPR for operation in SLO for all three units.
CTS Change 3 revises the corresponding Bases.
The proposed I
Unit 1 TS changes are marked-up against the existing TS pages which show 1.07 as the base SLMCPR for Unit 1.
I TS-377 (References 4 and 5) which has been previously submitted to NRC revises the Unit 1 base SLMCPR to be 1.10.
Therefore, the adjusted values for TS SLMCPR for Unit 1 SLO with the.02 adder would be 1.12 based on the changes requested in TS-377.
TS-377 has been approved for Units 2 and 3 and revised the SLMCPR to 1.10 which is also the same j
value for SLMCPR in the ISTS submittal, TS-362.
l l
In the plant, the process computer is adjusted to account for the change in SLMCPR by adding.02 to the Operating Limit MCPR (OLMCPR) when in SLO.
Existing CTS 3.5.K/4.5.K provide the requirements for monitoring MCPR and refer to the COLR for the actual OLMCPR limits.
To emphasize that an adjustment to the MCPR limit is required when in SLO mode, CTS change 11 adds a footnote to COLR TS 6.9.1.7.a indicating the adjustment will be included in the COLR.
In ISTS format the change in SLMCPR is addressed by change 12 by adding
.')2 to the value for two-loop operation in SL 2.1.1.2.
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l l
Core Operating Transients In Section 3 of NEDO-24236, the core wide transients potentially affected by single loop operation were evaluated.
Since single-loop operation involves operating at a reduced power level, the consequences of the abnormal operating transients (pressurization, flow increase, cold water injection events) were less severe than the same events analyzed for two-loop mode, and thus bounded by the two-loop UFSAR analyses.
Additionally, the current cycle-specific reload analyses on Units 2 and 3 have been performed by GE considering SLO as an operating flexibility option in accordance with the methods described in NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel."
It was concluded i
that the OLMCPR (with an SLO MCPR adder) as specified in l
the COLR provides adequate protection for transients initiated during SLO.
The Rod Withdrawal Error (RWE) is also evaluated each fuel cycle as part of the cycle-specific reload analysis, and RBM setpoints selected such that a rod block will occur prior to reaching the SLMCPR.
The same analysis applies to single-loop operation although, as discussed in Section 3.2 of NEDO-24236, an adjustment to the flow-biased rod block is required to account for the change in the relationship between the core flow and recirculation drive flow due to backflow through the inactive loop jet pumps when in SLO mode.
The appropriate recirculation drive flow correction factor is provided in Section 3.2 of the NEDO and, when applied to the RBM rod block equation, assures the SLMCPR would not be violated if a RWE event occurred daring SLO.
The same correction factor will be applied to the APRM scram and rod block lines since they are also flow-biased although direct credit is not taken for the APRM flow-biased scram and rod blocks in the UFSAR safety l
analyses.
l l
Changes 2 and 5 of the CTS adds the recirculation drive l
flow correction factor to the APRM scram line and provides l
a definition of the correction factor.
CTS Changes 6 and 7 l
add the same flow correction for the APRM and RBM rod block settings.
CTS Change 9 revises LCO 3.6.F (Recirculation Pump Operation) to add a requirement that the flow correction factor be applied in SLO in accordance with the i
I El-17
corresponding APRM and RBM instrumentation tables.
CTS Change 10 inserts a discussion in the Recirculation Pump Operation Bases on the same subject.
Since the actual rod block settings are documented in the COLR, CTS change 11 adds a footnote to the COLR TS requiring that the flow correction factor be included in the COLR.
For the above changes, a 6-hour allowance is provided to make the APRM scram and APRM rod block setpoint changes.
The APRM adjustment is performed by instrument technicians and is a simple procedure.
The proposed time allowance is the same as that allowed for the adjustment of APRM flow-biased setpoints in TS 4.5.L related to peaking factor, and is a similar type of adjustment.
Also, if SLO was entered as a result of a recirculation pump trip, it is likely that a 4.5.L peaking factor adjustment (6-hour LCO) would also be needed and would be implemented simultaneously with the SLO recirculation drive flow correction adjustment.
The RBM adjustment is more time consuming and a 24-hour allowance is appropriate as proposed in CTS change 7 The existing TS allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of operation without any instrument adjustments being required, therefore, a 24-hour allowance for the RBM setpoint change is no less stringent than current TS.
Also, a review of other plants SLO TS indicates several plants have a 24-hour grace period when in SLO prior to requiring instrumentation changes.
In ISTS format, changes 13 and 14 require the same drive flow correction factor be applied when in SLO for the flow-biased scram line and RBM rod block line.
Concerning the RBM changes, the proposed TS differs from BWR/4 ISTS which are written for plants that have implemented the ARTS improvement option which involves elimination of flow-biased RBM setpoints.
BFN has not yet implemented the ARTS improvements although, as requested in TS-353R1 and TS-353S1 (References 6 and 7), implementation is planned starting with Unit 2 in the next operating cycle (October 1997).
Pending incorporation of the ARTS improvements, in the BFN proposed SLO TS provisions are added which require resetting the RBM flow-biased upscale settings in accordance with Table 3.3.2.1-1 and the COLR.
As previously noted, the COLR will be modified to include the flow correction factor for rod block settings.
A BASES discussion of the flow correction factor applicability has also been added as shown in ISTS Changes 21 and 22 to provide an explanation of the subject changes.
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l ISTS do not have equivalent TS sections for APRM rod i
blocks, therefore, no TS changes are necessary in this regard.
Time requirements for changing instrument setting will be the same as currently in BWR/4 ISTS.
In summary, based on the evaluations provided in NEDO-24236 and the cycle-specific reload analyses, the SLMCPR would l
not be exceeded by reactor transients if experienced when l
in SLO.
This conclusion includes the consideration that j
the SLMCPR adder and RBM flow correction factor are implemented as described in the above discussion and in the proposed TS changes.
Accident Evaluation NEDO-24236, Section 5, provides an evaluation of the effects of SLO on the applicable accident analyses.
The LOCA and Recirculation Pump deizure events were evaluated since the remaining Chapter 14 UFSAR accidents are not significantly affected by the recirculation mode or are bounded by the two-loop accident analysis.
Section 5.1 of NEDO-24236 provides a summary of the LOCA analysis for SLO, and a set of figures is provided showing the transient response of the reactor and fuel.
As discussed in the NEDO, calculated core reflood times in some cases were longer than the two-loop limiting break resulting in the application of an APLHGR multiplier factor for specific fuel types used at the time as shown in Table 5.1 of the NEDO.
For the LOCA analysis, NEDO-24236 used the SAFE /REFLOOD computer codes.
BFN has since transitioned to the improved SAFER /GESTR codes for performing LOCA analyses.
Our February 23, 1996, submittal (Reference 8) to NRC provided notification of the change to SAFER /GESTR in accordance with 10 CFR 50.46 (a) (3) (ii).
A copy of the BFN LOCA results using the new codes was provided to NRC in Reference 9 as NEDC-32484P, Revision 1,
" Browns Ferry Nuclear Plant, Units 1, 2,
and 3, SAFER /GESTR-LOCA, Loss-of-Coolant Accident Analysis."
Section 5.3.4 of NEDC-32484P, Revision 1 provides the results of the LOCA results for SLO using SAFER /GESTR for the limiting single failure.
The analysis for the design basis LOCA assumes there is essentially no period of pump coastdown, and thus, dry out is assumed to occur l
simultaneously at all axial locations in the hot bundle El-19
l l
l very early in the event.
This assumption is very conservative and provides bounding results for the design i
NEDC-32484P, Revision 1 also concluded that the design basis accident (large breaks) are more affected than small break sequences and, therefore, the large break results are l
bounding for SLO.
Furthermore, with the application of a
.9 APLHGR n.ultiplier for the current fuel types, the LOCA peak clad temperature for SLO would always be lower than that for two-loop operation.
The SLO APLHGR multiplier is l
also provided in the cycle-specific reload analyses as
)
shown in Section 16 of the attached Unit 3 reload report.
Following approval of the proposed SLO TS, the APLHGR SLO multiplier will be documented in the COLR reports for each fuel cycle.
Changes 8, 9,
and 10 of the CTS are made to indicate that an APLHGR multiplier is applied while in SLO.
CTS Change 11 adds a footnote to the COLR TS which indicates the SLO l
APLHGR multiplier for LCO 3.5.I (APLHGR) limits must be documented in the COLR.
In ISTS format, changes 15, 16, 21, and 22 make the corresponding changes using the l
BWR/4 ISTS NUREG recommended wording.
In the plant, the i
process computer recognizes SLO and automatically applies the APLHGR multiplier when calculating core thermal limits.
Section 5.2 of NEDO-24236 provides a discussion of the analysis of seizure of the operating recirculation pump while in SLO.
The analyses referenced in the NEDO indicate that the MCPR will be greater than the SLMCPR, and that the event will terminate with the reactor continuing to operate in natural circulation (Reference 10),
t GE also has indicated that, based on the results of similar calculations of the recirculation pump seizure accident (PSA) for other BWRs with recent fuel designs (Gell, GE12, i
and GE13), it is clear that the PSA during SLO will not result in fuel entering boiling transition.
Consequently, radiological releases are avoided and there is no violation of 10 CFR 100 radiation release limits.
The reference analyses used current GE methodologies and included l
improvement programs such as ARTS / Maximum Extended Load l
Line Limit (MELLL) and thermal power uprate.
Based on the l
GE review, it was concluded that the PSA for SLO remains l
non-limiting as previously stated in NEDO-24236.
i El-20 1
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i Stability Requirements i
l l
The least stable power / flow conditions under normal conditions occurs in natural circulation with a high l
power / flow ratio.
Operation in two-loop operation is the i
most stable.
SLO is slightly less stable than two-loop l
operation, but as recirculation flow increases, the stability performance is similar to two-loop operation.
In l
SLO, measured neutron noise has been observed to increase i
over that experienced in two-loop operation due to higher noise in the recirculation flow associated with jet pump back flow in the inactive jet pump loop.
This phenomena is not related to core instability as was demonstrated in a set of single-loop operational tests conducted on BFN Unit 1 in February 1985 by Oak Ridge National Laboratory for the NRC (Reference 11).
Section 4 of NEDO-24236 provides a discussion of the SLO stability analysis and shows a graphical comparison of single and two-loop recirculation modes.
The thermal-hydraulic stability licensing criteria was later revised as documented in NEDE-240ll, Revision 6, Amendment 8,
" Thermal Hydraulic Stability Amendment to GESTAR II which was subsequently approved by NRC (Reference 12).
The methodology described in the NEDE also applies to SLO.
It is noted that cycle-specific stability analyses are, however, no longer performed since, as discussed in part 15 of the attached Unit 3 Reload Analysis, plants subsequently adopted stability monitoring provisions as elaborated below.
Because of generic stability concerns and the experience of some BWR reactors regarding the potential for power oscillations at low flow /high power operating map conditions, NRC issued Bulletin No. 88-07 which was supplemented in December 1988 (Reference 13).
Bulletin 88-07, Supplement 1, requested Licensees adopt the Boiling Water Reactor Owners Group (BWROG) stability monitoring guidelines similar to those originally issued by GE as Services Information Letter (SIL) 380 Revision 1 (Reference 14).
BFN responded to Bulletin 88-07, Supplement 1, by incorporating a stability power / flow monitoring TS along with associated surveillance requirements and action statements as shown in existing CTS Section 3.5.M and carried forth in the proposed TS-362 ISTS conversion package in section 3.4.1.
The CTS change was submitted to l
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-. ~ _ -
l l
NRC as TS-272 for Unit 2 in June 1989 (reference 15) and l
was approved in October 1989 (Reference 16).
These same TS l
changes were subsequently approved for Units 1 and 3.
l The stability monitoring provisions in the TS apply equally to two-loop or single-loop recirculation power operation.
NRC later requested Licensees address long-term solutions for concerns for thermal-hydraulic instabilities in Generic Letter (GL) 94-02 dated July ll, 1994 (Reference 17).
In the GL, NRC requested that utilities review operating procedures and training program for detection and suppression of power oscillations based on more recent operating experiences and revised BWROG Guidelines on stability monitoring.
BFN responded to the GL on September 8, 1994 (Reference 18) and provided commitments to revise operating procedures and training programs to be consistent with the most current BWROG stability monitoring guidelines.
No additional TS changes beyond those previously approved for TS-272 were needed to be responsive to the GL.
As noted in the previous paragraph, the stability monitoring provisions in the TS are applicable in two-loop or single-loop recirculation pump power operation.
i Regarding the proposed SLO TS changes, ISTS change 14 revises LCO 3.4.1, Recirculation Loops Operating, to invoke the current two-loop operation stability monitoring requirements for single-loop operation.
No additional changes are needed for the CTS since existing 3.5.M, Core Thermal-Hydraulic Stability Monitoring, is not recirculation mode specific and, thus, is also applicable to SLO.
In summary, the stability analyses provided in NEDO-24236 and NEDE-24011-P-A, coupled with existing stability monitoring TS satisfactorily address stability issues in SLO.
Jet Pump Surveillance In single-loop recirculation mode, increases in APRM noise and core delta p fluctuations have been observed in some plants while operating at high drive flows which may be l
associated with increased vibration at the active jet pumps.
BFN custom TS and the TS-362 ISTS already require daily monitoring of jet pumps operability, therefore, no new TS changes are needed to ensure SLO does not adversely l
j El-22
are 3.6.E, Jet Pumps, in custom format and LCO 3.4.2, Jet Pumps, in ISTS format.
ISTS change 26 does modify the jet pump Bases to clarify that during initial entry in SLO, baselining of jet pump performance may be necessary in order to apply the related surveillance requirements.
This change makes the jet pump Bases consistent with BWR/4 ISTS.
Vessel Internal Vibration During preoperational and startup testing of BEN Unit 1 between December 1972 and March 1974, extensive vibration data was recorded to demonstrate the adequacy of the 251-inch diameter vessel design.
The testing included operation in single-loop at a recirculation pump speed greater than 90%.
Confirmatory tests were later performed at Peach Bottom.
The BFN testing is documented in GE document 386HA219 entitled Browns Ferry-1 Vibration Measurements.
Pertinent portions of this document were previously submitted to NRC (Reference 19).
The test report concluded that all vibratory responses were within the acceptance criteria for all test conditions, and that vessel internals were not threatened by single-loop operation under the conditions which bound the values expected during operation.
Since the maximum recirculation pump speed during the test was 91%, BFN currently limits pump operation to 90% in the system operating instruction.
Also, as discussed in the previous section, BFN currently has TS provisions which require daily surveillance of jet pump operability.
This provides further assurance that no adverse effects result from vibration while operating in single-loop mode.
IV.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION TVA has conclbded that operation of BFN Units 1, 2,
and 3 in accordance with the proposed change to the TS does not involve a significant hazards consideration.
TVA's conclusion is based on its evaluation, in accordance with 10 CFR 50. 91(a) (1), of the three standards set forth in 10 CFR 50.92(c).
l l
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y A.
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
An analysis of the limiting operational transients has been performed by GE for BEN as documented in NEDO-24236 to demonstrate adequate margin to the Safety Limit Minimum Critical Power Ratio (SLMCPR).
In addition, SLO has been specified as a operating option for the transient and accident evaluations performed as part of the cycle-specific core reload analyses for Units 2 and 3 which ensure that operating
' limit Minimum Critical Power Ratios (OLMCPRs) for the current fuel types are established that maintain required margin to the fuel cladding safety limit.
A cycle-specific analysis with SLO will be performed for Unit 1 prior to restart and experience indicates similar results are expected as those for Units 2 i
and 3.
A review of the values used in the statistical l
analysis used in the basis of the fuel cladding safety 1
limit determined that, due to increased uncertainties in total core flow readings and Traversing In-Core Probe (TIP) readings during SLO, an increase in the SLMCPR of.02 is bounding when in SLO.
Therefore, l
while operating in single-loop mode, an additional.02 L
is added to the OLMCPR which maintains the same margin to the fuel cladding safety limit as that established for two-loop operation.
This is a conservative approach because the two-loop transients have been shown to be more severe than the equivalent single-loop events and, therefore, the OLMCPRs l
established for two-loop operation would always be bounding.
Thus, the margin of safety for fuel clad integrity is assured and the probability or consequences associated with reactor transients is not increased for SLO.
l SLO results in backflow through the jet pumps in the inactive recirculation loop which perturbs the relationship between the core flow and recirculation l
drive flow on which the flow biased Average Power Range Monitor (APRM) and Rod Block Monitor (RBM) setpoint equatior.s are based.
To compensate, the proposed TS changes modify the setpoint equations to correct for one-loop operation.
With this adjustment, j
the setpoint equations preserve the original i
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~
relationship between the setpoints and the effective recirculation drive flow such that the consequences of a RWE in SLO are bounded by the cycle-specific RWE analyses.
Therefore, these changes do not increase the probability or consequences of the RNE transient previously eTaluated.
Average Planar Linear Heat Generation Rate (APLHGR) limits are established to ensure the acceptance criteria for fuel and Emergency Core Cooling Systems established in 10 CFR 50.46 are met.
A SLO Loss of Coolant Accident (:LOCA) analysis was performed using the SAFER /GESTR computer code as documented in NEDC-32484P, Revision 1,
" Browns Ferry Nuclear Plant, Units 1, 2, and 3, SAFER /GESTR-LOCA, Loss-of-Coolant Accident Analysis."
The LOCA results for SLO using SAFER /GESTR showed that, with the application of an APLHGR multiplier as proposed in the TS change, the LOCA peak clad temperature for SLO will always be lower than that for limiting design basis pipe break for two-loop operation.
An APLHGR multiplier of 0.9 is applicable for all current fuel types being used.
This multiplier is documented in each cycle-specific reload analysis and included in the COLR.
NEDC-32484P Revision 1 also concludes that the design basis accident (large breaks) are more affected than small break sequences and, therefore, the large break results are bounding for SLO.
The Recirculation Pump Seizure event in SLO was evaluated in NEDO-24236 and shown to be a non-limiting event.
This conclusion is also supported by GE analyses on other BWRs.
In summary, based on the above discussion, the proposed changes for SLO do not increase the probability or consequences of an accident previously evaluated.
B.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Although the proposed change allows extended operation in a configuration that was previously allowed for a limited period, analysis has shown (as described in i
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.___-------_----q item A above), that operation with one recirculation pump out-of-service is within existing analyses based on the proposed TS requirements.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
C.
The proposed amendment does not involve a significant reduction in a margin of safety.
The proposed change to operate in single-loop recirculation mode has been analyzed in accordance with established transient and accident methodologies, and margins of safety for the design basis accidents and transients anslyzed in Chapter 14 of the BEN UFSAR have not been significantly reduced.
The basis for this conclusion is outlined in item A above.
Therefore, the proposed change does not involve a significant 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 any effluents that may be released off-site, 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.
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VI.
REFERENCES 1.
Letter from NRC to TVA, dated September 19, 1978, Amendment 41, Browns Ferry Nuclear Plant Unit 1, Single-Loop Operation 2.
Letter from NRC to TVA, dated September 29, 1978, Amendment 43, Browns Ferry Nuclear Plant Unit 1, 1
Single-Loop Operation 3.
NRC Generic Letter 86-09, Technical Resolution of Generic Issue B-59-(N-1) Loop Operation in BWRs and PWRs, March 31, 1986 4.
Letter from TVA to NRC dated June 21, 1996, Browns Ferry Nuclear Plant (BFN) - Units 1, 2,
and 3, Technical Specification (TS) 377 - Change in Safety Limit Minimum Critical Power Ratio (SLMCPR) and Revision to Bases Description of Residual Heat Removal Supplemental Fuel Fool Cooling Mode 5.
Letter from TVA to NRC dated February 7,
- 1997, Browns Ferry Nuclear Plant (BFN)- Units 1, 2,
and 3, Technical Specifications Change (TS) 377 - Change in Safety Limit Minimum Critical Power Ratio (SLMCPR) and Revision to Bases Description of Residual Heat Removal Supplemental Fuel Pool Cooling Mode - Supplemental Information 6.
Letter from TVA to NRC dated March 6, 1997, Browns Ferry Nuclear Plant (BFN) - Units 1, 2,
and 3,-
Technical Specifications (TS) Change 353R1 - Power Range Neutron Monitor (PRNM) Upgrade with Implementation of Average Power Range Monitor (APRM) and Rod Block Monitor (RBM) TS (ARTS) Improvements and Maximum Extended Load Line Limit (MELLL) Analyses -
Revision 1 7.
Letter from TVA to NRC dated April 11, 1997, Browns i
Ferry Nuclear Plant (BFN), Units 1, 2,
and 3,-
Technical Specifications (TS) Change 353S1 - Power Range Neutron Monitor (PRNM) Upgrade with Implementation of Average Power Range Monitor (APRM) and Rod Block Monitor (RBM) TS (ARTS) Improvements and Maximum Extended Load Line Limit (MELLL) Analyses -
Supplement 1 - Improved Standard Technical j
Specification (ISTS) Format l
i l
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i 8.
Letter from TVA to NRC dated February 23, 1996, Browns Ferry Nuclear Plant (BFN), Units 1, 2,
and 3, Adoption i
of the General Electric (GE) SAFER /GESTR Loss of Coolant Accident Methodology 9.
Letter from TVA to NRC dated March 11, 1997, Browns Ferry Nuclear Plant (BFN), Units 1, 2,
and 3, Adoption of the General Electric (GE) SAFER /GESTR Loss of Coolant Accident Methodology 10.
Letter from TVA to NRC dated September 28, 1978, Additional Information on Single-Loop Operation Analysis, Recirculation Pump Seizure and Idle Loop Startup i
11.
ORNL/TM-9601, Browns Ferry-1 Single-Loop Operation j
Tests, September 1985 12.
Letter from NRC to GE dated April 24, 1985, Acceptance for Referencing Licensing Topical Report NEDE-24011, Rev.
6, Amendment 8,
" Thermal Hydraulic Stability Amendment to GESTAR II" 13.
NRC Bulletin No. 88-07, Supplement 1: Power Oscillations in Boiling Water Reactors (BWRs),
December 30, 1988 14.
General Electric Nuclear Services Information Letter (SIL) No. 380, Revision 1, BWR Core Thermal Hydraulic Stability, February 10, 1984 15.
Letter from TVA to NRC dated June 20, 1989, TVA BEN Technical Specificaticn No. 272 - Thermal-Hydraulic Stability Section 3.5/4.5-M 16.
Letter from NRC to TVA dated October 5,
- 1989, Technical Specification Changes Involving Thermal-Hydraulic Stability, Section 3.5/4.5-M (TAC 73435)
(TS 272) - Browns Ferry Nuclear Plant, Unit 2 17.
NRC Generic Letter 94-02: Long-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors, July 11, 1994 El-28
i 18.
Letter from TVA to NRC dated September 8,
- 1994, Browns Ferry Nuclear Plant (BEN) - Units 1, 2, and 3, Response to Generic Letter (GL) 94 Long-Term Solutions and Upgrade of Interim Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors
)
l 19.
Letter from TVA to NRC dated January 6, 1983, Browns Ferry Nuclear Plant (BFN), Units 1, 2,
and 3, Response to Request for Additional Information, Single-Loop Operation l
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El-29 t
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Browns Ferry Nuclear Plants Units 1,2, and 3 Single-Loop Operation j
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!l BROWNS FERRY NUCLEAR PLANTS 1
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SINGLE-LOOP OPERATION ie
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NEDO-24236 81NED264 Class I May 1981 i
BROWNS FERRY NUCLEAR PLANTS UNITS 1, 2, AND 3 SINGLE-LOOP OPERATION NUCLEAR PO'AER SYSTEMS DivlSION e GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL h ELECTRIC
0-e NED0-24236 V'
.: n Si 1
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT (Please' Read Carefully)
This report was prepared by General Electric solely for the Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory J
Commission (USNRC) for supporting TVA's operating license of the Browns Ferry Nuclear Plants Units 1, 2 and 3.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General
)
Electric at the time this report was prepared.
1 The only undertakings of the General Electric Company respecting infor-mation in this document are contained in the General Electric Company Single-Loop Operation Analysis Proposal 414-TY37-ERO (GE letter No.
G-ER-9-34, dated April 27, 1979) and TVA Purchase Contract 79P64-164185, dated August 31, 1979. The use of this information except as defined by said proposal and contract, or for any pur. pose other than that for l
which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result
-from such use of such information.
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NEDO-24236 CONTENTS
[
I
[ ale 1.
INTRODUCTION AND SUINARY 1-1 2.
MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT 2-1 l
2.1 Core Flow Uncertainty 2-1 2.2 TIP Reading Uncertainty 2-4 l
3.
MCPR OPERATING LIMIT 3-1 31 Core Wide Transients 3-1 3.2 Rod Withdrawal Error 3-2 3.3 Operating MCPR Limit 3-3 4.
STABILITY ANALYSIS 4-1 5.
ACCIDENT ANALYSES 5-1 l
5.1 Loss-or-Coolant Accident Analysis 5-1
)
5.2 One-Pump Seizure Accident 5-9
-6.
REFERENCES 6-1 i
4 r.
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' NEDO.-24236 ILLUSTRATIONS Figure Title g
2-1 Illustration of Single Recirculation Loop Operation Flows 2-5 3-1 Main Turbine Trip with Bypass Manual Flow Control 3-4 4-1 Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 4-2 5-1 Browns Ferry 1 Suction Break Spectrum Reflood Times 5-3 5-2 Browns Ferry 1 Discharge Break Spectrum Reflood Times 5-4 5-3 Browns Ferry 2 Suction Break Spectrum Reflood Times 5-5 5-4 Browns Ferry 2 Discharge Break Spectrum Reflood Times 5-6 5-5 Browns Ferry 3 Discharge Break Spectrum Reflood Times 5-7 5-6 Browns Ferry 3 Suction Break Spectrum Reflood Times 5-8 TABLES Table Title M
5-1 Limiting MAPLHGR Reduction Factors 5-2 l
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NEDO-24236 1.
INTEIl0CTIM AND SINGERY The current technical specifications for the Browns Ferry Nuclear Plants, Units 1, 2, and 3 do not allow plant operation beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if an idle recirculation j
loop cannot be returned to service. Technical Specification 3.6.F.3 for each unit provides that if the pump cannot be made operable after this period of j
time, the plant must be placed in hot shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
+
The capability of operating at reduced power with a single recirculation loop t
is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pump or other component takes longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and one loop is rendered inoperative. To justify single-loop operation, the safety analyses documented in the Final Safety Evaluation Reports and Reference 1 were reviewed for one-pump operation.
Increased uncertainties in the core total flow and TIP readings result in an 0.01 incremental increase in the MCPR fuel cladding integrity safety limit during single-loop operation.
These uncertainties are compensated for by adding 0.01 to the MCPR operating limit for single-loop operation. No other increase in this limit is required, as core-wide transients are bounded by the rated power / flow analyses performed for each cycle, and the recirculation flow-rate dependent rod block and scram setpoint equations given in the technical specifications are adjusted for one-pump operation. The least stable power / flow condition (natural circulation) is not affected by single-loop operation. Under single-loop operatior., the flow control should be in manual, since control oscillations may occur in the recirc-ulation flow control system under abnormal conditiot.. Derived MAPLHGR reduction factors for single recirculation pump operation are tabulated below:
Fuel Type 7x7 8x8 8x8R P8x8R 0.70 0.83 0.82 0.82 The analyses were performed assuming the equalizer valve was closed, in accordance with normal valve lineup for operation at Browns Ferry. The discharge valve in
}
the idle recirculation loop should be operable and able to close in order to main-l tain the LOCA mitigating systems for Browns Ferry. Alternatively, suction valves in the idle recirculation loop may be closed to prevent the loss of low pressure coolant injection (LPCI) flow out of a postulated break in the idle suction line.
1-1/1-2
NEDO-24236 2.
ICPR FUEL CLADDIEG INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the statis-tical analysis to determine the MCPR fuel cladding integrity safety limit are not dependent on whether coolant flow is provided by one or two recirculation pugs. Uncertainties used in the two-loop operation analysis are documented c
in Table 5-1 of Beference 1 for reloads. A 65 core flow measurement uncertainty l
has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 2.
The random noise component of the TIP reading uncertainty was revised for single recirculation loop operation to reflect operating plant test results given in Subsection 2.2 below. This revision resulted in a single-loop operation process computer uncertainty of 9.1% for reload cores. A comparable two-loop process computer uncertainty value is 8.7% for reload cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.
2.1 CORE FLOW UNCERTAIETY 2.1.1 Core Flow Measurement During Single-Loop Operation The jet pump core flow measurement system is calibrated to measure core flow when both banks of jet pumps are in forward flow; total core flow is the sum of ',he indicated loop flows. For single-loop operation, however, the inactive jet puws will be backflowing. Therefore, the measured flow in the backflowing jet pumps must be subtracted from the measured flow in the active loop.
In add ition, the jet pump flow coefficient is different for reverse flow than for forward flow, and the measurement of reverse flow must be modified to account for this difference.
For single-loop operation the total core flow is derived by the following formula:
l
[ActiveLoop {
!InactiveLoop I Total CoreI
-C l
\\
Flow /
(Indicated Flowj (Indicated Flow) 2-1
l 1
NEDO-24236 where C (= 0.95) is defined as the ratio of " Inactive Loop True Flow" to "Inac-tive Loop Indicated Flow," and " Loop Indicated Flow" is the flow indicated by the jet pump " single-tap" loop flow summers and indicators, which are set to indicate forward flow correctly.
j The 0.95 factor was the result of a conservative analysis to appropriately modify the single-tap flow coefficient for reverse flow."
If a more exact,
less conservative core flow measurement is required, special in-reactor cali-l bration tests would have to be made. Such calibration tests would involve calibrating core support plate 6P versus core flow during two-pump operation along the 100% flow control line, operating on one pump along the 100% flow control line, and calculating the correct value of C based on the core flow derived from the core support plate AP and the loop flow indicator readings.
2.1.2 Core Flow Uncertainty Analysis f
The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump operation is essentially the same as for two-pump operation, except for some extensions. The core flow uncertainty analysis is described in Befer-ence 2.
The analysis of one-pump core flow uncertainty is summarized below.
1 For single-loop operation, the total core flow can be expressed as follows (refer to Figure 2-1):
WC:WA-WI where WC = total core flow; WA = active loop flow; and WI = inactive loop (true) flow.
f By applying the " propagation of errors" method to the above equation, the vari-ance of the total flow uncertainty can be approximated by:
l
'The expected value of the "C" coe fficient is JO.88.
i 2-2
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1 O
D l
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i l
C*
sys 1
Arand 1a Irand where owc uncertainty of total core flow;
=
s
(
Ow uncertainty systematic to both loops.
=
3y3 l',
OgA random uncertainty of active loop only*.
=
rand Owy random uncertainty of inactive loop only;
=
U uncertainty of "C" coefricient ; and j
C
=
l ratio of inactive loop flow (W ) to active loop flow (W )-
a
=
I A
1 Resulting from an uncertainty analysis, the conservative, bounding values of og O
and DC are 1.6%, 2.6%, 3.55, and 2.8%,
,OwI
- 8Y8, ggrand rand respectively.
Based on the above uncertainties and a bounding value of 0 36 for a "a," the variance of the total flow uncertainty is approximately:
2 f
(2.6)2
[1-0.36 C = ( 1.6)2 [
0.36 (3 5)2 + (2.8)2
( 5.0% )2, 2
1 c
=
W g1-0.36 j
(
When the effect of 4.1% core bypass flow split uncertainty at 12% (bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant flow uncertainty is:
ofetive = (5.0%)2 0.12 (4.1%)2 = (5.0%)2 2
coolant which is less than the 65 core flow uncertainty assumed in the statistical analysis.
2-3 i
I
NEDo-24236 In summary, ccre flow during one-pump operation is measured in a conservative way, and its uncertainty has been conservatively evaluated.
2.2 TIP READIEG'UNCERTAIETY To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was performed at a power level 59.3% of rated with a single recirculation pump in operation (com flow 46.3% of rated).
A rotationally symmetric control rod pattern existed prior to the test.
Five consecutive traverses were made with each of five TIP machines, giving a total of 25 traverses. Analysis of their data resulted in a nedal TIP noise of 2.855. Use of this TIP noise value as a component of the process computer total uncertainty results in a one-sigma process computer total uncertainty value for single-loop operation of 9.1% for reload cores.
i 2-4 l
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i NEDO-24236 I
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/
WC W
WA l
TOTAL CORE FLOW W
=
C ACTIVE LOOP FLOW W
4 W,
INACTIVE LOOP FLOW Figure 2-1.
Illustration Of Single Recirculation Loop Operation Flows 2-5/2-6
ICPR OPERATIEG LIMIT 31 CORE WIDE TRANSIENTS Operation with one recirculation loop results in a maximum power output which is about 30% below that which is attainable for two-pump operation. Therefore, the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-loop operational mode.
!e For pressurization, flow decrease, and cold water increase transients, previously i
transmitted Reload /FSAR results bound both the thermal and overpressure conse-quences of one-loop operation.
l l
Figure 3-1 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be seen, the consequences of the tran-l sient during one-loop operation are considerably less because of the associated reduction in operating power level.
1 l
The consequences from flow decrease transients are also bounded by the full power analysis. A single pump trip from one-loop operation is lese severe than a two-pump trip from full power because of the reduced initial power level.
Cold water increase transients can result from either recirculation pump speedup or restart, or introductim of colder water into the reactor vessel by events such as loss of feedwater heater. The Kr factors are derived assuming that both recirculation loops increase speed to the maximum permitted by the M-G 1et scoop tube position. his condition produces the maximum possibet power increase and hence maximum ACPR for transients initiated from less tha t rated power and flow. When operating with only one recirculation loop, the flow and power increase associated with the increased speed en only one M-G set
'will be less than that associated with both pumps increasing speed; therefore, the Kr factors derived with the two-pump assumption are conservative for single-loop operation. Inadvertant restart of the idle recirculation pump would result l
in a neutrm flux transient which would exceed the flow reference scram. The l
resulting scram is expected to be less severe than the rated power / flow case documented in the FSAR. he latter event, loss of feedwater heating, is gener-ally the most severe cold water increase event with respect to irscrease in l
core power. This event is caused by positive reactivity insertion from core 3-1 l
NEDO-24236 flow inlet subcooling; therefore, the event is primarily dependent on the initial power level.
The higher the initial power level, the greater the CPR change during the transient. Since the initial power level during one-pump operation will be significantly lower, the one-pump cold water increase case is conser-vatively bounded by the full power (two-pump) analysis.
From the above discussions, it can be concluded that the transient consequence from one-loop operation is bounded by previously submitted analyses.
l 3.2 ROD WITHDRAWAL ERROR j
The rod withdrawal error at rated power is given in the FSAR for the initial i
core and in cycle dependent reload supplemental submittals. These analyses are performed to demonstrate that, even if the operator ignores all instrument 1
indications and the alarm which could occur during the course of the transient, I.
the rod block system will stop rod withdrawal at a minimum critical power ratio
)
which is higher than the fuel cladding integrity safety limit. Correction of
{
the rod block equation (below) and lower power assures that the MCPR safety limit is not violated.
One-pump operation results in backflow through 10 of the 20 jet pumps while i
i the flow is being supplied. into the lower plenum from the 10 active jet pumps.
Because of the backflow through the inactive jet pumps, the present rod block equation was conservatively modified for use during one-pump operation because
{
the direct active-loop flow measurement may not, without correction, indicate actual flow above about 35% total drive flow.
A procedure has been established for correcting the rod block equation to account for the discrepancy between actual flow and indicated flow in the active loop.
This preserves the original relationship between rod block and actual effective drive flow when operating wid1 a single loop.
l f
The two-pump rod block equation is:
l RB = mW + (RB100 - m(100))
l l
3-2
r NEDO-216236 The one-pump equation becomes:
RB = mW + (RB100 - m(100)) - mow where 6W = difference, determined by utility, between two-loop and single-loop l
effective drive flow when the active loop indicated flow is the same; RB = power at rod block in %;
m = flow reference slope for the rod block monitor (RBM), and W = drive flow in 5 of rated.
RB100 = top level rod block at 100% flow.
}
If the rod block setpoint (RB100) is changed, the equation must be recalculated using the new value.
The APRM trip settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram trip settings are subject to the same procedural changes as the rod block monitor trip setting
{j discussed above.
l 33 OPERATING MCPR LIMIT For single-loop operation, the rated condition steady-state MCPR limit is in-creased by 0.01 to account for the increase in the fuel cladding integrity safety limit (Sec tion 2).
At lower flows, the steady-state operating MCPR limit is conservatively established by multiplying the rated flow steady-state limit by the Kr factor. This ensures that the 99.9% statistical limit require-ment is always satisfied for any postulated abnormal operational occurrence.
3-3
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1020 1000 980 RANGE OF EXPECTED 7
-4 MAXIMUM 1 LOOP POWER OPER ATION i
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20 40 60 80 100 120 140 l
POWER LEVEL (% NUCLEAR BOILER RATED) l h
Figure 3-1.
Main Turbine Trip with Bypass Manual Flow Control l
3-4
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i NEDO-24236 0
4.
STABILITY ANALYSIS l
l The least stable power / flow condition attainable under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This l
condition may be reached following the trip of both recirculation pugs.
As
(
shown in Figure 4-1, operation along the minimum forced recirculation line with one pump running at minimum speed is more stable than operating with natural cir-culation flow only, but is less stable than operating with both pugs operating at minimum speed. Under single-loop operation, the flow control should be l
in manual, since control oscillations may occur in the recirculation flow control system under abnormal conditions.
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4-1 1
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4 NEDO-24236 i
1.2 4
ULTIMATE STABILITY LIMIT 9
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SINGLE LOOP, PUMP MINIMUM SPEED a
- - BOTH LOOPS, PUMPS MINIMUM SPEED i
0.8 l
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0.6 c:
6 NATURAL CIRCULATION RATED FLOW W
CONTROL LIN E
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HIGHEST POWER ATTAINAB LE FOR SINGLE LOOP OPERATION 0.2 -
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20 40 60 80 100
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POWER (%)
Figure 4-1.
Decay Ratio Versus Power Curve for Two-Loop and Single-Loop Operation 4-2
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NED0-24236 5.
ACCIDErr ANALYSES l
l l
The broad spectrum of postulated accidents is covered by six categories of design basis events.
These events are the loss-of-coolant, recirculation pump seizure, control rod drop, main steamline break, refueling, and fuel assembly loading accidents. he analytical results for loss-of-coolant and recirculation pump seizure accidents with one recirculation pump operating are given below. The results of the two-loop analysis for the last four events conservatively bound one-pump operation.
f 5.1 LOSS-OF-COOLANT ACCIDENT AEALYSIS l
Single-loop operation analyses utilizing the models and assumptions documented in Reference 3 were performed for Browns Ferry Units 1, 2, and 3 Using this method, SAFE /REFLOOD computer code runs were made for both the suction and dis-charge side breaks.
Figures 5-1 through 5-6 show the variatice2 of reflooding time for the standard two-pump operation analysis and the one-pump operation analysis. We plots show that for Browns Ferry Unit 1, the one-pump operation analysis indicates reflooding times that are similar to the rerlooding times in two-pump operation.
Since the results are similar, the multipliers for the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for Browns Ferry Unit 1 were determined for the discharge and suction break Design Basis Accidents (DBA's) and for discharge breaks of maximum reflooding times as described in Section II.A.7.4 of Reference 3 The one-loop operation MAPLHGR's for each type of fuel can I
be determined by multiplying the standard two-loop operation MAPLHGR's by the appropriate multiplier. As discussed in Reference 3, the one-loop MAPLHGR's I
calculated in this manner result in conservative values.
For Browns Ferry Units 1, 2 and 3, the most limiting break for single-loop opera-i tion was found to be a discharge line break at 65% of the DBA.
For Units 2 and 3, the reflooding time was longer compared to the two-loop operation. Since, l
l for these two units, the reflooding times for the limiting breaks were considered significantly longer than the two-loop reflooding times, an analysis of MAPLHGR using the approved CHASTE model was performed for each unit. The analyses proved the procedure described in Section II. A.7.4 of Reference 3 still is 1
5-1
I
('
NEDO-24236 conservatively applicable to all three units, and analyses for each type of fuel were made at 1005 DBA suction and discharge breaks and at 65% DBA discharge breaks using the methods of Section II.A.7.4 of Reference 3 The limiting MAPLHOR reduction factors are provided in Table 5-1.
The analyses were done assuming the equalizer valve was closed and with the LPCI modified configuration. The discharge valve in the idle recirculation loop is normally closed or operable, but if its closure is prevented, the suction valve in the loop should be closed to maintain normal system alignment for low pressure coolant injection (LPCI).
Table 5-1 LIMITING MAPLHGR REDUCTION FACTORS Fuel Type Reduction Factor 7x7 0.70
)
8x8 0.83 8x8R 0.82 P8x8R 0.82 I
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O 1 LOOP OPER ATION O
i 190 O,
2 iso I
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170 8,
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a~
160 l
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t 150 1.0 ft 1.927 f 2 t
a
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140 10 20 30 40 50 60 70 80 90 100 110
[
BREAK AREA (% OF OBA) 1 Figure 5-2.
Browns Ferrv 1 Discharge Break Spectrum Ref'lood Times
),
NEDO-24236 150 0 1 LOOP OPERATION O 2 LOOP OPERATION 140 4
1 130 4
i 4
_}
w
- E E
E 120 i
5 i
O0 a
b 5
I 1
1 1
110 i
e i
l i
100 i
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1 t
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l I
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I I
l g
20 30 40 50 80 70 80 90 100 110 BREAK AREA (% OF DBA)
Figure 5-3 Browns Ferry 2 Suction Break Spectrum Ref100d Times 5-5
NEDO-24236 300 290 280 0 1 LOOP OPERATION 270 0 2 LOOP OPERATION 280 250 240 j
230 5
I 220 E
5 h
210 -
be 200 -
190 -
180 -
170
~
l l
180 150
-4 i
I I
l jy 50 00 70 80 90 100 110 BREAK ARE A (% OF DBA)
Figure 5-4 Browns Ferry 2 Discharge Break Spectrum Reflood Times 5-6
_ _, =
NEDO-24236 290 280 270 260 01 LOOP OPERATION 250 O 2 LOOP OPER ATION 240 -
230 -
I 220 2*
e 3
210 O
O 3
h 200 too 180 1M 187 150 140 -
_n n
m I
I
[
f l
330 50 80 70 80 90 100 110 BREAK AREA (% OF DBA)
Figure 5-5.
Browns Ferry 3 Discharge Break Spectrum Reflood Times 5-7 I
NEDO-24236 150 j
O 1 LOOP OPEP ATION O 2. LOOP OPERATION i
140 i
n i
13n
)
4 5
C l
w I
s-i e 120 l
E o
e o
3 e
110
)
100 1
I i
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1 l
l
,o 20 30 40 50 60 70 80 90 100 110 BREAK AREA (% OF DBA)
Figure 5-6.
Browns Ferry 3 Suction Break Spectrum Reflood Times 5-8
1 NEDO.-24236 5.1.1 amall Break Peak Cladding Temperature I
Section II.A.7.4.4.2 of Reference 3 discusses the relative insensitivity of the calculated peak clad temperature (PCT) to the assumptions used in the one-pump operation analysis and the duration of nucleate boiling. As this slight increase
(#500F) in PCT is overwhelmingly offset by the decreased MAPLHGR (equivalent l,
to <3000 to 5000F PCT) for one-pump operation, the calculated PCT values for small breaks will be significantly below the 2200 F cladding temperature limit 0
l*
specified in 10CFR50.46.
5.2 ONE-PIBF SEIZURE ACCIDENT l
The one-pump seizure accident is a relatively mild event during two recirculation puw operation as documented in References 1 and 2.
Similar analyses were per-fermed to determine the impact this accident would have on one recirculation pump operation. hese analyses were performed with the models documented in Reference l
1 for Browns Ferry Unit 1 (Reference 4).
The analyses were initialized fro'4 steady-i i
state operation at the following initial canditions, with the added condition of one inactive recirculation loop:
(1) thermal power = 755 and core flow = 585; and (2) thermal power = 825 and core flow = 565.
These conditions were chosen because they represent reasonable upper limits of single-loop operation within existing MAPLHGR and MCPR limits at the same maximum pump speed. Pump seizure was simulated by setting the single operating pump speed to zero instantaneously.
The anticipated sequence of events following a recirculation pug seizure which occurs during plant operation with the alternate recirculation loop out of ser-l vice is as follows:
i I
i 1.
We recirculation loop flow in the loop in which the pump seizure occurs drops instantaneously to zero, I
i 2.
Core voids increase which results in a negative reactivity insertion 4
and a sharp decrease in neutron flux.
}
1 3.
Heat flux drops more slowly because of the fuel time constant.
5-9
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.__..__-..._._.m.
i t
4.
Neutron flux, heat flux, reactor water level, steam flow, and feedwater flow all exhibit transient behaviors. However, it is not anticipated that the increase' in water level will cause a turbine trip 'and result in scram.
It is expected that the transient will terminate at a condition of natural cir-culation and reactor operation will continue. There will be a small decrease l
in system pressure.
I The minimum CPR for the pump seizure accident for Browns Ferry thit 1 was deter-
)
mined to be greater than the fuel cladding integrity safety limit; therefore, no fuel failures were postulated to occur as a msult of this analyzed event.
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These resultt; are applicable to Browns Ferry Units 2 and 3 l
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NEDO-24236 6.
REFERENCES 1
General Reload Fuel Application, General Electric Company, August 1979 (NEDE-24011 -P-A-1 ).
2.
General Electric BWR Thermal Analysis Basis (CETAB): Data, Correlation, and Design Application, General Electric Company, January 1977 (NEDO-10958-A).
3 General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFRSO Appendix K Amendment No. 2 - One Recirculation Loop Out-of-Service, General Electric Company, Revision 1, July 1978 (NEDO-20566-2).
4.
Enclosure to TVA Letter, J. E. Gilleland to T. A. Ippolito, dated September 28, 1978.
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