ML20050T802
| ML20050T802 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/29/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20050T794 | List: |
| References | |
| NUDOCS 8204150005 | |
| Download: ML20050T802 (16) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 51 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 3 DOCKET NO. 50-296 Author:
Dick Clark 1.0 Introduction By letter dated December 9,1981 (TVA BFNP TS 170) (1), the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifica-
.tions (Appendix A) appended to Facility Operating License No. DPR-68 for the Browns Ferry Nuclear Plant, Unit No. 3.
The proposed amendments and revised Technical Specifications were to:
- 1) incorporate the limiting.
conditions for operation associated with the fifth fuel cycle and 2) rcflect changes resulting from design, equipment and procedural modifications made during the current refueling outage.
2.0. Discussion and Evaluation
- 2.1 Reload Discussion Browns Ferry Unit No. 3 (BF-3) shutdown for its fourth refueling on October 30, 1981 with a scheduled restart date of mid-March 1982. The initial core loading for BF-3 consisted of 764 of the single water rod 8x8 fuel assemblies, l
each containing 63 fuel rods.
During the first refueling in September 1978, 208 of the fuel assemblies were replaced with 8x8R fuel assemblies con-l l
taining 62 fuel rods in each.
During the second refueling outage starting in August 1979, an additional 144 of the initial fuel bundles were replaced with P8x8R fuel assemblies, each containing 62 fuel rods.
During the third refueling outage, which extended from November 23, 1980 to January 17, 1981, an additional 124 of the original 8x8 fuel assemblies were replaced with a like number of new P8x8R fuel assemblies. The prepressurized fuel assemblies (P8x8R) are essentially ide.ntical from a core physics l
standpoint to the two water rod fuel assemblies (8x8R) except that they are prepressurized with about three atmospheres rather than one atmosphere I
of helium to minimize fuel clad interaction. During the. current refueling outage, an additional 272 of the P8x8R fuel assemblies will be loaded (160 of the P80RB299 and 112 of the P80RB284Z) along with 8 lead test assemblies (LTAs).
With this reload, all but 8 of the original one-water-rod fuel assemblies will be replaced with improved fuel bundles and these 8 will be symmetrically located at the four peripheral corners of the core.
In support of this reload application 'for BF.-3, TVA submitted yith its application of Decembpg)9,1981 a supplemental reload analysist21 and a revised ECCS analysisk prepared by the General Electric Company (GE) for TVA.
8204150005 820329 PDR ADOCK 05000296 P
e-2 This is the first reload for BF-3 in which the pressurization transients are calculated by GE's ODYN Code (in place of analyses previously performed by the REDY Code).
Our generic letters of November 4,1980 and January 29, 1981 (Reference 6) required that any reload submittals received after February 1,1981 must contain appropriate ODYN analyses.
(The most recent reload submittal for Browns Ferry Unit No. I was analyzed with the ODYN Code in accordance with our requirement. Unit 1 started up in Cycle 5 on October 1,1981 after a 6-month outage.)
As noted above, this reload involves loading of prepressurized GE 8 x 8 retrofit (P8 x 8R) fuel.
This is the same type of fuel as was loaded during the last reloads for all three Browns Ferry Units. The description of the nuclear and mechanical designs of 8 x 8 retrofit fuel is contained in Reference 4.
Reference 4 also contains a complete set of references to topical reports which describe GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and information regarding the applicability of these methods to cores containing a mixture of fuel. The use and safety implications of prepressurized fuel are presented in Reference 4 and have been found acceptable per Reference 5 (enclosed in Appendix C of Reference 4).
Values for plant-specific data such as steady state operating pressure, core flow, safety and safety / relief valve setpoints, rated thermal power, rated steam flow, and other design parameters are provided in Reference 4.
Additional p] ggt and cycle dependent information is provided in the reload application \\ / which closely follows the outline of Appendix A of Reference 4.
Reference 5 includes a description of the NRC staff's review, approval,d plant-and conditions of approval for the plant-specific data. The above-mentione specific data have been used in the transient and accident analyses provided wth the reload application in compliance with Reference 5.
s Our safety evaluation of the GE generic reload licensing topical report has also concluded that the nuclear, and mechanical design of the 8x8R and P8x8R fuels, and GE's analytical methods for nuclear and thermal-hydraulic calculations as applied to mixed cores containing 7x7, 8x8, 8x8R and P8x8R fuels, are acceptable. The staff's safety evaluation on the ODYN Code (Reference 6) concluded that this model more accurately and conservatively predicted pressure, neutron flux and aCPR during pressurization transients (e.g., turbine trip) than the REDY Code and was acceptable for analyzing these transients in i
supplemental reload licensing submittals.
2.2 Reload Evaluation Because of our previous review of a large number of generic considerations related to use of 8X8R and P8X8R fuels in mixed core load.ings, and on the basis of the evaluations which have been presented in Reference 4, only a limited number of additional areas of review needed to be included in this safety evaluation report. The areas evaluated were the proposed operating e
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3 limit minimum critical power ratios (0LMCPRs), the proposed MAPLHGR limits, the overpressurization analysis, the stability analyses, the control rod drop analyses, shutdown margins and the loss of coolant accident analyses.
For evaluations of areas not specifically addressed in this safety evaluation report, the reader is referred to Reference 4 For Cycle 5, 272 fresh pressurized type P8X8R fuel bundles will be loaded
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into the core. The remainder of the fuel bundles in the core will be a combination 8X8, 8X8R and P8X8R fuel bundles exposed during the previous four cycles.
The fresh fuel will be loaded and the previously peripheral fuel will be shuffled inward so as to constitute an octant-symmetric core pattern, which is acceptable.
2.2.1 Thermal Hydraulics Based on the data provided in Sections 4 and 5 of Reference 2, both the control rod system and the standby liquid control system will have an acceptable shutdown capability during Cycle 5.
As stated in Reference 4, for BWR cores which reload with GE's retrofit 8x8R fuel, the safety limit minimum :ritical power ratio (SLMCPR) resulting from either core-wide or localized abnormal operational transients must be equal to at least 1.07.
When meeting this SLMCPR during a transient, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.
Various transient events can reduce the MCPR from its normal operating level.
To assure that the fuel cladding integrity safety limit MCFR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed for this reload by the licensee, in order to determine which event results in the largest reduction in the minimum critical power ratio.
These events have been analyzed for both the exposed 8X8, 8X8R, and P8X8R fuel and the fresh P8X8R fuel. Addition of the largest reductions in critical power ratio to the safety limit MCPR establishes the operating limits for each fuel type.
The transient events analyzed were load rejection without bypass, feedwater controller failure, loss of 1000F feedwater heating, control rod withdrawal i
error and rotated bundle error. These events were analyzed with the ODYN Code.
l for this relcad.
l The calculated system responses and reductions in CPR during each of the i
operational transients have been provided in Sections 9 and 10 of the GE.
Supplemental Reload Licensing Submittal (Reference 2). On this topic, it is acceptable if fuel specific operating limits are established for prepressurized fuel (Appendix C, Reference 4).
On this basis, the transient analysis results i
are acceptable for use in the evaluation of the operating limit MCPR. Thus, when the reactor is operated in accordance with the proposed operating limit I
l
4 MCPRs, the 1.07 SLMCPR will not be violated in the event of the most severe abnormal operational transient. This is acceptable along with the proposed changes to the Technical Specifications which incorporate the new OLMCPRs calculated by GE to be necessary to protect the fuel during cycle 5 operation.
1 2.2.2 ECCS Appendix K TVA submitted errata and addenda to the BF-3 Loss of Coolant Accident Analysis (3).
with this reload application. The analyses were performed for TVA by GE and evaluate the new reload fuel as well as the 8 LTAs.
The Maximum Average Planar Heat Generation Rate (MAPLHGR) versus Planar Average Exposure for the most limiting break size were calculated by the CHASTE code.
This code is used, with suitable inputs from the other codes, to calculate the fuel cladding heatup rate, peak cladding temperature, peak local cladding oxidation, and core-wide metal-water reaction for large breaks.
The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction.
The empirical core spray heat transfer and channel wetting correlations are built into. CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly.
Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the requirements of 10 CFR 50.46 acceptance criteria.
The MAPLHGR values and peak clad temperature (PCTs) for each fuel type that will be in the BF-3 core during cycle 5 were presented in reference 3 and sub-mitted as proposed changes to the Technical Specifications in TVA's submittal (l).
0
.The maximum PCT calculated for any fuel assembly was only 1790 F - 4100F less than the 2200 F specified in 10 CFR Part 50.46. This maximum PCT is predicted to occur at 15,000 Mwd /t average planar exposure in an older 8X8 fuel bundle-loaded in the first reload. All other fuel assemblies, including the eicht LTAs, are calculated to have lower PCTs throughout the entire cycle.
In NUREG-0630.(" Cladding Swelling and Rupture Models for LOCA Analysis" issued April 1980) the staff recommended that all industry ECCS models be
- evised to incorporate proposed new cladding correlations resulting from the NRC's confirmatory research program.
On May 15,1981, G.E. submitted a generic sensitivity study of fuel rod cladding ballooning and rupture phenomenon during a postulated LOCA.
In the generic study, GE assessed the BWR ECCS sensitivity to rupture temperature by using three rupture temperature models:
(1) the GE CHASTE model, (2) the NUREG-0630 model, and (3) a proposed GE model termed the adjusted model.
For l
the 8X8 type two-water-rod fuel design, GE found that the use of the adjusted model, which may be the best of the three models and which is in fact a 0
combination of the CHASTE and NUREG models, gave a maximum impact on PCT of 110 F.
l l
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e 5
e Inasmuch as any cgmbined uncertainties in the-GE generic study are very much less than the 410 F minimum available margin in the highest PCT, we conclude that the issues of clad rupture and clad ballooning have been adequately accounted for in the LOCA analysis.
We have reviewed the analyses and information submitted for the reload and conclude that BF-3 conforms with all requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 when it is operated in accordance with the Technical Specifications we are issuing with this amendment.
2.2.3 Control Rod Drop Accident For Cycle 5, the key plant-specific and cycle-specific nuclear characteristics for the worst case control rod drop accident (CRDA) occurring during both cold and hot startup conditions are conservatively bounded by the values used in bounding CRDA analyses given in Reference 4.
The results of G.E.'s analysis are presented in Section 15 of the Supplemental Reload Licensing Submittal (2),
The bounding analysis, which includes the adverse effects of fuel densification l
. power spiking, shows that the peak enthalpy will not exceed the 280 ca /gm design limit.
Therefore, for Cycle 5 of BF-3, the peak fuel enthalpy associated with a CRDA from the hot and cold startup condition will also be within the 280 cal /gm design limit.
Thus, we conclude that the peak enthalpy associated with a control rod drop accident occurring from any in-sequence control rod movement will be below the 280 cal /gm design limit.
- 2.2.4 Overpressure Analysis For Cycle 4, the licensee has reanalyzed the limiting pressurization event (MSIV closure followed by neutron flux scram) to demonstrate that the ASME Boiler and Pressure Vessel Code requirements are met for BF-3 The methods used for this analysis, when modified to account for one failed safety valve, have also previously been approved by the staff. The acceptance criterion for this event is that the calculated peak transient pressure not exceed 100%
of design pressure,, i.e., 1375 psig. The reana]pis, which is presented in Section 12 of the supplemental reload submittal W/, shows that the peak pressure at the bottom of the reactor vessel does not exceed 1272 psig for worse case end-of-cycle conditions, even when assuming the effects of one failed safety valve. This is a decrease of 27 psig from the previous fuel cycle and is part of the reason for the changes on pages 30 and 225 of the proposed Technical Specifications. We conclude that there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one valve. Therefore, the limiting overpressure event as analyzed by the licensee is considered acceptable on the bases outlined in Reference 4.
2.2.5 Thermal Hydraulic Stability A thermal-hydraulic stability analysis was performed for this reload using the methods described in Reference 4 The results yhich are presented in Section 13 of the Supplemental Reload Licensing Submittal (21 show that the fuel dependent channel hydrodynamic stability decay ratios and reactor core stability decay ratio
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at the least stable operating state (corresponding to the intersection of the natural circulation power curve and the 105", rod line) are 0.29 (8X8R/P8X8R), 0.37
s..
6 (8X8) and 0.79 respectively. These predicted decay ratios are all Well below the 1.0 Ultimate Performance Limit decay ratio which we have found acceptable.
Prior to Cycle 3 operation, the staff as an interim measure, added a requirement to the BF-3. Technical Specifications which restricted planned plant operation in the natural circulation mode. Continuation of this restriction will also provide a significant increase in the reactor core stability operating margins during Cycle 5.
On the basis of the foregoing, the staff considers the thermal-hydraulic stability of BF-3 during Cycle 5 to be acceptable.
2.2.6 Lead Test Assemblies During the current refueling,eight lead test assemblies (LTAs) will be loaded in the core.
Four test assemblies have an average bundle enrichment of 2.83 weight percent (vs. 2.84 and 2.99 w/o in the regular P8X8R fuel being loaded in the core). The other four test assembl.ies have an average bundle enrichment of 3.14 weight percent.
One of each type will be loaded diagonally across from each other in the same cell in a symmetrical pattern near the center of the core.
The locations of the LTAs is shown in Figure 1 of Reference 2.
The LTA's incorporate the improved features of a third water rod, increased prepressurization to five atmospheres, larger pellet diameter / thinner cladding /
higher stack density, improved upper tie plate, improved spacer design, axial gadolinia distribution, and barrier fuel.
GE used NRC approved Codes, methods and procedures to evaluate the thermal-hydraulic and nuclear characteristics of the LTAs in this reload.
It was also demonstrated that these approved methods are applicable to the L: As.
The results are presented in the Supplemental Reload Submittal (Reference 2).
Present methods and procedures were considered to be adequate to evaluate the l
core response to the Control Rod Drop Analysis (CRDA), Local Rod Withdrawal Error (RWE) and Fuel Loading Errors (FLE) since sufficient nuclear inputs are available to represent the LTA bundles discretely.
For RWE and CRDA, because both analyses are performed for the most limiting error rod, the rod adjacent to the LTAs was also analyzed if its worth plus any additional effects so warranted.
For the CRDA evaluation, the rod drop response for the rod adjacent to the LTAs was significantly lower than the nominal error rod and thus LTAs had no impact on the results given in the reload license submittal (Reference 2).
For RWE, however, the LTA adjacent rod (although not the limiting rod) had a worth sufficient to warrant a second RWE analysis using the LTA adjacent rod as the error rod.
This second RWE analysis is given in an appendix to the reload license submittal (Reference 2).
An evaluation of the potential impact of a rotated bundle error was performed for both the standard reload bundle as well as the LTAs. The results are given in Section 14 and Appendix A of. the Supplemental Reload Submittal (Reference 2).
TVA has committed to perform additional surveillance during loading activities l
to preclude a mislocated or misoriented LTA so that the calculated aCPR for a I
l FLE on these fuel assemblies will not affect the operating limit MCPR. We find j
the analyses and special surveillance procedures on the LTAs to be acceptable.
.e 7
3.0 Chances to Technical Specifications - Reload Our evaluation of the specific changes to the Technical Specifications resulting from the current reload is presented below:
Pages 18, 24, and 178 - This is the first reload for BF-3 in which the j
transients were analyzed by GE's ODYN code as required by us. An additional citation is being added to the technical specifications to reference our approval' of this code for core reloads.
Page 29 - A reference to the NRC approved GE Generic Reload Topical Report, NEDE-240ll-P-A and Addenda, was added to provide further support for the Section 2.1 bc:;es.
Pages vii, 165,176,181,182 and 182a - During the reload 4 refueling, the last l
of the initial core type-2 fuel assemblies will be removed. Therefore, reference to this fuel type is being deleted.
In addition, the exposure limits for the presently installed fuel types have been extended as supported by NED0-24194A.
Finally, MAPLHGR tables have been added for the new fuel types and LTAs loaded in this cycle (i.e., Tables 3.5.I-4, 5 and 6) as discussed in Section 2.2.2 above.
Pages 11, viii,166,167,167a, and 182b - As supported by the reload submittal, the operating limit MCPRs are being changed. Since the MCPRs were determined by the ODYN code (rather than the REDY code), the OLMCPRs are now calculated from two curves rather than being a single value (or a ramp change with fuel exposure).
Our evaluation was covered in Section 2.2.1 above.
' Pages 123,124,126, and 133 - As discussed in Section 2.2.6 above, eight lead test assemblies (LTAs) will be loaded.
In order to obtain additional physics data, special cold criticality tests have been planned for this cycle. These criticality tests require suspension of the rod sequence control system (RSCS) constraints by means of the individual rod bypass switches. This testing is planned as part of the Lead Test Assembly Program in which TVA and GE are participating.
The RSCS is a backup to the Rod Worth Minimizer (RWM).
It independently imposes restrictions on control rod movement to mitigate the effects of a postulated rod drop accident. The RWM, in turn, serves as a backup to procedural controls to limit control rod worth during startup and low power operation.
During low power and startup operation, unrestrained rod patterns can create rods of sufficient worth to exceed design limits on.a Rod Drop Accident (RDA).
Neither the RSCS or RWM are reouired at high power, so both systems are bypassed at greater than 20% power. The RWM is a computer monitoring system which minimizes individual control rod worths by blocking rod movement if the existing control rod pattern deviates from a specific sequence. The sequences are developed by the Plant Nuclear Engineers and loaded into the RWM memory. Actual rod positions are obtained for comparison to the sequence from the Rod Position Information System.
Rod movement sequences are developed to limit rod worth to a i
level below which, if an RDA were to o.ccur at a free-fall rate limited by the velocity limiter, the fuel enthalpy from the transient" would be less than 280 cal /gm.
, - + -
8 The RSCS also imposes restrictions on control rod movements to reduce rod worths, thus reducing the consequences of a postulated RDA. As such, it is a backuo to and complements the RWM. The components of the RSCS are grouped as belonging' to the Sequence Control Mode or the group Notch Control Mode of operation.
The Sequence Control. Mode controls rod movement from rods full in to the 50% rod density by imposing rod select blocks.
The Group Notch Control Mode controls rod movement from the'50% rod density level to the 30% power bypass by imposing rod withdrawal and insert blocks.
The RSCS circuitry does allow ifmited manual bypass cabability.
In the Sequence Control Logic, the full in o_r fully out position for each rod can be bypassed (rod simulated as being full in or full out).
This is necessary for scram time surveillance and system surveillance.
The present Technical Specifications (Section 3.3.B.2) require that the RSCA shall be operable wnenever the reactor is in the startup or run modes below 20% rated power.
Since the RSCS is a backup to RWM, the present Technical Specifications permit (Section 3.3.3.3.C) a second licensed operator to verify that the operator at the reactor console is following the control rod program below 20% rated power if the RWM is inoperable (except during scram time testing).
As noted above, the proposed chang.e to the Technical Specifications to be able to test the LTAs is to permit the RSCS restraints to be suspended by means of the individual rod bypass switches for special criticality tests or control rod scram timing.
If the RSCS is bypassed, the RWM must be operable (i.e., a second licensed operator could not substitute for the RWM).
The purpose of both the RSCS and RWM is to mitigate the effects of a postulated rod drop accident.
Since the RSCS is a backup to the RWM, it is reasonable to permit one or the other to be bypassed under certain controlled conditions as long as the other system is operational.
Furthermore, the constraints imposed by the RSCS and RWM are the results of analyses programmed into the process computer.
For the LTAs, TVA performed an analysis to show that a postulated rod drop accident involving control rods withdrawn during the cold critical test would not exceed the peak fuel enthalpy design limit of 280 cal /gm. The rod worth minimizer (RWM) will be programmed to ensure adherence to the withdrawal sequence specified in the cold critical test procedure.
The RWM must be operable for this test; a second licensed operator may not be used in lieu of the RWM' for this testing.
Based on the analyses and the compensatory actions to be taken when the RSCS is byoassed (i.e., RWM system operable), the proposed changes to the Technical Specifications are acceptable. During the September 1951 reload for Browns Ferry Unit 1, four LTAs were placed in the core.
The proposed changes to the Technical Specifications for BF-3 are the same as those we approved for BF-1 by Amendment No. 76 to the Unit 1 license on September 15, 1981.
BF-1 started up on October 1,1981.
Our experience with the BF-1 startup program reinforces the acceptability of the proposed controls and changes to the BF-3 Technical Specifications.
4.0 Plant Modificat'.ons 4.1 Discussion BF-3 shutdown for the present refueling and maintenance outage on October 30, 1981 is projected to be down for over five months. The reason for the extended cutage is the time needed to complete a number of NRC required modifications as well as the inspections, repairs, surveillance, 'maintenhnce, and other activities r:rmally associated with a refueling outage.
During this shutdown, TVA expects
- cerclete 63 of the 300 plus modifications which NRC has proposed or recuired
9 for operating reactors such as Browns Ferry in various Bulletins, Orders, the TMI-2 Action Plan (NUREG-0737), new regulations, revisions to the Security Plan and Emergency Response Plan, resolution of generic issues,.etc. Some of these modifications require changes to the Technical Specifications prior to startup' and are included in this safety evaluation for convenience.
4.2 Evaluation i
(a) Torus Modifications On January 13, 1981 the Commission issued an Order modifying the BF-3 license to require TVA to promptly institute a reassessment of the containment design for suppression pool hydrodynamic loading conditions and to install any plant modifications needed to confom to the staff's Acceptance Criteria, which are contained in Appendix A to NUREG-0661 (" Safety Evaluation Report, Mark I Containment Long-Term Program" dated July 1980) by March 31, 1982.
This Order was subsequently modified by an Order dated January 19, 1982 extending the time to complete some of the modifications to the cycle 6 outage.
These modifications are required by NRC to restore the originally intended margins of safety in the containment design. The structural I
modifications to the torus containment include addition of torus tiedowns, addition of ring girder reinforcement and reinforcing attached piping nozzles.
Vent system modifications include shortening the downcomers, adding local reinforcement to the vent header, and adding new tie bars to the downcomers.
Attacned piping is being strengthened including modification of the ECCS header support. Many changes are being made to the safety relief valve (SRV) piping system including adding quencher arms to the ramshead, adding quencher am and ramshead supports, adding 10-inch vacuum valves, reinforcing the ring girder at the S.RV hanger attachment, rerouting of piping, and adding new snubbers and supports for the piping. These modifica~-
tions to the torus require changes to the Technical Specifications to account for water displaced by the additional structural steel and to reflect the plant unique analysis which TVA was required to perform to conform the design to the staff's Acceptance Criteria in NUREG-0661. The specific changes to the Technical Specifications are discussed below.
Pages 231 and 285 - The minimum torus water level limits in section 3.7. A.I.a and in the bases for this section are being changed from -7 inches (differential pressure control greater than 0 psid) to -6.25 inches and from -8 inches (0 psid differential, pressure control) to -7.25 inches; a change in each case of 0.75 inch. There are 15-inch by 15-inch sealed box beams being added as support for the safety relief valve lines and HPCI-RCIC internal supports.
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Addition of these supports will result in appreciable water displacement.
Calculations indicate that the box beams and HPCI-RCIC supports will increase the torus water level approximately 3/4-inch due to their presence. This rise in the torus water level is ' reflected in these revised Technica.1 Specification values. The changes, which we have reviewed and approved, are necessary to ensure that the minimum water volume is maintained in the torus for suppression of potential LOCA loads and are acceptable.
I Pages 246 and 286 - In section 3.7.A.6.a (and the bases thereto), the setpoint for the drywell-suppression chamb' r (wetwell) differential pressure control e
(z.P) is being changed from 1.3 psid to 1.1 psid. Downcomer water clearing loads are greatly reduced by physically shortening the downcomers (by almost one foot) and imposing a drywell-wetwell AP.
The Browns Ferry un Mue loads were determined by considering a differential pressure of 1.10 psid at the 3
maximum allowable torus water level.
In order to be consistent with this analysis. the Technical Specification associated with the 2P control has been estat'ished at 1.10 psid.
The char.ces to the Technical Specifications ccnform g
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e to the requirements in Section 2.16, " Differential Pressure Control Requirements,"
in Appendix A t'o NUREG-0661 and are therefore acceptable.
286 - In the bases for the limits established for primary containment,
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Page there is a discussion of steam condensing loads associated with relief valve opera tion. The peak temperature of the torus water used in the evaluation is being changed from 160 F to 200 F local temperature.
During the current refueling 0
0 outage, the T-quenchers are being added to the safety-relief valve discharge device.
In Section 2.13.8 of Appendix A to NUREG-0661 (" Suppression Pool 1 local Temperature Limits") the staff sgecified that "the suppression pot temperature shall not exceed 200 F throughout all plant transients volving The Technical Specifications are being changed to t;nform to SRV operations."
d the staff's acceptance criteria in NUREG-0661 to avoid excessive steam condensing loads and are therefore acceptable.
(b) Replacement of Safety Valves During the present refueling outage, the two presently installed main steam line safety valves are being replaced with two-stage Target Rock safety / relief
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valves (SRVs) identical to the other 11 SRVs.
Thus, BF-3 in the future will have 13 SRVs. The capacities of these valves were factored into GE's "0verpressurization Analysis" in Section 12 of the Supplemental Reload Submittal (Reference 2) as discussed in Section 2.2.4 of this Safety Evaluation.
The value of 83.77 percent total relief capacity is derived from the values of 77.33 p2rcent for 12 SRVs operable out of a total of 13 SRVs. The capacity of 77.33 percent of nuclear boiler rated steam flow, as listed in the BF-3 Reload 4 Supplemental Licensing Submittal, was calculated based on certified valve capacity for a 5.125-inch throat diameter valve (870,000 lbs/ hour at 1,090 +3 psig) issued by the ASME National Board of Boiler and Pressure Vessel The certified values are obtained by testing and are listed as Inspectors.
As noted in 90 percent of the measured capacity values for conservatism.
Section 2.2.4, the licensee's analysis of the limiting overpressure event is Since the number, type, and capacities of the SRVs are specified acceptable.
in the Technical Specificatio,ns (and bases thereto), ch'anges need to be made 26, 27, 30,192, 224 and 225 to reflect 13 rather than 11 SRVs and the on.pages 83.77 percent capacity.
(c) Containment Vent and Purae Modifications 29, 1978, September 27, 1979, and October 22,1979 to Our letters of November all licensees identified concerns regarding containment venting and purging All licensees were requested to implement certain during normal operation.
corrective actions and to evaluate their systems with respect to our positions in Standard Review Plan Section 6.2.4 Revision 1 and Branch Technical Position The TMI Action Plan Requirements, NUREG-0737, Item II.E.4.2, CSB 6-4 Revision 1.
" Containment Isolation Dependability," imposed additional requirements on the design of containment systems.
In our letter to TVA of December 17, 1981 we advised TVA that except for certain areas in which our review had not been completed (e.g., environmental qualification) TVA's proposed actions and modifications satisfactorily resolved Multiplant Action B-24 (Venting and Purging e
m
11
.e Containments While at Full Power) and NUREG-0737 Items II.E.4.2.1 through II.E.4.2.5,provided certain testing requirements were included in the Technical Specifications.
Some of these requirements were included in TVA's submittal of December 9, 1981.
In response to our generic letters of September 27, 1979 and October 22, 1979 to licensees of all light water reactors, TVA is modifying the containment purge system for BF-3 during this outage to satisfy applicable requirements of NRC Branch Technical Position CSB 6-4 regarding valve closure times and addition of debris screens. Pages 263 and 264 are being revised to ref. lect the significant reduction in the maximum allowable operating time.for the purge valves. On the nitrogen purge valves, the operating time is being reduced fece 10 seczds to 5 seconds and on the purge inlet and exhaust isolation valves, the operating time is being reduced frcm 100 and 90 seconds, respectively, to only 2.5 seconds. The faster valve closure times significantly reduce potential offsite doses.
The addition of the debris screens provides protection against foreign material entering the purge ducting and interfering with closure of the purge valves. The changes to the Technical Specifications are those specified in our letter of December 17, 1981 and are acceptable.
(d) Primary Containment Isolation Valves Tables 3.7. A through 3.7.H list the various valves associated with primary containment isolation. Specifically, Table 3.7. A lists the primary containment isolation valves that must be operable during reactor power operation (in accordance with Section 3.7.D of the Technical Specifications) along with the maximum operating tiues and nonnal positipn. Table 3.7.D lists the primary containment isolation valves on which local leak rate tests.must be performed each cycle in accordance with Section 4.7.2.g.
Tables 3.7.E, 3.7.F and 3.7.6 list the stop-check and check valves on the torus and drywell influent lines that must be similarly tested. As discussed below, TVA has proposed revisions to these tables to reflect plant modifications and the requirements in NUREG-0737 Item II.E.4.2.
Table 3.7.A Page 262 - FCV-1-55 and 1-56 drain valves are required to be open for extended periods during power operation. Therefore, these valves will be considered as normally open and technical specification surveillance requirement 4.7.D.l.b will apply'.
Page 263 - TVA has proposed to delete valve FCV-69-12 on the Reactor Water Cleanup System from Table 3.7. A.
This valve is not a containment isolation val ve.
Isolation is provided by check valves69-579 and 3-572.
Based on our review, we find this acceptable.
Page 263 - FCV-73-81, the bypass valve around the HPCI steam supply outboard isolation valve (FCV-73-3), was added to BF-3 during the 1980 refuel outage.
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During quarterly surveillance testing on HPCI isolation valve FCV-73-3 in which the valve is closed and reopened, the steamline downstream from FCV-73-3 is subject to thermal stresses from the closure and subsequent reopening.
This is a one-inch FCV-73-81 was added to relieve those thermal stresses.
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valve.
It is an isolation group 4 valve with a maximum closure time of 10 seconds.
Since the valve was added to the HPCI system, it must be periodically verified as being operable. The addition of this valve to Table 3.7.A is both necessary and acceptable.
Pages 263, 264 and 264A - During the 1980 refueling outage on BF-3, the hydrogen-oxygen analyzer system was replaced with the new Hays-Republic Hydrogen-0xygen analyzer system. Our safety evaluation for this new system was covered in Amendment No. 37 to Facility License No. DPR-68 for BF-3 issued January 12, 1981., All of the system 76 (containment inerting system) valves were installed in the plant during the last outage as part of the new hydrogen-oxygen monitoring system. These valves are being added to Table 3.7. A to require periodic verification that these valves are operable. We have reviewed the changes and find them acceptable.
Page 264 - System suction isolation valves to the drywell air compressors "A" and "B" trains, FCV-32-62 and FCV-32-63, have been installed in the plant as part of a system modification. The drywell air compressors supply air to the air-operated valves in containment.
These valves are being added to Table 3.7. A to reflect this modification and to insure that their operability is periodically verified. The proposed change is acceptable.
Page 264A - The nomal position of FCV-71-7A, 7B is closed rather than open.
The nomal position and action on receiving an initiating signal are being changed to show the correct positions.
Page 265 - The core spray discharge to reactor check valves FCV-75-26 and FCV-75-54 should be included in this table. They are primary containment isolation valves and were inadvertently omitted from this table.
Page 265 - It is proposed to add to Table 3.7. A valves FCV-64-139 and FCV-64-140.
These are the drywell t.P air compressor suction and discharge valves.
They are containment isolation valves and should be verified for operating time.
These valves are a part of the addition of the drywell pressurization system.
Their addition to Table 3.7. A is acceptable.
Page 265 - The following valves are being added to Table 3.7.A because these valves were inadvertently omitted in the original Technical Specifications.
These valves are all containment isolation valves and need to be included in Table 3.7. A to ensure that their required operating times are periodically i
tested.
FCV-90-254A and B, drywell CAMA suction valves FCV-90-257A and B, drywell CAM discharge valves FCV-90-255, drywell CAM suction valve Tables 3.7.0 throuch 3.7.G TVA is proposing to revise Tables 3.7.0 through 3.7.G to be more consistent with the BWR Standard Technical Specifications (NUREG-0123, Rev. 3 issued Fall 1980).
These tables presently list a " Test Medium" (i.e., air, water e
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or nitrogen) and a " Test Method" (i.e., the specific valves between which the test medium is to be applied). The Standard Technical Specificatio'ns (Table 3.6.3-1) do not specify a test method, since this is more appropriately left to the pump and valve testing procedures. TVA has proposed to include the test medium in the title of the tables (i.e.,
separate tables for those valves to.be tested by air vs. water or nitrogen) and to delete the test method. We find the. proposed changes are consistent with the BWR Standard Technical Specifications and are therefore acceptable.
Table 3.7.D (presently p277, proposed pgs 270-272)
The following valves have been added to table 3.7.D:
Valve 76-66 on the new Hays-Republic H -02 analyzer 2
Valve 73-81, bypass valve around HPCI outboard isolation valve 73-3 Test connections were added to BF unit 3 so that the following valves could be tested.
FCV-2-1192 Service Water FCV-2-1383 Service Water FCV-33-1070 Service Air FCV-33-785 Service Air FCV-68-508 CRD to RC Pump Seals FCV-68-523 CRD to RC Pump Seals FCV-68-550 CRD to RC Pump Seals FCV-68-555 CRD to RC Pump Seals
" Valves 74-54 and 74-68 have been added to table 3.7.D due to inadve.rtently omitting them from this table. They are tested and should be included in the Technical Specifications.
Valves76-215 to 76-254 have been deleted from table 3.7.D due to the replacement s # the H -02 analyzer system. Valve 64-141 has also been 2
deleted because it is not an isolation valve and is not tested. Valve 85-573 has been omitted from this table due to a plant modification that eliminated the containment penetration for this CRD return line.
Table 3.7.E (p 279)
Table 3.7.E has been revised to include valve 2-1143 on the demineralized water line into the torus, which is now required to be tested, since.it is an isolation valve.
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NUREG-0737, Item II.K.3.15 TMI Action Plan item II.K.3.15 requires licensees of BWR's to modify pipe-break-detection circuitry so that pressure spikes resulting from HPCI and RCIC initiation will not cause inadvertent system isolation. TVA elected to employ the BWR Owner's Group modification which incorporates a three-second time delay relay (TDR) to prevent spurious isolation.
In our letter to TVA of 4
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l'4 October 13, 1981 we reouested the licensee to provide certain analyses and to " propose the appropriate Surveillance Requirements and Limiting Conditions of Operation for the HPCI and RCIC systems which address this item." The safety evaluation was provided by TVA's letter of December 16, 1981. All of the Browns Ferry units have had a three-second TDR on the HPCI systems.
During the current outage for BF-3, a TDR was added to the RCIC system. The proposed changes to the Technical Specifications requiring calibration and surveillance of the time delay relays was submitted with TVA's application of December 9,1981. Table 4.2.B (p95) is being modified to require a logic system functional test, including calibration, of the RCIC and HPCI system isolation. logic. The changes to the Technical Specifications reflect the surveillance requirements requested in our letter of October 13,1981 on item II.K.3.15 and are acceptable.
5.0 Administrative Chances Browns Ferry and other BWRs are not presently permitted to operate in the natural circulation mode (i.e., without one of the recirculation pumps in operation).
This restriction is presently a paragraph in Section 2.1 (top of og 19) wnich contains the bases for the " Limiting Safety System Settings Related to Fuel Cladding Integrity." This paragraph is being moved, verbatim, to the bases for recirculation pump operation on page 227, which is a more appropriate location. There is no safety significance to this reformatting of the Technical Specifications. Also, the title of Section 3.6.F (p 195), which is presently entitled " Jet Pump Flow Mismatch," is being changed to " Recirculation Pump Operation" which is what the section encompasses.
Section 3.5. A.l.(2) of the Technical Specifications on the Core Spray System (page 146) and Section 3.5.B.l.(2) on the Residual Heat Removal System (page 149) presently contain references to non-applicable sections of the Technical Specifications.
Since these changes do not affect any actual limiting conditions for operation, plant safety is not affected and the non-c l
applicable references are being removed.
Section 3.5.L of the Bases specifies the reporting requirements if any of l
the thermal-hydraulic limits associated with fuel rod inteority (e.g., MAPLHGR, LHGR or MCPR limits) are exceeded. The present Technical Specifications (pages 177 and 178) require that "Each event involving steady state operation beyond a specified limit shall be logged and reported quarterly." Actually, TVA has been notifying NRC promptly of any such incident and has been filing a 30 day LER as for other incidents. This section is being changed to require i
a report within 30 days. Also, on page 178, an additional reference is being added to reflect that this reload was analyzed by the ODYN Code.
6.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in i
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15 any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 151.5(d)(4),
that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
7.0 Conclusion tle have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: March 29,1982 M
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References 1.
Letter, L. M. Mills, TVA to Harold R. Denton, NRC, dated December 9,1981.
2.
" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 3,. Reload 4 (Cycle 5)," Y1003J01 A30, dated November 1981.
3.
Errata and Addenda sheets dated September 1931 to NED0-24194A, " Loss of Coolant Accident Analysis for Browns Ferry Nuclear Plant Unit 3, dated July 1979.
4.
" General Electric Boiling Water Reactor Generic Reload Application,"
NEDE-240ll-P-A, August 1979.
5.
Letter, T. A. Ippolito (USNRC) to R. Gridley (GE), April 16,1979, and enclosed SER.
6.
Generic letter 81-08 to all hol.ders of Construction Permits and Operating Licenses for Boiling Water Reactors,
Subject:
ODYN Code, dated January 29, 1981 forwarding the staff's " Safety Evaluation for the General Electric Topical Report, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, NED0-24154 and NEDE-24154-P, Volumes I, II and II'," dated June 1980.
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