ML20042G639
| ML20042G639 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 05/11/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20042G638 | List: |
| References | |
| NUDOCS 9005150191 | |
| Download: ML20042G639 (9) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.140 TO FACILITY OPERATING LICENSE N0. OPR-77 TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT. UNIT 1
- r DOCKET N0. 50-327 1.0 INTRODUCTI,0_N By the two applications dated January 12, 1990, the Tennessee Valley Authority (TVA) proposed to-modify the Sequoyah Nuclear Plant, Unit 1 Technical Specifications (TSs) to remove the up(BIT).per head injection system (VHIS) and to deactivate the boron injection-tank The modifications were done at Unit 1 in the current Unit 1 Cycle 4 refueling outage. The modifications to Unit-2 will be done in the upcoming Unit 2 Cycle 4 refueling outage. This evaluation will ~ address only the changes to the Unit 1 TSs; however, the evaluation applies to both units.
For the UHIS removal, TVA proposed the following changes to the TSs: TS 3/4.5.1.2 on the.UHIS would be deleted; Tables 3.4-1, 3.6-1, and 3.6-2 for the reactorcoolantsystem(RCS)pressureisolationvalves, penetrations,and containment isolation valves would be revised; Limiting Condition for Operation-(LCO) 3.2.2.and Surveillance Requirement (SR) 4.2.2.2 would be revised in
- terms of the-peaking factor limit and, as a result of the peaking factor limit revision, Figure-3.2-2 would be revised; LCO 3.5.1.1 for the cold leg injection accumulators would reflect new values for the volume of water and nitrogen cover pressure; and SR 4.5.2.h would be revised for new minimum flow rate
. values for emergency core cooling systems pumps. This is TVA TS Change Reouest 89-25.
For the BIT deactivation, TVA proposed the following changes to the TSs:
the refueling water storage tank boron concentration would be changed in LCOs.3.1.2.5 ano 3.5.5, the volume of the boric acid storage system and the boron concentration of the refueling water storage tank would be changed in
.LCO 3.1.2.6, the reference to boron injection throttle valves will-be changed to charging pump injection throttle valves in SR 4.5.2.g, TSs 3/4.5.4.1 and 3/4.5.4.2 for the boron injection system would be deleted, and the boron concentration for the cold leg injection accumulators would be changed in LC0 3.5.1.1. This is TVA TS Change Request 89-26.
There are also proposed changes to the affected TSs listed above and to the index of the TSs.
2.0 EVALUATION
'The evaluations of the proposed changes to the TSs will be given below in Sections 2.1 and 2.2. for the removal of the UHIS and the deactivation of the BIT, respectively.
9005150191 900511 PDR ADOCK 05000327 P
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3m 2.1 - UHIS Removal The UHIS has been the subject of regulatory concerns at Sequoyah (SQil),_
includinganintegrateddesigninspectionandtwolicenseeeventreports.Thb-system was designed to provide additional core cooling during reactor blowdown following a large break loss-of-coolant accident (LOCA).
It was also modeled into the secondary side depressurization transients.c Experience has demonstrated that the UHIS adds to the complexity-.of plant operation, requires additional maintenance and generally reduces plant availability.
Because:of-this, plants such as McGuire and Catawba, and now Sequoyah, have proposed removal of the UHIS, By letter-dated flovember 3,1986, TVA. committed to removei the UHIS before restart from the Unit 1 'and Unit 2 Cycle 4 refueling outages.
In a follow-up letter dated January 12, 1990, TVA proposed an amendment to the Technical
. Specifications (TSs) which would reflect the removal of the UHIS.
To support the request for-UHIS removal, TVA has reanalyzed the following postulated events without credit for core cooling from the UHIS: (1) large'and small break LOCA, (2) transients for a steamline break, and (3) the largest single failed-open steam generator relief, safety, or dump valve.
in performing-these analyses, TVA has considered the effects of the following modifications which will 'also'be implemented during the Cycle 4 refueling outages for Units 1 and 2:
0 !mplementation of the Eagle 21 digital protection system 0 Deactivation of the boron injection tank (see Section 2.2 below) 0.lmplementation of the Vantage 5 hybrid fuel 0 Use of;a new steamline break protection 0 Elimination of a reactor trip on steam flow / feed flow mismatch The staff evaluation of the analyses which support the proposed facility modification for the UHIS' removal and associated TS changes is described in the following sections.
2.1.1 Large Break LOCA Evaluation TVA provided the results of a large break LOCA analysis supporting the request for removal of the UHIS.
In the licensee's submittal only the double end cold leg guillotine (DECLG) breaks were analyzed since they were icentified previously(as limiting cases that result in the highest peak cladding temperature PCT). The DECLG break analysis was performed with a total peaking factor of.2.32, 102% of the core power of 3411 Mwt, and an assumed loss.of offsite power at the beginning of the accident. The effect of varying break discharge coefficients on the peak cladding temperature was studied.
The results of the study showed that the DELCG break with a discharge 2001.2 fent of 0.6 is the worst large break LOCA case resgiting in a PCT of coeffic F which is below the acceptance criterion of 2200 F.
The analysis was performed using Westinghouse Emergency Core Cooling System (ECCS) evaluation models (Ref. 1).
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, In their review of the analysis, the staff found that approved analytical
' methods and computer codes were used to perform calculations, and that the-results showed that the PCT, clad oxidation and hydrogen generation are within the acceptance criteria imposed in 10 CFR 50.46-for LOCA analysis.
LTVA provided a revised table of mass and energy rates used for the containment
-backpressure calculation'as a function.of time during blowdown in the large break LOCA. Removal of the UHIS was included in the containment /LOCA analysis that was submitted by TVA in its letter dated January 12, 1990 for its TS change Request 90-05. The peak containment pressure is 10.9 psi following the large break LOCA. This peak pressure is:below the design value of 12 psi and the staff accepted the containment analysis in its letter dated March 2,1990 approving Amendments 131 and 118 for Units 1 and 2, respectively.
2.1.2 - Snell Break LOCA Evaluation The small break LOCA analysis was performed with the approved computer codes, i.e., (1)-the 110 TRUMP (Refs. 2 and 3) code for the calculation of the transient depressurization of the reactor. coolant system, core power, water-steam mixture height and steam flow'past the uncovered portion of-the core and (2) the LOCTA-lY (Ref. 4)~ code for the PCT analysis. The analysis was done assuming 102% of the core power of 3411 Hwt and a total peaking factor of 2.7.
The total' peaking factor of 2.7 is conservative in. comparison to the proposed TS value of 2.32.
Various break sizes were assumed and the results showed that
.the worst break size is a 3-inch diameter break This break size results in 0
the highest peak gladding temperature of 2105.5 F which is below the acceptance criterion of 2200 F.
The staff concludes that the small break LOCA analysis is acceptable since the approved method was used to show the analytical results to be-within the acceptance criteria in 10 CFR 50.46.
2.1.3 Transient Evaluation TVA used the approved LOFTRAN code (Ref. 5) to reanalyze two plant
.transientswhichwere(1)asteamlinebreakand(2)the-largestsingle
-falled-open steam generator relief, safety, or dump valve.
The THINC code (Refs, 6 and 7.) was used to determine if departure from nucleate boiling (DNB)
= occurred in the core for the steamline break. For the failed open valve transient, the results of the LOFTRAN analysis were evaluated to determine if DNB occurred. The results confirmed that no DNB occurred for either the steamline break or the failed open valve and thus assured no fuel damage resulting from the transients. The THINC code has been used in prior Final Safety Analysis and is therefore acceptable.
The staff concludes that the licensee's transient analysis is adequate and acceptable since an approved method was used and TVA demonstrated that specified acceptable fuel design limits would not be exceeded.
2.1.4 Technical Specification Changes The following is a list of the proposed changes to the TSs associated with the removal of the UHIS in the opplication dated January 12, 1990.
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TS 3/4.2.2, Heat Flux Hot Channel Factor-F (Z) 9 The proposed change would revise the total peaking factor from 2.15 to.2.32 I
and also replace figure 3.2-2, the K(z), i.e., the normalized F (Z) curve.
q The staff finds this acceptable since the ECCS analysis was performed using approved methods and gave acceptable resul'cs for the higher total peaking factor, b.
TS 3/4.4.6.2, Table 3.4-1, Reactor Coolant System Pressure Isolation Valves UHIS valves identified as87-558, 87-559,87-560, 87-561,87-562, 87-563, FCV-87-7, and FCV-87-8 were deleted from the table.
Because the removal of the UHIS results in the deletion of tne UHIS reactor coolant system pressure isolation valves, this change is acceptable.
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-TS 3.5.1.1, Cold Leg injection Accumulators
-The proposed change revises the operable range of water volume between 7615 and 8094 gallons and increases the operable range of nitrogen
'i cover-pressure between 600 and 683 psig.
The changes are consistent with the assumptions of the LOCA analysis supporting the request for removal of the UHIS.
The changes are acceptable.
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TS 3/4.5.1.2, ECCS, Upper Head Injection The specifications associated with the operability and maintenance of the UHIS are being deleted because this system is to bc removed before the j.
restart of the current Cycle 4 refueling outage. This is acceptable, e.
TS 3/4.6.1, Containment Penetration Valves, Table.3'.6-1 The proposed changes reflect the sealing of one UHIS penetration and the reclassification of the remaining UHIS bypass leakage paths to maintenance L
penetrations. Making these changes does not affect the requirements on containment integrity and containment leakage in TS 3/4.6.1.1 and TS 3/4.6.1.2. These changes reflect the removal of the UHIS system and are, therefore, acceptable, f.
TS 3/4.6.3, Containment Isolation Valves, Table 3.6-2 The proposed change reflects the removal of containment isolation valves associated with the UHIS containment penetration.
These changes are acceptable since the UHIS system is to be removed.
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TS 3/4.5.2, ECCS Subsystems, Tavg Greater Than or Equal to 350 F
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The minimum flow rate for all four cold leg injection lines decreases from 3976-to 3931 gallons per minute.
The changes are consistent with the results of the LOCA analysis supporting the request for removal of the UHIS and are, therefore, acceptable.
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TS Bases.
TVA also proposed a change to the TS bases. The staff finds that the change to the bases of the TSs are consistent with the proposed changes to the TSs and is, therefore, acceptable.
L2.1.5 UHIS Removal Conclusions t
-The staff has evaluated TVA's request to remove the UHIS and change the 4
associated TSs. Based on its review of the results of the LOCA and transient analyses provided by TVA, the staff has concluded that there is reasonable assurance that the-ECCS without the UHIS satisfies the performance requirements of 10 CFR 50.46 for Sequoyah. The staff, therefore, concludes that operation without the UHIS poses no undue risk to the public health and safety and is acceptable. The TS changes relating to the UHIS removal are consistent with the analytical results and the removal of the UHIS and.thus are acceptable.
2.1.6-References 1.
"The 1981 Version of the Westinghouse ECCS Evaluation Model Using BASH",
WCAP-11524-A, -Revision 2 (Non-proprietary), March 1987.
2.
"NOTRUMP, A Nodal Transient Small Break and General Network Code", WCAP-10080-A, August 1985.
j 3.
" Westinghouse small Break ECCS Evaluation Model Using the NOTRUMP Code",
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l WCAP-10081-A, August 1985.
4.
"LOCTA-IV Program: Loss-of Coolant Transient Analysis,"WCAP-8305. (Non-Proprietary), WCAP-8301 (Proprietary) June,1974.
5.
"LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A l
(Non-Proprietary), April 1984.
l 6.
" Application of the THINC Program to PWR Design," WCAP-7359-L, August, 1969, (Proprietary), WCAP-7838, January,1972.
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7.
" Application of the THINC IV Program to PWR Design," WCAP-8054, October, 1973, (Proprietary), WCAP-8195, October,1973.
1 2.2 BIT Deactivation In a letter dated January 12, 1990, TVA proposed changes to the TSs for the deactivation of the BIT and the deactivation or removal of its heat tracing..
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.w The' BIT and piping from the charging pump to the reactor coolant is not being removed but the tank will not contain a high concentration of boron as now required by TS 3/4.5.4.1. The heating tracing as required by TS 3/4.5.4.2 will be removed or deactivated. The tank and piping will continue to serve as part of the high head / low flow emergency core cooling injection path to the reactor coolant system using the charging pumps but the boron Will be that of the charging flow and no greater than the concentration of boron in the-borated water in the refueling water storage tank. The BIT bypass line is also being 3
-removed in response to the concerns addressed in NRC Bulletin 88-08 on thermal stresses in-piping connected to the reactor coolant system.
The tank inlet and discharge double isolation valves remain. These valves will continue to be normally; closed and to be opened on a safety injection signal. These valves will ensure no cold water slug from the normal charging system will thermally' stress any piping connected to the reactor coolant system. The' tank and its associated piping will be filled and vented prior to being declared operable for power operation and are isolated from sources of noncondensables.
In performing the evaluation for removing the DIT function, TVA also considered
-the effects of the fol_ lowing planned modifications during the Cycle 4 refueling outages for Units 1 and 2:
Elimination of the resistance temperature detector bypass
- Implementation of the Eagle 21 digital protection system Removal of the upper head injection (see Section 2.1 above)
Implementation of the vantage S hybrid fuel
- Use of a new steamline break protection Elimination of the reactor trip on steam flow / feed flow mismatch The staff's evaluation of these proposed TS changes for deactivating the BIT is described in the following section.
F 2.2.1 Evaluation The BITS were originally incorporated into Westinghouse-designed plants as a means of mitigating the consequences of accidental depressurization of the main steam system events. The sole purpose of the BIT, as a component of the safety. injection system, is to insert concentrated boric acid (i.e., 20,000 ppm) into the reactor vessel and thus create a negative reactivity during accidents.
Problems and safety concerns associated with the BIT were identified in NRC Generic Letter 85-16. The high concentration of boric acid imposes operational
' and maintenance problems that adversely affect plant availability)such as (1) minimum volumes and concentrations in boric acid system tanks, (2 heat tracing malfunctions, (3) BIT valve testing, and (4) recovery f rom an inadvertent safety injection. The high boric acid concentrations also cause a safety concern involving boric acid solidificotton that renders emergency core cooling
. inoperable. Therefore, many plants such as Beaver Valley, Byron /Braidwood, Turkey Point, McGuire, and Catawba have removed the BIT.
TVA has decided to deactivate the BIT and the associated heat-tracing systems from SQN Units 1 and 2 during the Cycle 4 refueling outages for each unit.
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.r:3 TVA performed safety analyses for (1) a steamline break, with or without
- offsite power available, for the largest double-ended rupture of a steam pipe upstream and downstream of a flow restrictor, and (2) the largest single failed-open steam generator relief, safety, or dump valve with or without offsite power available. The staff's acceptance criterion for a main steamline break is that the radiological release should not exceed the limits-set forth in 10 CFR Part 100.
The stuck open relief valve. analysis is an event in w'hich the plant may return to criticality with the acceptance criterion being that the specified acceptable fuel design limits should not be violated.
The analyses were performed by TVA using the NRC-approved method and the computer code 10FTRAN. To minimize. future TS changes. TVA selected the highest possible boron concentration that would (1) acconmodate the removal of the BIT, (2)'acconnodate removal of the. upper head injection, (3) meet the requirements
-for the post-loss of coolant accident sump potential hydrogen-ion activ.ity, and (4)' provide the maximum available margin for future reloads. The BIT was assumed in the analysis to be at o zero ppm concentration without heat tracing.
As boron was injected from the refueling water storage tank, the BIT acted in the analysis as a dilution volume, reducing the effectiveness of the boron in.
~ the refueling water storage tank (RWST).
The heat tracing for the BIT was used only to keep the temperature of the water in the BIT high enough to keep the high concentraticn of boron in solution.
With the BIT boron concentration being reduced to at or below the proposed concentration in the refueling water storage tank, no heat tracing is needed
.for the BIT.
'TVA stated that the design basis for the departure from the nucleate boiling ratio would be met for all cases analyzed. No fuel failures were predicted.
Thus, the releases resulting from the stuck open relief valve analysis would comply. with the 10 CFR, Part 20 criteria. Even for the ANS Condition IV main steamline break event, using the final safety analysis report criteria of a conservative fuel failure level of 1 percent, the radiological consequence complies,with.10 CFR Part 100 criteria.
TVA stated that no valves in the safety injection line through the BIT are being replacea. There are also no changes needed to be made to the engineered safety feature response times specified in the TSs due to the deactivation of the BIT.
2.2.2 Technical Specification Changes The following are the proposed changes to the TSs associated with the deactiva-tion of the BIT and the removal of its associated heat tracing.
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Index and TS 4.5.2.g.2 and TS 3/4.5.4 The proposed changes are editorial and reflect the deactivation of the BIT and the deactivation or removal of its heat tracing.
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TS 3.1.2.5.b.2 - RWST, Modes 5'ond 6 4
1The minimum boron concentration is increased from 2000 to 2500 ppm.
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-TS 3.1.2.6.b.2 and TS 3.5.5.b - RWST, Moces 1,2,3, and 4 The boron ccacentration range of 2000 to 2100 ppm is increased to a range of 2500 to 2700 ppm..
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TS 3.1.2.6.a.1 - Boric Acid Storage System The minimum volume of the borated water _is increased from 6542 to 7176
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gallons #
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TS 3.5.1.1.c - Cold-Leg injection Accumulators The boron concentration is increased from a range of 1900 to 2100 ppm to a' range of 2400 to 2700 ppm.
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TS Bases TVA also proposed a change to the TS bases. The change is consistent with the other proposed. changes to the TSs and the deactivation of the-BIT.
2.2.3 BIT Deactivation Conclusions The staff has-reviewed TVA's justification for deactivation of the BIT, o,
I deactivation or removal of its_ associated heat tracing, and the proposed TS changes. As NRC ap; mved methods were used for the analysis of the BIT deactivation and the results conform to the acceptance criteria, the proposed TS changes-aro' acceptable.
2-3 Conclusions Based on the staff's review of the-two applications dated January 12, 1990 for changes to the Unit 1 TSs to reflect (1) the removal of the UHIS and (2) deactivation of the BIT at-Unit 1 in the current Unit 1 Cycle 4 refueling outage, the staff concludes that these two actions and the proposed TS changes associated with these actions are acceptable. The proposed changes for the Unit 2 TSs will issued during the upcoming Unit 2 Cycle 4 refueling outage, i
The evaluation for Unit I applies to Unit 2.
The other modifications being proposed for the Cycle 4 refueling outage are
~ being reviewed separately by the staff; and the staff approval of the UHIS removal and the BIT deactivation does not in itself constitute acceptance of these other modifications.
3.0 ENVIRONMENTAL CONSIDERATI0J p
This amendment involves a change to a requirement with respect to the L
installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.
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1 The staff has determined that the amendment involves no significant increase
' n the amounts, and no significant change in the types, of any effluents that -
i may be released offsite, and-that there is no significant increase in individual-i or cumulative occupational radiation exposure. The Commission has previously-issued a proposed finding that this emendment involves no significant hazards consideration and there has been no public coment c3 finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR:51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact
- statement.nor environmental assessment need be prepared in connection with the-
- issuance of this' amendment..
- 4' 0 CONCLUSION-
-The Commission made proposed determinations that the amendment based-
-on TVA's-applications TS 89-25 and TS 89-26 involves no significant hazards consideration. These determinations for TS.89-25 and 89-26 were published in-
~ the Federal Register (55 FR 4279 and 55 FR 4280, respectively) on FebruaryL7, t
1990.. The Commission-consulted'with the State of Tennessee.
No public comments were received and the State of Tennessee did:not have any comments.
The staff has conc'luded, based on the considerations-discussed'above, that:
(1) there is reasonably ast;urance that the health and safety of the will not'be endangereo by operation in the proposed manner, and (2) public such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will not be inimical to the common defense-
-and ' security or to the health and safety of the public.
Principal Contributor:
D. Katze and J. Donohew Dated: May 11, 1990 4
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