ML20248D502
| ML20248D502 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 08/03/1989 |
| From: | Black S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248D507 | List: |
| References | |
| NUDOCS 8908110062 | |
| Download: ML20248D502 (35) | |
Text
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'o UNITED STATES
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4 NUCLEAR REGULATORY COMMISSION 7;
- E WACHINGTON, D. C. 20565 N,t....,p I.
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i TENNESSEE VALLEY AUTHORITY
'i DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 170 License No. DPR-33 l
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee)datedMay 15, 1989, omplies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
.There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance :cith the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
P.
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L 2.
Accordingly, the. license is amended by changes to the Technical L
Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-33 is hereby amended to read'as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.170, are hereby incorporated in the l
license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION f,
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yv,J zaline black Assistant Director for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 3, 1989 I
i I
i
__.._-____---__________-_-_-._A
ATTACHMENT TO LICENSE AMENDMENT NO. 170 s
FACILITY OPERATING LICENSE NO. DPR-33 DOCKET N0. 50-259 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE INSERT 3.6/4.6-1 3.6/4.6-1*
3.6/4.6-2 3.6/4.6-2 3.6/4.6-26 3.6/4.6-26*
3.6/4.6-27 3.6/4.6-27 3.6/4.6-28 3.6/4.6-28 3.6/4.6-29 3.6/4.6-29 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3.6/4.6-31*
J
L 3.6/4.6-PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY l
i Aeolicability
)
ADolicability i
Applies to the operating status Applies to the periodic I
of the reactor coolant system.
examination and testing
}
requirements for the reactor
{
coolant system.
Obiective Obiective To assure the integrity and safe To determine the condition of j
operation of the reactor coolant the reactor coolant system and system.
the operation of the safety devices related to it.
Specification Specification A.
Thermal and Pressurization A.
Thermal and P g3Jurization Limitations t
Limitations 1.
The average rate of 1.
During heatups and e
reactor coolant temperature cooldowns, the change during normal heatup following parameters or cooldown shall not exceed shall be recorded and 100*F/hr when averaged over reactor coolant a one-hour period.
temperature determined at 15-minute intervals until 3 successive readings at each given location are within 5'F.
a.
Steam Dome Pressure (Convert to upper vessel region temperature) b.
Reactor bottom drain temperature c.
Recirculation loops A and B d.
Reactor vessel bottom head temperature i
e.
Reactor vessel shell adjacent to shell flange BFN 3.6/4.6-1 Unit 1 1
J
p.
3.6/4.6 FRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.A..
Thermal and Pressurization 4.6.A.
Thermal and Pressurization Limitations Limitations 2.
During all operations with 2.
Reactor vessel metal a critical core, other than temperature at the for low-level physics tests, outside surface of the-except when the vessel is bottom head in the vented, the reactor vessel vicinity of the control shell and fluid temperatures rod drive housing and shall be at or above the reactor vessel shell temperature of curve #3 of adjacent to shell Figure 3.6-1.
flange shall be recorded at le.tst every 15 minutes during inservice hydrostatic or leak testing when the vessel pressure is > 312 psig.
f 3.
During heatup by 3.
Test specimens
=
nonnuclear means, except representing the reactor when the vessel is vented vessel, base weld, and or as indicated in 3.6.A.4, weld heat affected zone during cooldown following metal shall be installed nuclear shutdown, or in tue reactor vessel during low-level physics adjacent to the vessel tests, the reactor vessel wall at the core midplane temperature shall be at or level. The number and above the temperatures of type of specimens will curve #2 of Figure 3.6-1 until be in accordance with CE removing tension on the head report NEDO-10115. The stud bolts as specified in specimens shall meet the
'3.6.A.S.
intent of ASTM E 185-82.
1 BFN 3.6/4.6-2 Unit 1 Amendment No. 170
m i (, '
3.6'/4.6 EASES 3.6.A/4.6.A Thermal and Pressurization Limitations The vessel has been analyzed for stresses caused by thermal and pressure transients. Heating and cooling transients throughout plant life at uniform rates of 100*F per hour were considered in the temperature range of 100 to 546*F and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (65 Edition includins Summer 1966 addende.).
Operating limits on the reactor vessel pressure and temperature during-normal heatup and cooldown, and during inservice hydrostatic testing, were established using Appendix G of the Summer 1972 Addenda to Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition, as a guide.
These operating limits assure that a large postulated surface flaw, having a depth of one-quarter of the material thickness, can be safely accommodated in regions of the vessel shell remote from discontinuities..For the purpose of setting these operating limits the reference temperature, RTNDT, of the vessel material was estimated from impact test data taken in accordance with requirements of the Code to which this vessel was designed and manufactured (65 Edition to Summer 1966 addenda).
The fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the RPV.
Two types of information are needed in this analysis:
- 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and 2) a measure of the neutron fluence at the point of interest in the RPV wall.
A relationship between neutron fluence and change in Chalpy V, 30-foot pound transition temperature has been developed for SA302B/SA533 steel based on at least 35 experimental data points as shown in Figure 3.6-2.
In turn, this change in transition temperature can be related to a change in the temperature ordinate shown in Figure G 2110-1 in Appendix G of Section III of the Boiler Code.
The neutron fluence at any point in the pressure vessel vall can be computed from core physics data.
The neutron fluence can also be measured experimentally on the ID of the vessel wall. At present valid experimental measurements can be made only over time periods of less than five years because of the limitations of the dosimeter materials.
This causes no problem because of the exact relationship between thermal power produced and the number of neutrons produced from a given core geometry.
A single experimental measurement in a time period of one year can be used tu predict the fluence for the life of the plant in terms of thermal power output if no great changes in core geometry are made.
1 BFN 3.6/4.6-26 Unit 1
n f-3'.6/4.6 BATTER 3.6.A/4.6.A (Cont'd) l The vessel pressurization temperatures at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Figure 3.6-2.
For heatup or cooldown and core operation, see curves Nos. 2 and 3 on Figure 3.6-1.
During the first fuel cycle, only calculated neutron fluence values can be used.
At the first refueling, neutron dosimeter wires which are installed adjacent to the vessel vall can be removed to verify the calculated neutron fluence..
As more experience is gained in calculating the fluence the need to verify it experimentally will disappear. Because of the many experimental points used to derive Figure 3.6-2, there is no need to reverify if for technical reasons, but in case verification is required for other reasons, three j
1 sets of mechanical test specimens representing the base metal, weld metal i
and weld heat affected zone metal have been placed in the vessel.
These can be removed and tested as required 2 TVA letter dated May 15, 1987, proposed to withdraw the first set of reactor surveillance specimens from each reactor vessel at the end of each hait's cycle which most closely approximates 8.0 EFPY of operation.
The reasoning was the development of an integrated surveillance program related to estimated fluence obtained from reactor vessel specimens prior to 8.0 EFPY would be premature because it would be based only on extrapolations of limited dosimetry measurement taken from unit i during the first cycle of operation.
Dosimetry measurements for 8.0 EFPY would be more credible than cycle 1 dosimetry data.
NRC letter dated December 2, 1988, stated that BFN could withdraw the first reactor vessel specimen from each reactor vessel at the end of each unit's cycle of operation that most closely approximates 8.0 EFPY of operation.
After withdrawal of each unit's first sample, the remaining specimens will be withdrawn every 6.0 EFPY thereafter.
As described in paragraph 4.2.5 of the Safety Analysis Report, detailed stress analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material fatigue. The reaults of these analyses are compared to allowable stress limits.
Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and i
bottom head region are acceptable.
The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow.
This colder water is forced up when recirculation pumps are started.
This will not result in stresses which exceed ASME Boiler and Pressure Vessel Code,Section III limits when the temperature differential is not greater than 145'F.
1 The requirements for full tension boltup of the reactor vessel closure are based on the NDT temperature plus 60*F.
This is derived from the requirements of the ASME code to which the vessel was built. The NDT temperature of the closure flanges, adjacent head, and shell ma'terial is a maximum of 40*F and a maximum of 10*F for the stud material.
Therefore, the minimum temperature for full tension boltup is 40*F plus 60*F for a BFN 3.6/4.6-27 Unit 1 Amendment No. 170
3'.6/4.6 BASES 3.6,A/4.6.A (Cont'd) total of 100*F.
The partial boltup is restricted to the full loading of eight studs at 70*F, which is stud NDT temperature (10*F) plus 60*F.
The neutron radiation fluence at the closure flanges is well below 1017 nyt 1 1 Mev; therefore, radiation effects will be minor and will not influence this temperature.
3.6.B/4.6.B Coolant Chenistry Materials in the primary system are primarily 304 stainless steel and the Zircaloy cladding..The reactor water chemistry limits are established to prevent damage to these materials. Limits are placed on conductivity and chloride concentrations. Conductivity is limited because it is continuously measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant.
Chloride limits are specified to prevent stress corrosion cracking of stainless steel.
Zircaloy does not exhibit similar stress corrosion failures.
- However, there are some operating conditions under which the dissolved oxygen content of the reactor coolant water could be higher than.2.3 ppm, such as reactor.STARTUP and Hot Standby.
During these periods, the most restrictive limits for conductivity and chlorides have been established.'
When steaming rates exceed 100,000 lb/hr, boiling deaerates the reactor water. This reduces dissolved oxygen concentration and assures minimal chloride-oxygen content, which together tend to induce stress corrosion cracking.
When conductivity is in its normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range.
When conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values.
This would not necessarily be the case.
Conductivity could be high due to the presence of a neutral salt which would not have an effect on pH or chloride. In such a case, high conductivity alone is not a cause for shutdown.
In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.
In the case of BWRs, however, where no aeditives are used and where near neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water. Significant changes therein provide the operator with a warning aschanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to j
the operator for correcting the off-standard condition include operation j
of the reactor cleanup system, reducing the input of impurities and placing the reactor in the Cold Shutdown condition.
The major benefit of Cold Shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the reactor coolant.
The conductivity of the reactor coolant is continuously monitored.
The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a j
reference for calibration of these monitors and is considered adequate to assure accurate rec *ings of the monitors.
If conductivity is within its BFN 3.6/4.6-28l Unit 1 Amendment No. 170
3'.6/4.6 BASES
' 3.6.B/4.6.B (Cont'd) normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content.
Daily sampling is performed when increased chloride concentrations are most probable. Reactor coolant sampling is increased to once per shift when the continuous conductivity monitor is unavailable.
The basis for the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 36 rem at the exclusion distance during the two-hour period following a steam line break.
This dose is computed with the conservative assumption of a release of 140,000 lbs of coolant prior to closure of the isolation valves, and a X/Q value of 3.4 x 10-4 Sec/m3 The maximum activity limit during a short term transient to established-from consideration of a maximum iodine inhalation dose less than 300 rem. The probability of a steam line break accident coincident with an iodine concentration transient is significantly lower than that of the accident alone, since operation of the reactor with iodine levels above the equilibrium value is limited to 5 percent of total operation.
The sampling frequencies are established in order to detect the occurrence of an iodine transient which may exceed the equilibrium concentration limit, and to assure that the maximum coolant iodine concentrations are not exceeded. Additional sampling is required following power changes and off-gas transients, since present data indicate that the iodine peaking phenomenon is related to these events.
3.6.C/4.6.C Coolant Leakare Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite ac power.
The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits.
The behavior of cr~.s in piping systems has been experimentally and analytically L.vestigated as part of the USAEC sponsored Lc:ctor Primary Coolant System R2pture Study (the Pipe Rupture Study). Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth.
This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.
However, the establishment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the data presently available would be premature because of uncertainties associated with the data.
For leakage of the order of five gpm, as specified in 3.6.C, the experimental and analytical data BFN 3.6/4.6-29l l
Unit 1 Amendment No. 170
og 3 6/4.6 BASES 3.6.C/4.6.C (Cont'd)
I
+
suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The two spm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC (Reference 2).
This limit applies only during1the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm, Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
4 REFERENCE
- 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
- 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow at a reference pressure of (1,105 + 1 percent) psig.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief valve operation shews that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the i i percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MM to l
establish that they will open and pass steam.
I BFN 3.6/4.6-30 l j
Unit 1 Amendment Nos. 159, 170 I
l
1 3.6.'D/4.6.D~(C:nt'd)
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
1 REFERENCES 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) l i
2.
Amendment 22 in response to AEC Question 4.2 of December 6, 1971.
3.
" Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 4.
Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 5.
Generic Reload Puel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6,E/4.6.E Jet Pumos Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.
The detection technique is as follows. With the two recirculation pumps balanced in speed to within i 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation.
If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the potential blowdown flow area is increased, the system BFN 3.6/4.6-31 Unit 1 i
_____________-__.-_..___--__-----a
/
UNITED STATES 8
')
NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D C. 20555
~ %...+ /
TENNESSEE VALLEY _ AUTHORITY DOCKET NO. 50-260 BROWNS FERRY NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.170 License No. DPR-52 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated May 15, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of tnis amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
l
~
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. OPR-52 is hereby amended to read as follows:
(2)' Technical Specifications The Technical Specifications contained in Appendices.A and B, as revised through Amendment No.170..are hereby incorporated in the
' license. The licensee shall operate the facility in accordance with the Technical Specifications.
l 3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION r-,
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Fins'uza ne Black, Assistant Director 3
'for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 1, 1989 1
p.
ATTACHMENT TO LICENSE AMENDMENT NO.170 FACILITY OPERATING LICENSE NO. DPR-52 DOCKET NO. 50-260 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.
REMOVE INSERT 3.6/4.6-1 3.6/4.6-1*
3.6/4.6-2 3.6/4.6-2 3.6/4.6-26 3.6/4.6-26*
3.6/4.6-27 3.6/4.6-27 3.6/4.6-28 3.6/4.6-28 l
3.6/4.6-29 3.6/4.6-29 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3.6/4.6-31*
1 1
1 3.6/4.6 PRIMARY SYSTEM BOUNDARY 5
LIMITING CONDITIONS FOR OPERATION SURVEIL. LANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Anolicability Aeolicability i
l Applies to the operating status Applies to the periodic of the reactor coolant system.
examination and testing requirements for the reactor coolant system.
Obiective Q jective To assure the integrity and safe To determine the condition of operation of the reactor coolant the reactor coolant system and I
system.
the operation of the safety i
devices related to it, j
Specification SoecificatinD A.
Therrai and Pressurization A.
Thermal and Pressurization Limitations Limitatigna i
1.
The average rate of 1.
During heatups and
\\
reactor coolant temperature cooldowrA, the change durir.g normal heatup followfag parameters or cooldown shall not exceed shall be recorded and 100*F/hr when averaged over reactor coolant a one.-hour period, temperature determined at 15-minute intervals i
until 3 successive j
readings at each given location are within 5'F.
a.
Steam Dome Pressure (Convert to upper vessel region temperature) b.
Reactor bottom drain temperature c.
Recirculation loops A and B d.
Reactor vessel bottom head temperature e.
Reactor vessel shell adjacent to shell flange BFU 3.6/4.6-1 Unit 2 f
I w
t
'3 3.6/4.6 pkIMARY SYSTEM BOUNDARY-
. LIMITING CONDITIONS FOR OPERATIOh' SURVEILLANCE REQUIREMENTS 3.6.A.
Thermal and Pressurization 4.6.A.
ThermsL.RDA Pressurization Limitations Limitations 2.
During all operations with 2.
Re&ctor vessel metal-a critical core, other than temperature at the for low-level physics tes'.2, outside surface of the except when the vessel is bottom head in the vented, the reactor vessel vicinity of the control
-shell and fluid temperatures rod drive housing and shall be at or above the -
reactor vessel shell temperature of curve #3 of adjacent to shell Figure 3.6-1.
flange shall be recorded at least every 15' minutes during inservice hydrostatic or leak testing when the vessel pressure is >:312 psig.
3.
During heatup by 3.
Test specimens nonnuclear means, e2 cept' representing the reactor when the vecsel is vented vessel, base veld, and or as indicated in'3.6.A.4, weld heat affected zone during cooldown following metal shall be installed nuclear shutdown, or in the reactor vessel during low-level physics adjacent to the vessel tests, the reactor vessel wall at the core midplane 1
temperature shall be at or level. The number and above the temperatures of type of specimens will curve #2 of Figure 3.6-1 until be in accordance with GE removing tension on the head report NEDO-10115. The stud bolts as.specified in specimens shall meet the 3.6.A.S.
intent of ASTM E 185-82.
BFN 3.6/4.6-2 Unit 2 Amendment No. 170 j
)
I 3.6/4.6. BASES
'3.6.A/4.6.A Diermal and Pressurization Limitations The' vessel has been analyzed for stresses caused by thermal and pressure transients.. Heating and-cooling transients throughout uant life at L
uniform rates of.100*F per hour were considered in the temperature range l
of 100 to 546*F and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (65 Edition including Summer 1966 addenda).
t Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic testing, were established using Appendix G of the Summer 1972 Addenda to Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition, as a guide.
These operating limits assure that a large postulated surface flaw, having a depth of ora-quarter of the material thickness, can be safely accommodated in regions of the vessel shell remote from discontinuities.
For the purpose of setting these operating limits the reference temperature, RTNDT, of the vessel material was estimated from impact test data taken in accordance with requirements of the Code to which this vessel was designed and manufactured (65 Edition to Summer 1966 addenda).
The fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons above a threthold value, and it is prudent and conservative to account for this in the operation of the RPV.
Two types of information are needed in this analysis:- 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and 2) a measure of the neutron fluence at the point of interest in the RPV wall.
A relationship between neutron fluence and change in Charpy V, 30-foot pound transition temperature has been developed for SA302B/SA533 steel based on at least 35 experimental data points as shown in Figure 3.6-2.
In turn, this change in transition temperature can be related to a change in the temperature ordinate shown in Figure G 2110-1 in Appendix G of Section III of the Boiler Code.
The neutron fluence at any point in the pressure vessel wall can be compuced from core physics data. The neutron fluence can also be measured experimentally on the ID of the vessel wall. At present valid experimental measurements can be made only over time periods of less than five years because of the limitations of the dosimeter materials.
This causes no problem because of the exact relationship between thermal power produced and the number of neutrons produced from a given core geometry.
A single experimental measurement in a time period of one year can be used to predict the fluence for the life of the plant in terms of thermal power output if no great changes in core geometry a.e made.
BEN 3.6/4.6-26 Unit 2
3.'6/4.6 BASES 3.6.A/4.6.A (Cont'd)
The vessel pressurization temperatures at any time period ten be determined from the thermal power output of the plant and its relation to the neutron fluence and from Figure 3.6-2.
For heatup or cooldown and core operation, see curves Nos. 2 and 3 on Figure 3.6-1.
During the first fuel cycle, only calculated neutron fluencs values can be used. At the first refueling, neutron donimeter wires which are installed adjacent to the vessel wall can j
be removed to verify the calculated neutron fluence. As more experience is gained in calculating the fluence the need to verify it experimentally will i
disappear.
Because of the many experimental points used to derive Figure 3.6-2, there is no need to reverify if for technical reasons, but in case verification is required for other reasons, three sets of mechanical test specimens representing the base metal, weld metal and weld heat affected zone metal have been placed in the vessel.
These can be removed and tested as required.
l TVA letter dated May 15, 1987, proposed to withdraw the first set of reactor surveillance specimens from each reactor vessel at the end of each unit's cycle which most closely approximates 8.0 EFPY of operation.
The reasoning was the development of an integrated surveillance program related to estimated fluence at this time would be premature because it would be.
based only on extrapolations of limited dosimetry measurements taken from unit 1 during the first cycle.
Donimetry measurements for 8.0 EFPY vould be more credible than cycle 1 dosimetry data. NRC letter dated December 2, 1988, agreed and stated that BFN could withdraw the first specimen from each reactor vessel at the end of each rnit's cycle of operation most closely approximates 8.0 EF,*Y of operation. After withdrawal of each unit's first sample, the remaining specimens will be withdrawn every 6.0 EFPY thereafter.
As described in paragraph 4.2.5 of the Safety Analysis Report, detailed stresa analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material fatigue.
The results of these analyses are compared to allowable stress limits.
Requiring the coolant temperature in an 16;e recirculation loop to be within 50*F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow.
This colder water is forced up when recirculation pumps are started.
This vill not result in stresses which exceed ASMI Boiler and Pressure Vessel Code,Section III limits when the temperature differential is not greater than 145'F.
The requirements for full tension boltup of the reactor vessel closure are based on the NDT temperature plus 60*F.
This is derived from the requirements of the ASME code to which the vessel was built. The NDT l
temperature of the closure flanges, adjacent head, and shell material is a maximum of 40*F and a maximum of 10*F for the stud material.
Therefore, the minimum temperature for full tension boltup is 40*F plus 60*F for a total of 100'F.
The partial boltup is restricted to the full loading of BFN 3.6/4.6-27 Unit 2 Amendment No. 170
4
'3.6/4.6 BASES
~
3.6.A/4.6.A (Cont'd) eight studs at 70*F, which is stud NDT temperature (10'F) plus 60*F.
The neutron radiation fluence at the closure flanges is well below 1017 nyt i 1 Mev; therefore, radiation effects will be minor and will not influence this temperature.
3.6.B/4.6.B Coolant Chemistry Materials in the primacy system are primarily 304 stainless steel and the Zircaloy cladding. The reactor water chemistry limits are established to prevent damage to these materials. Limits are placed on conductivity and chloride concentrations. Conductivity is limited because it is continuously measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant.
Chloride limits are specified to prevent stress corrosion cracking of stainless steel.
Zircaloy does not exhibit similar stress corrosion failures. However, there are some operating conditions under which the dissolved oxygen content of the reactor coolant water could be higher than.2.3 ppm, such as reactor startup and hot standby.
During these periods, the most restrictive limits for conductivity and chlorides have been established.
When steaming rates exceed 100,000 lb/hr, boiling deaerates the reactor water. This reduces dissolved oxygen concentration and assures minimal chloride-oxygen content, which together tend to induce stress corrosion cracking.
When conductivity is in its normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range. When conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values. Thin would not necessarily be the case.
Conductivity could be high due to the presence of a neutral salt which would not have an effect on pH or enloride.
In such a case, high conductivity alone is not a cause for shutdown.
In come types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.
In the case of BWRs, however, where no tdditives are used and where near neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water.
Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system, reducing the input of impurities and placing the reactor in the Cold Shutdown condition. The major benefit of Cold Shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the reactor coolant.
The conductivity of the reactor coolant is continuously monitored.
The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors.
If conductivity is within BFN 3.6/4.6-28 l Unit 2 Amendment No. 170
._-_A
3.'6/4.6 BASES 3.6.B/4.6.B (Cont'd) its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Daily sampling is performed when increased chloride concentrations are most probable. Reactor coolant sampling is increased to once per shift when the continuous conductivity monitor is unavailable.
The basis fsr the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 36 rem at the exclusion distance during the two-hour period following a steam line break.
This dose is cceputed with the conservative assumption of a release of 140,000 lbs of coolanc prior to closure of the isolation valves, and a X/Q value of' 3.4 x 10-4 Sec/m3, The maximum activity limit during a short term transient is established from consideration of a maximum iodine inhalation dose less than 300 rem. The probability of a steam line break accident coincident with an iodine concentration transient is significantly lower than that of the accident alone, since operation of the reactor with iodine levels above the equilibrium value is limited to 5 percent of total operation.
The sampling frequencies are established in order to detect the occurrence of an iodine transient which may exceed the equilibrium concentration limit, and to assure that the maximum coolant iodine concentrations are not exceeded. Additional sampling is requ2 red following power changes and off-gas transients, since present data indicate that the iodine peaking phenomenon is related to these events.
3.6.C/4.6.C Coolant Leakaze Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite ac power. The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits. The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth.
This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.
However, the establishment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the data presently available would be premature because of uncertainties associated with the data.
For leakage of the order of BFN 3.6/4.6-29l Unit 2 Amendment No. 170
C 3.'6/4.6 BASES f
3.6.B/4.6.C (Cont'd) five spa, as specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid
' propagation. Leakage'1esa than the magnitude specified can be detected
]
reasonably in a matter of a few hours utilizing the available leakage
]
detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to. allow further I
investigation and corrective action.
The 2 gpm limit for coolant leakage-rate increases over any 24-hour
)
period is s' limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity _ of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment. sump pump is also 50 gpm. Removal of 25 gpm.
from either of these sumps can be accomplished with considerable margin.-
REFERENCE
- 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
- 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position, scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 7sig.
Experience in relief valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief valves are benchtested every second operating cycle to ensure that their setpoints are within the i i percent tolerance.
The relief valver re tested in place in accordance l-with Specification 1.0.MM to establish that they will open and pass l
steam.
BFN 3.6/4.6-30l Unit 2 Amendment Nos. 155, 170
3.6*/4.6 BASES
+
^
3.6.D/4.6.D (Cont'd)
The requirements established above apply when the nuclear system can be pressurized above ambient conditions. These requirements are applicable at nuclear system pressures below normal operating pressures because abnormal operational transients could possibly start at these conditions such that' eventual overpressure relief would be needed. However, these transients are much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
REFERENCES 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.
Amendment 22 in response to AEC Question 4.2 of December 6, 1971.
3.
" Protection Against Overpressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 4.
Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. G111 eland to F. E. Kruesi, August 29, 1973 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda 3.6.E/4.6.E Jet Pumos Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or riser, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height. level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.
The detection technique is as follows. With the two recirculation pumps balanced in speed to within i 5 percent, the flow ratec in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser and nozzle assembly integrity has been verified.
{
If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived frca the measured values of ?acp flow to core flow correlation.
If the difference between measured and derived core flow rate is 10 percent or mor:
- ith the derived value higher) diffuser measurements will be taken ts define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the potential blowdown flow area is increaaed, the system BFN 3.6/4.6-31 Unit 2
)
uj[
v *""*%o j
UNITED STATES F
.8%
7n NUCLEAR REGULATORY COMMISSION I:
5 E.
WASHINGTON, D. C. 20555 l
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.141 License No. DPR-68 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (the licensee) dated May 15, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the at.tivities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
f
- l l
1, 2.
Accordingly.'the license is amended by changes to the Technical f
. Specifications as indicated in the attachment.to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-68 is hereby-amended to read as follows:
i (2) Technical Specifications
-l The Technical Specifications contained in Appendices A and B, as revised through Amendment No.141, are hereby incorporated in the license. The licensee shall operate the facility in accordance.with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION c
~
~
FrinSudnb6'hck,AssistantDirector for Projects TVA Projects Division Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
August-3m 1989 l
l
l J
ATTACHMENT TO LICENSE AMENDMENT NO. 141 j
1 i
~
l FACILITY GPERATING LICENSE NO. DPR-68 d
1 L
DOCKET NO. 50-296 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal i
lines indicating the area of change. Overleaf pages* are provided to maintain document Completeness.
REMOVE INSERT 3.6/4.6-1 3.6/4.6-1*
3.6/4.6-2 3.6/4.6-2 3.6/4.6-26 3.6/4.6-26*
3.6/4.6-27 3.6/4.6-27 3.6/4.6-28 3.6/4.6-28 3.6/4.6-29 3.6/4.6-29 3.6/4.6-30 3.6/4.6-30 3.6/4.6-31 3.6/4.6-31 3.6/4.6-32 3.6/4.6-32 3.6/4.6-33 3.6/4.6-33*
i.
l e
3.674.6 PRIMARY SYSTEM BOUNDARY
$ LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Aeolicability Aeolicability Applies to the operating status Applies to.the periodic of the reactor coolant system.
examination and testing requirements for the reactor coolant system.
Obiective Objer Qyg To assure the integrity and safe To determine the condition of operation of the reactor coolant the reactor coolant system and system.
the operation of the safety devices related to it.
Specification Specification A,
Thermal and Pressurization A.
Thermal and Pressurization Limitations Limitations 1.
The average rate of 1.
During heatups and reactor coolant temperature cooldowns, the change during normal heatup following. Parameters or cooldown shall not exceed shall be-recorded and 100*F/hr when averaged over reactor coolant a one-hour period, temperature determined at 15-minute intervals until 3 successive readings at each given location are within 5'F.
a.
Steam Dome Pressure (Convert to upper vessel region temperature) b.
Reactor bottom drain temperature c.
Recirculation loops A and B d.
Reactor vessel bottom head temperature e.
Reactor vessel shell adjacent to shell flange BFN 3.6/4.6-1 Unit 3
. a., -
i
),6/4.6 PRIMARY SYSTEM BOUNDARY-
' LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RETIREMENTS 3.6.A.
Thermal and Pressurization 4.6.A.
Thermal and Pressurization Limitations Limitations 2.
During all operations with 2 '.
Reactor vessel metal a critical core,.other than temperature at the for low-level physics tests, outside surface of the a
except when the vessel is bottom head'in the vented, the reactor vessel vicinity of the control shell and fluid temperatures rod drive housing and shall be at or above the reactor vessel shell temperature of curve #3 of-adjacent to shell Figure 3.6-1.
flange shall be recorded at least'every 15 minutes during inservice hydrostatic or leak testing when the vessel pressure is > 312 psig.
3.
During heatup by 3.
Test specimens nonnuclear means, except representing the reactor when the vessel is vented vessel,' base weld, and or as indicated in 3.6.A.4, weld heat affected zone during cooldown following metal shall be installed
. nuclear shutdown, or in the reactor vessel during low-level physics adjacent to'the vessel tests, the reactor vessel wall at the core midplane temperature shall be at or level.
The number and above the temperatures'of type of specimens will curve #2 of Figure 3.6-1 until be in accordance with GE removing tension on the head report NEDO-10115. The stud bolts as specified in specimens sha11' meet the 3.6.A.S.
intent of ASTM E 185-82.
l BEN 3.6/4.6-2 Unit 3 Amendment No. 141
j i
1 3.614.6 DMf]
J.6.A/4.6.A Thermal _ and Pressurization Limitations The vessel has been analyzed for stresses caused by thermal and pressure transients. Heating and cooling transients throughout plant life at uniform rates of 100*F per hour were considered in the temperature range of 100 to 546'T and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (65 Edition including Summer 1966 Addendt).
l Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic testing, were established using Appendix G of the Summer 1972 Addenda to Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition, as a guide. These operating limirs assure that a large postulated surface flaw, having a depth of one-quarter of '.he material thickness, can be safely accommodated in regions of the vessel shell remote from discontinuities.
For the purpose of setting these operating limits the reference temperature, RTNDT, of the vesse7 material was estimated from impact test data taken in accordance with requirements of the Code to which this vessel was designed and manufactured (65 Edition to Summer 1966 Addenda).
The fracture toughness of all ferritic steels gradually and unifornly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the RPV.
Two types of information are needed n this analysis:
- 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and 2) a measure of the neutron fluence at the point of interest in the RPV wall.
A relationship between neutron fluence and change in Charpy V, 30-foot pound transition temperature has been developed for SA302B/SA533 steel based on at least 35 experimental data points as shown in Figure 3.6-2.
In turn, this change in transition temperature can be related to a change in the temperature ordinate shown in Figure G 2110-1 in Appendix G of Section III of the Boiler Code.
The neutron fluence at any point in the pressure vessel vall can be computed from core physics data.
The neutron fluence can also be measured experimentally on the ID of the vessel wall. At present valid experimental measurements can be made only over time periods of less than five years because of the limitations of the dosimeter materials.
This causes no problem because of the exact relationship 'oetween thermal power produced and the number of neutrons produced from a given core geometry.
A single experimental measurement in a time period of one year can tse used to predict the fluence for the life of the plant in terms of thermal power output if no great changes in core geometry are made.
BFN 3.6/4.6-26 Unit 3
l 4.
3.4/4.6 BASES 3.0.A/4.6.A (Cont'd)
The vessel pressurization temperatures at any time period can be determined from the thermal power output of the plant and its relation to the neutron-fluence and from Figure 3.6-2.
For heatup or cooldown and core operation, see curves Nos. 2 and 3 on Figure 3.6-1.
During the first fuel cycle, only calculated neutron fluence values can be used. At the first refueling, neutron dosimeter wires which are installed adjacent to the vessel wall can
.be removed to verify the calculated neutron fluence.
As more experience is gained in calculating the fluence the need to verify it experimentally will disappear.
Because of the many experimental points used to derive Figure 3.6-2,.there is no need to reverify if for technical reasons, but in case verification is required for other reasons, three sets of mechanical test specimens representing the base metal, weld metal and weld heat affected zone metal have been placed in the vessel.
These can be removed and tested as required.
TVA letter dated May 15, 1987, proposed to withdraw the first set of reactor surveillance specimens from each reactor vessel at the end of each unit's cycle which most closely approximates 8.0 EFPY of operation.
The reasoning was the development of an integrated surveillance program related to estimated fluence at this time would be premature because it would be, based only on extrapolations of limited dosimetry measurements taken from unit 1 during the first cycle.
Dosimetry measurements for 8.0 EFPY would be more credible than cycle 1 dosimetry data.
NRC letter dated December 2, 1988, agreed and stated that BFN could withdraw the first specimen from each reactor vessel at the end of each unit's cycle of operation most closely approximates 8.0 EFPY of operation. After withdrawal of each unit's first sample, the remaining specimens will be withdrawn every 6.0 EFPY thereafter.
I As described in paragraph 4.2.5 of the Safety Analysis Report, detailed stress analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material fatigue. The results of these analyses are compared to allowable stress limits.
Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow.
{
This colder water is forced up when recirculation pumps are started. This will not result in stresses which exceed ASME Boiler and Pressure Vessel L
Code,Section III limits when the temperature differential is not greater than 145'F.
The requirements for full tension boltup of the reactor vessel closure are based on the NDT temperature plus 60*F.
This is derived from the requirements of the ASME code to which the vessel was built.
The NDT temperature of the closure flanges, adjacent head, and shell material is a i
maximum of 40*F and a maximum of 10*F for the stud material.
Therefore, I
the minimum temperature for full tension boltup is 40*F plus 60'F for a total of 100*F.
The partial boltup is restricted to the full loading of BFN 3.6/4.6-27 Unit 3 Amendment No. 141 1
3.6/4.6 BASES 3.6.A/4.6.A (Cont'd) eight studs at 70*F, which is stud NDT temperature (10*F) plus 60*F.
The neutron radiation fluence at the closure flanges is well below 1017 nyt i 1 Mev; therefore, radiation effects will be minor and will j
not' influence this temperature.
{
I 3.6.B/4.6.B Coolant Chemistry Materia 3i in the primary system are primarily 304 stainless steel and l
the Zircaloy cladding. The reactor water chemistry limits are j
established to prevent damage to these materials. Limits are placed on
/
conductivity and chloride concentrations. Conductivity is limited i
because it is continuously measured and gives an indication of abnormal conditions and the presence of unusual materials in the coolant.
Chloride limits are specified to prevent stress corrosion cracking of stainless steel.
Zircaloy does not exhibit similar stress corrosion failures. However, there are some operating conditions under which the dissolved oxygen content of the reactor coolant water could be higher than.2.3 ppm, such as reactor startup and hot standby.
During these periods, the most restrictive limits for conductivity and chlorides have been established.
When steaming rates exceed 100,000 lb/hr, boiling deaerates the reactor water. This reduces dissolved oxygen concentration and assures minimal chloride-oxygen content, which together. tend to induce stress corrosion cracking.
When conductivity is in its normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range. When conductivity becomes abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values.
This would not necessarily be the case.
Conductivity could be high due to the presence of a neutral salt which would not have an effect on pH or chloride.
In such a case, high conductivity alone is not a cause for shutdown.
In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.
In the case of BWRs, however, where no additives are used and where near neutral pH is maintained, conductivity provides a very good measure of the qualJty of the reactor water.
Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system, reducing the input of impurities and placing the reactor in the Cold Shutdown condition.
The major benefit of Cold Shutdown is to reduce the temperature dependent corrosion rates and provide time for the cleanup system to reestablish the purity of the reactor coolant.
The conductivity of the reactor coolant is continuously monitored.
The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readinan of the monitors.
If conductivity is within BFN 3.6/4.6-28l Unit 3 Amendment No. 141
3 6/4.6 EASES 3.6.B/4.6.B (Cont'd) its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to determine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content. Daily sampling is performed when increased chloride concentrations are most probable. Reactor coolant sampling is increased to once per shift when the continuous conductivity monitor is unavailable.
The uasis for the equilibrium coolant iodine activity limit is a computed dose to the thyroid of 30 rem at the exclusion distance during the two-hour period following a steam line break.
This dose is computed with the conservative assumption of a release of 140,000 lba of coolant prior to closure of the isolation valves, and a X/Q value of 2.9 x 10-4 3
Sec/m.
The maximum activity limit during a short tens transient is established from ccusideration of a maximum iodine inhalation dose less than 300 rem. The probability of a steam line break accident :oincident with l
an iodine concentration transient is significantly lower.han that of the accident alone, since operation of the reactor with iodine levels above the equilibrium value is limited to 5 percent of total operation.
The sampling frequencies are established in order to detect the occurrence of an iodine transient which may exceed the equilibrium concentration limit, and to assure that the maximum coolant iodine concentrations are not exceeded. Additional sampling is required following power changes and off-gas transients, since present data indicate that the iodine peaking phenomenon is related to these events.
3.6.C/4,6.C Coolant LeakaRe Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite ac power.
The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits.
The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress cerrosion cracking or some other mechanism characterized by grcdual crack growth.
This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.
However, the establishment of allowable unidentified leakage greater than that given in 3.6.C on the basis of the dcea presently available would be premature because of uncertainties associated with the data.
For leakage of the order of five gpm, as specified in 3.6.C, the experimental and analytical data i
BFN 3.6/4.6-29l Unit 3 Amendment No. 141
3,6/4.6 BASES
'3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
Leakage less than the magnitude specified can be detected reasonably in.a matter of a few hours utilizing the available leakage detection schemes, and if the origin carnot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The two gpm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified,.which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
References 1.
Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10) 2.
Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 83.77 percent of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations.
The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the i i percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they will open and pass steam.
1 BFN 3.6/4.6-30l Unit 3 Amendment Nos. 130, 141 1
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3.6.D/4.6.D (Ctnt'd) s
,The requirements established above apply when the nuclear systea_can be
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pressurized above ambient conditions. These requirements are applicable at
-nuclear system pressures below normal operating pressures because abnormal l
operational transients could possibly start at these conditions such that eventual overpressure relief would be needed. However, these transients are i
much less severe, in terms of pressure, than those starting at rated conditions. The valves need not be functional when the vessel head is removed, since the nuclear system cannot be pressurized.
Enferences 1.
Nuclear System Pressure Relief System (BFNP FSAR Subsection 4.4) 2.
" Protection Against Over;ressure" (ASME Boiler and Pressure Vessel Code,Section III, Article 9) 3.
Browns Ferry Nuclear Plant Design Deficiency Report--Target Rock Safety-Relief Valves, transmitted by J. E. Gilliland to F. E. Kruesi, August 29, 1973 3.6.E/4.6.E, Tat Pumos Failure of a jet pump nozzle assembly holddown mechanism, nozzle assembly and/or ri9er, would increase the cross-sectional flow area for blowdown following the design basis double-ended line break. Also, failure of the diffuser would eliminate the capability to reflood the core to two-thirds height level following a recirculation line break. Therefore, if a failure occurred, repairs must be made.
The detection technique is as follows.
With the two recirculation pumps balanced in speed to within 5 percent, the flow rates in both recirculation loops will be verified by control room monitoring instruments.
If the two flow rate values do not differ by more than 10 percent, riser ar.d nozzle assembly integrity has been verified.
l If they do differ by 10 percent or more, the core flow rate measured by the jet pump diffuser differential pressure system must be checked against the core flow rate derived from the measured values of loop flow to core flow correlation.
If the difference between measured and derived core flow rate is 10 percent or more (with the derived value higher) diffuser measurements will be taken to define the location within the vessel of failed jet pump nozzle (or riser) and the unit shut down for repairs.
If the potential blowdown flow area is increased, the system resistance to the recirculation pump is also reduced; hence, the affected drive pump will "run out" to a substantially higher flow rate (approximately 115 percent to 120 percent for a single nozzle failure).
If the two loops are balanced in flow at the same pump speed, the resistance characteristics cannot have changed. Any imbalance between drive loop flow rates would be indicated by the plant process instrumentation.
In addition, the affected jet pump would provide a leakage path past the core BFN 3.6/4.6-31 Unit 3 Amendment No. 141
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'3.6.E/4'6.E (Crnt'd) thus riducing tha coro flow rata. Tha rcv:rco flow thrcugh the inactiva jot pumb would etill ba indicated by a positiva differential pressure but the net
,effect'would be a slight decrease (3 percent to 6 percent) in the total core flow measured.
This decrease, together with the loop flow increase, would result in a lack of correlation between measured and derived core flow rate.
Finally, the s.ffected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.
A nozzle-riser system failure could also generate the coincident failure of a jet pump diffuser body; however, the converse is not true.
The lack of any substantial stress in the jet pump diffuser body makes failure impossible without an initial nozzle-riser system failure.
3.6.F/4.6.F Recirculation Pumo Ooeratio.n Steady-state operation without forced recirculation will not be permitted for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. And the start of a recirculation pump from the natural circulation condition will not be permitted unless the temperature difference between the loop to be started and the core coolant temperature is less than 75'F.
This reduces the positive reactivity insertion to an acceptably low value.
Requiring the discharge valve of the lower speed loop to remain closed until the speed of the faster. pump is below 50 percent of its rated speed provides assurance when going from one-to-two pump operation that excessive vibration of the jet pump risers will not occur.
3.6.G/4.6.G Structural Integrity The requirements for the reactor coolant systems inservice inspection program have been identified by evaluating the need for a sampling examination of areas of high stress and highest probability of failure in the system and the need to meet as closely as possible the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code.
The program reflects the built-in limitations of access to the reactor coolant systems.
It is intended that the required examinations and inspection be completed during each 10-year interval.
The periodic examinations are to be done during refueling outages or other extended plant shutdown periods.
BFN 3.6/4.6-32 Unit 3 Amendment No. 141
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'3.6.G/4.6.G (Csnt'd) i 2
Only proven nondestructive testing techniques will be used,
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'More frequent inspections shall be performed on certain circumferential pipe
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' welds as listed in Section 4.6.G.4 to provide additional protection against pipe whip. These welds were selected in respect to their distaneit from hangers or supports wherein a failure of the weld would permit tha unsupported segments of pipe to strike the drywell wall or nearby auxiliary systems or control systems. Selection was based on judgment from actual plant observation of hanger and support locations and review of drawings.
Inspection of all these welds during each-10-year inspection interval vill
-result in three additional examinations above the requirements f Section XI of ASME Code.
References 1.
Inservice Inspection and Testing (BFNP FSAR Subsertion 4.12) 2.
Inservice Inspection of Nuclear Reactor. Coolant Systems,Section XI, ASME Boiler and Pressure Vessel Code 3.
ASME Boiler and Pressure Vessel Code Section III (1968 Edition) 4.
Anserican Society for Nondestructive Testing No. SNT-TC-1A (1968 Edition)
BFN 3.6/4.6-33 Unit 3
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