ML18033A755
| ML18033A755 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/15/1989 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18033A754 | List: |
| References | |
| NUDOCS 8905220137 | |
| Download: ML18033A755 (23) | |
Text
ENCLOSURE-1 Proposed Technical Specification Browns Ferry Nuclear Plant Units 1, 2 and 3 (TVA BFN TS 270)8905220i37 S905i5 PDR ADOCK 05000259 P PDC
.6/4.6 PRIMARY SYSTEM BOUNDARY LIMITING CONDITIONS FOR OPERATION 3.6.A.Thermal and Pressurization Limitations SURVEILLANCE REQUIREMENTS 4.6.A.Thermal and Pressurization Limitations 2.During all operations with a critical core, other than for low-level physics tests, except when the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve¹3 of Figure 3.6-1.2.Reactor vessel metal temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to shell 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 nonnuclear means, except when the vessel is vented or as indicated in 3.6.A.4, during cooldown following nuclear shutdown, or during low-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve¹2 of Figure 3.6-1 until removing tension on the head stud bolts as specified in 3.6.A.5.3.Test specimens representing the reactor vessel, base weld, and weld heat affected zone metal shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane level.The number and type of specimens will be in accordance with GE report NEDO-10115.
The specimens shall meet the intent of ASTM E 185-82.BFN Unit 1 3.6/4.6-2 3.6/4.6 BASES 1 3.6.A/4.6.A (Cont'd)BFN Unit 1 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,>ield 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 obtained from reactor vessel specimens prior to 8.0 EFPY would be premature because it would be based only on extrapolations of limited dosimetry measurements taken from unit 1 during the first cycle of operation.
Dosimetry me'asurements 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 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 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 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 3.6/4.6-27 3.6/4.5 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 10nvt>1 Mev;therefore, radiation effects wi 11 be minor and will not influence this temperature.
3.6.8/4.6.B Coolant Chemistr 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 restri cti ve limits for conductivity and chlorides have been established
.Hhen 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.Nhen conductivity is in its normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range.Hhen 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 BHRs, however, where no additives 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 puri ty of the reactor coolant.BFN Unit 1 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 its 3.6/4.6-28 1
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" Sec/m','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'aximum 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..6.I.6.6~tk 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).Hork 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 Unit 1 3.6/4,6-29 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 cannot 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'nly 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-'nidentified, 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 wi th 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 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+1 percent tolerance.
The relief valves are tested in place in accordance with Specification 1.0.MN to establish that they will open and pass steam.BFN Unit 1 3,6/4.6-30 DESCRIPTION AND JUSTIFICATION BROWNS FERRY NUCLEAR PLANT (BFN)Units 1, 2 and 3 REASON FOR CHANGE BFN units 1, 2, and 3 technical specification (TS)4.6.A.3 and BASES section 3.6/4.6, Thermal and Pressurization Limitations, is being updated in accordance with 10 CFR 50 Appendix H and NRC letter to TVA dated December 2, 1988.DESCRIPTION AND JUSTIFICATION FOR THE PROPOSED CHANGE EXISTING SURVEILLANCE REQUIREMENT 4.6.A.3 READS: The specimens shall meet the intent of ASTM E 185-70.Samples shall be withdrawn at one-fourth and three-fourths service life.PROPOSED CHANGE TO SURVEILLANCE REQUIREMENT 4.6.A.3 The specimens shall meet the intent of ASTM E 185-82.JUSTIFICATION FOR CHANGE TO SURVEILLANCE 4.6.A.3 This change is an administrative change in that it updates surveillance requirement 4.6.A.3 to comply with the 10 CFR 50 Appendix H, II.B.1.This section of Appendix H requires that the testing procedures and reporting requirements for each capsule (specimen) withdrawn from the reactor vessel after July 26, 1983 must meet the requirements of ASTM E 185-82 to the extent practical for the configuration of the specimens in the capsule.By letter dated December 2, 1988, NRC approved BFN to withdraw the first set of surveillance specimens from each reactor vessel at the end of each unit's cycle which most closely approximates 8.0 effective full power years (EFPY)of operation.
After the first capsule from each unit has been withdrawn, the subsequent specimens will be tested at a 6.0 EFPY interval as required by Appendix H.PROPOSED CHANGE TO BASES SECTION 3.6/4.6 NOULD READ: 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 obtained from reactor vessel specimens prior to 8.0 EFPY would be premature because it would be based only on extrapolations of limited dosimetry measurements taken from unit 1 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 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'first sample, the remaining specimens will be wi thdrawn every 6.0 EFPY thereafter.
JUSTIFICATION FOR CHANGE: Changing the BASES section as proposed provides clarification as to the agreed upon specimen withdrawal program between NRC and TVA as documented in the mentioned letter transmittals.' DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BROHNS FERRY NUCLEAR PLANT (BFN)UNITS 1, 2, AND 3 DESCRIPTION OF PROPOSED TECHNICAL SPECIFICATION AMENDMENT The proposed amendment would change the BFN technical specifications (TS)for units 1, 2, and 3 to update surveillance requirement 4.6.A.3 and revise the BASES section 3.6/4.6 to comply with 10 CFR 50 Appendix H for reactor vessel test specimen testing frequency.
BASIS FOR PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION NRC has provided standards for determining whether a significant hazards consideration exists as-stated in 10 CFR 50.92(c).A proposed.amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1)involve a significant increase in the probability of consequences of an accident previously evaluated, or (2)create the possibility of a new or different kind of accident from an accident previously evaluated, or (3)involve a significant reduction in a margin of safety, The proposed change does not involve a significant increase in the probability or consequence of any accident previously evaluated.
This is an administrative.
change in that it only updates the BFN technical specification to comply with the 10 CFR 50 Appendix H.This proposed amendment does not change or modify any safety related equipment, its operation, or safety analysis in which BFN is licensed for.By updating the TS to ASTM E 185-82 increases the frequency for reactor vessel specimen withdrawal from 8 effective full power years (EFPY)to 6 EFPY.This increase in frequency does not involve any safety issue.The procedures and methods of withdrawing these specimens will remain the same.2.3.The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change is administrative in that it only updates BFN reactor vessel specimen program to ASTM E 185-82.Implementation of this change does not change any equipment or modify any actions required for mitigation of any accident currently analyzed in the BFN FSAR~This change does not create any additional radiation release pathways to the environment.
The proposed amendment does not invol.ve any significant reduction in a margin of safety.The change updates the BFN reactor vessel specimen withdrawal program with that in 10 CFR 50 Appendix H.BFN has agreement with NRC to withdraw the first specimen from each unit after 8'EFPY.After the first specimen is pulled from each unit, subsequent specimens will be pulled at a 6.0 EFPY.
3.6/4.6 PRIMARY SYSTEM BOUNDARY l LIMITING CONDITIONS FOR OPERATION 3.6.A.Thermal and Pressurization Limitations SURYEILLANCE REQUIREMENTS 4.6.A.Thermal and Pressurization Limitations 2.During all operations with a critical core, other than for low-level physics tests, except when the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve¹3 of Figure 3.6-1.2.Re ac tor ve s se 1 me ta1 temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to shell 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 nonnuclear means, except when the vessel is vented or as indicated in 3.6.A.4, during cooldown following nuclear shutdown, or during low-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve¹2 of Figure 3.6-1 until removing tension on the head stud bolts as specified in 3.6.A.5.3.Test specimens representing the reactor vessel, base weld, and weld heat affected zone metal shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane"level.The number and type of specimens will be in accordance with GE report NEDO-10115.
The specimens shall meet the intent of ASTM E 185-82.BFN Unit 2 3.6/4.6-2
3.6/4.,6 BASES 3.6.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 influence 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 influence 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 influence.
As more experience is gained in calculating the influence the need to verify it experimentally wi 11 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 influence 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.
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 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 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 Unit 2 3.6/4.6-27
~4 I 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 10" nvt>1 Mev;therefore, radiation effects will be minor and will not influence this temperature.
3.6.8/4.6.B Coolant Chemistr 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~teel.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'hen 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 chlorides 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 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 Unit 2 3.6/4.6-28 3.6/4.5 BASES 3.6.8/4.6.8 (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 consideredadequate 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 valve of 3.4 x 10'ec/m'.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 equi librium 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.6.6.CI6.6.C C~l"t 6 k 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).Hork 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 Unit 2 3.6/4.6-29
3.6/4.6 BASES\3.6.B/4.6.C (Cont'd)five gpm, 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 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 2 gpm limit for coolant leakage rate increases over any 24-hour 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.REFERENCE 1.Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)2.Safety Evaluation Report (SER)on IE Bulletin 82-03 es,.~~'.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 psig.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+1 percent tolerance, The relief valves are tested in place in accordance with Specification 1.0.MM to establish that they wi 11 open and pass steam.BFN Unit 2 3.6/4.6-30 3.6/4.6 PRIMARY SYSTE BOUNDARY'LV LIMITING CONDITIONS FOR OPERATION 3.6.A.Thermal and Pressurization Limitations SURVEILLANCE REQUIREMENTS 4.6.A.Thermal and Pressurization Limitations 2.During all operations with a critical core, other than for low-level physics tests, except when the vessel is vented, the reactor vessel shell and fluid temperatures shall be at or above the temperature of curve¹3 of Figure 3.6-1.2.Reactor vessel metal temperature at the outside surface of the bottom head in the vicinity of the control rod drive housing and reactor vessel shell adjacent to shell 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 nonnuclear means, except when the vessel is vented or as indicated in 3.6.A.4, during cooldown following nuclear shutdown, or during low-level physics tests, the reactor vessel temperature shall be at or above the temperatures of curve¹2 of Figure 3.6-1 until removing tension on the head stud bolts as specified in 3.6.A.5.3~Test specimens representing the reactor vessel, base weld, and weld heat affected zone metal shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane--'level.The number and type of specimens will be in accordance with GE report NEDO-10115.
The specimens shall meet the intent of ASTM E 185-82.BFN Unit 3 3.6/4.6-2 3.6/4.0 BASES 3.6.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 influence 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 influence 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 influence.
As more experience is gained in calculating the influence 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 influence 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 wi 1'1'e 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 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 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 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 I BFN Unit 3 3.6/4.6-27 I~r, I>V I l, 3.6/4.5 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 10nvt>1 Mev;therefore, radiation effects will be minor and will not influence this temperature.
3.6.B/4.6.8 Coolant Chemistr 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 fai lures.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.-
Nhen 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.Nhen conductivity is in its normal range, pH and'chloride and other impurities affecting conductivity must also be within their normal range~Nhen 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 BNRs, however, where no additives 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 Unit 3 3.6/4.6-28 3.6/4.$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)on 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 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 lbs of coolant prior to closure of the isolation valves, and a X/Q value of 2.9 x 10'ec/m'.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 equi librium 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.CI.6.C C~lt 3 3 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 five gpm, as specified in 3.6.C, the experimental and analytical data BFN Unit 3 3.6/4.6-29 3.6/4.$BASES~'.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 cannot 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 1~2.Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)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 re]ect 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+1 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.BFN Unit 3 3.6/4.6-30