ML20009C888
| ML20009C888 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/10/1981 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20009C886 | List: |
| References | |
| TAC-43333, TAC-46676, TAC-46677, TAC-51387, TAC-52156, TAC-54259, NUDOCS 8107220105 | |
| Download: ML20009C888 (28) | |
Text
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1 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) r
- c. The Auxiliary Feed Pump Turbine Steam Generator level Control System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
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Auxiliary Feed l Pump Suction Pressure Interlocks ,and Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days, and a CHANNEL CALIBRATION at least once per 18 months.
DAVIS-8 ESSE, UNIT 1 3/4 7-5 8107220105 810710 ~
Docket No. 50-346 License No. NPF-3 ,
Serial No. 731 July _10, 1981 Attachment 2 I. Change to Davis-Besse Nuclear Power Statien Unit 1, Appendix A Technical Specifications 3.4.2, 3.4.3, 4.4.3 and dases concerning the'setpoint index for pressurizer electromatic relief and code safety valve.
A. Time required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 79-348)
To reduce the probability of opening pressurizer electromatic relief cnd code safety valve during a transient C. Safety Evaluation See attached 4
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Safety Evaluation For a RPS high pressure trip setpoint of 2300 psig, the maximum over-shoot of the Reactor Coolant System
- pressure for a loss of feedwater (LOFW) event would be to 2350 psig. Also, the LOFW is the maximum over-Pressure anticipated transient. The string inaccuracies and drift for the RPS high pressure trip are 15.29 psi, or 16 psi conservatively.
The inaccuracies and drift for the string that controls the electromatic relief valve for the pressurizer are 16.75 psi, or 17 psi conservatively.
Included in this value is an inaccuracy of 4 psi and a drift of 7.5 psi for the transmitter. The 4 psi and 7.5 psi were combined by taking the square root of the sum of the squares, giving 8.5 psi. Subtracting 4 psi from 8.5 psi gives a value of 4.5 psi that is attributable to only the drift. The 8.5 psi was then added to inaccuracy and drift values for other components in the string to obtain a total of 15.75 psi.
The allowable value of 2 2431 psig is obtained by subtracting 4.5 psi due to the drift from the trip setpoint of 22435.5 psig. The minimum lift pressure for the pressurizer electromatic relief valve is then (2435.5-
- 17) psig = 2418.5 psig. Consequently, the resultant margin between the maximum pressure peak of 2366 psig and minimum lift pressure of 2418.5 psig for the pressurizer electromatic relief valve following an anticipated transient is 52.5 psig.
The above values for the pressurizer electromatic relief valve in conjunc-tion with a 2300 psig RPS high pressure trip setpoint will avoid actuation of the pressurizer electromatic relief valve during anticipated transients.
All safety analyses for Davis-Besse Unit I assume that the vent capacity of the pressurizer electromatic relief valve will not be available; thus, these analyses are unchanged by an increase in its setpoint.
The pressurized code safety valves must be set such that the peak reactor coolant system pressure does not exceed 110% of design system pressure or 2750 psig. The peak reactor coolant system pressure aas determined from the control rod group withdrawal accident from low power at beginning-of-life conditions. The analysis was done for a high pressure trip of 2300 psig. In the analysis, 30 psid was used to account for the instrument string inaccuracy and draft. This is consistent with the value in FSAR. However, the 30 psid is conservative, because the actual, as built instrumentation has 15.3 psid inaccuracy and drift. The pres-surizer electromatic relief valve was assumed not to open. In the analysis, the code safety valve set pressure plus 3% was used, based on subsection 7614.1 of the ASME code. Also, the analysis was based on a maximum reactivity insertion rate of 1.655 x 10 -44 k/k/sec. The peak reactor coolant system pressure calculated is 2716 psig. This is accept-able, according to the above criteria.
2-Teledyne Engineering Services reviewed the setpoint changes and have concluded that the increase in setpressures will have no significant effect on the imposed limit on the number of discharge events of these valves.
The changes do not constitute an unreviewed safety question because:
- 1. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, previously evaluated in FSAP, has not been increased.
- 2. The possibility of an accider.t or malfunction of a different type other than any evaluated previously in the FSAR has not been created.
- 3. The margin of safety as defined in the basis for any technical specification has not been reduced.
pp b/7-9 l
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REACTOR COOLANT SYSTEM UN ER NRC REVIEW 3 SAFETY VALLES - SHUTDOWN ADDlIl0NAL CHANGES PREV 100 Sty PROPOSED By LEUER Seria' No._66 9 g,,, , ,]-
LIMITING CONDITION FOR OPERATION ji 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of IBrooPSIG + 1%.*
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APPLICABILITY: MODES 4 and 5.
- ACTION: -
With no pressurizer code safety valve OPERABLE, fr. mediately suspend all operations invol'ving positive reactivity changes and place an OPERABLE DHR loop into operation in the shutdown cooling mode.
SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirtments other than those ' required by Specification 4.0.5.
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- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
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DAVIS-BESSE, UNIT 1 3/4 4-3 I
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s REACTOR COOLANT SYSTEM SAFETY VALVES
?:.RATI.3 AND ELECTEoAATic AE.t.tEF VALVE - oPFRATie LIMITIPIG CONDITION FOR OPERATION 3.4.3 All pressurizer code safety valves shall be OPER 3LE with a lift settin feU*G vcke.g of Acve.
skli 2Foo PSIG + 1%.*
a, tete e,e4 Whe n,{ tsalSeci, Re pr<uoriu.e ekJ<- -(-.'c APPLICABILITY: MODES 1, 2 and 3. 7 ..4 4 p ggy .py g pg aglw,Qe src.lu e o E A a.4 3 / T m t 5 . -*
- ACrION:
With one pressurizer code safety valve inoperdble, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTOOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS For %e. pressad.er eacie cedtSt vslacs, %c o.<e ~o ,,
4.4.3 .% additional Surveillance Requirements other than those r,equire by Specification 4.0.5.
Fe< %e press o<i 2 .ee elec4<.uc+ic re(te s vcIVe_
a eLw<l esl.&dk cAeA z kalf 6e pu +cined eveig It-k.
' are litt setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
Allowcble v4lve h CAgnacf cal br a/lh cdec.f<.,
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OAVIS-BESSE, UNIT 1 3/4 4-4 m ._
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REACTOR COOLANT SYSTEM BASES For a RPS high pressure trip setpoint of 2300 psig, tne maximum overshoot of the Reactor Coolant System pressure for a loss of feedwater (LOFW) event would be to 2350 psig. Also, the LOFW is the maximum over-pressure anticipated transient.
The string inaccuracies and drif t for the RPS high pressure trip are 15.29 psi, or 16 psi conservatively. The maximum pressure peak for an anticipated transient is then 2366 psig.
The inaccuracies and drift for the string that controls the ciectromatic relief valve for the pressurizer are 16.75 psi, or 17 psi conservatively. Included in this value is an inaccuracy of 4 pai and a drift of 7.5 psi for the transmitter.
The 4 psi and 7.5 psi were combined by taking the square root of the sum of the squares, giving 8.5 psi. Subtracting 4 psi from 8.5 psi gives a value of 4.5 psi that is attributable to only the drift. The 8.5 psi was then added to inaccuracy and drift values for other components in the string to obtain a total of 16.75 psi.
The allowable value of h2431 psig is obtained by subtracting 4.5 psi due to the drift from the trip setpoint of 22435.5 psig. The minimum lift pressure for the pressurizer electromatic relief valve is then (2435.5-17) psig = 2419.5 psig.
Consequently, the resultant margin between the maximum pressure peat. of 2366 psig and minimum lift pressure of ?418.5 psig for the pressurize electromatic relief valve following an anticipated transient is 52.5 pai.
Thus, a 2300 psig RPS high pressure trip setpoint and the above values for the pressurizer electromatic relief valve will avoid actuation of the pressurizer i electromatic relief valve during anticipated transients.
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Docket No. 50-346 License No. NPF-3 Serial No. 731 July 10, 1981 Attachment 3 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 4.8.1.2.2.C.2 and 4.8.1.2 concerning Emergency Die'sel Generator (EDG) surveillance requirements.
4 A. Time required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-270A)
The Technical Specification changes for the EDG are to correct typographical errors in Section 4.8.1.2 per request r in NRC Inspection Report 50-346/80-29 dated December 23, 1080 (Log No. 1-456). Section 4.8.1.1.2.C.2 requires the EDG to demonstrate the capability to reject a load of 6480 KW without tripping. Instead the load shedding capability of the largest single emergency load should be tested.
C. Safety Evaluation See attached l
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Safety Evaluation T1.e first change is to correct a typographical error in Tech. Spec.
4.8.1.2. Surveillance Requirement 4.8.1.2, applicable in modes 5 and 6 currently require the Emergency Diesel Generator to be demonstrated as OPERABIE by the performance of each of the Surveillance Requirement of 4.8.1.2.2 except for 4.8.1.2.2.a.5. The reference to 4.8.1.2.2.a.5 as an exception in not correct. The correct exception to be referred is 4.8.1.1.2.a.7. Surveillance Requirement 4.8.1.1.2.a.5 concerns the synchronization and loading of the Emergency Diesel Generator and opera-tion for 60 minutes. This requirement should be performed, not exempted because one Emergency Diesel Generator has to be OPERABLE in modes 5 and
- 6. Surveillance Requirement 4.8.1.1.2.a.7 concerns the verification of the operability of the Safety Features Actuation System (SFAS) automatic load sequence timer. This is the requirement which should be exempted in modes 5 and 6 as the SFAS instrumentation Tech. Spec's. 3/4.3.2, Table 3.3-2, item 4 is only applicable in modes 1, 2, 3 and 4.
The second change is to correct an error in Surveillance Requirement 4.8.1.1.2.C.2 which requires the EDG to demonstrate the capability to reject a load 6480 KW without tripping. Instead the load shedding capability of the Emergency Diesel Generator to reject the largest single emergency load connected to it should be tested.
Correction of these errors does not create a safety question, therefore, this is not an unreviewed safety related issue.
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ELECTRICAL POWER SYSTEMS SHUTOOWN ,
LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
- a. One circuit between the offsite transmission network and the onsite Class lE distribution system consisting of;
- 3. One OPERABLE 13.8 KV bus, and
- b. One diesel generator with:
- 1. Day fuel tank containing a minimum volume of 4000 gallons of fuel,
- 2. A fuel storage system containing a minimum volume of 32,000 gallons of fuel, and
- 3. A fuel transfer pump.
APPLICABILITY: MODES 5 and 6.
ACTION:
With less than the akte minimum required A.C. electrical power sources f OPERABLE, suspend all operations involving CORE ALTERATIONS or positive l
' reactivity changes until the minimum required A.C. electrical power sources are restored to OPERABLE status.
l SURVEILLANCE REQUIREMENTS l
4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance j
Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.f.
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Y DAVIS-BESSE, UNIT 1 3/4 8-5 l
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- I ELECTRICAL POWER SYSTEMS bv SURVEILLANCE REQUIREMENTS (Continued) 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. At least once per 31 days on a STAGGERED TEST BASIS by:
- 1. Verifying the fuel level in the day fuel tank,
- 2. Verifying the fuel level in the fuel storage tank,
- 3. Verifying the fuel transfer pump can be started and
. transfers fuel from the storage system to the day tank, Verifying the diesel starts from ambient condition and
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accelerates to at least 900 rpm in 1 10 seconds,
, 5. Verifying the generator is synchronized, loaded to > 1000 kw, and operates for > 60 minutes, and
- 6. Verifying the diesel generator is aligned to pruvide standby power to the associated essential busses.
Verifying that the automatic load sequence timer is O' 7.
OPERABLE with each load sequence time within + 10% of its required value.
- b. At least once per 92 days by verifying that a sampld of diesel fuel from the fuel storage tank is within the acceptable limits specified in Table 1 of ASTM D975-68 when checked for viscosity, water and sediment.
- c. At least once per,18 months during shutdown by:
- 1. Subjecting the diesel to an inspection in accordance with
- procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service, 64 f.,
Verifying the generator capability to reject a load of 7 l . 2.
l g4 1 480 kw without tripping, J MN 3. Simulating a los's of offsite power in conjunction with a safety injection actuation test signal, and:
Verifying de-eargization of the essential busses a) and load shedding from the essential busses, O DAVIS-BESSE, UNIT 1 3/4 8-3 l .
l PROPOSED R E VIS E D TE C H. SPEC.
T S. 4. 8. I. l. 2. C. 2.
VE R I F3 I N G THE EMERGENC S DIESCL G GNERATOR 'S CAPABILIT.Y TO REJECT A LOAD EQU AL To THE L. ARGEST SINGL E EMERG[NC y LO AD SUPPLIED 8 3 THIS G E NER A TO R W ITH O U T TRIPPING.
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Docket No. 50-346 License No. NPF-3 Serial No. 731 July 10, 1981 Attachment 4 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 3.1.3.6, 3.2.1, 3.2.5 and Figure 3.1-3a concerning Regulating Rod insertion limits, Axial Power Imbalance and DNB Parameters.
A. Time required to Implement This change is to be effective upon NRC approval B. Reason for Change (Fac.lity Change Request 79-088B)
To correct administrative errors in Technical Specifications and revised safety evaluation for Amendment request submitted February 11, 1980 (Serial 590).
C. Safety Evaluation See attached l
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Safety Evaluation This license amendment request proposes changes to the Davis-Besse Unit 1 Technical Specifications as amended by License Amendment No. 33 dated October 1, 1980. Similar changes to Technical Specifications 3.2.1 and 3.2.5 had been previously submitted to the NRC as part of the reload report submittal for Davis-Besse Unit 1, Cycle 2; but were not granted for lack of bases. This safety evaluation provides the bases for the same.
The safety function achieved by Technical Specification 3.2.1 is to ensure that axial power imbalance is maintained within limits when oper-ating above 40% power. Action Statement b of this Technical Specifica-tion presently requires that the unit be brought to hot standby mode if the axial power imbalance exceeds the allowable value and is not restored within limits within 15 minutes. The requirement of bringing the unit to hot standby reduces unit availability and is excessive because the limiting condition of operation is not applicable below 40% of rated thermal power. The proposed change deletes the shutdown requirement.
With the proposed change, thermal power is required to be reduced to less than 40% of rated thermal power or until imbalance limits are met.
Due to the nature of the axial power imbalance limit envelopes, a reduc-tion in power may bring the imbalance within the limit because the limits are wider at lower power levels. If the power reduction does not bring the axial power imbalance within the limits, further power reduc-tion to less than 40% rated thermal power is required. The limiting condition for operation is then no longer applicable. If the Action Statements (a and b) are not satisfied, provisions of Specification 3.0.3 are applicable requiring that unit be taken to hot standby mode in one hour. The proposed change offers clarity and flexibility to the operator in the event that axial power imbalance limits are exceeded.
The proposed change also reduces the requirements of rapid shutdown of the unit which, in turn, reduces adverse impact on components and equip-ment. The axial power imbalance can be readily restored within limits or the thermal power reduced within the specified time. The possibility of a rapid shutdown in this case is therefore eliminated.
As noted above, this license amendment request also modifies the action statement for Technical Specification 3.2.5. This specification sets limits on the DNB related parameters of reactor coolant pressure, flow and hot leg temperature. According to the present Technical Specifi-cations, if any of these parameters falls oucside the prescribed limits, the parameter is required to be brought within limits within two hours ,
or thermal power reduced to less than 5% of rated thermal power within the next four hours.
The safety function achieved by this Technical Specification is to ensure that DNB limits are no: tiolated. The present Technical Speci-fication action statement requirement is too restrictive for reactor coolant flow. The measured system flow at Davis-Besse is 111.4% of
88,000 gpm/ reactor coolant pump (after taking into account the 2.5%
measurement uncertainty) whereas Technical Specification 3.2.5 calls for 110% flow. This license amendment request proposes that for every 1%
flow that is below the limit, thermal power be reduced by 2% of rated thermal power. B&W has performed calculations to determine the DNBR margin gain for the proposed flow and power tradeoff. The CHATA and TEMP codes were used to determine hot bundle flow, and subchannel flow and minimum DNBR, respectively. A factor N was defined as the percen-tage reduction in thermal power. The design overpower case of 112% of 2772 MWt and 110% of design reactor coolant system flow was used as the basis for comparison. An increase in DNBR margin was observed when power was decreased per the proposed change when flow was off-limit.
Figure 1 provides a graph of this increase in DNB margin. B&W has also concluded that this analysis is bounding for the case of three reactor coolant pumps operating and adequate DNB margin will also be usined in that case if the modified action statement is followed. Based on the above, it is concluded that by following the revised action statement, adequate margin to DNB is gained thereby enhancing the safety function achieved by this Technical Specification. In addition, it further reduces the possibility of an undesired unit shutdown ani increases unit availability.
Two other changes proposed by this license amendment request are admin-istrative in nature and relate to Technical Specification 3.1.3.6.
Namely, in the note for Technical Specification 3.1.3.6, an improper section of Technical Specifications is referenced. The correct section to be referenced is 3/4.1.1.1. The correct reference was originally submitted to the NRC with the reload report submittal, but was apparently inadvertently changed when the license amendment was granted. Simi-larly, Figure 3.1-3a (regulating rod group position limits for three pump operation) is for 0 to 150 1 10 EFPDs. The present figure in the Technical Specification states that the figure is valid for 0-125 i 10 EFPDs. This is contrary to the relesd report submittal to the NRC and needs revised as attached. Since t se two changes are administrative, no safety concern is involved.
i l Based on the above, it is concluded that the changes to the Technical l' Specifications proposed by this license amendment request do not involve an unreviewed safety question.
4 pp.b/12-14 1
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O REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a and -2b and 3.1-3a and -3b, with a rod group overlap of 25 + 5% between sequential withdrawn groups 5, 6, and 7.
APPLICABILITY: MODES 1* and 2*f.
ACTION:
With the regulating rod groups inserted beyond the above insertion limits (in a region other than acceptable operation), or with any group sequence or overlap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:
- a. Restore the reg 61ating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or ,
- b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the rod group position using the above figures within 2 ho.urs, or ,,
- c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
NOTE: If in unacceptable region, also see Section 3/4. l. l. I'.
- See Special Test Exceptions 3.10.1 and 3.10.2.
- With K,ff >,1.0. -
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DAVIS-DESSE, UNIT 1 3/4 1-26 Arendment No.)a',3 3 i
, 3/4.2. POPER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on ,
Figures 3.2-1 and 3.2-2.
APPLICABILITY: MODE 1 above 40% of RATED THERMAL POUER,*
ACTION:
With AXIAL POWER IMBALANCE exceeding the limits specified above, either a= Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes,
,to Ice _.s fhan '4of et RATEDTHER. MAL POWER 4
- b. , Reduce ~ poyerTuntil,,imbalan_ce limit [s_ are , met _Wilhin Cildckjh' [ _ . -
SURVEILLANCE REQUIREMENTS 4.2.1 The AXIAL POWER IMBALANCE shall be determined 'to be within limits at
- least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED THERMAL POWER except when the AXIAL POWER IMBALANCE alarm is inoperable', then calculate the AXIAL POWER IMBALANCE at least once per hour.
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- See Special Test Exception 3.10.1.
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l DAVIS-BESSE, UNIT 1 3/4 2-1 D
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POWER DISTRIBUTION LIMITS .
5- DNB PARAMETERS .
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i, LIMITING CONDITION FOR OPERATION t- .
3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:'
- a. Reactor coolant hot leg temperature. ,
- b. Reactor coolao'. pressure.
- ~ .
,, c. Reactor coolant flow rate.
&Y APPLICABILITY: MODE 1.
- a. .
ACTION: -
en. BOW ,
Ef parameter a or bg exceeds its 112 nit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce 'nlERMAL POWER to less than,5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. .
If . parameter e exceeds its limit, either:
- 1. Restore the parameter to within its liniit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or
- 2. ' Limit THERMAL POWER at least 2% below RATED THERMAL POWER for each 1% para-meter c is outside its limit for four pump operation within the r ext 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or limit THERMAL POWER at least 2% below 75%.of. RATED THERMAL POWrR for each .
1% parameter e is outside its 112 nit for. 3 pump operation within the next 4 .
hours.
SURVEILLANCE REQUIREMENTS
.. 4.2.5.1 , Each of the parameters of Table.3.2-l_shd11 be_ verified to be within .
their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. .
4.2.5.2 The reactor coolant system total flow rate shall be determined to be -
within its limit by measurement at least once per 18 months.
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DAVIS-BESSE, UNIT 1 ,
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9
4 Docket No. 50-346 License No. NPF-3 Serial No. 731 July 10, 1981 Attachment 5 I. Change to Davis-Besse Nuclear Power Staiton Unit 1, Arpendix A Technical Specifications 6.5.2.0, Figures 6.2-1 and 6.2-2 concerning CNRB membership and organizational changes.
A. Time required to Implement This change is to be effective upon NRC approval B. Reason for Change (Facility Change Request 80-2490, D) To reflect organizational changes at Toledo Edison C. Safety Evaluation See attached i
r sn , , -- * - , .---.r- ,, e = - m-e--,e ~-- - -w e my en
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- 1 Safety Evaluation The Technical Specification change requested reflects organizational changes. The safety function of the CNRB is to perform: ,
A. Review per Section 6.5.2.7 of Technical Specifications
- 1. Review Safety Evaluations.
- 2. Unreviewed Safety Questions.on procedures, equipment, systems, test or experiments.
- 3. Changes to. Facility Operating License.
- 4. Violecions having nuclear safety significance.
- 5. Abnormalities or deviations that affect nuclear safety.
- 6. Events requiring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification.
- 7. Unanticipated deficiency in safety related structures, systems 1 or components. l
- 8. SRB Minutes.
B. Audits of facility activity per Secti-n 6.5.2.8 of Technical Specifi-cations will remain unchanged.
The organizational charges to the CNRB will only change some members and not the function of the board. The changes improve independence and expertise on the board in facilitating their function.
- The safety function of the station organization is to show lines of responsibility for overall facility operation and maintenance of the station in a safe, reliable and efficient manner. The operating license requirement has been deleted from the Assistant Shift Supervisor. This position is an administrative function to aid the Shift Supervisor. The number of licenses in the crew composition is governed under Section 6, Table 6.2-1. Any license required functions by the control room organization shall be dispatched in accordance with 10 CFR 50.54 and Part 55.
There are no physical or safety related functional changes there-fore, this is not an unreviewed safety related issue.
pp b/15-16
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DAVIS-SESSE. UNIT 1 6-2 Amendment No.
- i. .
STATifN SUPERINT D DDT E i *< 8 **
*Y *
- OFFICE A D>f1NISTRAT IV E
-E SUPERVISOR COORDINATOR jU .
'. c - i STAT 14;4 REVIEW 4 BOARD ASSISTAWT STATION 7
$UPERINTENDENT OPERATIONS CH DIST AND MAlWT ENA NCE SOL DeCINEEg HEALTH ENGINEER FHYSICIST \ * - , STAFF l ,
TEQlNICAL w ' i - I ENCINEER CHDitCAL AND LEA 0 OPERATIONS eA01AT10N MAINTENANCE LEA D Sol. SUPERVISOR DOTECTION SUPPORT l&C ENGINEEg ENCINEER DCIN EER I . 1 l l OIDilSTRY & RADIO. NM W DANCE Sul. SHIFT IbC NEEAR AND SUPERVISOR CHDt1STRY DHYSICS SUPERVISOR FORDtA N P E RFORM/..tC E SUPERVISOR SUPERvlSCR ENCINEER A S$1s tAKT SHIFT k I SUPERVISOR C6HF C&llP MAINTO4ANCE ILC OL. ofg,A 3 FOMN FOR N N MRMN MECMNICS
- r i i ;
1 l STAFF I REPAIRMA N m TESTERS TEST ER S 1 ** D2u t rMt'NT AND g OPERA TORS 0.ECT R ICIA NS DAVIS BESSE NUCLEAR POWER STATION STATION ORCANIZATION i! CUKE 6.2=1 ADDill0NAL CHANGES PREVIOUSLY
- PR0f0 SED BY LETTER l UNDER NRC REVIEW Serial No. 66 9 Datdll24[FO
~ f I
/
L '. ' 5 {UNDER NRC REVIEWj' ADMINISTRATIVE CONTROLS ADDill0NAL CHANGES PREVIOUSLY
- PROPOSED BY LETTER ~
COMPOSITION SerialNo. h6 9 Date r 2/76 0 *
, , I 6.5.2.2 .The Company Nuclear Review Board shall be compos'ed of the:
Chairman: Director, Fossil Facilities Engineering and Construction Member: General Superintendent, Transmission and Substations Member: Superintendent, Davis-Besse Station Member: Director, Nuclear Services Member: Manager, Nuclear Engineering - Member: Director, Quality Assurance Member General Superintendent, Fossil Generation Facilities ' Member: Director, Nuclear Safety Member: Manager, Facility Engineering Member: Others as deemed advisable by the CNRB Chairman
- ALT ERNATES 6.5.2.3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis; however, no more than two alter-nates shall participate as voting members. in CNRB activities at any one time.
CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB - Chairman to provide expert advice to the CNRB. MEETING FREQUENCY 6.5.2.5 The'CNRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter. l 000 RUM 6.5.2.6 A quorum of CNRS shall consist of the Chairman or his designated alternate and at least half of the appointed CNRS members , or their alternates. No more than a minority of the cuorum shall have line responsibility for operation of the facility. i
*0thers as deemed advisable by the CNRS chairman, who are appointed to the Company Nuclear Review Board shall have an academic degree in an Engineering or Physical Science Field; and in addition, shall have i
a minimum of five years of technical experience, of which a minimum of l three years shall be in one or more of the areas specified in Specification 6.5.2.1. DAVIS-BESSE, UNIT 1
'6-9 Amendment No. J2, 27
e PLANT SYSTEMS . 3/4.7.10 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.10 All fire barrier penetrations (including cable penetration barriers, firedoors and fire dampers) in fire zone' boundaries protecting safety related areas shall be functional. , APPLICABILITY: At all times. ACTION:
- a. With one or more of the above required fire barrier penetrations non-functional, within one hour either, establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the non-functional fire barrier and establish a hourly fire watch patrol.
Restore the non-functional fire barrier penetration (s) to functional status within 7 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the non-functional pene-tration and penetration (plans s) to and status. functional schedu'.a for restoring the fire barrier
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIP.EMENTS 4.7.10 The above required penetration fire barriers shall be verified to be functional:
- a. . At least once per 18 months by a visual inspection.
- b. Prior to returning a penetration fire barrier to functional status following repairs or maintenance by performance of a visual inspection of the affected penetration fire barrier (s).
CAVIS SESSE, UNIT 1 3/4 7-47 Amendment No. ?
u Docket No. 50-346 License-No. NPF-3 Serial No. 731 July 10, 1981 Attachment 6 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications 3.7.10 concerning fire barrier penetration. A. Time required to Implement This change is to be effective upon NRC approval L. Reason for Change (Facility Change Request 80-263A) To comply with Mr. R. Reid's letter dated September 23, 1980 (log 609) and completion of Toledo Edison application for Amendment request submitted on December 26, 1980 f. Serial No. 669). C. Safety Evaluation The function of a # ire barrier penetration is to protect safety related areas. Section 3.7.10 establishes a continuous fire watch or verification of the operability of the fire detectors on at least one side of the nonfunctional fire barrier and establish an hourly fire watch patrol. This action will detect fires, suppress those fires that may occur and ensure a safe cold shutdown state can be achieved. The change proposed has been evaluated as an integrated part of the Davis-Besse' Nuclear Power Station Unit No. 1 Fire Hazard Analysis Evaluation and enclosed in Amendment 18 to the Facility Operating License NPF-3. The change does not affect the safety function of the fire barrier for safety related areas. Therefore, this is not an unreviewed safety issue. PP b/17 i i9p - g -w -- g m p .-- -y F-N- yg - ----
PLANT SYSTEMS . 3/4.7.10 FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION 3.7.10 All fire barrier penetrations (including cable penetration barriers, firedoors and fire dampers) in fire zone boundaries protecting safety related areas shall be functional. APPLICABILITY: At all times.
~
ACTION:
- a. With one or more of the above required fire barrier penetrations non-functional, within one hour either, establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the non-functional fire barrier and establish a hourly fire watch patrol.
Restore the non-functional fire barrier penetration (s) to functional status within 7 days or, in lieu of any other report required by Specification 6.9. 1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 Within the next 30 days outlining the action taken, the cause of the non-functional pene-trati6n and plans and schedule for restoring the fire barrier penetration (s) to functional status.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.7.10 The above required penetration fire barriers shall be verified to be functional:
- a. At least once per 18 months by a visual inspection.
- b. Prior to returning a penetration fire barrier to functional
, status following repairs or maintenance by performance of a visual inspection of the affected penetration fire barrier (s).
CAVIS-BESSE, UNIT 1 3/4 7-47 Amendment No. 9
. - -}}