ML20049A784

Proposed Tech Specs 3/4.3,4.4.6.2.1,3.4.8 & 3/4.4.7 Re Instrumentation Surveillance Requirements,Limiting Condition for Operation & Bases,Respectively

ML20049A784
Person / Time
Site:	Davis Besse
Issue date:	09/25/1981
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20049A770	List: ML20049A769 ML20049A776 ML20049A784
References
TAC-47040, TAC-47041, TAC-47043, NUDOCS 8110020270
Download: ML20049A784 (13)

Text

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3/4.3 INSTRUMENTATION BASES - _ -

REMOTE SHUTDOWN INSTRUMENTATION (Continued) -

HOT STANDBY of the facil. i ty from locations outside of the control room.

This capability is required in the event control room habitability is lost.

3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the posc-accident instrumentatiqn ensures that suffi-cient information is available on selected plant parameters to monitor and assess these variables following an accident.

3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the chlorine detection systems ensures that an acci-dental . chlorine release will be detected promptly and the . control room ant' 80 ut e,arne- emersancy ventilatica systam will automatically isolated.the centrol ennm sea v untMud-3and iaitiate its operatier. in the recirculation mode to provide the

gan Wa GE required protection. The chlorine detection systems required by this TauALLY specification are consistent with the recommendations of Regulatory iTURTED Guide 1.95, " Protection of Nuclear Power Plant Control Room Operations Against an Accidental Chlorine Release," February 1975.

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION

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Operability of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detec. tion of fires.  ;

This capability is required in order to detect and locate fires in i their early stages. Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperabib, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

8110020270 810925 PDR ADOCK 05000346 P PDR DAVIS-BESSE, UNIT 1 B 3/4 3-3 Amendment No. 9 1

Docket No. 50-346 License No. NPF-3 Serial No. 738 September 25, 1981 Attachment 3 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A, Technical Specifications 4.4.6.2.2 and 6.9.1.9.

A. Time Required to Implement This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request 81-143)

To combine leakage test without opening the motor operated containment isolation valves snd delete the reporting require-ments when any part of the system is inoperable only for surveillance testing, instrument calibration or preventive l maintenance purposes.

C. Safety Evaluation l See attached.

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l Safety Evaluation This Amendment Request proposes changes to Sections 4.4.6.2.3 and 6.9.1.9 of the Davis-Besse Technical Specifications. These Technical Specifica-tions relate to the reactor coolant system pressure isolation valves CF30, DH,76 (CF31, DH77) and the thirty day reporting requirements respec-tively.

The safety function of the pressure isolation valves CF30, DH76 (CF31, DH77) and containment isolation valve DHIA (DH1B) is to provide a pressure isolation barrier between the high pressure reactor coolant system inside the containment and the low pressure Decay Heat Remcval System outside ths cont.-inment. Whenever the integrity of a pressure isolation i valve listed in Table 3.4-2 cannot be deconstrated, Toledo Edison will determine the integrity of the high pressure line by performing either a leakage test of the remaining pressure isolation valve or a combined leakage test of the remaining pressure isolation valve in series with the closed motor operated containment isolatica valve. The combined leakage test can be done when the reactor is in mode 3 during shut down or startup. If at least cne of the two pressure isolation valves, DH76 or CF30 (DH77 or CF31) is not leaking and the motor operated containment isolation valve, DHIA (DH1B) is closed before the combined leakage test is performed, the integrity of the low pressure line is not challenged.

Moreover, the plant is only allowed to stay in this configuration (i.e.

with DH1A or DH1B closed) for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per Technical Specifications Sec-tion 3.5.2. Therefore, it is concluded that the combined leakage test will not compromise the safety of the plant.

The safety function of the affected reporting requirements is to make the Nuclear Regulatory Commission (NRC) aware of any situations which-require entry into the action statements per Technical Specifications.

During routine surveillance testing, instrument calibration, or preventive maintenance the plant configurations are sometimes rendered less conser-vative than those established by the Technical Specifications, or the station is made to operate in a degraded mode permitted by a limiting condition for operation. When one of these two conditions exist, the plant will be placed in the action statements. Consequently, a Licensee Event Report (LER) will have to be written and submitted to the NRC per Technical Specification Section 6.9.1.9.

Toledo Edison feels that this reporting requirement is not applicable for the purpose of surveillance testing, instrument calibration, or preventive maintenance. Regulatory Guide 1.16 also supports this excep-tion to the reporting requirement. This Amendment request proposes to incorporate this Regulatory Guide 1.16 exemption clause into the Technical Specifications. This change proposes to delete the reporting requirement when the plant is placed in the action statements due to routine surveil-lance testing, instrument calibration, or preventive maintenance.

Compliance with other existing reporting requirements is not affected by this change.

Pursuant to the above, it is concluded that the proposed changes do not 0 invsive an unreviewed safety question.

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. 's REACTOR C00LAtlT SYSTEM SURVEILLANCE REQUIREMErlTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by: *

a. Monitoring the containment atmosphere parbiculate radioactivity

, monitor at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. Measurement of the CONTROLLED LEAXAGE to the reactor coolant pump seals to the makeup system when the Reactor Coolant System pressure is 2185

+

, _,20 psig at least once per 31 days. -

d. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation.

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:

a. After each refueling outage,

b. Whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or more, and if leakage testing has not been performed in the previous 9 months, and

c. Prior to returning the valve to service following maintenance, repair or repleement work on the valve, 4.4.6.2.3 Whenever integrity of a pressure isolation valve. listed in Table 3.4-2 cannot be demonstrated, the integrity of the remaining pressure isolation valve3 in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the posi-tion of the seer closed alve located in the high pressure piping shall be recorded daily

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! DAVIS-BESSE, UNIT 1 3/4 4-16 Order dtd. 4/20/81 j

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' ADMINISTRATIVE CONTROLS ,

[ e. Failure or malfunction of one or more components which prevents or could preven,t, by itself/ the fulfillment of the fun::tional requirements of system (s) used to cope with accident sr.alyzed in the SAR. .

Personnel error or p'rocedural inadequacy which prevents or could f.

prevent, by itself, the fulfillment of the functiona'l require-ments of systems required to cope with acciecnts analyzed in the SAR.

3 Conditions arising from natural or man-made events that, as a "

direct result of the event require plant shutdown, operation of safety systems, or other protective measures required by technical specifications. '

h. Errors. discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety g! analysis report or in the bases for the technical specifications that have or could have permitted reat. tor operation in a manner less conservative than assumed in the analyses.

i. Per omance of structures , systems, or components that requires reme. 'ial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifica-tions bases; or discovery during plant life of conditions not specifically consid.ered in the safety analysis report or technical specifications that require remedial action or cor-rective measures to prevent the existence or development of an unsafe condition.

THIRTY DAY WRITTEN rep 0RTS5

\, k 6.9.1.9 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty ' days of occurrence of the ,even'.. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the ,

circumstances surrounding the event.' '

a. Reactor protection system or engineered safety feature instru- I ment settings which are found to be less conservative than {

those established by the technical specifications but which do i not prevent the fulfillment of the functional requirements of affected systems. ~

DAVIS-BESSE, UNIT 1 6 17 Jcendment No. 12 p

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Docket No. 50-346 License No. NPF-3 Serial No. 738 September 25, 1981 Attachment 4 I. Change to Davis-Besse Nuclear Power Station Unit 1, Appendix A, Technical Specifications 3.4.8, Table 4.4-4, Figure 3.4-1 and Bases.

A. Time Required to Implement This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request 81-163, Rev. A)

The Technical Specifications values for iodine are currently based upon a parametric evaluation by the NRC of typical site location. The values are conservative in the specific site parameters of the Davis-Besse site. Operating experience shows these. values to be . conservative and when the plant trips, the iodine spike exceeds the limit for Dose Equivalent I-131. Changing the Dose Equivalent limit for I-131 will not exceed 10CFR100 dose limits and will reduce Licensee Event Reports on a normal reactor trip event.

C. Safety Evaluation See attached, bj a/20

Safety Evaluation This amendment request proposes a change to the dose equivalent I-131 concentration in the reactor coolant system from 1.0 uci/gm to 3.2 uci/gm.

The Technical Specifications limitation on the Dose Equivalent I-131 concentration in the primary coolant is established to insure that the resulting two hour doses at the site boundary will not exceed a small fraction of the 10CFR100 limit following a design basis steam' generator tube rupture accident in conjunction with an assumed steady state primary i to secondary leakage ra';e of 1.0 gpm. The two hour dose at the site boundary for 3.2 uci/gn dose equivalent I-131 has already been analyzed in FSAR Section 15.4.2.3 for the design basis steam generator tube rupture accident. The resultant 0.23 Rem whole body and 27.1 Rem thyroid dose as presented in FSAR Table 15.4.2-3 are below one tenth of the 10CFR100 limits of 25 Ree for whole body and 300 Rem for thyroid. The FSAR doses are calculated assuming 1% defective fuel rods oparated at steady state with 1 gpm primary to secondary leakage rate prior to the postulated double-ended rupture accident of one steam generator tube.

The resulting site boundary doses from a less severe rupture should be much lower than the case mentioned above since the activity released is far less in magnitude. Based on the above, it is concluded that the original premise of doses being a small :. a. tion of 10CFR100 limits is still met.

Following the implementation of the proposed change, the primary coolant specific activity sampling analysis program as defined in Technical Specifications Table 4.4-4 will still be complied with as required.

Therefore, it is concluded that no unreviewed scfety question is involved, i

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McQh'qs c f CN'@ e.s t% M, REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY

L'IMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. 1 pCi/ gram DOSE EQUIVALENT I-131, and [

b. 1 100/f PCi/ gram APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1, 2 and 3*.

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a. uCi/ gram- )

With 00SE the specific activity EQUIVALENT I-131 butofwithin the primary coolant the allowable limit > g(below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that operation under these i

circumstances shall not exceed 10% of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

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b. With the specific activity of the primary coolant >g2 uCi/ gram I

I DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T,yg < 530*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With the specific activity of the primary coolant > 100/E j pCi/ gram, be in at least HOT STANDBY with T avg < 530*F wit,hin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

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a. With the specific activity of the primary coolant > 3/fuCi/ gram DOSE EQUIVALENT I-131 or > 100/E pCi/ gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE shall be prepared and submitted to the Commission pursuant to Specification 6.9.1.

This report shall contain the results of the specific activity analyses together with the following inf>rmation:

• With T avg > 530*F.

DAVIS-BESSE, UNIT 1 3/4 4-20

REACTOR COOLANT SYSTEM f

ACTION: (Continued) t

1. . Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the

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first sample in which the lim,it was exceeded, 4

2. Fuel burnup by core region,

3. Clean-up flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded,

4. History of de-gassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded, and

5. The time duration when the specific activity of the primary coolant exceeded M uci/ gram DOSE EQUIVALENT I-131. 9,4 SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analys'is

- program of Table 4.4-4.

DAVIS-BESSE, UNIT 1 3/4 4-21

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AND ANALYSIS PROGRAM e

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1 El TYPE OF MEASUREMENT -

! I' SAMPLE AND AND ANALYSIS MODES IN WHICH SAMPLE C ANALYSIS FREQUENCY AND ANALYSIS REQUIRED

{ [} 1. Gross Activity Determination

'At least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 t

2. Isotopic Analysis for DOSE,

.! 1 per 14 days 1 I EQUIVALENT I-131 Concentration i

,, 3. Radiochemical for EI Determination 1 per 6 months

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) i' 4. Isotopic Analysis for Iodine l a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever # #

l,2,3,4,5 # # #

E3 Including I-131, I-133, and 1-135 the specific activity exceeds

l}l2, jpr6'pci/ gram DOSE EQUIVALENT

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3 I-131 or 100/E pCi/ gram, and

i. b) One sample between 2 and 6 1, 2, 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following a THERMAL

. POWER change exceeding 15 per-cent of the RATED THERMAL i

POWER within a one hour period.

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1. ntil the specific activity of the primary coolant system is restored within its limits ,

• Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed sinc reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > mci / gram Dose Equivalent 1131 l

1 3.'2-DAVIS-BESSE, UNIT 1 3/4 4-23 l ._ __ _ __ _

REACTOR COOLANT SYSTEM O

BASES-3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits shown on Table 3.4-1 provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that con-centrations in excess of the limits will be detected in sufficient time to take corrective action.

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! 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant l ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of the Part 100 limit following a steam generator tube rupture accident in conjunction with an assumed steady, state primary-to-secondary steam generator leakage rate of 1.0

- GPM. The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typ'iccl site locations. These values are conservative in the specific site para-meters of the site, such as site boundary location and meteorological conditions, were not considered in this evaluation. The NRC is finaliz-ing site specific criteria which will be used as the basis for the reevaluation of the specific activity limits of this site. This reevaluation may result in higher limits.

DAVIS-BESSE, UNIT 1 B3/44-5

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r tREACTOR COOLANT SYSTE_M BASES 3

The ACTION statement pemi ing POWER OPERATION to continue for limited time periods with the rimary coolant's soecific activity > 3,2 l pCi/ gram DOSE EQUIVALENT I-1 , but within the allowable limit show on Figure 3.4-1, acconnodates ossible iodine spiking phenomenon which may occur following changes i HERMAL POWER. Operation with specific ac-tivity levels exceeding pCi/ gram DOSE EQUIVALENT I-131 but within the l limits shown on Figure 3.4-1 must be restricted to no more than 10 percent of the units yearly operating time since the activity levels allowed by Figure 3.4-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing T to R 530*F prevents the release of activity should a a

steam generator dbe rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be de-tected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses follow-ing power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary are established in accordance with the requirements of Appendix G to -

10 CFR 50 and with the themal and loading cycles used for design purposes.

l The limitations prevent non-ductile failure during nomal operation,

. including ant'icipated operational occurrences and system hydrostatic tests. The limits also prevent exceeding stress limits during cyclic operation. The loading conditions of interest include:

1. Normal operations, including heatup and cooldown,

2. Inservice leak and hydrostatic cests, and

3. Reactor core operation.

The major components of the reactor coolant pressure boundary have been inalyzed in accordance with Appendix G to 10 CFR 50. The closure head region, reactor vessel outlet nozzles and the beltline region have been identified to be the only regions of the reactor vessel, and con-sequently of the reactor coolant pressure boundary, that detemint the pressure-temperature limitatiuns concerning non-ductile failure.

DAVIS-BESSE, UNIT 1 B 3/4 4-6 i

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