ML20205C876

Proposed Tech Specs Extending Interval for Subj Surveillance Requirement to Scheduled Refueling Outage

ML20205C876
Person / Time
Site:	River Bend
Issue date:	03/10/1987
From:	GULF STATES UTILITIES CO.
To:
Shared Package
ML20205C727	List: ML20205C876 RBG-25461, Application for Amend to License NPF-47,changing Tech Specs to Extend Surveillance Interval for Surveillance Requirements to First Refueling Outage Scheduled to Begin on 870915.Fee Paid
References
TAC-65129, NUDOCS 8703300283
Download: ML20205C876 (42)

Text

ATTACHMENT 1 GULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET 50-458/ LICENSE NO. NPF-47 SAFETY / RELIEF VALVES LICENSING DOCUMENT INVOLVED: TECHNICAL SPECIFICATIONS PORTIONS: 4.4.2.1.l(b) Page 3/4 4-7 REASON FOR REQUEST:

A permanent change is being requested in accordance with 10CFR50.92 and 10CFR50.12 (a) (2) (V) . GSU has and will make a good faith effort to conduct this surveillance on the current frequency if an outage of sufficient duration occurs. In order to test this component, the plant must be in a shutdown condition. To require the plant to shutdown solely to perform surveillances would cause an unnecessary thermal transient on the plant. GSU requests to amend the subject Technical Specifications contained in Appendix A to the River. Bend Station (RBS) Operating License (see Justification) to perform the subject test during a scheduled refueling outage. Should these proposed changes not be granted in a timely manner, GSU will be forced to implement an outage during the first cycle at a large expense to the company and its ratepayers.

DESCRIPTION The RBS Techical Specifications require many surveillance tests be performed every eighteen (18) months (plus a maximum extension defined by Specification 4.0.2). This proposed change is a request to extend the interval for the subject Surveillance Requirement to the scheduled refueling outage (09-15-87).

Tech Spec 3/4.4.2.1 requires a minimum number of Safety Relief Valves be operable in modes 1, 2 and 3. Additionally, the acoustic monitor for each operable valve shall be operable. The Surveillance requirements for this Tech Spec state a calibration of the acoustic monitors must be performed at least once per 18 months. Paragraph 4.0.2 does apply. The calibration testing requirements, including entry into the drywell, result in the need for a unit outage. As a result of the first cycle operation the 18 months, including the tolerance of 4.0.2, will be exceeded on 9/3/87, prior to the scheduled refueling outage with the result being the plant will suffer a reduced capacity to perform the calibration test. Extension of the surveillance period to 9/15/87 will not affect the intent of the test period or the probability of the test failing the required acceptance criteria to allow continued operation.

8703300283 870310 PDR ADOCK 05000458 P PDR i

Jp

~9 NUREG-0737' Item-II.D.3 requires a reliable . indication of valve

-position or flow in the safety relief valve discharge pipe be provided in;the control 1 room. The basic . requirement is to provide the operator unambiguous indication.of flow in the SRV discharge pipe so that appropriate actions'may be taken. Present River Bend Station design as- identified in the FSAR Section

-5.2.2.12 . includes SRV Discharge. Pipe Temperature, SRV

-pilot-actuation. indicating lights.(which show that an actuation signal is present) and SRV Discharge pipe acoustic monitoring.

The present design of the--acoustic monitors and associated calibration. requirements confirm the acoustic monitor sensor as

-installed has no time related drift error. The acoustic monitor sensor,-a piezo electric device, .provides a highly. reliable indication of ' flow in the SRV discharge pipe with a-setpoint of 1% rated flow. Discussions with the vendor also confirmed the frequency of the test is discretionary and extension to the calibration period is acceptable. It can be concluded that the 18 month calibration frequency specified in Item 4.4.2.1 is based on the expected refueling outages and is consistant with vendor recommendation. An extension of the calibration frequency of up to.30 months has been evaluated by GSU with vendor concurrence and found not to reduce the margin of safety as previously analyzed in the safety analysis.

For that portion of the instrument loop, exclusive of the piezo electric device, which may exhibit instrument related drift GSU performs a monthly channel functional test. A review of the present system operation indicates all acoustical. monitor's are operable as confirmed by technical specification- 4.4.2.1.1.a, Channel Functional test, review of the results of this STP indicates that over the past year no adjustments have been required and all surveillance are current as of the date of this submittal. Additional indication of SRV discharge flow is provided by .the tailpipe temperature monitors which are operational and providing clear and consistent information to the

. operators.

Based on the review of the subject documentation GSU concludes that the intent of technical- specification 4.4.2.1.1.b is to allow for testing during a refueling outage, therefore, an extension of the surveillance period is acceptable. This position is consistent with the requirements of NUREG's 0737 (TMI Action Plan), 0898 (RBS-SER) 0800 (SRP), RBS-FSAR and RG-1.45.

The plant design provides additional indication of valve position and SRV discharge. Also, monthly surveillance tests provide confirmation of acoustical monitoring operability.

....-...-se.ed

c  %

~3  :..

"W Y

s SIGNIFICANT HAZARDS CONSIDERATION As : . discussed' [in 10CFR50.92, the following -discussions are' provided to support the NRCfStaff in 'its determination of. "no significant hazards.' considerations".

.a. lThe. proposed: change'does not involve a significant increase inithe probability or consequences ofJan accident -previously evaluated- because the. change in the surveillance interval will'not. result in:a decrease- in the instrument accuracy.

This:Eposition 'is supported by.the manufacture's information thatJLthere are no . time related. drift effects. on the instrument, as- a result the extension will not result'in a decrease in the Safety Relief ~ Valve acoustic monitor-accuracy and therefore, the response to previously evaluated' events will be unchanged. Since'the revision does not involve a design, configuration or operational change to the plant and the response.to events.is unchanged. There is no increase in the. probability or consequence of any accident previously evaluated.

b. The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because'the change in the frequency-of calibration of. the acoustic monitors is consistent with the design.

specification for the system and the safety analysis, and does not involve a design change'or physical change, and' therefore does ~ not. alter the design response of the instrumentation. Thus, no new accident scenario 'is introduced by this revised frequency of calibration of the acoustic monitors.

c. The proposed change to 'the acoustic monitors surveillance period does not'involvefa significant' reduction in a margin of safety because the change in the frequency of calibration for the acoustical monitor sensor is consistent with .the design specification of the system and the sensor design is not sensitive to the surveillance' period. Additionally, as delineated in the justification, the proposed' change will not effect the performance requirements in the Limiting Conditions of Operation contained in the Technical Specification. Thus, the margin of safety is not impacted.

This change is not considered, as stated above, to increase the probability or consequences of a previously analyzed accident or reduce safety margins; further,the results of the change are clearly within an acceptable criteria with respect to the system of component specified in the Standard Review Plan. The basis for 'this conclusion is that calculated drift for the requested frequency will remain within the allowable value discussed in above. Therefore, the criteria for system performance as discussed in the FSAR have not been affected and continue to agree with the Standard Review Plan.

r.i.

Since the proposed- amendment does not change any previously revised and approved description or analysis described in the-FSAR, the proposed amendment.does not create the possibility of a-new or different type of accident, and the proposed change does not involve a significant reduction in a margin of safety. GSU proposes that no significant hazards considerations are involved.

REVISED TECHNICAL SPECIFICATION The requested revision is provided in the Enclosure.

SCHEDULE FOR ATTAINING COMPLIANCE r As indicated above, River Bend Station is currently in compliance

,= with the applicable Techical- Specification. This Technical Specification revision is required prior to August, 1987 to avoid a unit outage- to conduct the required surveillance test as discussed above.

NOTIFICATION OF' STATE PERSONNEL A copy of the amendment application and this submittal has been provided to the state of Louisiana, Department of Environmental Quality - Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL Revision of this Technical Specification does not result in an environmental impact beyond that previously analyzed. Therefore, the approval of this amendment does not result in a significant environmental impact nor does it change any previous environmental impact statements for River Bend Station.

I f

f

. - , - - - - , - - , , - - - - , .,,,--~,-,,n-- -, --,,,,- - - , , ,,.-,,--,,,,,_,-n--,,,-e,,m--,a-,--,.-,

- - - --- ,,-,--,,,-~,-m-,,,,,- , - - ,

W- ^~ . .,_

e

t ENCLOSURE t-i 1

f

REACTOR COOLANT SYSTEM-SURVEILLANCE REQUIREMENTS 4.4.2.1.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

b. CHANNEL CALIBRATION at least once per 18-months,*

refueling 4.4.2.1.2 The relief valve function pressure actuation instrumentation shall be demonstrated OPERABLE by performance of a:

a. CHANNEL FUNCTIONAL TEST, including calibration of the trip unit set-point, at least once per 31 days.

b. CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least once per 18 months.

• The provisions of Specification 4.0.4 are n t applicable provided the surveil-lance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.

RIVER BEND - UNIT 1 3/4 4-8

s ATTACHMENT 2 GULF STATES UTILITIES COMPANY RIVER BEND STATION DOCKET NO. 50-458/ LICENSE NO. NPF-48 DRYWELL BYPASS LEAKAGE LICENSING DOCUMENT INVOLVED TECHNICAL SPECIFICATIONS PORTIONS: 4.6.2.2 Page 3/4 6-19 REASON FOR REQUEST:

A one-time exemption is being requested in accordance with 10CFR50.92 and 10CFR50.12 (a) (2) (V) . GSU has and will make a good faith effort to conduct this surveillance on the current frequency if an outage of sufficient length occurs. In order to test this component, the plant must be in a shutdown condition.

To require the plant to shutdown solely to perform surveillances would cause an unnecessary thermal transient on the plant. GSU requests to temporarily amend the subject Technical Specifications contained in Appendix A to the River Bedn Station (RBS) Operating Licensing for up to 3 days. Assuming approval of the change to the Technical Specifications submitted to the Nuclear Regulatory Commission (NRC) on January 23, 1987 (RBG-25,278) is approved, the subject test will be performed during the first refueling outage. Should these proposed changes not be granted in a timely manner, GSU will be forced to implement an outage at a large expense to the company and its rate payers.

DESCRIPTION The RBS Technical Specifications require many surveillance tests be performed every eighteen (18) months. This proposed changethe is a one-time exemption request to extend the interval for subject flow rate Surveillance Requirement for Cycle 1, to the first refueling outage (09-15-87).

FSAR Section 6.2.1.1.3.4 discusses steam bypass of the suppression pool from the drywell to the containment. The analysis establishes the design value of A/k FK as 1 sq ft based on a small steam leak in the drywell.

Technical Specification 3/4.6.2.2 requires that drywell bypass leakag_e shall be less than or equal to 10% of the acceptable A/ TK design value of 1.0 sq. ft. when Drywell Integrity is required. Furthermore, the surveillance requirements for this technical specification require that a test be performed at least onceper18monthsatadifferentialpressurebetweenthedgrwell and containment of 3.0 psid to determine the value of A/? K for conformance to the requirements of the specification. It is I

proposed that for the first cycle, this drywell bypass leakage test surveillance frequency be extended to coincide with the f

l

refueling outage. It will be shown, that by so doing neither the intent of the test frequency will be affected nor will there be an increased probability of the test failing the required acceptance criteria of.the specification. If the proposed change is approved River Bend Station will be allowed to continue operation:without performing the leak test until September 15, 1987, which is the scheduled date for the first refueling outage.

The pre-operational test for measuring drywell bypass leakage as part of the design verification was performed on February 10 and 11, 1985. The test results have shown that at a test pressure of 3-psid the measured leakage was 647 SCFM compared to the limit of 4011 SCFM _(ev 16%) . At a design pressure of 25 psig the measured leakage was 2425 SCFM compared to the limit of 9660 SCFM (^/25%).

Since then, a drywell bypass leak test was performed on October 25, 1985 just prior to initial criticality utilizing the surveillance test procedure. During the test the A/ff value was computed at 0.014 sq. ft., well within the required value of 0.1 sq. ft. (corresponding to only 14% of the value of the allowed value). The only maintenance performed since which required a recomputation of the A/7F R value was to an electrical penetration in the drywell wall. The leak test performed on the penetration showed an insignificant change in the A/TF E value with the result that the value remains at 0.014 sq. ft.

As stated earlier, the current requirement to test for Drywell leakage is conducted at 18 month intervals. This extension can be justified by the passive nature of the drywell, where it is expected that no dramatic changes in the overall leak rate occurs over extended time intervals. Of those systems which can actively communicate with the drywell, the drywell hydrogen mixing valves are only expected to be operated as part of routine surveillances while operation of the drywell purge ventilation dampers is prohibited when at power. The remaining valves and the access doors are generally not operated except during shutdown conditions. As a result of the passive nature very little system degradation is expected and hence leak testing of the drywell is permitted to be conducted at 18 month intervals.

Extending the test interval a minimal period beyond the guidelines of specification 4.0.2 will not result in any appreciable change in drywell leakage.

Providing this additional surveillance frequency tolerance is consistent with the passive nature of the drywell where little change in the leak rate resulting in suppression pool bypass is expected. Assurance of drywell integrity is demonstrated by conducting the Technical Specification Surveillance Requirements which monitor the configuration of possible drywell bypass pathways. The surveillance requirements and associated periods are summarized as;

- Specification 4.6.2.1 verifies drywell penetrations have operable automatic isolation or are secured in the closed

7 .

4 14E .

r i

F position every 31-days .and ' verify the personnel door .is operable-every 7Ldays.

=-- . Specification- 4.6.2.3 verifies the --drywell air ~1ock seal system as operable every 7 days and tested within 72~' hours

following its use.

Specification 4.6.2.7 ' verifies. the drywell vent and purge system is sealed closed every 31~ days.

Specification 4.6.3.1 verifies the suppression. pool is within

' allowable. volume limits every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4 Specification. 4.6.6.2 . verifies every 7 days the-containment /drywell hydrogen mixing system is open less than 5 hours in. conditions 1 and 2 and less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> in=

condition 3 per 365 days.. Also, only one pathway-may be open-within the above limits.

The _drywell. to- containment openings above are. the- only.

significant-suppression pool bypass pathways which could Ebe

- opened. As'noted, each offthe pathways are verified to be in the-correct configuration at intervals not exceeding 31 days.

As previously mentioned, drywell structural leakage integrity is not expected to dramatically change over the extended test interval.. Finally, prior testing' -conducted on the drywell indicates leakage as'only a small fraction of the current total

' leakage allowed by technical specifications. Allowing operation

-of this facility; through September 15, 1987 without performing

- the: leak test will allow operational flexibility consistent with the. maximization of the plants capacity _ factor as intended by the original surveillance frequency of'18 months. Normally, an_ 18 month surveillance frequency is sufficient to allow operation

from refueling outage to refueling outage. However, startup

- testing' and non scheduled outages have extended the first cycle of River Bend Station so that the refueling outage, where the test-is normally performed, is beyond the allowed interval. This

one time extension, as discussed above, does not affect the g drywell's ability to meet the' safety analysis leakage

. specification.

i I' ,.

i I

SIGNIFICANT HAZARDS CONSIDERATION The proposed amendment to the Technical Specifications would not involve a significant increase in the probability or consequences of an accident previously evaluated because there is no change in the design or performance of plant systems or components from those evaluated in the Final Safety Analysis Report (FSAR). The proposed revision is consistent with the accident analyses described in the FSAR. Due to the near passive nature of the

.drywell structure, allowing the Drywell Bypass Leak Test to be performed at the refueling outage will not result in any additional loss of structural integrity. Test data previously obtained reveals that the current drywell bypass leakage is a small fraction of the allowed leakage hence removing further the possibility of exceeding the design analysis. No increase in the probability or consequences of an accident, therefore, exists.

The proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated because this change does not involve a design change or involve a change in the operating mode of existing equipment. Thus, no new accident scenario is introduced.

The proposed change does not involve a significant reduction in the margin of safety because the margin of safety discussed in the FSAR Section 6.2.1.1.3.4 assumes the drywell bypass test to be conducted at each refueling outage. Allowing the technical specification to agree with the statement in the FSAR does not, therefore, cause a reduction in the margin of safety previously evaluated. Item one above furthermore, discussed the passive nature of the drywell and it is because of this passive nature that it can also be stated that no reduction in the margin of safety exist.

Since the proposed amendment does not change any previously revised and approved description or ana_ysis described in the FSAR, the proposed amendment does not create the possibility of a new or different type of accident. Finally, GSU concludes that the proposed change does not involve a significant reduction in a margin of safety because there will be no detrimental changes in the availability, operability or reliability of the systems.

REVISED TECHNICAL SPECIFICATION The requested revision is_provided in the Enclosure.

SCHEDULE FOR ATTAINING COMPLIANCE As~ indicated'above, River Bend Station is currently in compliance with the applicable Technical Specification. This Technical Specification revision 11s required prior to September, 1987 to avoid a unit outage to conduct the required surveillance test as discusssed above.

' NOTIFICATION OF STATE PERSONNEL A copy of the. amendment application and this submittal has been provided .to the state of Louisiana, Department of Environmental Quality - Nuclear Energy Division. ,

ENVIRONMENTAL IMPACT APPRAISAL Revision of this Technical Specification-does not result in an environmental impact beyond that previously analyzed. Therefore, the approval of this amendment does not result in a significant environmental impact nor does it- change any previous environmental-impact statements for River Bend Station.

i A

i i

a o n-, n -- A 1

e O

ENCLOSURE

CONTAIMENT SYSTEMS DRYWELL BYPASS LEAKAGE LINITING CONDITION FOR OPERATION 3.6.2.2 Drywell bypass leakage shall be less than or equal to 10% of the acceptable A/8 design value of 1.0 ft8 APPLICA8ILITY: When DRYWELL INTEGRITY is required per Specification 3.6.2.1.

ACTION:

With the drywell bypass leakage ' greater than 10% of the acceptable A/8 design value of 1.0 ft8 , restore the drywell bypass leakage to within the limit prior to increasing reactor coolant system temperature above 200'F.

SURVEILLANCE REQUIREMENTS 1

4.6.2.2 At least once per 18 months

• the drywell bypass leakage rate test _

shall be conducted at an initial differential pressure of 3.0 psid and the A//k shall be calculated from the measured leakage. One drywell airlock door shall remain open during the drywell leakage test such that each drywell door is leak tested during at least every other leakage rate test.

a. If any drywell bypass leakage test fails to meet the specified limit, the schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the limit, a test shall be performed at least every 9 months until two consecutive tests meet the limit, at which time the 18 month test schedule may be resumed.

-br---The-provisions-of-Spesification-4rgr2-are-not-applicabler Change submitted i under RBG-25278 1-23-87.

For the first cycle only, this may extended to coincide with the refueling outage, but not beyond 9-15-87.

RIVER BEND - UNIT 1 3/4 6-19

e l

p .

p ATTACHMENT 3

[

t GULF STATES UTILITIES COMPANY RIVER BEND STATION i DOCKET 50-458/ LICENSE NO. NPF-47 PRIMARY CONTAINMENT /DRYWELL HYDROGEN MIXING LICENSING DOCUMENT INVOLVED: TECHNICAL SPECIFICATIONS PORTIONS: 4.6.6.2(c) Page 3/4 6-63 REASON FOR REQUEST:

A one-time exemption is being requested in accordance with 10CFR50.92 and 10CFR50.12 (a) (2) (V) . GSU has and will make a

good faith effort to conduct this surveillance on the current frequency if an outage of sufficient length occurs. In order to test this component, the plant must be in a shutdown condition.

To require the plant to shutdown solely to perform surveillances would cause an unnecessary thermal transient on the plant. GSU requests to temporarily amend the subject Technical Specifications contained in Appendix A to the River Bend Station (RBS) Operating License for up to 43 days (see Justification) to perform the subject test during the first refueling outage.

Should these proposed changes not be granted in a timely manner, GSU will be forced to implement an outage at a large expense to the company and its ratepayers.

DESCRIPTION The RBS Technical Specifications require many surveillance tests be performed every eighteen (18) months (plus a maximum extension defined by Specification 4.0.2). This proposed change is a one-time exemption request to extend the interval for the subject flow rate Surveillance Requirement for Cycle 1, to the first refueling outage (09-15-87).

Technical Specification 3/4.6.6.2 requires that two primary containment /drywell hydrogen mixing systems be operable during operational conditions 1, 2 & 3 with the inlet and outlet valves closed except during the limited time that one inlet or outlet valve is open for controlling drywell pressure. The surveillance requirements of 4.6.6.2.c specify that the flow rate for each redundant hydrogen mixing system train is verified at least once per 18 months. The testing necessary to meet this surveillance requirement requires a unit outage since only a single inlet or outlet valve may be opened during operational modes 1, 2, or 3 for pressure control without exceeding the design basis drywell bypass leakage allowable of A/1 F i=1.0 sq. ft.

As a result of the first cycle operation, the 18 month surveillance interval, including the tolerance of paragraph

4.0.2, will be exceeded prior to the scheduled refueling outage.

It will be shown that an extension of the surveillance interval until the scheduled refueling outage will not affect the intent of the test interval or the probability of failure to meet the test acceptance criteria. The proposed change will allow continued operation until a scheduled outage in September, 1987 in lieu. of an additional outage in August, 1987 to perform the subject surveillance test.

Regulatory Guide 1.7, " Control of Combustible Gas Concentration in Containment Following a Loss-of-Coolant Accident", requires that a means be provided to control combustible gases (i.e.,

hydrogen) within the primary containment following a design basis Loss of Coolant Accident (LOCA). The basic requirement is to maintain the hydrogen concentration below the lower limit for l upward flame propagation (4 percent by volume.) The present j design of the RBS Combustible Gas Control System, presented in l FSAR section 6.2.5 complies with this regulatory guidance by including subsystems to monitor the hydrogen concentration, to l mix the drywell/ containment atmospheres, and to reduce the hydrogen concentration within the primary containment as required i by RG 1.7 and a 10CFR50.4 Part 44. In addition, the RBS design

( includes a backup containment hydrogen purge system to aid in

long-term post-LOCA cleanup and a hydrogen igniter system to l provide a method of controlling the hydrogen concentration by deliberate ignition for degraded core events which produce hydrogen in excess of the amount generated during a design basis event.

Regulatory Guide 1.7 is concerned with the possibility of the l

rapid reaction of hydrogen with oxygen producing high l temperatures and/or significant overpressure of the containment

! resulting is a breach of containment or damage to systems and l components essential to the continued control of the post-LOCA l conditions. .

Evaluation of the drywell response to degraded core accidents, (a 75% metal water reaction) performed by the Hydrogen l Control Owners' Group (HCOG) indicates that this is not a concern l even for much more severe accidents than described in Regulatory l Guido 1.7.

l The design basis of the Containment /Drywell hydrogen mixing i system is to reduce the post-LOCA drywell hydrogen concentration l below 4 percent by volume by dilution with the containment l atmosphere. The containment hydrogen concentration, in turn,

! will be reduced by recombination with oxygen in the redundant l Hydrogen Recombiners. Either primary containment hydrogen l recombiner in concert with either containment /drywell hydrogen mixing system is capable of controlling the hydrogen generation associated with ( 1) Zirconium-metal water reactions (MWR) involving 1% (Design Basis) of the active fuel cladding (2) radiolytic decomposition of water and (3) corrosion of metals within the containment and drywell. Current operating procedures require initiation of the containment /drywell hydrogen mixing l

l I

i

system when the drywell hydrogen concentration reaches 3.5 volume percent (approximately 2 1/2 hrs. post-LOCA). The containment hydrogen recombiners are also initiated at 3.5 percent by volume in containment (14 days post-LOCA). Operation of one recombiner and one train of the drywell mixing system will maintain the drywell and containment hydrogen at or below 3.5 percent by volume at all times. Therefore, there is margin between the design basis for these system and their actual capability.

The containment /drywell hydrogen mixing system flow rate surveillance interval of 18 months as given in Technical Specification 3/4.6.6.2.c is based on the expected refueling interval rather than any system technical requirements. This conclusion is supported by the fact that this system is not an active system. Its operation is required only during the post LOCA period. Degradation of the fan performance (i.e., flow rate) between surveillances is not expected because the fgn motor is qualified to operate in a normal environmental of 140 F (100%

humidity) and an accident environment of 212 F (steam) while the safetg analysis requirements are limited to a normal environment of g0 F (50% humidity) and a design basis event environment of 165 F (100% humidity). The qualified life of this equipment is 40 years normal environment concluding with a 100 day event.

This system employs pipe rather than ductwork so that flow reduction due to deformation of ductwork is not a concern. Based on the system design and the conservatism of its design there is no expectation of degradation of flow capacity as a function of time.

The system flow rate has been verified to be in excess of 600 cfm during the startup test program and performance of STP-254-0601 which was conducted on 9/15/85. In addition, the surveillance required by 3/4.6.6.2.b, system functional test which is conducted during cold shutdown if not performed within the previous 92 days has been satisfactory for the previous 6 performances for the time period of 9/8/85 through 1/13/87. This has demonstrated that the system is both reliable and operational. Therefore, there is no reason to believe that this system will not perform its intended function when required.

The margin of safety for this system concerns the ability to maintain the post-LOCA drywell hydrogen concentration below 4 percent by volume. Current analysis demonstrates that one containment /drywell mixing system operating at a flow rate of 600 cfm in conjunction with one containment hydrogen recombiner will maintain the drywell hydrogen concentration at or below 3.5 percent by volume at all times post-LOCA. Thus, based on the startup and test performance and subsequent surveillance test which demonstrated a flow rate of greater than 600 cfm and coupled with the 100% pass rate for functional testing GSU had concluded that this liydrogen Mixing system will be capable of performing its intended function when required.

Additional margin was provided by other conservative analysis assumptions. First, the hydrogen generated was assumed to remain in the drywell until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the accident. In reality, the drywell hydrogen concentration would be reduced during this time interval due to bypass leakage flow or flow through the LOCA vents due to the drywell pressure exceeding the containment pressure. Secondly, the zinc corrosion rate used in the analysis conservatively assumed that zine corrosion is a function of temperature only. This assumption does not take credit for the buildup of an oxide layer which would inhibit further corrosion.

As discussed above, the containment /drywell hydrogen mixing system is not sensitive to any time related degradation of flow capacity. The intent of Technical Specification 3/4.6.6.2.c was to allow for testing during a refueling outage. Therefore, an extension of this surveillance interval is acceptable. This position is consistent with the requirements of Regulatory Guide 1.7, 10CFR50.44, NUREG-0989 (RBS SER sections 6.2.5 and SER 2, 4,

& 5), and the intent of Technical Specification 3/4.6.6.2.c.

SIGNIFICANT HAZARDS CONSIDERATION As discussed in 10CFR50.92, the following discussions are provided to support the NRC Staff in its determination of "no significant hazards considerations."

The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated because the change in the frequency of flow rate tests is considered within the design specification for the Primary Containment /Drywell Hydrogen Mixing system and the safety analysis. This is supported by the lack of mechanistic means of system flow degradation and the considerable conservatism in the system as designed. Also functional testing successfully conducted six times in the previous 15 months confirms system reliability and functions. This test verifies valve and form operability. Since this proposed change does not involve a design change or physical change to the plant, it does not increase the probability or consequence of any accident previously evaluated.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the change in the frequency of flow rate testing, is considered within the design specification for the hydrogen mixing system and the safety analysis, and does not involve a design change or physical change, and therefore does not alter the singic failure design. Thus, no now accident scenario is introduced by this revised frequency of flow rate testing.

The proposed change to the surveillance period does not involve a significant reduction in a margin of safety because the change in the frequency of rate testing is considered within hydrogen

f f

mixing system and the safety analysis since there is no credible mechanism for degradation of actual capability due to extension of the surveillance interval, this margin of safety should be maintained.

This change is not considered, as stated above, to significantly increase the probability or consequences of a previously analyzed accident or reduce safety margins; further the results of the change are within acceptable criteria specified in the Standard Review Plan. The basis for this conclusion is that the system flow rate specified (600 cfm) will remain equal or greater.

Therefore, the critoria for transients and accidents discussed in the FSAR have not been affected and continue to agree with the Standard Review Plan.

Since the proposed amendment does not change any previously revised and approved description or analysis described in the FSAR, the proposed amendment does not create the possibility of a new or different type of accident, and the proposed change does not involve a significant reduction in a margin of safety. GSU proposes that no significant hazards considerations are involved.

REVISED TECHNICAL SPECIFICATION The requested revision is provided in the Enclosure.

SCHEDULE FOR ATTAINING COMPLIANCE As indicated above, River Bend Station is currently in compliance with the applicable Technical Specification. This Technical Specification revision is required prior to August, 1987 to avoid a unit outage to conduct the required surveillance test as discussed above.

NOTIFICATION OF STATE PERSONNEL A copy of the amendment application and this submittal has been provided to the state of Louisiana, Department of Environmental Quality - Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL Revision of this Technical Specification does not result in an environmental impact beyond that previously analyzed. Therefore, the approval of this amendment does not result in a significant environmental impact nor does it change any previous environmental impact statements for River Bend Station.

I i

F:

ENCLOSURE

CONTAlle4ENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.6.2 Each primary containment /drywell hydrogen mixing system shall be demonstrated OPERA 8LE:

a. At least once per 7 days, by determining the cumulative time that:

1. The hydrogen mixing system inlet or outlet line has been open during OPERATIONAL CONDITIONS 1 and 2 during the past 365 days, and

2. The hydrogen mixing system inlet or outlet line has been open during OPERATIONAL CONDITION 3 during the past 365 days.

b. During each COLD SHUTDOWN, if not performed within the previous 92 days, by:

1. Starting the system from the control room, and .

2. Verifying that the system operates for at least 15 minutes,

c. At least once per 18 months *by verifying a system flow rate of at least 600 cfm. ,

• May be extended to the refueling outage during the first cycle but no later than 9-15-87.

RIVER BEND - UNIT 1 3/4 6-63

i 1

ATTACHMENT 4 GULF STATES UTILITIES CONPANY RIVER BEND STATION i

DOCKET 50-458/ LICENSE NO. NPF-47 LEAK RATE TESTS l LICENSING DOCUMENT INVOLVED: TECHNICAL SPECIFICATIONS Items: a. 4.4.3.2.2a Page 3/4 4-12

b. 4.6.1.3d Page 3/4 6-5

c. 4.6.1.3f Page 3/4 6-5

d. 4.6.1.31 Page 3/4 5-6

e. 4.6.2.3.d.2 Page 3/4 6-21

f. 4.6.2.1.e.2 Page 3/4 6-18 REASON FOR REQUEST:

A one-time exemption to Technical Specification Leakage Testing is being requested in accordance with 10CFR50.92 and 10CFR50.12 (a) (2) . GSU has and will make a good faith effort to conduct these surveillances within the current frequency if an outage of sufficient length occurs. In order to test certain components, the plant must be in a shutdown condition. To require the plant to solely to perform surveillances would cause an unnecessary thermal transient on the plant and result in unnecessary exposure to workers. GSU requests to temporarily amend the Technical Specifications contained in Appendix A to the River Bend Station (RBS) Operating License, NFP-47, for a maximum of 41 days to perform the subject leak rate tests during the first refueling outage on a total of fifty-two (52) valves. Should these proposed changes not be granted, GSU will be forced to implement an unscheduled outage no later than August 5, 1987.

DESCRIPTION The RBS Technical Specifications require many surveillance tests be performed every eighteen (18) months (plus a maximum extension as defined by Specification 4.0.2). This proposed change is a one-time exemption request to extend the interval for the leak rate Surveillance Test Requirements for the items listed below, to the scheduled start of the refueling outage, 09/15/87. The types of leak rate tests involved are 10CFR50, Appendix J, III.D.3 (Type C), Reactor Coolant System boundary valve tests (10CFR50, Appendix A, General Design criteria 54), and tests of air systems providing a seal against Drywell Bypass Leakage. The affected surveillance tests are described as follows:

1

(

~

F a- .

2; f

SURVEILLANCE REQUIREMENT DESCRIPTION 4.4.3.2.2a' This surveillance requires each reactor coolant. system-. pressure isolation valve'specified' in ' Table 3.4.3.2-1 shall' be demonstrated OPERABLE by' leak testing pursuant to.

Specification 4.0.5 including.

paragraph IWV-3427(B) of 'the .ASME , _

j. Code 'and- verifying . the leakage of

' each- valve to be- within 'the specified limit at least once'per'18 months.

[_

4.6.1.3d This surveillance requires that Type B and C containment leakage rate tests :be conducted at - intervals no

_greaterit han 24 months.

l 4.6.1.3f This surveillance requires a total sealing air leakage into the primary containment shall be determined by-test at'least once;per 18 months..

4.6.1.31 This surveillance ~ requires ' primary containment isolation valves in-p hydrostatically tested lines- :per

Table 3.6.4-1 to be leak tested at

( least once per 18 months.

4.6.2.3d.2 This surveillance requires.the 4.6.2.le.2 drywell airlock and the personnel door in the drywell equipment hatch.

inflatable seal system be OPERABLE l by conducting at least once'per 18 L

months a seal pneumatic system leak

test and verifying that system

! pressure does not decay ~more than 0.67 psig from 75 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

- In addition to the exemptions to the referenced specification sections, a temporary exemption is also requested to Section III.D.3 of 10CFR50 Appendix J. This section requires Type C

- testing of isolation valves at intervals not to exceed 2 years,

. which is the basis of Section 4.6.1.3.d of the Technical Specifications.

l Similar precedents for the proposed leak rate test extensions exist. The following documents reference exemption requests granted by the NRC to allow a nuclear station to continue operation to their next scheduled outage:

L :.

F'

-s 1, I

, . 4

• p 't

"}

\,'

tt <

1. ' NRC April 7, 1.986 approval in response to a Februar;% 27...

-1986 request, Docket No. 50-302; i '

i'

2. NRC ; March 18', 1986 approval in response to a January 17,

'1986 request,~-Docket No. 50-397; " Y u

- 3. NRC March 3,1986 approval' in- response to a December 18, , F

~.1985 request, Docket JJo. 50-352. ,,7 4

[y ', [ '

'N

. .; , 't,

/ '. t t \'

u .

s 10

% ( )

s a f r

\

i w $ \

g j

t f1 D 1m )" y

'.y 5 g 1 (

t i 6, #

xs

\

r

~

h

-.e j

' , b

/ 1 o' I

_____.__.__________.___

\ 'l e

Item a.

Tech Spec 4.4.3.2.2a j.

RCS BOUNDARY LEAKAGE 'N s' )

Surveillance requirement 4.4.3.2.2a for RCS. pressure boundary

• provides added assurance;of valve integrity, thereby reducing the probability' of gross valve failure and consequent intersystem-LOCA. l

\

] Reactor Coolant System (RCS) leakige Tech Spec 4.4.3.2.2a

', requires 18 month, testing of 15 valves. RBS is requesting for 13 valves a maximum extension of 40 days. Of these valves only two a showed any leakage at the last testing,.both 0~.0.L gpm or less, as 3 all others had 0.0 gpm leakage. Technical Specification 3.4.3.2 allowable leak rate is 0.5 gpm per nominal inch of valve size up

,to a maximum of 5 gpm per valve at RCS pressure of 1025+ 15 psig

-. ' pursuant to Technical Specification 4.0.5 and ASME Code IWV-3427

{, '[ (B). Bdsed on the valve sizes the allowable' leakage rates., range -

~, . from a minimum of three (3) gpm to a maximum of five (5) gpm.

t. . . ' The aforementioned results of 0.01 gpm leakage is well below the />

allowable leakage rates. .

x T

k; i, S s'

s o

/'

\

e

'i .

i "t'

d

. k

't

, ,. . . , , ,_~. , ._. -_~.

Item b.

Tech Spec 4.6.1.3d APPENDIX J, TYPE C, OUTLEAKAGE TESTING GSU successfully completed the preoperational Type A integrated leak rate testing (ILRT) on 04/06/85 and the Type B and C local leak rate testing (LLRT) on 04/07/85. The LLRT results were submitted to the NRC via RBG-21814 from J. E. Booker to H. R.

Denton dated 08-07-85. Both the LLRT ar.d ILRT results demonstrated a low level of leakage.

The performance of the preoperational ILRT/LLPJF testing was scheduled to be- consistent with a projected fuel load date of April 1985. The intent of the scheduling was to allow adequate time for the first cycle of operation so as to satisfy the 24 month Type B and C testing requirements of Surveillance Requirements 4.6.3.1.d at the first refueling outage. Due to various factors, the low power license of River Bend was issued on August 29, 1985 and commercial operation of the facility occurred in June, 1986. Therefore, GSU is faced with an extended outage late in the fuel cycle in order to perform Type C testing which was not able to be scheduled during various midcycle outages. Containment leakage rate is primarily affected by equipment wear and maintenance. Isolation valves typically see little usage except for periodic operability testing. This translates to little, if any, expected increase in leakage rate.

From the time of performance of the ILRT/LLRT testing to receipt of the low power license in August 1985, Type C equipment was subject to only minimal wear.

One hundred sixty-six valves at River Bend are associated with Surveillance requirement 4.6.1.3.d and 10CFR50 Appendix J Type C outleakage testing. Only 5 valves tested under these requirements require extension. Total leakage for valves requiring extension was found to be 307.5 standard cubic feet per day (scfd) at the most recent LLRT. The as-left leakage for the entire scope of type C tested valves is presently 2114.64 (scfd).

Currently 3412 scfd is allowed by the Tech Spec leaving a margin of 1297 scfd. Using a straight line extrapolation for the first available outleakage total since fuel load to the latest available as-found leakage, the leak rate as of 9-15-87 is expected to be 3080.69 scfd. Additionally, Technical Specification 4.6.1.3.d and Section III.D.3 of Appendix J require a 24. month test interval with RBS requesting an extension ranging from 5 to 29 days.

., . . . _ . . . - _ . . _. _ ___ - m._ > . . . _ . . . _.

_m a.

. ~ .f,.  ;

'e-

?

hn

,4' :

{

Item c. i

i ~

Tech Spec 4.6.1.3f i CONTAINNENT INLEAKAGE TESTING l l- (Leakage-l tests' are. performed on each. containment isolation valve  ;

sealed lby Main Steam Positive Leakage' Control System (MS-PLCS)

,i , and . Penetration Valve Leakage Control System.(PVLCS) similiarly.

Jto 10CFR50, Appendix J,. procedures'for Type C valves.

-In addition'to the. identified. 10CFR50 Appendix J and 10CFR50.

~

Appendix A' .L'eak Test Valves,. Technical Specification 4.6.1.3.f .

' requires testing _of valves sealed by the MS-PLCS and PVLCS- for

-containment ~inleakage. .; The design and operation of these systems- ,

n :are' addressed in_the FSAR, Sections _6.7-and.9.3.6. This test is to ensure ' leakage from the associated pressurized systems into  !

~ containment'will;be limited to 50% (7.5 psig) . of the containment j Ldesign ' pressure: -(15 psig) during a 30 day period following an l event. -The staff's evaluation.of this design ~is contained- in Section- 6'.2.3 .and 6.7.3 of the. Safety: Evaluation. Report;  ;

(NUREG-0989) . Technical Specification Bases 3/4.6.1.5 (MS-PLCS),_ j

,. .and. 3/4.6.1.10 (PVLCS) and Reg. Guide 1.96 indicate.the. purpose  :

.of these' systems -are to minimize the _ leakage of'_ untreated ,

containment atmosphere.

! been addressed. through. Containment conservativepressurization. analysis. concerns The have resulting r L -Technical Specifications restrict- the pressurization due to_ .

F

~

inleakage to 50% of design pressure over a period of 30 days.with ru) .. operator intervention. Inleakage testing set forth in

_ Technical, Specification _4.6.1.3.f,-requires an extension beyond i Tech. Spec 4.'O.2' limits of 6 to 33 days for 25 valves. A total of 39 valves (64% of total) encompass the inleakage. criterion. The valves ~ -to be extended constitute as-left leak rates of less than 1 one-third of allowable inleakage rates. . Those valves not included in this request, including Main Steam Isolation Valves,

. were previously tested during the October - November 1986 outage.

L1 .As/previously submitted to the staff (RBG 19,995 from J. E.

,- . Booker to H. R. Denton, dated 01/28/85, Gulf States Utilities' y, , Company (GSU) response to the River Bend Station (RBS) Safety 4  ; Evaluation Report '( SER) - Confirmatory Item #15 - Containment

- Repressurization: SER Sections 6.2.3, page 6-22 and 6.7.3, page 6-53), an analysis has been performed to show all sources of L

!-- containment inleakage, i.e. the ?enetration Valve Leakage Control

' ' System (PVLCS) and the Main Steam - Positive Leakage Control ,

' System (MS-PLCS), do not repressurize the containment to more

, Lthan 50 percent of the containment design pressure during the 30

~ day-period following a LOCA. The analysis determined that 425 SCFH is -the maximum amount of acceptable containment inleakage

^

s from the PVLCS and MS-PLCS. However, as a safety margin, the allowable containment inleakage is set at 80 percent of the

acceptable inleakage, i.e. at 340 SCFH, in the RBS Technical

! Specifications.

e v

1 , - - - , ,. .-,, ,--,,.- -,,.. ..,,,, , - ...,. ,., _ ,_.. - _ _ _ _ _ _ . _ _ _ ,._ ,~. _ _ _ -- _ _ _

Additionally, if containment inleakage should exceed that allowed by Technical Specifications due to the manual initiation of MS-PLCS and/or PVLCS after a. DBA LOCA, sufficient diverse indications exist to alert the plant operations staff. of the containment pressurization. Furthermore, sufficient and diverse containment pressure control methods and the Emergency Operating Procedures governing their use are available for the operations staff to ensure containment integrity can and will be-maintained.

GSU has determined that sufficient technical justification exist to merit the requested extension of 6 to 33 days.

i.

L

Item d.

Tech Spec 4.6.1.3i UNIDENTIFIED LEAKAGE - HYDRO TEST Hydrostatically tested. containment isolation valves are indicated in table.3.6.4-1 of the Technical Specifications and Table 6.2-40 of- the FSAR. " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," 10CFR50 Appendix J, allows . testing with ' fluid .at 1.10 Pa, as' long as acceptance is' based ~on Technical Specification. fluid leakage limits and sufficient fluid inventory is available to provide a water seal for 30 days at a pressure of 1.10 Pa. NUREG 0800, The Standard ' Review Plan, states, " Hydrostatic testing of containment isolation valves is permissible if the line is not a potential containment atmosphere leak path and may be found acceptable if it can be demonstrated in accordance with the requirements of Section III.6 of. Appendix J, that a liquid inventory is available to maintain a water seal

. . . during the post accident period." Due to the location of these valves and their elevation in relation to the suppression

' pool, the requirements of Appendix J and NUREG-0800 are satisfied. Review of the River Bend SER shows that ". . . the applicant indicated that a liquid inventory will produce a water seal during the' post accident period and only liquid leakage.from the containment will occur. The combined leakage from all these valves will satisfy the acceptance criteria of 10CFR100 regarding the site radiological safety analysis and will be included in the Plant Technical Specifications."

Review of the regulatory guidance provided indicates that' valves may be water tested if they will remain below the elevation of the top of the suppression pool during a design basis accident and do not provide a secondary bypass path. This insures that these- lines can only leak water, and not containment atmosphere, into the secondary containment. No gaseous release is possible with the maximum suppression pool temperature of 185 degrees fahrenheit, so 10CFR100 and Appendix J are not compromised.

Therefore, these valves are excluded from the 0.6La leakage total of 10CFR50 Appendix J, with a slight penalty of testing at 1.10 Pa. The leakage amounts presented in the Technical Specifications are based on limits derived from 10CFR100. The valves involved are ECCS, RCIC and DER isolation valves and are not generally utilized except for periodic operability testing.

Leakage from these valves has been very low and no rework has been required to meet the imposed leakage criteria.

Of the 24 valves covered by Technical Specification 4.6.1.3.1, only nine valves require extension of their 18 month frequency.

Allowable leak rate for these components is 1 gallon per minute per valve. All these valves have been previously tested with several as-left leakages of 0.0 gpm. Total leak rate for all 9 valves to be extended is 0.319 gpm with 9 gpm allowable.

Extensions required range from 32 to 41 days. This is less than a 6% extension. This test is performed at 1.1 Pa, 8.36 psig.

l 4

During the latest outage.several valves with this- requirement were tested with no failures, and no rework required.

Extension of testing frequencies for these valves is not expected i to increase overall leakage rate in excess of allowable due to l the excellent results from previous testing. In addition, the  ;

subject valves are a minority of the_ total ~and as described above '

do not provide a secondary bypass path. GSU request the subject i surveillance extension. Also, note the annulus bypass leakage and secondary containment bypass leakage are not affected by this request.

, - - -g -- .- e - .,,,,,m - - , , ----un y -

~p: --ey

Item e.

Tech Spec 4.6.2.3.d.2 and 4.6.2.1.e.2 DRYWELL PERSONNEL DOOR AND EQUIPMENT DOOR LEAK TEST Also being requested is relief from requirements to perform leak tests on the air system on the drywell airlock and the personnel door.in the drywell equipment hatch. These air - systems, which are tested- at 18 month intervals, supply air to the inflatable door seals. The inflatable seals, along with all other drywell boundary seals, provide a barrier to any leakage from the drywell to the containment, assuring all steam resulting from line breaks is routed through the suppression pool for condensation. The Technical Specification bases require only one drywell airlock door to be closed and operable. The drywell bypass test verifies that one door is adequate by opening one door for the performance of the test utilizing only one door to limit bypass leakage.

The drywell airlock and personnel doors, due to their location',

are seldom operated and see little wear to their mechanical components or inflatable seals. Historically testing has shown little leakage from either of the air systems or from the pressure tests conducted between the door seals or on the airlock as a whole. The air systems operate at a relatively low pressure, 75 psig, and the incoming air is regulated at the door which lessens the effects of any instrument air system pressure fluctuations or spikes. These items contribute to the good test values found in the past. The drywell airlock inner door was tested satisfactorily on 01-14-87 during the last outage when maintenance was performed on the air system. The Technical specifications on the drywell airlock are satisfied with one operable door closed and locked. The exemption period requested for the drywell outer door and the personnel door is 21 days.

The air systems are both redundant and monitored. Each seal has an individual air system, with a dedicated accumulator and check valve. The air systems are designed and installed to ASME III.

Upon loss of incoming instrument air or seal leakage to below 75 psig, an alarm is received in the Control Room. This insures that if a leak occurs, operators are aware and can take appropriate action.

In summary, the maximum exemption period requested is 21 days.

Historical test data of the seal leakage STP, as well as the saveral other leakage STPs, indicate low values due in part to low usage. The redundancy and supervision in the air nystem - dual seals, accumulators isolation valves, control room alarms - provide adequate reliability and ability to detect and respond to problems if they develop. .

SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Commission had provided guidance concerning the application of standards' in 10 CFR 50.92 for determining whether license amendments involve significant hazards consideration by providing certain examples which were published in Federal Register on April 6, 1983 (48 FR 14870) . One of the examples (vi) of an action involving no significant hazards consideration is a change which may in some way reduce a safety margin, but where the results of the change are clearly within all acceptable criteria.

The foregoing requested change and exemption fits this example.

Postponing the aforementioned leak rate tests until an outage commencing on or before September 15, 1987 would allow for continued operation of the plant and would have little or no effect on containment integrity as discussed in the Technical Justification and for the following additional reasons:

1. For RCS boundary leakage, which cover 13 valves for a maximum extension of 40 days, demonstrated as left leakage was only 0.01 gpm or less for two valves and 0.0 gpm leakage for all others. This is well below the allowable leakage rates of 3 to 5 gpm depending on valve size.

2. Redundant primary containment isolation valves are provided for each process line penetrating containment, i.e. two isolation valves in series. Consequently, a reduction in the effectiveness of one valve to provide a seal would not in itself compromise containment integrity. Deterioration in the overall integrity of the containment penetrations is normally a gradual process. Considering the redundancy of the isolation barriers and the short duration of the requested extension of the testing interval, any reduction in containment integrity during the extension period would be negligible.

The intent of the Technical Specifications and Section III.D.3 of Appendix J to 10CFR50 is to require testing of the isolation valves once every fuel cycle. A normal reactor fuel load is designed to provide an 18 month cycle with approximately 16 months of full power operations.

Consequently, the primary containment isolation valves are normally exposed to 18 months of rated temperature conditions between each Type C test. Those valves for which we are requesting extension will have been subjected to rated temperature conditions for a time frame consistent with the above assumptions. The extension in the Type C test interval does not appear to be inconsistent with the intent of the test schedule specified by the Technical Specifications and Appendix J.

3. -Containment:Inleakage. Testing - LeakageLtests are performed

~on-the process ~1ine isolation valves sealed by.the MS-PLCS cur PVLCS similiarly;to.10CFR50, Appendix J procedures for Type C

, valves. _ This leakage is excluded from the Type C. leakage rate , total 'per Paragraph III.C.3 of Appendix J. . The combined j leakage of valves requiring . extensions constitute as-left leakage rates?of.less than one-third of Lallowable inleakage

-rates, sand ~'those valves. requiring extensions are,'for the most.part, valves which have in the past demonstrated a' high

degree- of reliability. Also,,the MSIVs which are the major sources of' leakage in this system have been . satisfactorily  ;

tested .during' the. October-November 1986 outage. Based on  !

these considerations, GSU has determined .that sufficient technical justification exists to merit the requested extension of 6 to 33 days.

Additionally, if as a result of unexpected degradation or failure of the- subject isolation valves, containment inleakage. .should exceed that allowed by . Technical Specifications due to the manual initiation of MS-PLCS and/or

-PVLCS 20 minutes after a DBA LOCA, sufficient diverse indications exist to alert the plant operations staff of the containment pressurization. Furthermore, sufficient and diverse containment pressure control methods and the Emergency Operating Procedures governing their use are available for! .the operations staff to ensure containment integrity can and will be maintained.

Finally, as previously submitted to the staff (RBG 19,995, J.

E. Booker- to H. R. Denton, dated 01-28-85, Gulf States L ' Utilities Company (GSU) response to the River Bend Station (RBS) Safety' Evaluation Report (SER) Confirmatory Item

1. 15;- Containment Repressurization: SER Sections 6.2.3, page 6-221and 6.7.3, page 6-53), an analysis has been performed to show. all sources of containment inleakage, i.e. the Penetration Valve Leakage Control System (PVLCS) and the Main Steam - Positive Leakage Control System (MS-PLCS), do not repressurize the containment to more than 50 percent of the containment design pressure during the 30 day period following a LOCA. The analysis determined that 425 SCFH is the maximum amount of acceptable containment inleakage from the PVLCS and MS-PLCS. However, as a safety margin, the E allowable containment inleakage is set at 80 percent of the acceptable inleakage, i.e. at 340 SCFH, in the RBS Technical Specifications.

4. Testing per RBS Technical Specification Section 4.6.1.3.i, for which extension is requested, is for hydrostatically tested valves in which leakage is included from the Type C leakage rate per paragraph III.C.3 of Appendix J. Allowable leak rate for these components is one (1) gallon per minute per valve. All these valves have been previously tested with

~

pu g,y

.'several as-left'lhakages-of 0.0 gpm. . Total _ leak, rate for:all

.9 valves'.-to-be extended was 0.319 gpm with 9 gpm allowable.

Extensions required range from 32 to 41 days. :This is less than a.10%. extension. During Lthe; latest outage- several valves with this requirement were tested with no failures.and with no: rework required.

-5.- Also being' requested is relief from~ requirements to perform leak ; tests on the air system for the drywell airlock and the personnel door in the drywell equipment hatch. These air

-systems, which are- tested at. 18 ' month. intervals, supply-instrument air to the. inflatable door seals. The inflatable seals, along-with all other drywell boundary seals, provide a barrier to any leakage from the drywell to the containment, assuring. all- steam _resulting from line breaks is routed through.the suppression pool for condensation._.The Technical Specification bases require only one drywell airlock door'to be closed and operable. The drywell . bypass test verifies that one door- is adequate by _ opening one door for the performance of the test, ucilizing only one door to limit bypass leakage..

The drywell airlock and personnel doors, due to .their location, are seldom operated and see little wear to their

-mechanical components or inflatable . seals. Historically, testing has shown little leakage from either _of the air systems;or from the pressure tests conducted between the door seals or on the airlock as a whole.

-Additionally, the preoperational test for measuring drywell-bypass leakage was performed on February 10 and 11,'1985.

The test results have shown.that at a test pressure of 3 psid the measured leakage. was 647 SCFM compared to the limit of 4011 SCFM-(m/ 16%). At a design pressure of 25 psig the measured leakage was 2425 SCFM compared to the limit of 9660

-SCFM ( es 25 %) . Since then, a drywell bypass leak test was performed on October 25, 1985 just prior to initial criticality utilizing the surveillance test procedure.

During the test the A/[5'value was computed at 0.014 square feet, well within the required value of 0.1 square feet

(corresponding to only 14% of the value of the allowed leak i- rate).

6.- The change 'to the surveillance frequency for the above mentioned four Technical Specifications are a one-time extension only and the maximum extension requested is forty-one days.

i-i :For 'these reasons, the proposed amendment to the River Bend

' Station License does not constitute a significant hazards

! consideration in that it would not:

I l

u

1. Involve. a significant . increase in the probability or the consequences of an accident previously evaluated results from:

this change because:

a. The valves.were last tested satisfactorily and due_to the short period of the extension, no significant increase in the probability of equipment 1 failure is postulated.

b. The LLRT- testing provides verification of-valve seating -

' integrity and does not provide assurance of_the actuation of the valve 'when called oon to perform its_ isolation function. The increase in surveillance frequency does not affect the probabilities of the' valve actuating when called upon to perform its required isolation function.

Isolation function testing are satisfactory and current.

I

c. The change increases allowable surveillance interval less than 6% .beyond the current conservative surveillance requirements -and has no affect on the assumptions of valve leakages assumed in the present accident analysis.

2. This change would not create _the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change _ introduces no new systems, modes of operation, failure modes or other changes to any equipment. The proposed change does not change the

-system functional analysis and therefore, new accident scenarios are not credible based on scheduling of testing alone.

3. This change would not involve a significant reduction in the margin of safety because, based on the enclosed technical justification which indicates the- number of valves and

' penetrations involved, their current leakage rates and estimated leakage rates at proposed 09-15-87 refueling date, the following can be stated:

a. Those valves for which extensions are being requested have for the most part, based on initial and subsequent LLRT results, exhibited a high degree of leak tight reliability.

b. Overall ILRT shows a very tight containment,

c. The drywell airlock and the personnel door in the drywell equipment hatch have demonstrated a high degree of leak tight reliability and are infrequently used.

d. The requested extension does not significantly increase the allowable frequency interval provided in the Technical Specifications (The maximum increase is approximately 6%).

e. ,

s -

-e.- There :will~. be no' identified, increase in postulated-individual offsite or cumulative occupational radiation exposure- as .a. result _ of the requested amendment which merely' requests to delay. testing.

f. The requested . amendment concerns schedule relief for surveillance- testing of a limited number of containment-isolation, valves and.drywell access doors will not result

-in a- significant . change. in the amounts or types of-effluents that may be released:off-site.

REVISED: TECHNICAL 1 SPECIFICATION I The: requested revision is provided-in the Enclosure.

SCHEDULE FOR ATTAINING COMPLIANCE g i

LAsfindicated above, River Bend Station is currently in compliance with. the applicable Technical- Specification. .This Technical Specification: revision is required prior to August, 1987 to avoid

- a _ unit -outage to conduct the required- surveillance 1 test as discussed above.

NOTIFICATION-OF STATE PERSONNEL A copy of the' amendment application and this' submittal 'has been provided to the state of Louisiana, Department of Environmental Quality - Nuclear Energy Division.

ENVIRONMENTAL IMPACT APPRAISAL Revision of this Techical Specification does not result in an

- environmental' impact beyond that previously analyzed. Therefore, the approval of this amendment does not result in a significant environmental- impact nor .does it change any previous.

' environmental impact.statemnets for River Bend Staiton.

i l'

l l

[ :'

l

e O

ENCLOSURE i

s f

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. If any periodic Type A test fails to meet 0.75 La, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 La, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 La, at which time the above test schedule may be resumed. .

c. The accuracy of each Type A test shall be verified by a supplemental test which: ,

1. Confirms the accuracy of the test by verifying that the differ-ence between the supplemental test data and the Type A test .

data is within 0.25 La. The formula to be used is: [Lo + Lam  !

- 0.25 La] $ Lc $ [Lo + Las + 0.25 La] where Lc = supplemental

test results; Lo = superimposed leakage; Lam = measured Type A leakage.

2. Has duration sufficient to establish accurately the change in

< 1eakage rate between the Type A test and the supplemental test.

3. Requires the quantity of gas, injected into the primary containment or bled from the primary containment during the supplemental test, to be between 0.75 La and 1.25 La.

Md. Type B and C tests shall be conducted with gas at Pa, 7.6 psig*, at intervals no greater than 24 monthi*except for tests involving:

1. Air locks,

2. Main steam positive leakage control system (MS-PLCS) valves and PVLCS valves,

3. Penetrations using continuous leakage monitoring systems, 4

4. Primary containment isolation valves in hydrostatically tested lines per Table 3.6.4-1 which penetrate the primary containment, and

5. Purge supply and exhaust isolation valves with resilient material seals.

e. Air locks shall be tested and demonstrated OPERABLE per Surveillance

^

Requirement 4.6.1.4.

' f. Total sealing air leakage into the primary containment, at a test l

pressure of 11.5 psid for MS-PLCS valves and 33 psid for penetration leakage control system sealed valves, shall be determined by test at least once per 18 months. This leakage may be excluded when determin-ing the combined leakage rate, 0.6 La.

• Unless a hydrostatic test is required per Table 3.6.4-1.

• • This test may be performed during the refueling outage following the first cycle only RIVER BEND - UNIT 1 3/4 6-5 but nn later than 9-15-87.

4

- - - , . - -- , w .,---n,.

. , - - . - - - , , - , ,.-,,..----n , , _ , , , - - , - - , , . _ , - - - - . - - , . . - - . , - - - . n,,-.. . . - , -- ,,-- . - -

f REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:

a. Monitoring the drywell atmospheric particulate radioactivity 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 sump flow rates 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. Monitoring the drywell air coolers condensate flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

d. Monitoring the reactor vessel head flange leak detection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 including paragraph IWV-3427(B) of the ASME Code and verifying the leakage of each valve to be within the specified limit:

a. At least once per 18 months *, and

b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.

4.4.3.2.3 The high/ low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints per Table 3.4.3.2-2 by performance of a:

a. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

b. CHANNEL CALIBRATION at least once per 18 months.

i

• This test may be performed during the refueling outage following the first cycle only but no later than 9-15-87.

I RIVER BEND - UNIT 1 3/4 4-12

e

~

' CONTAIPMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

• g. Type B tests for electrical penetrations employing a continuous leakage monitoring system shall be conducted at Pa, 7.6 psig, at intervals no greater than once per 3 years, rh. Leakage from isolation valves that are sealed with the PVLCS shall be tested once per 24 months with the valves pressurized to at least Pa, 7.6 psig. This leakage may be excluded when determining the combined leakage rate, 0.6 La.

i. Primary containment isolation valves in hydrostatically tested lines per Table 3.6.4-1 which penetrate the primary containment shall be leak tested at least once per 18 months.

j. Purge supply and exhaust isolation valves with resilient material seals shall be tested and demonstrated OPERABLE per Surveillance Requirement 4.6.1.9.3.

k. The provisions of Specification 4.0.2 are not applicable to Specifications 4.6.1.3.a. 4.6.1.3.d 4.6.1.3.g, and 4.6.1.3.h.

l l

• This test may be performed during the refueling outage following the first cycle only but no later than 9-15-87.

RIVER BEND - UNIT 1 3/4 6-6

fJ l CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.2.3 The drywell air lock shall be. demonstrated OPERABLE:

-a. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing,.except when the air lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, by verifying seal leakage rate less than or equal .to 4.05 scf per hour when the gap between the door seals is pressurized to 3.0 psid.

b. By conducting an overall air lock leakage test at 3.0 psid and veri-fying.that the overall air lock leakage rate is within its limit:

1. Each cold shutdown if not performed within the previous 6 months # .

2. Prior to establishing DRYWELL INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.

c. By verifying, prior to drywell entry if not performed within the past 18 months, that only one door in the air lock can be opened at a time.

~

d. By verifying the door inflatable seal system GPERABLE by:

1. At least once per 7 days verifying seal. air flask pressure to be greater than or equal to 75 psig.

N

2. At least once per 18 months conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 0.67 psig from 75 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I

1. At least once per 18 months, the air lock will be pressurized to 19.2 psid prior to conducting the overall air lock test.

• This test may be nerformed during the refueling outage following the first cycle only but no later than 9-15-87.

I RIVER BEND - UNIT 1 3/4 6-21 ,

l l

4 CONTAINMENT SYSTEMS

}

SURVEILLANCE REQUIREMENTS (Continued)

e. By verifying the personnel door inflatable seal system OPERABLE by:

1. At least once per 7 days verifying seal air flask pressure to be greater than or equal to 75 psig.

2. At least once per 18 months

• conducting a seal pneumatic system leak test and verifying that system pressure does not decay more than 0.67 psig from 75 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l t

• This test may be performed during the refueling outage following the first cycle only but no later than 9-15-87.

RIVER BEND - UNIT 1 3/4 6-18

_ - _ _