ML19345E499

Comments on Util Request for Exemption from ECCS Requirements.Responds to NRC 760405 Notice of Exemption Request.Recommends Exemption Be Granted Subj to Stated Conditions

ML19345E499
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/18/1976
From:	Rusche B Office of Nuclear Reactor Regulation
To:	NRC COMMISSION (OCM)
Shared Package
ML19345E497	List: ML19345E495 ML19345E499
References
NUDOCS 8102040686
Download: ML19345E499 (26)

Text

~O

\\./

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

)

IF

)

CONSUMERS POWER COMPANY

)

Docket No. 50-155

?-

)

'I (Big Rock Point Nuclear Power Plant)

COMMENTS BY THE DIRECTOR, NUCLEAR REACTOR RECULATION RELATING TO THE REQUEST FOR EXEMPTION OF THE BIG ROCK POINT fiUCLEAR POWER PLANT FROM THE REQUIREMENTS OF 10 CFR SECTION 50.46 On December 31, 1975, the Commission granted two limited exemptions from the Commission's ECCS Acceptance Criteria in 10 CFR Part 50, 50.46 and Appendix !

to Comsamers Power Company (CPCo) for its Big Rock Point Nuclear Power Plant, subject to certain specified conditions. The first exemption permitted reactor operation until March 1,1976, pending completion of the installation of the s

Reactor Depressurization System.. the second exemption allowed Consumers Power Company (CPCo) until March i,1076 to provide more information supporting its request for an exemption from the ECCS single-failure criterion of 10 CFR Part 50.46 and Appendix K, Paragraph I.D.1, as applied to a loss-of-coola.t accident j

(LOCA) caused by a break in a core spray line and a concurrent eingle failure in the remaining core spray system. In addition, the Commission requested additiona~

staff comments concerning a possible further exemption from the Commission's ECCS single-failure criterion.

r 7L The staff comments submitted on January 7,1976, by the Director of Nuclear Reactor Regulation, outlined certain additional analyses that shculd be performed

~

by CPCo and possible system modific6tions to enhance operating reliability.

l Noso10GHe l

t 2_

i

~

On February 27, 1976, in response to opportunity provided' by Commis-

'N

't sion Memorandum and Order of December 31, 1975, CPCo submitted further sf -information to support its exenption request in a report entitled " Report on Evaluation of Aiequacy of Emergency Core Cooling System" and reyaested 3

further ce nsideration of its life-time exemption request. The Commission publisbad notice of this additional exemption request in the Federal Register and provided an opportunity for public comment (41 F.R.10969, March 5, 1976).

Af ter obtaining additional information from CPCo, on March 26, 1976, the Director of Nuclear Reactor Regulation submitted 4

certain additional staff comments on the pending exemption request and requested an extension of time until April 19, 1976, in which to complete the review of all ECCS related issues for Big Rock Point.(1)

The Commission granted this request on April 51976.

CPCo's original request for exemption from the single failure require-ments of 10 CFR Part 50, Appendix K, Paragraph I.D.1, identified the case of a core spray line break and the failure of a valve in the redundant core spray system as the specific instance of nonconformance with single failure requirements. Further analysis carried out by CPCo i

.I (1)

"Additionr1 Comments by the Director, Nuclear Reactor Regulation I7 and Request for Extension of Time for Staff to Respond to the Request for Exemption of the Big Rock Point Nnelear Pouer Plant From the Requirements of 10 CFR Section 50.46" m

pursuant to the Commission's December 31, 1975 Memorandum *and Order I identified additional equipment which would not conform to the single Ly failure criterion in the case of a core spray line break.

During the course of the staff's review of the Big Rock facility, the staff identified additional aspects of ECCS performance which required further evaluation. These items were identified in the staff's March 26, 1976 Additional Comments.

Further, in the course of providing more information CPCo has very recently (on April 15,1976) identified areas which have required additional staff evaluation of the ECCS performance capability for the Big Rock facility.

The various ECCS related issues are as follows:

1.

CPCo's Exemption Recuest Relaq1ng to Single Failures of Core Spray Valves As noted in the Staff's March 26, 1976 Additional Comments, CPCo's estimates of the failure probabilities for the combination of events resulting in a core spray line break and a failure of the other core j

-8 spray system, were approximately 2 x 10 per year for a break in the nozzle spray line and 4 x 10 per year when the break is in the ring

-7 spray line.

Although the staff has based its recommendation upon technical judgements J*b relating to the performance capability of the entire ECCS and existing design margir.s at Big Rock, it has also performed an independent assess-ment of the reliability of certain systecs installed at Big Rock to provide core cooling in the event of a LOCA resulting from a break in

4 -

)

one of the core spray lines. As part of this assessmadt, the staff

~f estimated the failure probability of the valves in the core spray

-L{[

lines, combined with estimates of spray line break probability and failure probability of other portions of plant systems that can provide core cooling. The use of failure estimates was one aspect of the staff assessment used to provide a better understanding of system relidbility and plant safety.

Staff failure probability estimates are generally higher than those used by CPCo. These higher values give more appropriate consideration to service conditions, to test anc maintenance programs, and to operating experient at Big Rock. For pipe bresia in the nozzle spray line the staff has used the value of 0 x 10-5 per year. For a break in the ring core spray line, the failure probability used is 10-3 per year. This higher value reflects the existence of four uninspectaible welds in this system.

In the event of a break in one core spray line, water can be supplied to the core through the other core spray line. Since each of the redundant core spray lines (the ring spray and nozzle spray) has two actuating valves in series, the single

g failure of one of these valves would render the system inoperative.

..7

}'

The failure probability in the core ring spray system is approxi-mately 0.02 per demand with the CPCo proposed annual testing of the valves.

valves in the ring spray system are powered from the de station battery bus. The valves in the nozzle spray system are not

3 connected to the station battery. They are Nwered frEn the emergency

~

'A

'I diesel generator or by off-site power. The failure probability of M

these valvec is also about 0.02 per deaand since the failure proba-bility is dominated by valve failure when offsite power is considered.

Using these values. the probability fo'. a nozzle spray line break and failure of a valve in the ring spray system is about 2 x 10 per year (with annual testing). Tae probability for a ring spray line break and a failure of a valve in the nozzle spray systeu is about

~

7 x 10 per year if off-site power is assumed unavailable and about

~

2 x 10 per year with off-site power availability (all with annual te sting). More frequent testing or other enhanced surveillance, woJ.d increase reliability.

Independent of the core spray systems, adequate cooling capacity is available for core spray breaks by use of the feedwater system.

The supply of water to the feedwater system, when used for this purpose, is the stored condensate. This supply is supplemented by a six inch line from the fire protection system. The feedwater system, in conjunc'.: ion with adequate water sources, can provide cooling for an extendee period until the contain:2ent core spray recirculation system

e is placed in opocation to cool the core. The feedwater system 57 j

is powered by the offsite power systems. As discussed in item 6, l

offsite power has a high reliability and therefore the failure probability for the feedwater system is dominated by necessary operator actions and the operabilicy of the valves which must function to provide a

. ~'

water source to the feedwater system.

Since neither the staff "s

'I nor CPCo has explicity evaluated the performance of this system, 1!

if it is not clear what failure probabilities can be assigned to this course of events. However, since this system is used during

~

normal' operation, its f ailure probability for the short period of time it would be required ie believed to,be small.

However, it is i tear that this alternative source of cooling water supply in he event of a spray line break enhances overall ICCS reliability so that failure to provide adequate cooling to the core as a result of a break in one core spray line and a single failure in the other spray system is acceptably low.

Accordingly, the staff believes that the requested lifetime exemp-tion from the single-failure requirements of the Commission's ECCS Acceptance Criteria in 10 CFR Part 50, 50.46 and Appendix K,

- Paragraph I.D.1 for the case of a pipe break in one spray line system, should be granted. For other reasons which are discussed in the items below, the exemptions should be conditioned.

For example, the staff evaluation discussed above 'is based on adequate spray performance by the spray nozzle. However, as discussed in item 5, recent information has raised some question as to the at adequacy of nozzle spray performance. For this reason, the i.

~

'f exemption requested by CPCo should be conditioned as set

~

-t ji[

forth in item 5.

Other conditions which should be imposed as part of the exemptions, are set forth in each appropriate -

Jitem and the conclusion. These conditions set forth certain requirements which must be met prior to the return of Big Rock to power operation and other requirements which must be met prior to its return to operation af ter tl.c refueling shutdown presently scheduled for the Spring of 1977.

i 2.

Use of 900 lb Class Valves in the Ring Spray Line As discussed in the staff's March 26, 1976 Additional Comments, two valves in the ring spray system are 900 lb class valves.

Since the pressure which these valves can withstand is related to service temperatures, the staff investigated the adequacy of these valves to withstand the expected normal operating pressure of the reactor coolant system of 1335 psig at 600 F.

Our review has verified that these valves, the core spray valve (MOV-7051) and check valve (CV-7) have en allowable pressure-temperature of 1640 psig at I

600 F.

Since 1640 psig is 300 psi above the reactor coolant system normal operating pressure of 1335 psig, we have concluded

.c that an adequate safety margin is available relative to CPCo's ECCS exemption analysis. Existing safety margins provide adequate assurance of the inuegrity of these valves for the period of time required for CPCo to obtain and for the staff to review the

4 f

. 1 1

modified overpressure protection analysis of the reactor

's pressure boundary to consider the lower operating pressure rating if of the valves. CPCo has agreed to furnish this information within six months, No Commission action relating to CPCo's ECCS exemption request is required for resolution of this matter.

3.

ECCS Water Supply System The staff has reviewed the potential for both active and passive failure of the fire protection system, which for Big Rock Point is a vital portion of the ECCS, as well as other safety systems, used for dissipation of heat from the core.

The active portions of the fire protection system are either redundant or are adequately protected against single failures.

Passive piping failures have been considered for both the short term cooling requirements associated with injecting water into the core and for long term cooling requirements for removing core decay heat.

For short term cooling requirements, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA, the probability of both-a LOCA and an unrelated passive failure of a

r the piping in the fire protection system during this limited period

$7 of time is sufficiently low to be acceptable. An important factor in reaching this conclusion is the short period of time during which the fire water system is required to operate.

9-The, situation-for long term core cooling is quit

.itterent.

As the

~&

plant is currently designed, the fire protection system is needed for an indefinite period during the long term cooling mode to provide cooling water for the containment heat renoval heat exchanger.

Since this cooling water is needed almost continuously, the resultant failure probability can be related directly to tha failure -

probability of the. fire system piping. That failure probability of that piping alone has an upper bound value of 2.6 x 10-3 per year

]

9 3

e

[- t 4

i 1

! and a median value of 9 x 10-5 per year. Because of the limited

'A

-t inspectability due to the inaccessability of the underground

-L k[

piping of the fire protect.on system, the staff believes that its evaluation should be based on the upper bound value.

Since this system is an essential component of the ECCS system cooling water supply for long term cooling for all LOCA events, not solely those involving a core spray line break, and the system is vital in achieving safe shutdown for many other conditions, the si. if believes that this annual f ailure proba-bility is not acceptable for a period covering the remaining life of the plant. However, for a period of another fuel cycle, some 12 to 16 months, the probability of a LOCA or other event requiring the fire protection system and concurrent failure of the fire protection system is sufficiently low to permit operation with the current design.

WHile assessment of the adequacy of the fire protection system as part of the ECCS is not lLnited to *he case of sirgle failure events involving the requested exemption, the staff believes that fire system reliabil'ity has a vital bearing on the overall safety of the plant. This in turn is of primary importance in assessing the l

b effect of deviations from the Commission's ECCS performance require-ments.

, i Consequently, the staff recommends that granting of the

,[

exemption requested by CPCo be conditioned to require, prior to I

startup following the refueling outage scheduled for Spring 1977, modification of the facility such that long term cooling can be accomplished without relying on the underground portion of the fire protection system.

4.

ECCS Performance tyvi luation Model With respect to other aspects of the ECCS performance evaluations submitted by CPCo for Big Rock Point, the staff review has uncovered no other outstanding issues and no issues relating to the subject matter of CPCo's requested exemptions other than those discussed in these Staff Comments. Consequently, no Commission action is required with respect to other aspects of the ECCS performance evaluati.ons for Big Rock Point.

5.

Effectiveness of Nozzle Core Spray The single overhead spray nozzle in the. nozzle spray system at Big Rock Point is a Spraying System's Company "Distriboj et," Catalog No. 4R80160. It is a one piece cast type nozzle with non-removable internal vanes. The nozzle, 5" in diameter by 81/8" high, is to

[- 8 distributes cooling water to all the fuel bundles in the core should 7!I

' the core become uncovered. The spray characteristics are described by the manufacturer as "Large capacity, f ull flow deluge type sprays.

1 12 -

t Nozzles designed to pass a large liquid volume in a full cone LI pattern." Neither the General Electric Company (GE) nor its overseas; if partners have tested the spray distribution of this type nozzle under expected usage conditions.

However, GE has tested a Type VNC nozzle without a deflector which is somewhat similar in configuration although of much smaller diameter (1 inch). Reported test data show that such a nozzle " exhibits a spray width reduction of approximately 50%

when spraying into atmospheric pressure saturated steam."

Such conditions (atsosphere pressure and saturated steam) are typical of those under which the nozzle spray system is required to operate at Big Rock. Thus, a width reduction of this magnitude at Big Rock Point would result in incomplete core spray coverage. Imcomplete core spray coverage could affect the ability of the nozzle spray system to adequately cool the core should it become uncovered.

~

Without representative tests on an identical nozzle confirming full core spray coverage under expected usage conditions, there is inadequate assurance that the core spray nozzle system would cool the core. GE agrees that a test on the 4R80160 nozzle in a steam environment would be desirable. However, the GE test facility does not have a sufficiently

. r large steam supply to perform such a test. GE is surveying other

.7 domestic and foreign companies in an attempt to find an adequate test facility.

D'ww

. 1 i

}

-e e

The spray nozzle is a vital element for adequate ECCS performance. For

.g a large number of the potential LOCA break sizes and locations, the if core could be uncovered. Cooling is provided in these cases by spray cooling from either the core ring spray system or the core nozzle spray system which were each designed to provide adequate spray cooling for the entire core, thus providing redundant spray capacity. For a break in the core ring spray system, the staff believes the core can be reflooded by the incomin feedwater system and by the water from the nozzle. Consequently, in this case, even though the spray nozzle distribution alone may not be suf ficient to provide adequate core cooling, the combination of flooding from the feedwater system and spray from the nozzle spray system may be sufficient to adequately cool the core. However, a specific evaluation has not yet been performed to verify the adequacy of feedwater systec reflood capability.

For breaks at locations for which reflooding of the core is not possible, the ability to provide adequate core cooling depends on the reliability of the core ring spray system if the nozzle spray system does not perform as designed. Specifically, the reliability of the valves in the core ring spray system become limiting in terms of LOCA protection for such breaks. While the reliability of the core spray line valves 1:

is sufficient in the case where both spray systems each provide adequate spray cooling, as designed, we do not believe the reliability of the core ring spray valves is presently sufficient in the event that only the ring spray system provides adequate cooling.

Augmented surveillance of the core ring spray valves ca'n enhance

'n-i reliability sufficiently for a limited period of operation such as

'd[

the next refueling period.

However, for the longterm operation the staff believes that redundant spray cooling systems should be provided to protect against failu.t in the in-core spray cooling sparger.

For the foregoing reasons, unless the adequacy of the nozzle spray distribution has been demonstrated, the staff recommends that the exemption be granted for the Big Rock ECCS system only on the follow-ing conditions:

1.

Before return to operation, CPCo shall provide the staff with an analysis of ECCS perforrance which properly demonstrates that in the event of break in a core ring spray line, the feedwater system and the flow through the core spray nozzle will reliably provide suf ficient cooling water to the core so that the peak clad temperature does not exceed the Commission's Acceptance Criteria set forth in

~

10 CFR 50.46.

2.

Before return to operation, provision will be made for enhanced reliability of the core ring spray system by augmented surveillance

.c((

of the valves and valve actuating circuits, or by other modifications or procedural changes which provide reasonable assurance that the core ring spray system can, by itself, provide reliable and adequate core cooling in the event of a LOCA at a location where reflooding does not provide adequate core cooling.

~

.3.

Modify the nozzle spray system to provide adequate

't

..g spray distribution, prior to return to operation after the

'if refueling shutdown presently' scheduled for the ' Spring of 1977.

6.

Electric Power Supply The Fire Protection System has redundant pumps available to provide short term core cooling.

One pump is an elecric drivent pump powered from of fsite power or from the onsite emergency diesel generation; the other pump is driven directly by a separate diesel engine. Ilowever, for long term cooling, in addition to the proper operation of the fire protection system, operation of electrically driven recirculation water pumps is required to supply water to the core and remove heat from the containment.

Saveral older plants such as Big Rock Point were

~

con'etructed with redundant offsite power circuits but only one onsite emergency diesel.

Consequently, a single failure of the only emergency diesel generator would leave the station without onsite power for long-term or recirculation rode core cooling.

Such cooling is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LSe LOCA.

CPCo has advised the staff that a suitable portable diesel 4

generator could be obtained and made operable within the J'

f(

required time interval.

m*

t Without reliance on offsite power supplies, the single failure of the ansite diesel generator adversely affects the performance -

of the ECCS system.

This circumstance also deviates from the

't Commission's ECCS performanca requirements in 10 CFR Part 50,

~.

I 50.46 and Appendix K.

CPCo indicates that an exemption for the failure of the diesel gen erator is encompassed within their pending request which applies to all single failure in one, core spray system which do not conform to ECCS perfromance requirements in the event of a break in the other spray system.

Although single failures in the diesel generator, in the-absence of offsite power, can adversely affect systens performance for a number of events, the staff agrees that this single. failure is encompassed in CPCo's pending request.

To aid in reaching the staff recommendation in this area, it investigated the reliability of the existing offsite and onsite power systems.

The offsite system consists of two offsite power circuits originating at the same substation.

These circuits share the same right-of-way from the substation to a point ab out a mile from the plant where the circuits separate and enter the plant on separate rights-of-way.

Generally, two types of events can cause loss of offsite power, trips of the plant due to internal causes and external events such as ice storms.

lL The staff has assessed the offsite poeer unavailability resulting

-3 from plant trip to be 1 x 10 per event. This appears l

6 conservativa in view of the relatively small size of dhis plant (70 WMe) compared with the system capacity.

In addition, operating

~~

.y experience confirms that the unavailability is small since none of

~ b

?

the 11 plant trips from above 25 EWt have resulted in loss of at offsite power.

The plant has experienced one complete loss of offsite power in eight years.

This was due to severe weather conditions.

Therefore, based on this experience, the unavailability of offsite power due to causes external to the plant is equal to 1/8 year or about 0.125 per year (this is consistant with the WASH 1400 Report).

Offsite power is required as a redundant ECCS power source for the LOCA.resulting from a break in the spray ring line only for the time required to open the nozzle spray valves, i.e., probably less than a minute.

Thus, assuming a time fraction of six hours for availability of offsite power is conservative. For this time period, the unavaila-bflity of offsite power reduces to:

-5 0.125 X 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 8.6 I 10

=

year 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> / year Thus, the controlling or larger of the two offsite power unavailabilities 2i={

immediately follwoing a LO:A is that associated with a plant trip, i.e.,

~

1 x 10 rer event and not that associated with external events. This is the value used in the staff's independent reliability analysis described in item 1.

r --

., In reviewing the existing onsite power system, the staff. observed

~

that some of the emergency diesel generator protective trips I

'are set to shut down the unit, even in an emergency, for condicions L

[

which may result in economic loss but would not of themselves cause a loss of power.

Since such trips could significantly reduce the avail-ability of the diesel generator during a LOCA removal or revision of these trips vould significantly enhance the reliability of the onsite power system.

Based on the high availability of offsite power and the conditions which are noted below for the onsite power system, the staff believes that the probability of a LOCA at this facility coincident with a sustained

~ 1oss of offsite power is sufficiently low to be acceptable for long term core and the exemption should be granted subject to the following

^

conditions:

1.

Prior to plant startup from the current refueling outage CPCo modify the emergency procedures (a) to assure that in the un-likely event of a LOCA a second dedicated emergency diesel generator can and will be obtained expeditiously as a backup to the existing plant diesel generator can and will be obtained expeditiously as a backup to the existing plant diesel generator and (b) required that such dedicated emergency diesel generator must be fully operational at Big Rock Point within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA.

i 2.

Before return to power after the refueling scheduled for Spring i

1977, CPCo provide modifications to the emergency diesel generator 1

protective power circuitry to allow bypassing protective trips J,

during the demand for emergency power except forethe retention I!b of the engine overspeed and the generator differential trips, or as an acceptable alternate CPCo may make modifications to provide at least two independent measurements for any other protective trips con 'dered necessary by CPCo. Trip logic for any other prott*.tive trips retained would require specific coincidence logic.

Simi' ar design changes should also be made to improve the reliability of the diesel driven fire pump.

l

.- i l

7.

Other Electrical Failures

^f The staff has continued its review of the electrical single failures

-L((

identified by CPCo in its February 27, 1976 submittal and in the staff comments dated March 26, 1976.

Additionally, we have reviewed the CPCo comments of March 26, 1976, relating to other staff identified electrical single failures and electrical system operational concerns.

CPCo has initiated action to correct all electrical single failures which would have a direct consequence during a LOCA except those discussed in item 1 above. This includes reconnection of the Emergency Diesel Generator starting circuit from the station battery to an independent uninterruptable power source thus making these two ECCS power sources truly independent. CPCo's actions will rectify the identified electrical single failures prior to return to operation.

CPCo has also identified other electrical single failures which do not directly affcet ECCS actuation but which cculd disable indication and annunciation channels required for accident and post-accident monitoring. Although these channels do not result in a failure of automatic controls, the indication and annunciation functions are of importance to the operator in monitoring conditions in the facility in i

the event of an accident. For chis reason, the staf f has informed CPCo that these failures must be corrected prior to return to power from the next refueling outage scheduled for the Spring of 1977.

e

.. No action is r'equired by the Commission on these items -in connection lf with the pending exemptions request.

8.

Surveillance Testability of ECCS The present design does not provide on-line ECCS actuation system testability. The staff has concluded that complete ECCS testability is necessary to substantiate the reliability analysis described in item 1 above. Therefore, the grant of the requested exemption should be conditioned to require:

1.

Prior to startup, modifications or procedural changes, acceptable to the staff, should be made to augment surveillance of ECCS availability including the ECCS actuation system, and the Teciaical Specifications shall be revised to require such surveillance.

2.

Prior to retern to operation after the next refueling outage scheduled for Spring 1977, modifications shall be made to the ECCS system to provide complete on-line testablity of the ECCS including the actuation system.

e b

i 1

- 21 9.

Submerged Components Affecting ECCS

's 1

There has been a series of correspondence between CPCo 1!

I and the staff on the subject of potentially flooded ECCS equipment without any recent indication by CPCo that further problems-exist.(I) k

1. In a letter dated April 1,1975, the staff required the submis-sion of the results of ECC instrument tests and analyses. On May 2, 1975, CPCo submitted Special Report 20.21 in response to that letter.

Item III D of that report addressed the effects of containment flooding due to LOCA on the core spray system motor-operated valves.

No nention was made of other possible safety consequences of such flooding.

On March 17, 1976, the steff specifically requested identification of all electrical equipment that might be submerged as a result of a LOCA. On March 26, 1976, CPCo responded that " Submerged ECCS equipment was identified and corrective actions taken as appropriate during the five-month plant sh itdown in 1975." 7n that letter, CPCo also identified a 125 Vdc motor control center, 125 Vdc distribution bus and a 480 Vac bus, all of which were stated to er have adequate tripping devices for the loads that.ould be submered such that the center and the two buses anuld not be lost due to a fault on submerged circuits. We found this to be acceptable.

l 1.

• n

,]

However, the staff was in' formed only recently in the course of its if review of CPCo's exemption request of the existence of additional control, indication and annunciation features, at least some of which are ECCS related, that can be lost due to containnhnt flooding.

On April' 15, 1976, representatives of CPCo orally informed the staff that the following five additional functions or panels could be lost due to containment flooding.

1.

Station Service Annunciator Panel 2.

Nuclear Steam Supply Annunciator Panel 3.

Fire System Annunciator Panel 4.

Containment Isolation Valve Indication 5.

Core spray valves control and indication The fifth item affects the controls for the valves discussed in item 1 above and must be rectified prior to return to operation.

i 5'

• l

r.

23 -

~~

Similarly, all other' items which would adversely affect the control

'I of the 10CS system must be rectified prior to return to operatior..

if For those; items which affect only indication and annunciation, but provide the only intelligence to the plant operator needed to take appropriate corrective actions during the course of a LOCA, rectification also must be made prior to the return to power.

For this reason, the granting of the requested exemption should be conditioned to require the following prior to return to operation:

1.

The core spray valves control and indication circuitry must be protected against the consequences of. flooding cs a result of a LOCA, during the entire course of events for which they may be required to function.

2.

All other ECCS control circuits which may be required to function must.be protected against the consequences of flooding as a result of a LOCA during the entire course cf events.

3.

All ECCS inoication and annunciation circuits the failure of which could adversely affect the ability of the plant operator to take corrective action during the course of a LOCA, as determined i-by the staff, shall be protected against the consequences of flooding as a result of a LOCA, during the entire course of

~

events.

t r

w

-,.v

4 4.

Resolution of the foregoing items shall be proposed.by CPCo

.and approved by the staff prior to power operation of the Big Rock facility.

M Conclusion in connection with its assessment of the exemption requested bf CPCo from the Commission ECCS performance requirements set forth in 10 CFR Part 50, 550.46 and Appendix K, the ataff has carefully assessed the capability of the entire ECCS installed at Big Rock Point to provide" adequate cooling water to protect the core in the event of a LOCA.

As discussed in items 1 through 9 above, this evaluation has used certain failure probability estimates as a valuable tool to aid in understanding certain aspects of system performance. These estimates, when coupled with our overall assessment of additional protection afforded by multiple backup systems, and the margins provided by conservative analytical techniques, assumptions and performance requirements provide reasonable assurance that the ECCS will perform adequately as discussed above.

In response to the Commissioners Memorandum and Order of October 31, 1975, CPCo has estimated in its October 27, 1976 submittal the various costs c

involved in modifying the facility for full compliance with the single

- r failure requirements of 10 CFR 50.46. These estimates appear reasonable.

Since the s taff concludes that the present system, under the conditions

' set forth herein, provides a high degree of safety, we bell' eve that there is good cause to warrant the exemption requested.

e o

25 -

~

Accordingly, the staff concludes that in accordance with 10 CFR 50.46 the Commission should grant an exemption from the single ' failure require-ments of 10 CFR Part 50 550.46 and Appendix K Paragraph I.D.1 as applied i

L!

to a LOCA caused by a break in a core spray line and a concurrent single if-failure in the remaining core spray system i cluding the failure of the onsite diesel generator subject to the following conditions:

1.

Prior to return to operation, CPCo shall:

a) Provide an analysis of the ECCS performance which properly demonstrates that in the event of a break in core ring spray line, the feedwater system and the flow through the core spray nozzle will reliably provide sufficient core cooling water unless adequate spray distribution of the nozzle has been demonstrated..

b) Enhance the reliability of the core ring spray system by augmented surveillance to provide reasonable assurance that the core ring spray system can, by itself, provide reliable and adequate core cooling for a LOCt not allowing reflooding unless adequr.te spray distribution of the nozzle has been demonstrated.

c) Modify the emergency procedures to assure a second emergency diesel will be obtained and operational within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA.

d) Augment the surveillance of ECCS to enhance its reliability in a method acceptable to the staff.

e) Protect the controls, indication and annunciation circuitry associated with the ECCS, including the core spray valves, as approved by the staff, against the consequences of flooding following a LOCA which affect the ability of the ECCS or plant

(

operator to take corrective action during the course of a LOCA.

~ :

e' t

~

ll i

26 -

2.

Prior to return to operation following the refueling outage cu:rantly g

t.

{

scheduled for Spring 1977, CPCo shall:

sr i

a) Modify the fire protection system such that long tenn cooling can be accomplished without relying on portions of its underground piping.

b) Provide test data showing the adequacy of the nozzle spray distribution during expected usage conditions or modify the nozzle spray system to provide adequate spray distribution.

c) -Modify the emergency diesel generator and diesel driven fire pump to bypass protective trips during accident conditions execpt for retention of engine overspeed and generator differential trips unless additional trips are approved by the staff.

d) Provide complete on-line testability of the ECCS including'the actuation system.

t' ? s'/

4

. G (,c. -

s Ben C. Rusche, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this';;,

.. a e

e 4

i Il 1

e l

j l

- -.,