ML20027A435

Responds to 780522 NRC Ltr Requesting Addl Info Relating to Engineered Safety Features Available to Mitigate Consequences of Postulated Fuel Handling Accident Inside Containment at Subj Facility.Related Correspondence

ML20027A435
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	11/29/1978
From:	Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Ziemann D Office of Nuclear Reactor Regulation
References
TASK-15-20, TASK-RR NUDOCS 7812040220
Download: ML20027A435 (12)

Text

{{#Wiki_filter:, l 1 I I l Consurners P0'#er Company l@:.. Wm u t..-, o..., m w., ou, 4.e-- m. wa. wc ~ pn men. 4. c >oe so mosso November 29, 1978 s \\ Director, Nuclear Reactor Regulation i Att: Mr Dennis L Ziemann, Chief Operating Reactors Branch No 2 US Nuclear Regulatory Commission Washington, DC 20555 \\ i DOCKET 50-155'- LICENSE DPR BIG ROCK POINT PL\\NT ' ADDITIONAL INFORMATION RELATIVE TO FUEL }{ANDLING ACCIDENT IN CONTAINMENT Your letter dated May 22, 1978 requested Consumers Power Company to provide additional information relating to the engineered safety features which are available to mitigate the consequences of a postulated fuel handling accident inside containment (FRAIC) at the Big Rock Point Plant. The purpose of this letter is to provide the requested information. Your letter requested the following four items: 1. Provide system descriptions (including P& ids and control system logic and schematic diagrams) and analysis to demonstrate the extent to which existing systems required to function during the FHAIC comply with the criteria established in the Hazards Summary Report for engineered safety features. 2. Provide a description of the extent to which these systems will comply with the current NRC criteria for engineered safety feature systems which l are listed in the NRC Standard Review Plan, NUREG-75/087. 3. Explain why it is not necessary to have these systems meet the current NRC criteria for ESF systems. 4 Information regarding proposed changes to the existing plant should be provided. This information should be as described in Regulatory Guide 1.70. 0\\ (0 42 de Q) l

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l 2 Response 1 System descriptions are provided via the following documents: 1 0740G30114 Sheet 2 Schematic Diagram - Air, Screen, (Attachment I) Fire and Post-Incident Systems 0740G40125 Rev P Reactor Building Ventilating, Heating (Attachment II) and Cooling System P&I Diagrams Sketch No 1 Logic Diagram for Air Supply Valves, (Attachment III) Scheme 8501 Sketch No 2 Logic Diagram for Exhaust Vent Valves, (Attachment IV) Scheme 8512 Analysis No I Written System Logic Description (Attachment V) An analysis of the existing containment isolation svstem shows that rccent modifications to the system (ie, automatic isolation oa high radiation, vacuum relief through the 24-inch supply and exhaust lines) have enhanced the system's ability to comply with the criteria established in the Hazards Summary Report. The Hazards Summary Report describes two 24-inch ventilation openings; one for supply, the other for exhaust, which are closed automatically within six seconds af ter any scram signal or loss of power. As described in the abcve documents, these two closing features are ",1! present with the additional feature of high radiation closing also < 3 kle. The Hazards Summary Report s also describes a vacuum relief line via'h 1 intended to prevent excessive external pressure from causing dar :1 containment's integrity. The vacuum relief modification has als. ea-

• uis system's ability to comply with the vacuum relief criteria est;blished t; the FHSR by providing two independent vacuum relief lines.

It is, therefore, concluded that the existing containment isolation system that is required to function during the FHAIC is in compliance with the criteria established in the Hazards Summary Report for engineered safety features. Response 2 The Containment Isolation System (CIS) is the only engineered safety feature system that is required to function to mitigate the radiological consequences l of an FHAIC. An analysis of the NRC Standard Review Plan, NUREG-75/087, has shown that the following current NRC criteria need to be considered regarding the Containment Isolation System: 10 CFR Part 50 Appendix A, Seneral Design Criteria, Criterion 23 - Protection System Failure Modes l l l

i i i 3 a i 1 10 CFR Part 50 Appendix A, General Design Criteria, Criterion 56 - Primary j Containment Isolation Regulatory Guide 1.53, Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems ] IEEE Standard 379-1977, Standard Application of the Single Failure Criteria to Nuclear Power Generating Station Class IE Systems 1 Criterion 56 requires that lines connecting directly to the containment atmosphere through the containment be provided with two containment isolation valves, one inside containment, the other outside containment. The criterion i further requires that valves outside containment be located as close to the containment as practical and that the automatic isolation vlaves shall be 4 " fail-safe" upon loss of actuating power. The Big Rock Point Plant has two such lines penetrating the containment. They are the air supply and exhaust lines. Each line has two isolation valves in series (see 0740G40125, Attachment II) that are both located outside of containment, as close to the containment as practical. The redundant isolation valves are automatically controlled via electric signals controlling air to the pneumatic valve operators. The valve operators are " spring to close" which allow automatic closure on loss of air and/or electric power. The existing system meets Criterion 56 with the exception of valve location. The remaining criteria concern the single failure and the failure mode. It is required that the components in the CIS have a failure mode that is " fail-sa fe. " It is also required that t: a redundant valves have control circuitry that meets the single failure criterion. As it applies to this case, the single failure criterion means that containment isolation shall not be prevented due to the failure of any single component in the scheme. (See schemes on Attachments III and IV.) Analysis II (Attachment VI) has been performed to provide a single failure analysis of the containment isolation system schemes. The analysis has been performed in accordance with IEEE I Standard 379-1977. Response 3 As noted in Response 2, the containment isolation system design complies with 1 Criterion 56 except for the location of the isolation valves. The existing j design at Big Rock Point places both isolation valves outside the containment j whereas the current criterion requires that one of these valves be located j inside containment. Aie existing isolation valves are located as close to the f containment as pra tical. It is not credible to assume that the conditions inside containmen during fuel handling activities could be severe enough to compromise the phys. al integrity of the ventilation piping between the containment and the ft mt isolation valve. Consumers Power Company, therefore, considers that the existing valve location design provides adequate j assurance that Part 100 limits will not be exceeded as a result of a fuel handling accident inside containment.

4 Analysis II (Attachment VI) identifies the fact that the existing ventilation isolation system control circuitry does not meet the single failure criteria j as established by IEEE Standard 379-1977. The postulated failures that do not meet single failure criteria are " stuck" contacts, mechanical failure of the solenoid valves and " hot shorts." l The probability of each of these failures is low. The probability (per Rasmussen) of a manual switch contact failing to transpose is 1x10-3/ demand. There is also a very low probability that the contacts on a relay will stick or " weld" closed. The probability (per Rasmussen) of the solenoid valve failing to operate is 1x10-3/ demand. This probability includes both electrical and mechanical failures. Since this application is concerned only with mechanical failures, and since they are less likely to occur than electrical failures, the probability of mechanical failure is considered extremely low. The probability (per Rasmussen) of " hot shortr" or shorts to power is 1x10-3/ hour, t Three events are necessary in connection with fuel handling in order to exceed Part 100 limits. First, a postu12ted failure of the crane is required while moving a fuel transfer cask over the reactor vessel with the head removed. Second, a postulated failure of the safety brake is also required, before the cask could possibly be dropped into the reactor. And third, the postulated failure of one of the ventilation isolation control valves, for at least fourteen minutes, is also required. The need for these three events to occur simultaneously further reduces the probability of exceeding Part 100 limits. As a result of the above analysis, Consumers Power Company considers that it is not necessary to have the containment isolation system meet current NRC criteria for ESF systems in order to mitigate the radiological consequences of a fuel handling accident inside containment. Response 4 No changes to the existing plant are planned. David A Bixel (Signed) David A Bixel Nuclear Licensing Administrator CC: JGKeppler, USNRC b

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ATTACHMENT V ANALYSIS I System Logic Description I. Air Supply Valves Logic Description The Logic Diagram for the air supply valves, Scheme 8501, is shown on Sketch No 1, Attachment III. The two series air supply valves (see 0740G40125, Attachment II) to the containment are labeled CV-4096 and CV-4097. One valve is a butterfly, the other is a check. These are pneumatically operated valves that require air to open but are spring to close. This feature makes the isolation valves " fail-safe" on loss of instrument air. The CVs can be opened with air from either of two parallel solenoid valves labeled SV-9151 and SV-9152. Air is supplied to these solenoid valves via an instrument air line or a connection to a bank of nitrogen bottles. Power to the solenoid valves is supplied from 125 V d-c BKR #72-ID26 via parallel contacts SVXI and SVX2. The closure of either of these contacts will energize the SVs, permitting the CVs to open. SVXI relay is energized via the proper alignment of contacts from the SS (closed under normal conditions, open during scram), HS 9001 (closed with switch in "open" or " normal af ter open" position), and SVX3 contact (closed on normal radiation levels, open on high). An open contact on any one of these devices provides an actuation signal that closes the CVs. The SVX2 relay is energized on a vacuum relief signal which causes the SVX2 contact to close which in turn energizes the solenoid valves and opens the CVs. The relay is energized on increasing vacuum at -1.00 psig and is de-energized on decreasing vacuum at -0.70 psig. The auxiliary relays PISX1/173 and PISX2/173, as shown on 0740G30114, Attachment I, provide the necessary contacts for the desired deadband and annunciation. II. Exhaust Vent Valves Logic Description The Logic Diagram for the exhaust vent valves, Scheme 8512, is shown on Sketch No 2, Attachment IV. The logic description is identical to the one above for air supply valves with the exception of the appropriate equipment numbers and the auxiliary relays on the vacuum relief scheme. These relays are not required inasmuch as no additicnal annunciation is required for this scheme. oc1178-037Sb-43

ATTACEMENT VI ANALYSIS II I. Single Failure Analysis for Air Supply Valves, Scheme 8501 A basic requirement in the design of a Class IE system is that no single failure of a component will interfere with the proper operation of an independent redundant counterpart or system. The redundant couaterparts in this case are the series isolation valves CV-4096 and CV-4097. The point of this analysis is to determine if these valves will have the proper failure made in the event of a single failure cf a component. At this point it is appropriate to emphasize tbst the failure mode of the CVs is in the closed position. The closed position is required for containment isolation. The valve operator is air to open and spring to close. 1 Independence and redundancy are the principal means of meeting the single failure criteria. The redundancy and independence of the CVs allows them to meet the single failure criteria as one set of components. Air is supplied to the CVs via parallel solenoid valves, SV-9151 and SV-9152. The solenoids are energized to supply air to the CVr. to maintain them in an open position. A common type of solenoid failure would involve having a coil wire "open" causing the solenoid to de-energize, thereby allowing the CV to " fail-safe." A less common but credible type of failure would involve a mechanical failure of the core assembly that prevents the solenoid frc:a cycling to vent the air to the control valve when the solenoid is de-energized. This type of failure would not allow the CV to close. Inasmuch as either SV can supply air l to both CVs and either SV is postulated to failure, the single failure criteria is not met at this point. The effect of interfacing systems on the CIS must also be analyzed ior single failure. The instrument air supply system is one such interfacing system. Inasmuch as air is not required to perform the containment isolatica function, the instrument air supply does not have to meet single failure. Relay contact SVX2 will energize the SVs on a high vacuum pressure signal. Energizing the SVs opens the CVs. Opening the CVs, during this condition only, allows air into the containment. As the vacuum condition is eliminated the CVs will again close to mitigate the release of contaminants to the atmosphere. The relay is normally de-energized and cannot be expected to fail electrically. A second type of failure mode to be considered is when the relay is energized during a vacuum relief signal and then the signal is removed. It can be postulated that i 1

l l ATTACHMENT VI ? t 1 the relay contact could fail closed, thereby allowing the CVs to remain open after the vacuum is gone. If the contact is postulated to stick i closed the single failure criteria is not met. j Relay SVXI is normally energized to close the SVX1 contact which energizes the SVs and allows the CVs to open. The relay is energized j through the closed contacts of the SS (Reactor Protection System), HS-j 9001, and relay SVX5. Opening any of these contacts causes relay SVX1 to de-energize, thereby de-energizing the SVs and closing the CVs. The two postulated failure modes of the relay SVX1 are the same as relay SVX2 above. If the coil fails electrically the contact will open and circuit will " fail-safe." If the contact is postulated to stick closed, the power to the SVs would be maintained and the single failure criteria I would not be met for this component. 4

j As with the above contacts, it can be postulated that the contacts for the HS-9001 or SVX5 could fail in a closed position, thereby not permitting the appropriate isolation signal to de-energize SVX1 which de energizes the SVs and allows the CVs to close. However, since these contacts represent redundant methods of providing the necessary 1

actuation signal, a failure. of either would not prevent closure of the j isolation valves. 1 The SS signal is provided through two series contacts, thereby providing ) redundancy and meeting the single fail-ure criteria. The high radiation signal to the SVX5 relay is provided through tuo series contacts from independent area monitors, thereby providing redundancy and meeting the single failure criteria. 3 i l The power supply for the isolation scheme is a 125 V d-c breaker. A postulated loss of power would de-energize the SVs and allows the CVs to " fail-safe." Inasmuch as power is not required to provide containment isolation, the single failure criteria for this component is met. In su= mary, the single failure criteria is not met due to the configuration of the following components: SV-9151, SV-9152, SVX2 contact, and SVX1 contact. A failure of either SV to cycle properly l could prevent the air to the CV from being vented and the CVs would remain open. A failure of the contacts to open in the remainder of the j above components could allow the SVs to remain energized and the CVs would remain open. 4 As a result of the lack of reduadancy in the control wiring to the j redundant CVs, it is also necessary to consider a postulated " hot short" in the wiring that could allow the SVs to remain or become energized independenc of the ventilation isolation actuation signal. This failure j also results in a nonconformance with the single failure criteria. 1 i 4 I 1 i

) ATTACHMENT VI l II. Single Failure Analysis for Exhaust Vent Valves, Scheme 8512 The exact similarity between Schemes 8501 and 8512 would result in an j identical analysis for both. Therefore, the analysis for Scheme 8512 will not be separately detailed. III. Conclusions 5 In performing a systematic single failure analysis in the format suggested by IEEE Standard 379-1977, the following has been determined: 1. The required protective function is containment isolation during a postulated fuel handling accident inside containment. 2. The required protective action is the closure of the containment 1 isclation valves. -t 3. The closure of the isolation valves is the only system available to provide the protective function. 4. Redundant isolation valves exist in the system, but there is no clearly defined independence or redundancy in the control circuitry that operates the isolation valves. 1 5. After conducting a systematic evaluation of potential failures, the single failure criteria is not met for the scheme as a whole. (Several components do, however, meet the single failure criteria separately.) The following assumptions were made in this single failure analysis: i i 1. There are no identified nondetectable failures inherit to the system ) design. l l; 2. The system is qualified to withstand the affects of a seismic event l without failure to any of its components. l It is, therefore, not required to analyze the system in the presence of event-caused failures and/or identified nondetectable failures coincident to any single failures. I 2 ) i I s -.}}