ML20213G742
| ML20213G742 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20213G733 | List: |
| References | |
| NUDOCS 8611180306 | |
| Download: ML20213G742 (8) | |
Text
,
l.
.autuq%
l UNITED STATES I
I' NUCLEAR REGULATORY COMMISSION o
{
E WASMNGTON, D. C. 20555
\\...../
SAFETY EVALUATION BY THE OFFICE Or NUCLEAR REACTOR REGULATION S!!PPORTING AMENDMENT NO. 112 TO PPOVISIONAL OPERATING LICENSE NO. DPR-16 GPti NttCLEAR CORPORATION AND
.1EPSEY CENTRAL POWER & LIGHT COMPANY OYSTER CPEEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 1.0 INTR 0DitCTION By letter dated September 11, 1986, GPU Nuclear (the licensee) requested an amendment to Provisional Operating license No. DPR-16 for the Oyster
(
Creek Nuclear Generating Station (0yster Creek). This amendment would authorize changes to the Appendix A Technical Specifications (TS).
j These changes would (1) increase the high drywell pressure trip setpoint limit to 3.5 psig, (2) add a bypass to the high flow trip to the "B" Isolation Condenser when initiating the alternate shutdown and (3) revise the appropriate Bases of the TS. These are changes to Table 3.1.1, Protective Instrumentation Requirements, and the Bases of Section 3.1, Protective Instrumentation, of the TS.
2.0 DISCUSSION AND EVALUATION The proposed amendment would (1) increase the high drywell pressure trip setpoint from not greater than 2.4 psig to not greater than 3.5 psig for the instrumentation to actuate its associated trip systems and (2) add a bypass to the high steam line flow trip and high condensate return line flow trip for isolation of the "B" Isolation Condenser for when the alternate shutdown panel is initiated. The associated trip systems for the high drywell pressure instrumentation cause reactor scram, core spray initiation, containment spray initiation, containment isolation, automatic reactor vessel depressurization and reactor building isolation and standby gas treatment system (SGTS) initiation.
For the bypass, the licensee is proposing to add a footnote "hh" stating that the trip function is bypassed upon initiation of the alternate shutdown panel to prevent a spurious trip of the "B" Isolation Condenser in the event of fire induced circuit damage.
The instrument setpoint for the High Drywell Pressure TS limit of 2.4 psig was found by the licensee to be unacceptable to maintain and achieve safe shutdown conditions during a postulated Appendix R event. This event involves a fire. For this event, the drywell cooling fans are assumed to be lost due to the fire and the reactor is being cooled by the Isolation Condenser with no feedwater flow. Thus, it is essential that the Automatic Depressurization System (ADS) logic is not actuated by high drywell pressure to further reduce reactor water level.
8611180306 861031 DR ADOCK 0500 9
. The licensee concluded in its analysis for this event that the drywell pressure could exceed the current high drywell pressure instrument setpoint when you take into account the required adjustments to the instrumentation to have its setpoint less than the existing 2.4 psig TS limit. The actual setpoint incorporated in the high drywell pressure instrumentation setpoint is lower than the TS limit to account for possible instrument error and drift. Th6refore, the three ADS actuation logic signals (low-low-low reactor water level, high drywell pressure, and core spray pump discharge pressure) may be satisfied with the existing TS instrument setpoint of 2.4 psig for high drywell pressure and, therefore, result in an inadvertent initiation of ADS. To prevent such an inadvertent actuation of ADS during a postulated Appendix R event and to minimize spurious trips caused by instrument drift, a revised TS limit of 3.5 psig has been requested by the licensee for high drywell pressure.
3.0 EVALUATION 3.1 High Drywell Pressure Setpoint The licensee has evaluated the acceptability of increasing the dryweil pressure TS limit to 3.5 psig and determined the effect of the increased TS limit on anticipated plant operational occurrences and design basis accidents. Each of the protective functions listed below was examined by the licensee to determine how each function would be altered by the new TS limit and how this altered protective function response would affect the plant design response to the transients and accidents evaluated on the Oyster Creek docket. The following are the conclusions by the licensee:
1 Reactor Scram The high drywell pressure scram function is p ovided to shut down the core following a loss-of-coolant accident (LOC.1).
For most LOCA events, this function will precede a low reactor water 'evel scram signal.
However, the Oyster Creek LOCA analyses which were submitted by the licensee in response to 10 CFR 50.46 and 10 CFR Part 50, Appendix K, demonstrate for large breaks that shutdown will occur as a result of excessive voiding, not high drywell pressure.
For small breaks, scram will occur at 0.3 second as a result of the loss of o'fsite power.
Therefore, the scrams which would have been associated with high drywell pressure were not the determinant scrams.
Small breaks without a loss of offsite power would be less severe due to feedwater availability. The core is adeouately protected by a scram on low reactor water level and a scram associated with main steam isolation valve (MSIV) closure at low-low reactor water level.
In any case, the scram delay associated with a drywell pressure TS limit increase to 3.5 psig is minimal and is more than compensated for by the conservative scram reactivity curves used in the analyses.
Further, the normal operating drywell pressure is typically greater than atmospheric pressure which is assumed in the analyses, and thus the actual pressure difference and associated time delay is less than in the analysis.
. For large steam region breakt with feedwater available, the scram associated with high main steam line flow, low system pressure and low reactor water level would occur at approximately the same time i
as the high drywell pressure.
In these cases, the effect of a TS limit increase to 3.5 psig would be negligible on the transient behavior. For i
small steam line breaks with feedwater, high drywell pressure is the only scram function.
In these cases, the small increases in the setpoint would have a nealigible effect on the transient severity.
In these cases, the vessel pressure and level remain within nonnal bounds so that the core is cooled in the normal manner.
In this way, even a large delay in scram time on high drywell prew:'re would not have any impact on this LOCA. The operator could, in fact, proceed with an orderly shutdown if the scram does not occur.
Core Spray Pump Start The core spray pumps will automatically start on either low-low reactor water level or high drywell pressure.
Depending on the nature of the LOCA, either one or both of these signals will be available.
In all i
cases analyzed for Oyster Creek, the time required to depressurize the system to the 285 psig core spray permissive pressure is limiting with respect to core spray initiation.
Thus, the core spray flow will begin at the same time for either a 2.4 or 3.5 psig high drywell pressure TS limit.
I Containment Spray The containment spray system will actuate automatically upon indication i
of high drywell pressure and low-low reactor water level. Depending upon the size and location of the break and whether or not feedwater is j
available, the high drywell pressure signal will occur either alone or in conjunction with low-low reactor water level. For all break sizes either above or below the core, without feedwater, high drywell pressure will I
occur prior to low-low reactor water level. A high drywell pressure TS limit increase from 2.4 to 3.5 psig will not change this result.
For large breaks with feedwater, this conclusion is also valid. For small breaks with feedwater, low-low reactor water level may not occur and i
operator action will be required to initiate the sprays. Again, the l
increased hiah drywell pressure setpoints will not affect this conclusion. There are no LOCA events analyzed on the docket for which the increase of 1.1 psig in the TS limit will prevent or delay the automatic initiation of the containment spray system.
Primary Containment and Reactor Puilding Isolation Primary and secondary containment isolation results automatically from high drywell pressure or low-low reactor water level. The arguments presented above regarding the negligible delay in high drywell pressure indication associated with a 1.1 psig TS limit increase are applicable here. Coupling this with the fact that high drywell pressure precedes low-low reactor water level for all break sizes and locations provides assurance that fuel damage will not have occurred as a result of a LOCA prior to isolation of the containment. The TS limit increase
_4 will not alter the order of signal initiation for all breaks analyzed.
As indicated previously, even for very small steam breaks, the time delay associated with the 1.1 psig TS limit {ncrease is negligible (approximately 40 seconds of a 0.01 ft main steam line break).
Automatic Depressurization System (ADS)
The actuation of the ADS, which is required during a small break LOCA to depressurize the vessel and permit low pressure core spray flow, is limited in its initiation by the time required to reach low-low-low reactor water vessel level.
For small breaks with or without feedwater flow, high drywell pressure will be reached within seconds even for the smallest break analyzed on the Oyster Creek docket. The time required to reach low-low-low reactor water level for these cases is much longer.
If feedwater is available, low-low-low reactor water level may not be reached in some cases and will be delayed in all cases. Thus, a high drywell pressure TS limit increase of 1.1 psig will not result in a change in the initiation time of ADS for any small break analyzed.
Standby Gas Treatment System (SGTS) Initiation The SGTS treats and exhausts the atmosphere of the reactor building to the stack during containment isolation conditions. This prevents around level leakage of fission products from the reactor building. This system is initiated by high drywell pressure or low-low reactor water level analogous to primary and secondary containment isolation.
The arguments pertaining to reactor building isolation are all applicable to the SGTS. Because both are initiated by the same signals, the SGTS will be available to perform its intended function simultaneously with isolation of the reactor building which is its mode of operation in response to an accident.
Jime Delay to Reach 3.5 psig Drywell pressure This change results in the initiation of automatic protection actions to protect the drywell at a later time in an accident involvino loss of 4
coolant and an increase of pressure in the drywell. As described above, the setpoint is still low enough that the trips initiated by the affected instruments will occur in time to ensure that they will perform their protective function. Based on the design basis LOCA (a large break) inside the drywell in Section XIII-2.4 of the Oyster Creek Unit No.1 Facility Description and Safety Analysis Report, the new setpoint will initiate trips less than 0.1 second later in time. The analysis in Section XIII-2.4 also ignores the trip that occurs at high drywell pressure because it is based on the initiation of automatic protective i
functions with the later trip at low-low reactor water level in the core.
Therefore, this analysis does not change with this proposed action to raise the trip setpoint for high drywell pressure from 2.4 to 3.5 psig.
. 3.2 Rvpass For the "B" Isolation Condenser Isolation Trip The following is related to the proposed bypass for "B" Isolation Condenser isolation. There is no bypass for the "A" Isolation Condenser. The alternate shutdown capability is being provided to assure safe shutdown and cooldown of the reactor in the event of a fire causing evacuation of the control room or the loss of control room function due to fire damage in the cable spread rooms. The design of the alternate shutdown system includes bypassing the high flow trip function for high flow in the steam line to and condensate return line from the "B" Isolation Condenser.
These trips are to isolate the condenser for a break in either line when the condenser should be isolated to prevent loss of water from the reactor coolant system.
The design of the alternate shutdown system including this bypass was revieued and approved by the staff in Sections 8.1.4 and 8.2.4.3 of its Safety Evaluation (SE) dated March 24, 1986, based on the licensee's submitt31s on its Fire Protection Plan dated April 3, July 12 and October 9, 1985. This capability utilizes the isolation condenser for decay heat removal and reactor cooldown to establish a safe shutdown condition. Since a fire affecting cabling associated with the high flow isolation condenser trip function could result in a spurious isolation of the "B" Isolation Condenser, the design includes a bypass of the trip function upon initiation of the alternate shutdown panel.
In Sections 8.1.4 and 8.2.4.3 of its SE, the staff stated that the control logic circuits of the isolation condenser valves will be modified and cables rerouted to allow control of the isolation condenser from the remote shutdown panel, that the devices whose inadvertent operation by spurious signals could affect safe shutdown have been identified as shutdown circuits and are included in the separation analysis and that the licensee will provide isolation and transfer switches for all shutdown circuits as needed to prevent spurious operation. This SE is based on Table D2, page D-22 and on the section, Deca Heat Removal, in Appendix A which discusses the use of the Isolation on ensers to provide hot shut-down capability during a fire. These are in the licensee's Fire Protection Plan. The Table D2 lists the isolation valves V-14-32, V-14-33 V-14-35 and V-14-37, main steam line and condensate return line of the "B" Isolation Condenser, respectively, as sub,iect to possible spurious closure which must be prevented.
In Appendix A, the licensee states that the control of the cooldown using the Isolation Condensers reouires the operator to keep the steam line valves and one of the condensate return valves open and cycle the other condensate return valve open and close.
The occurrence of a fire which requires initiation of the alternate shutdown panel in conjunction with a line break accident is not considered a credible event. The bypass becomes effective only when the alternate shutdown capability is initiated and this is done by operators at the alternate shutdown panel. The panel is initiated at the panel through transfer switches which are key locked. The keys are locked away in the control room. This is to prevent inadvertent actuation of the panel and inadvertent initiation of the bypass. There is also an alarm in the i
I
~
. control room when the panel is actuated to inform the operators that the panel has been actuated.
In addition, a single failure of the switch will not preclude operation of the Isolation Condenser high flow trip in the event of a line break accident.
If the Isolation Condensers should not be available or the isolation valves are tripped, the alternate hot shutdown capability in the licensee's Fire Protection Plan is the Electromatic Relief Valve Path.
The operators at the alternate shutdown panel have parameters before them as reactor coolant system water level and pressure to indicate the shutdown capability of the Isolation Condensers. The operators can isolate the Isolation Condensers and switch to the Electromatic Relief Valve Path if this is needed.
3.3 Conclusions The staff has reviewed the licensee's justification in Section 3.1 for increasing the high drywell pressure setpoint to less than or equal to 3.5 psig in Table 3.1.1 of the TS. The staff concludes that this increase will significantly reduce the possibility of unneeded ADS actuation during an event while not changing the accident analysis based on this setpoint. Therefore, the staff concludes that this change and the changes to the Bases describing this change are acceptable.
The staff has also reviewed the justification in Section 3.2 for the high flow trip bypass. The staff has accepted the design of the alternate shutdown system including bypassing the high flow trip function for the "B" Isolation Condenser in its SE dated March 24, 1986.
Using this bypass prevents a possible spurious isolation of the "B" Isolation Condenser during a fire. The transfer switches for alternate shutdown are keylocked to prevent inadvertent actuation of the alternate shutdown panel and bypass and no single failure of the transfer switch will preclude operation of the Isolation Condenser high flow trip.
Therefore, the staff concludes that this change is acceptable.
3.4 Consultation With The State On October 24, 1986, the staff's Project Manager consulted with Ms. R. Green and Mr. K. Tosch of the State of New Jersey, Bureau of Radiation Protection. This consultation, in accordance with 10 CFR 50.91(b)(4), was concerned with the staff's intent to issue these TS changes. The State stated that it did not have a concern with the TS change to increase the high drywell pressure setpoint limit but it did have a concern with the bypass for the "B" Isolation Condenser high flow trip.
The concerns of the State were that the Isolation Condenser piping, to and from the reactor vessel, is susceptible to intergranular stress corrosion cracking (IGSCC), has suffered some cracking in the past and is partly located outside the drywell where leaks from these pipes, from small breaks which may not be noticed by the operators, could cause severe environmental consequences offsite.
In addition, the isolation valves on the steam lines from the reactor vessel to the Isolation Condensers are both outside the i
. drywell. The State expressed the concern that the staff was not being sufficiently conservative in accepting this bypass in the proposed TS change instead of having the licensee reroute or fix the cable to prevent the chance of a spurious signal in a fire.
The high flow trip function is provided to isolate the system in the event of a line break. The occurrence of a major firc requiring the evacuation of the control room to the alternate shutdown panel and a line break accident is not required by the staff's implementation of Appendix R and is not considered sufficiently credible to be designed for. The Isolation Condenser system was inspected for IGSCC indications by the licensee in the current Cycle 11R outage and the results were submitted to the staff in the licensee's letter dated October 3, 1986. The licensee concluded, on the 18 structural weld overlays on the Isolation Condenser System, that there were no IGSCC indications and the overlays are acceptable for continued operation. The staff will complete its evaluation of these ins)ections as a separation action, before the plant is restarted from the current outage.
The concern that both containment isolation valves on the steam lines to the Isolation Condensers are on the outside of the drywell is Topic III-5.B. Pipe Break Outside Containment, of the staff's Systematic Evaluation Program (SEP). This is also a separate review by the staff for which the licensee has provided fracture mechanics analyses that the licensee has stated demonstrates that through-wall cracks in the Isolation Condenser steam pipe would open up, yet remain stable, under severe pipe pressure loading and rotation stresses. The analysis concludes that no instantaneous pipe break would occur and the estimated pipe leakage for these cracks would be less than 1 gpm. These lines are in the licensee's inservice inspection program and the inspections are in accordance with Section XI of the ASME Code. The lines are considered adequately sound for continued plant operatiun until the SEP Topic is resolved.
The proposed TS change en the bypass for the "B" Isolation Condenser high flow trip does not affect how the Isolation Condensers may be used.
Therefore, it is not related to the question of cracking in the Isolation Condenser system; however, this system has been inspected durina the current outage and the licensee stated it found no IGSCC indications in the weld overlays inspected. Therefore, subiect to the staff's separate action on this inspection, the system is considered adequately sound.
Appendix R only reouires the evaluation of a loss of offsite power concurrent with the fire and does not require that other unlikely events such as pipe breaks be considered. Therefore, the staff did not require the licensee to reroute or replace the applicable cable to prevent the spurious signal.
However, if a line break should occur after initiation of the alternate shutdown panel, the bypass will affect only the "B" Isolation Condenser.
The reactor will already have been scrammed before the break and, as a minimum, the radiation monitors in the stack for the Reactor Building
r
. ventilation will isolate the Reactor Building and start up the Standby Gas Treatment System (SGTS) to filter the release from the building to the environment. The release will be within acceptable limits and the operators will have sufficient information to isolate the Emergency Condensers if the high flow trip did not and use the Automatic Depressurization System (ADS) to achieve and maintain hot shutdown. The licensee's Fire Protection Plan dated August 25, 1986, has the Isolation Condensers and the ADS to achieve and maintain hot shutdown.
4.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change to a reouirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no sionificant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the elicibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement nor environmental assessment need be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the will not be endangered by operation in the proposed manner, and (2) public such activities will be conducted in compliance with the Connission's regulations and the issuance of this amendment will not be inimical to the common defense and security nor to the health and safety of the public.
6.0 REFERENCES
1.
Letter from J. A. Zwolinski (NRC to P. B. Fiedler (GPUN), Exemptions from Requirements of Appendix R to 10 CFR Part 50, Section III.G.2 and the Post Fire Shutdown Capability, dated March 24, 1982.
2.
Letter from P. B. Fiedler (GPUN) to J. A. Zwolinski (NRC), TSCR No. 147, dated September 11, 1986.
Principal Contributor:
J. Donohew Dated: October 31, 1986.
-