ML18052A405
| ML18052A405 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 04/28/1986 |
| From: | Kuemin J CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8605020043 | |
| Download: ML18052A405 (4) | |
Text
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/llllCHlliAN'S PROliRESS General Offices: 1945 West Parnell Road, Jackson, Ml 49201 * (517) 788-0550 April 28, 1986
- Director, Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT -
PROCEDURE AND TRAINING IMPROVEMENTS RESULTING FROM SINGLE FAILURE ISSUE FOR MAIN STEAM AND FEEDWATER ISOLATION VALVES The NRC letter of February 28, 1986 enclosed a safety evaluati.on on Main Steam Line Break - Single Failures and requested within 60 days that Consumers Power provide plans for addressing the items in section 6.5 of the safety evaluation.
Section 6.5 concludes that procedural and operator training improvements are required in four areas.
Consumers Power Company is currently in the process of completely rewriting our Emergency Procedures.
It is our intent to address the four areas of concern with these new procedures.
The operator training improvements requested will then automatically be addressed in the associated training program which will familiarize the operators with the new procedure set.
The training program is currently scheduled to start in August 1986 and the new procedures are to be implemented by December 1986.
As of April 1, 1986, most of the items in the four areas of concern were already addressed in the new Emergency Operating Procedures (EOPs).
A few additions have been initiated to ensure all areas of concern are fully addressed.
Each of the four areas of concern is discussed below.
Item 1 NRC Concern:
The emergency procedures dealing with secondary line breaks, EOP-6 and EOP-7, and the procedures for normal reactor trip, EOP-1, do not provide definitive guidance regarding maintenance of a heat sink (use of steam generator level or other secondary system parameter for feedback) even if the operator perceives that both steam generators may be faulted.
In fact, in EOP-1 it is noted that if dryout occurs, the affected steam OC0486-0084-NL04
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Director, Nuclear Reactor Regulation
Palisades Plant Procedure & Training Improvements April 28, 1986 2
generator is to be considered inoperable.
Additionally, the absence in the procedures of any recognition of overcooling expected to occur in these events enhances the potential for inappropriate spontaneous operator action.
Procedures based on systems analysis should be developed to enable the operator to cope with these events and to achieve a controlled cooldown.
The staff considers that analyses are needed to identify a control strategy that considers the decay heat generation rate, auxiliary feedwater flow requirements, instrumentation and controls available to the operator and how the response changes over the course of the cooldown.
CPCo Response:
All new EOPs provide definitive guidance regarding maintenance of the heat sink.
This includes monitoring of specific parameters including*
wide range steam generator levels, core temperature indications, steam generator pressures, and steam generator feed flow.
All new EOPs have safety function status checks which include the safety functions of core heat removal and PCS heat removal.
In all new EOPs, emphasis is placed on continuous feed to at least one steam generator, even if both S/Gs are "dried-out", until shutdown cooling or long term-post PCS break-core cooling is sufficient for decay heat removal.
It should also be emphasized that in all new EOPs, confirmation of adequate core heat removal requires the confirmatory indication of at least in-core temperature.
In the Functional Recovery Procedure, if in-containment indications are considered lost, the continuing actions for PCS Heat Removal and Core Heat Removal include maximizing feed flow to available steam generators.
The new EOPs are structured such that,* if in-containment vital instrumentation is considered lost then the Functional Recovery Procedure is implemented.
If a decision is made to not rely on any in-containment instrumentation, the result is that 1) the PORVs are opened and all available safety injection (SI) pumps started and SI valves open; 2) auxiliary feed flow is maximized and directed to at least one S/G; 3) containment air coolers are placed in emergency configuration and containment spray manually maximized; and 4) all available service water and component cooling water pumps operated.
Item 2 NRC Concern:
Because some instrumentation inside containment may be affected by the two steam generator blowdown environment, operator training/procedures to cope with possible loss of information or misinformation is needed.
OC0486-0084-NL04
Di~ector, Nuclear Reactor Regulation
'Palisades Plant Procedure & Training Improve~ents April 28, 1986 CPCo Response:
3 Cau~ion notes will be added to the beginning of ope~ator actions and Safety Function Status checks of the optimal recovery procedures for LOCA and Excess Steam Demand Event (ESDE-the events in which degraded containment atmospheric conditions are expected).
These notes will state that due to the pote~tial adverse effects of high containment temperature/pressure on in-containment instrumentation, the operator should not rely on any s.ingle instrument, but rather observe confirmat-ory indications; and significance should be placed heavily on trending and less significance on specific instrument readings.
In light of these notes, the existing guidance in the LOCA and ESDE draft.
procedures for maintaining S/G feed flow would prevent deliberately stopping auxiliary feedwater flow.
Additionally, the SI pump throttling criteria, which require a high operator confidence level in primary temperature and pressure indications, would **correctly.prevent any operator action limiting SI flow.
The acceptance criteria for containment atiµosphere.conditions in the LOCA an.d ESDE procedures will be changed to have an. upper limit of design pressure and temperature.
Therefore, these conditions in safety function status checks should "flag" operators into.using the Functional Recovery Procedure.
Item 3 NRC Concern:
The ADV open automatically on high RCS temperature. *The licensee noted in its evaluation that if there is a path from the RCS or the containment to the steam generator secondary system, suc.h as would occur from steam line breaks inside containment or SGTR, a release path would exist.
In the PRA the licensee assumed a high likelihood that the operators would close the valves.
To validate this assumption, the procedures and __
training should address the need to manually close these valves to*.*
provide containment integrity under such circumstances, especially since the emphasis for the ADV in procedures may be on keeping the valves open as a heat sink.
CPCo Response:
Steps which will 'require the ADVs to be selected to MANUAL control will be added to Steam Generator Tube Rupture, Excessive Steam Demand, and Functional Recovery Procedures.
This will ensure that the valve*s will be opened only when needed for core cooling.
Throughout the new EOP set, when a choice is unavoidable between a radio-active release to the environment and maintenance of core cooling, core cooling is maintained.
OC0486-0084-NL04
I.,
V Dir.ecror, Nuclear Reactor Regulation Palisades Plant
.Procedure & Training Improvements April 28, 1986 Item 4 NRC Concern:
4 A steam generator tube rupture with MSIV failure results in a loss of RCS i~ventory into the secondary side until system pressure is reduced.
Recognition of this possibility and contingency measures should be addressed.
CPCo Response:
The new Steam Generator Tube Rupture EOP specifically requires reduction of PCS pressure to within 100 psi of secondary pressure.
This will gr~atly reduce primary to secondary flow and primary. makeup consumption.
Also, a* caution will be inserted into the SGTR procedure which will advise the operators that steaming a steam generator which has a ruptured tube will cause a loss of primary makeup inventory.
They will also be advised to monitor SIRWT level and to initiate makeup as necessary in order to maintain SIRWT. level within Technical Specification requirements.
- ~x.~
?'iames L Kuemin Staff Licensing Engineer CC Administrator; Region III, USNRC NRC Resident Inspector - Palisades OC0486-0084-NL04