ML18348A868

Letter Reactor Vessel Overpressurization

ML18348A868
Person / Time
Site:	Palisades
Issue date:	03/08/1977
From:	Hoffman D Consumers Power Co
To:	Schwencer A Office of Nuclear Reactor Regulation
References
Download: ML18348A868 (10)

Text

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General Offices: 212 West Michigan Avenue, Jackson, Michigan 49201

• Area Code 517 788-0550 March 8, 1977 Director of Nuclear Reactor Regulatio Att: JY.ir Albert Schwencer, Chief Operating Reactor Branc.h No 1 US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-255 - LICENSE DPR PALISADES PLANT - REACTOR VESSEL OVERPRESSURIZATION As requested in your January 10, 1977 letter, one of two proposed implementation options for providing overpressure protection must be selected. As indicat§:d in our Decem-ber 6, 1976 letter, plant modifications will be completed. during our next refueling outage, presently scheduled to start in August 1977. This

• plan is contingent upon a plant specific analysis that indicates that the exist-ing PORVs provide acceptable relief capacity for all potential overpressuriza-tion inc id.en ts.

The design criteria developed during the November 3, 1976 mee-cing with the CE uwn-er' s group and published with* your January 10, 1977 letter seems to be in slight disagreeme:::i.t with our notes of the meeting. The "Single Failure Criteria" and "Seismic Design and IEEE-279 Criteria" are requirements that even the staff were in disagreement on. These requirements if followed completely may lengthen the time to complete long--term fixes. The results of our plant specific analysis and proposed modifications will consider the criteria and provide justification for any variances.

Attacb.."Ilent A is a summary of the questions/answers requested in your LTanuary 10, 1977 letter. Eome of the detailed information cannot be provided prior to com-pletion of the plant specific a~alysis.

It shouid be emphasized that Consumers Power is making every effort to comply with the Tu"R.C req_uests on potential overp:ressuriz.3.tion incidents and has taken corrective actions to eliminate such occurrences by modifying plant operating

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• corrective actions to eliminate such occurrences by modifying plant operating procedures and providing extensive training of plant operating personnel.

David P Hoffman Assistant Nuclear Licensing Administrator CC: JGKeppler, USNRC

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• Question l

-1 ATTACHMENT A The staff considers it essential that all plant operators (ie, reactor operators, equipment operators, Instrument & Control personnel) be made awe.re of the details of the pressure transients which have taken place at all PWR facilities.

POSITION: Formal discussions should be held with the operator to review the causes of past pressure transients that have occurred at other operating PWR facilities. Your discussions should include the plant conditions at the time, the mitigating action that could have been or was taken; and the preventive measures that could have been taken to avoid the event and the steps taken to prevent similar, further occurrences. Plant similarities and distinctions should be identified along with how these relate to plant start-up, shutdown, and test-ing operations. With regard to this position, you are requested to provide the following information:

a. If you have not already completed the required formal discussion, when will you do so?

b. How will the discussion be held?

c. Of the past PWR Appendix G violations that have occurred at PWR facilities and which are described in Licensee Event Reports, identify which are not credible in your plant due to equipment differences. Provide a description of the distinctions.

d. Describe, in detail, how you are reducing the likelihood of the other remain-ing credible events. Furnish schematics, diagrams or procedural summaries necessary to support the effectiveness and reliability of these measures.

Answer l Informal discussions were held with shift supervisors and licensed operators at the time procedural changes (November l976) were made to minimize overpressuriza-tion incidents. Beginning in February l977 formal training sessions, as part of the normal operator requalification program, were started to review all past PWR overpressurization incidents which might be applicable to the Palisades Plant.

In addition, particular Palisades Plant conditions and operations which might be susceptible to overpressurization incidents were identified and preventive measures discussed. The training sessions will be completed by March l5, 1977 and will have included Palisades Plant auxiliary operators, instrumentation

• technicians, licensed operators and shift supervisors.

i

• The "Generic Report - Overpressurization Protection for Operating CE NSSS" dated December 3, 1976 has identified potential credible events which might cause over-pressurization events in the Palisades Plant. The potential of these events was significantly reduced by procedural changes identified in our November 5, 1976 letter. The plant specific analysis, now in progress, will more completely identify events that the Palisades Plant might be susceptible to.

Question 2 The majority of the reported pressure transients f=Vents have occurred while the plants were operating in a water solid condition.

POSITION: The staff will require that operations <luring which the plant is maintained in a water solid condition be minimized or, if possible, eliminated.

Those operations in which the plant is in a water solid condition must be fully justified. Accordingly, please provide the following information:

a. Describe the procedures, evolutions or situations that require the plant be maintained in a water solid condition. Also, provide reasons why a nitrogen, air or steam bubble cannot be maintained in these situations .

• b. Include sufficient background or supplementary information such as system diagrams, procedure summaries and descriptions of equipment operation to justify your need for operating the plant in a water solid condition.

Answer 2 The Palisades Plant operating procedures were reviewed and modified on November 1, 1976 to minimize initiating conditions that could lead to overpressurization. As identified in the CE generic report dated December 3, 1976, the following actions lead to the most severe transients:

a. A primary coolant pump (PCP) start with hot steam generators.

b. Spurious safety injection signal.

c. Inadvertent start of a high-pressure safety injection (HPSI) pump.

As previously identified, the Palisades Plant Operating Procedures required that one or more PCP be operated until the steam generators are in equilibrium with the rest of the primary coolant system (PCS). The safety injection bottles are isolated and the fuses for the HPSI pump motors are removed during plant cooldown

• from hot standby (at 1400 psia on PCS).

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• Alternatives to water solid operation have been identified in the CE generic report and will be further addressed in the final fixes. At present, operation with a nitrogen or air bubble does not appear to be feasible due to chemistry and waste gas handling problems. The plant specific analysis will identify those conditions for which a steam bubble can be maintained.

Question 3 The inadvertent operation of SIS components during the cold shudown conditions has been responsible for a major portion of the overpressure incidents.

POSITION: Based on the licensee submittals, the recent November 3-5, 1976 meet-ings, and discussions with NSSS vendors, the staff will require the de-energizing of SIS pumps and closure of SI header/discharge valves during cold shutdown oper-ations.

In your November 5, 1976 letter to NRC staff, Paragraph 2, you state that the plant HPSI pumps will be disabled and returned to service at the same time as the safety injection tanks. Based on our review of this proposal, and the position stated above, the following additional information is requested:

• a.

b.

c.

A schematic diagram of the SIS showing the flow paths into the RCS .

The head flow characteristics of each of the SIS pumps.

Identify on the schematic diagrams the pumps and the valves to be closed and disabled.

d. How you intend to "disable" a component, and*if this is not de-energized, why power cannot be removed.

e. Your time schedule for implementing these administrative and operating pro-cedural changes to meet this position.

f. Indicate all circumstances for which the SIS pumps and valves may not be isolated and disabled, and for those situations, describe the manner in which SIS injection would be prevented.

g. If it is your intention to place component control switches in the "lockout" or "pullout" position, the location of these switches. Can the components be controlled from any other location? How?

h. The location of the component supply breakers, and the places from which they can be controlled.

i. Describe the position indication and status signals which would be lost if the components were (1) de-energized or (2) disabled by placing control switch in "pullout" or "lockout."

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• • j. Describe in detail; the administrative procedures which will be used to assure proper equipment alignment and the supervisory personnel respon-sible for maintaining control.

k. Indicate the RCS temperature/pressure conditions for which the accumulator isolation valve will be closed. .How will this valve be disabled and if the breaker is not opened, why it cannot be. The locations from which the breaker can be controlled.

1. Describe the impact on overall plant operations if you routinely lowered accumulator nitrogen pressure when in a cold shutdown.condition.

Answer 3 Although our November 5, 1976 letter stated that our HPSI pumps are disabled during cold shutdown, our procedures, in fact, require that the fuses for the HPSI pump motors be pulled. This became part of the Palisades Plant Operating Procedures.on November 1, 1976. Detailed answers to Question 3 will be provided with the plant specific analy~is.

Question 4

• The staff has noted that several Appendix G violations have occurred during component or system tests while in cold and shutdown conditions. In this regard, please address the following questions.

a. What components or systems that could cause overpressure transients are routinely tested while in a cold shutdown condition?

b. What extra measures are taken to prevent in overpressure event during these tests?

Answer 4 The Palisades Plant's required Safety Injection System (SIS) tests are performed while the plant is at power or in a hot standby condition. The testing includes tests for the correct operation of *all SIS pumps and their control circuits, all

• actuator-operated valves and their control circuits, and operability of numerous check valves as described in Palisades Plant Operating Procedure B3.5. The performance of SIS testing while at power or in a hot standby condition eliminates potential overpressurization incidents during component and system testing .

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• Question 5 The staff believes that a high-pressure alarm used during low RCS temperature operations is an effective means to attract the operator's attention to a transient in progress.

POSITION: The staff will require that if it is not presently installed, such an alarm must be installed as soon as possible. Accordingly, please furnish the following information:

a. Your method to provide the alarm, and the associated time schedule,_ or your justification for why this cannot be done.

b. A synopsis of the system modifications that are necessar~.

c. The alarm set point, mode of annunciation and sensor.

d. How you ensure that the alarm is available and operating properly during all water solid operations and how you minimize its downtime for all other cold shutdown conditions.

Answer 5

• The Pali.sades Plant does not currently have fhe capability of providing a low-pressure alarm during low PCS temperature operation. It is intended as part of the.final design modification for pressure protection to provide a single set

• point alarm at approximately 400 psi when the PCS is below 200°F. The final design details will be provided With the completion of the plant specific analysis.

Question 6 The RHR (or SCS) is normally connected to the RCS and operating when the plant is in a cold shutdown condition. The inadvertent isolation of the RHR system while water solid has caused a number of overpressure transients, and the RHR safety valve has actually terminated others. The RHR (or SCS), therefore, plays an important part in the initiation and possible mitigation of the PWR overpres-surizations. Accordingly, we retj_uest the following additional information.

a. RHR (or SCS) design pressure.

b. A description of the system isolation valves and their arrangement (eg, number and configuration of valves installed, and pneumatic or motor operated).

c. Interlocks, interlock set points and alarms associated with each isolation valve.

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• d.

e.

Nominal stroke time of isolation valves.

The set point ~nd capacity of RHR (or SCS) relief and safety valves.

f. All pressure alarms, set points and associated annunciation for the system.

Answer 6 The Palisades Plant does not have a pressure interlock on increasing pressure while operating in the shutdown cooling mode. The following information is provided for the shutdown cooling system.

Shutdown Cooling System

1. Maximum operating pressure = 270 psia and maximum operating temperature =

325°F per Plant Operating Procedures, Section B3.4.l.2.b. Also, see FSAR Table 6-3 on Page 6-16 for design data on SDC heat exchangers. See P&ID Drawing M-204 for SDC system schematic.

2. SDC relief valves.

NOTE: SDC uses LPSI pumps for coolant circulation, returns cooled water to the PCS through safety injection lines.

a. Suction relief valve:

- RV3164 on P&ID M-204, drawing location D&E-2.

Set to relieve @ 300 psig.

- Relief flow capacity - 133 gpm .

. b. Discharge relief valve (this also LPSI relief valve):

- RV3162 on P&ID M-203, drawing location C&D-1.

- Set to relieve @ 500 psig.

- Relief flow capacity =5 gpm.

3. Yes* - The SDC suction valves have an isolation interlock. This interlock only operates when pressure is decreasing, not when it is increasing. There are two such valves in'series, MOV-3015 and MOV-3016, as shown on P&ID drawing M-204 at drawing location D&E-1&2.

a. The suction valve interlock is automatic and it operates off of PS-0103 which is on the pressurizer per P&ID drawing M-201 @ location F-3.

b. The isolation valve interlock pressure set point is 260 psig on PS-0103.

This interlock only operates on decreasing pressure.

• c. The closure time of the MOVs is 45 seconds .

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• 4.

5.

The LPSI pumps shutoff head (psig) = 410 ft= 177.7 psig:

- Per FSAR Table 6-1 on Page 6-14.

The containment spray pumps shutoff head (psig) = 500 ft = 216.8 psig:

- Per CP Co File M8, Sheets 23, 25 and 27, which are from the preoperation pump test characteristic *curves.

Question 7 We have reviewed your letters dated October 7 and November 5, 1976, and the interim measures described. Paragraph 1 of your November 5 letter states your 1

requirement to run at least one PCP until the loops are essentially isothermal and the steam generator shell pressure is essentially atmospheric. Regarding this procedure:

a. What is the status of primary coolant flow in the other loops if only one PCP is in operation?

b. Provide diagrams and schematics to show that the steam generator shell side cannot be subsequently heated after completion of cooldown. Signify the valves that are closed, pumps that are disabled, and other flow or tempera-ture limiting devices .

• c.

d.

What instrumentation is available to measure the RCS temperature conditions?

What methods are used to prevent a nonisothermal RCS from developing?

Provide system diagrams and schematics to illustrate the technique, and include procedure descriptions or summaries.

Answer 7 The actual primary to secondary temperature and pressure differences for a PCP start during water solid conditions will be addressed in the plant spec~fic analysis.

As relates to PCP operation, the following information is provided:

a. With all 4 PCPs operating, the average flow per loop/pump is approximately 85,000 gpm.

b. With only 1 PCP operating, the flow through the operating pump is approximately 123,000 gpm. The nonoperating pump on the loop with the operating pump ex-periences approximately 20,000 gpm reverse flow resulting in a flow of ap-proximately 103,000 gpm through the steam generator. The backflow through

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• each of the other nonoperating pumps is about 4,700 gpm.

through the nonoperating steam generator is_ about 9, 400 Therefore, the flow gp~. The flow paths, equipment and instrumentation are detailed on Palisades Plant P&ID Numbers M-20l ~nd_M-207.

Question 8 We have reviewed your proposed reliance on the Shutdown Cooling System(SCS) relief valve during an inadvertent .isolation of the letdown valve. The following additional information is requested.

a. What alarms are available to the control room.personnel that indicate the SCS relief's opening?

b. What are the testing requirements of this valve to ensure its proper opera-tion?

c. What is the anticipated RCS pressure overshoot while the safety valve is opening and the solid system is being pressurized?

Answer 8

• As indicated in our November- 5, 1976 letter to the NRC, a pressure transient caused by an inadvertent isolation of the letdown control valve while operating in a shutdown cooling mode would be limited to approximately 300 psi. Relief capacity of 133 gpm, equivalent to the capacity of all three charging pumps, is provided by the Shutdown .Cooling Relief Valve (RV-3164). This relief valve is a totally enclosed, bellows type valve and the set point is checked and reset, if necessary, as part of the Palisades Plant Inservice Testing of Plant Valves (Engineering Manual Procedure EM-09-02 dated October 29, 1976).

Additional analysis of PCS conditions during a water solid transient will be provided with the plant specific analyses .

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