ML18092A745
| ML18092A745 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/24/1978 |
| From: | Christiana J, Leitz L, Wroblewski J Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18092A739 | List: |
| References | |
| CD-S-10, NUDOCS 8508280293 | |
| Download: ML18092A745 (36) | |
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7-19-78 ATTACHMENT D PUBLIC SERVICE ELECTRIC AND GAS COMPANY ENGINEERING DEPARTMENT CONTROLS DIVISION SALEM NUCLEAR GENERATING STATION UNIT NO. 1 AND 2 FUNCTIONAL SPECIFICATION:
CD-S-10 TITLE:
PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
Prepared By:
J. J. Wroblewski f)iw~
J. J. Wroblewski
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Reviewed By:
( 8508280293 g;gg5~72
~DR
- ADOCK PDR Approved:
Chief Controls Engineer
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II.
III.
IV.
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VI.
VII.
VIII.
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SALEM NUCLEAR GENERATING STATION FUNCTION SPECIFICATION CD-S-10 Purpose Scope References PRESSURIZER OVERPRESSURE PROTECTION SYSTEM '(POPS)
CONTENTS Background Information Design Criteria Functional Requirements System Operation Miscellaneous Requirements APPENDIX A -
USNRC Letter to PSE&G (2-4-74)
APPENDIX B - Design Drawings, Unit No. 1 APPENDIX C - Design Drawings, Unit No. 2
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- i. i t PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
I.
Purpose This functional specification describes the general design concept and applicable design criteria for a system to auto-matically open the pressurizer power operated relief valves lPRl and 1PR2 (Unit #1) during certain pressure transients at low temperature operating conditions.
The Unit #2 design uses electrically operated solenoid valves 2PR47 and 2PR48, which are piped in parallel with the normal relief valves.
It also serves as input for revised FSAR information.
II.
Scope This document includes requirements for the instrumentation and controls associated with the system stated in Section I and is designated as the POPS.
III.
References IV.
A.
IEEE 279-1971, "Criteria for Protective Systems for Nuclear Power Generating Stations".
B.
IEEE 344-1971, "Trial Use Guide for Seismic Qualification of Class I Electric Equipment for Nuclear Power Generating Stations:
C.
Salem FSAR - Sections 7.1 and 7.2.
D.
USNRC letter dated February 4, 1977 to General Manager -
Electric Production (Attached as Appendix A).
Background Information The need for a POPS has been determined by the NRC as stated in reference III.D which is attached to this document as Appendix A.
In summary, the system is provided to prevent pressure tran-sient during low temperature operation which would exceed the limitations of plant technical specifications (i.e., Appendix G to 10CFR50).
The overall protection against these pressure transients is provided by a combination of administrative controls and the automatic operation of power relief valves lPRl and 1PR2 (Unit #1) or by the solenoid relief valves 2PR47 and 2PR48 (Unit #2)
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PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
Design Criteria The basic design criteria for the POPS has been provided by the USNRC in reference III.D.
These include the re-quirements of references III.A and III.B.
The description of the normal plant protection system cited in reference III.C is also generally applicable to the POPS design.
The main design criteria applicable to the POPS are summarized below:
A.
Conformance to IEEE-279 and the Single Failure Criterion The details described in Sections VI through VIII illustrate a conceptual design which will satisfy the criteria.
B.
Conformance to Seismic I Requirements All equipment in the POPS shall be Seismic I except alarm system components.
All Class I components shall consist of types previously qualified for Seismic Class I, or shall be qualified to meet Class I requirements in accordance with reference III.B.
C.
The POPS design shall be based on the concept that the operator does not respond for ten minutes following the initiation of a pressure transient.
D.
Testability The POPS shall include provisions for testing the system on a schedule consistent with the expected operating frequency of the system.
Provisions for testing are described in Section VII of this document.
VI.
Functional Requirements A.
General The POPS design is based on the assumption that either of the two relief valves *1 is. suff'ici'e11t-*.,.* ~, p~ e::,,._...... I to alleviate the effects of a low-temperature pressure transient.
The POPS equipment used to actuate these valves shall be designed for independence and separation in accordance with the power supplies assigned to ~he
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PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
VI.
Functional Requirements (Cont'd)
The POPS shall be designed as an administr~tively initiated system.
The operator shall be required to activate the POPS prior to reaching 312°F reactor coolant temperature during a plant shut-down, and shall leave the POPS in service until reactor coolant temperature exceeds 312°F during a plant startup.
B.
IEEE-279 Since the POPS is classified as a "protection-grade" system, it shall be designed to meet the requirements discussed below.
(1)
Automatic Initiation The system shall operate automatically to mitigate the pressure transients associated with low temperature operation whenever it has been activated by controlled means dur-ing low temperature conditions.
(2)
Single Failure The system shall be designed such that any failure capable of causing a pressure tran-sient shall not d~feat protection against that transient even if a coincident failure were to occur in the POPS.
The POPS equip-ment design shall be such that a credible failure within the POPS shall not prevent initiation of the system, however, less likely failures which could conceivably initiate the system need not be precluded by the POPS design.
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VI.
PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
Functional Requirements (Cont'd)
B.
IEEE-279 (Cont'd)
( 3)
Quality The equipment u~ed in the POPS design shall be subject to a quality assurance p~ogram, which is compatible with requirements impose*d on the plant protectiorr system.
(4)
Equipment Qualification The POPS equipment shall meet the require-ments for Seismic I service and be qualified in accordance with Section V.B.
The equipment shall also be capable of meeting the environ-
~ental qualification requirements associated with the plant protection system except for post~LOCA considerations.
(5)
Channel Integrity All POPS equipment shall be capable of main-taining its functional capability during expected extremes of condit~ons which relate to environment, energy supply and malfunctions.
The channels shall be powered by separate vital busses.
(6)
Channel Independenc~
(7)
(8)
The POPS channels shall be independent and separated.
System Interfaces The POPS shall include provisions for alarms l
as shown in Appendi.ces B and C.. The. alarm functions f shall be equipped with appr'opriate isolation devices to assure completion of the system's function even with.failures in the alarm circuitry.
System Inputs The POPS is to be designed so that action is taken when coolant system pressure approaches the pressure allowed by the Technical Speci-fication (i.e., Appendix G to 10CFR50).
This
PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
VI.
Functional Requirements (Cont'd)
B.
IEEE-279 (Cont'd)
(8)
System Inputs (Cont'd) pressure shall be determined as a function of coolant system temperature.
Two independent pressure and temperature channels sh~ll be used to achieve proper separation and redun-dancy.
These channels are to be designed in accordance with Appendix B to this document.
(9)
Sensor Checks Means shall be provided to check the opera~
tional availability of the sensors during reactor operation.
This may be the same method used for sensors in the plant pro-tection system.
(10)
Testing and Calibration The channels and devices used in the POPS shall be provided with capability for test-ing and calibration.
System testing is addressed in Section VII.
(11)
Channel Bypass During Power Operation Since the POPS is used only during shutdown and startup conditions, this requirement does not apply.
(12)
Operating Bypasses The requirement for automatic removal of a by-pass condition shall not be met.
Due to the short time interval of POPS system operation in any given operating period, the alarms pro-vided in the POPS and administrative controls imposed on its operation provide adequate assurance that the POPS will be* operable when required.
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PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
VI.
Functional Requirements (Cont'd)
B.
IEEE-279 (Cont'd)
(13)
Indication of Bypasses If the POPS is bypassed during operating con-ditions which require it to be operable, an alarm shall be provided which will continu-ally alert the operator to the bypass condition.
(14)
Access to Bypass The devices used to bypass the POPS shall be subject to administrative control.
(15)
Completion of Protective Action The POPS design shall provide for actuation of relief valves such that automatic re-closure of the valves requires pressure to decrease below the actuating setpoint.
The ability to manually reclose the valves shall be provided to the operator in the main control room.
(16)
Manual Initiation The system design is predicated upon no credit for operator action during the first ten minutes of a pressure transient.
However, the existing manual control functions for PRl and PR2 shall remain as a manual means to initiate pressure relief.
(17)
Access The POPS design shall allow for administrative control of access to setpoint adjustments, calibration and test points.
(18)
Identification of Protective Actions, Information Read-Out, and Identification The POPS design shall include provisions for indication of pertinent parameters as shown in Appendix B to this document.
The equipment and cables associated with the POPS shall be identified distinctly as being part of the pro-tection system and shall distinguish between the redundant portions of the system. I
VII.
PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
System Operation (Refer to Appendix B)
The POPS shall use independent sensors, logic and relief valves in a basic two-division arrangement.
Each relief valve is to be actuated by its own logic output relay which is energized by a bistable device.
The bistable is to be energized if the reactor coolant system pressure exceeds a setpoint based on the allowable 10CFR50 Appendix G pressure.
Existing pressure and temperature sensors shall be used to develop the POPS signals.
The operation of the POPS is to be governed by two administra-tively controlled, keylocked pushbuttons, which will perform three functions.
When reactor coolant temperature is below 312°F the "ON" pu.shbuttons shal.l be depressed.
This action will open the MOV upstream of the relief valve, provide an opening permissive for the relief valve, and provide an alarm permissive to indicate that the POPS is armed if temperature exceeds 312°F.
In this mode the relief valv~s will be opened if the coolant system pressure exceeds the allowable value of 375 psi.
Operation of the system shall be alarmed in the main control room.
I When coolant system temperature is above 312°F, the "OFF" pushbutton is depressed.
This action will remove the opening l' permissive from the relief valve, remove the opening signal from the upstream MOV, and provide an alarm input to indicate tha~ the POPS is disarmed if temperature should decrease below 312 F.
If either relief valve is opened by the POPS, it will remain open until the coolant system pressure falls below the pressure set-I point for valve opening.
This feature requires a sufficient supply of air or nitrogen on Unit #1 to be available for sub-sequent valve openings if the coolant pressure cycles around the relief. setpoint.
The air supply should be capable of operatin the valve for the number of operating cycles expected during the first ten minutes of the transient.
This requirement does not apply to Unit #2.
The relief valves can be tested for proper functioning by I I depressing the "TEST" pushbutton.
This will cause the relief valve to open if its associated upstream MOV has been closed *. i I
VII.
VIII.
PRESSURIZER OVERPRESSURE PROTECTION SYSTEM (POPS)
System Operation (Cont'd)
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- Any portion of the POPS can be tested at any time prior to system activation.
System testing during conditions of "ON" operation is not required.
The POPS shciuld be tested prior to entry into a condition requiring its safety function.
Miscellaneous Requirements The POPS is to be provided with the following indication on a channel basis:
(1)
Reactor coolant system pressure (0-600PSI).
(2)
Reactor coolant system temperature (wide
- range).
An alarm shall also be provided to warn the operator of an approach to the POPS operation setpoint.
The purpose of this alarm is to allow the operator some opportunity to correct a slow pressure transient prior to relief valve operation.
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APPENDIX A USNRC Letter to Public Service Electric and Gas Company - Dated F~bruary 4, 1974
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 FEB 0 4 1977 A??EA!blX,. A-
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Docket No. 50-272 Public Service Electric & Gas Company ATIN:
Mr. F. P. Librizzi General Manager - Electric Production Production Department 80 Park Place, Room 7221 New~rk, New.JP.rsey 07101 Gentlemen:
SUBJECT:
SALEM l\\"UCLEAR GENERATING STATION, UNIT NO. 1 - OVERPRESSURE PROI'ECTION In August of this year we sent letters to you and licensees of other operating P\\\\R facilities which expressed our concern regarding the m.nnber of reported instances of reactor vessel overpressurization.
l'!e requested that an analysis be provided of the reactor coolant systeJT!.
response to pressure transients and that any design modifications be identified that were detennined to be necessary to preclude exceeding the limits of Appendix G to 10 CFR Part SO.
We. also requested that, if the design changes could not be made within a few months, interim measures be implemented immediately to reduce the likelihood of over-pressurization events.
Your letter of October 25, 1976 identified the interim measures that
- were being implemented at your facility.
We have completed our review of your submittal and have concluded that add~_ti~n-~.l.. inf.QJJWl.t.iPn.Jd_lL f (
be required for us to detennine the adequacy of these interim measures.
O.ir revie\\*: of youT submittal u.nd the responses from other Pl'.'R licensees has identified certain measures l~hich in our opinion are significant enough to be required in most Pl\\'R facilities. These measures have been identified as staff positions and are included with the additional information requests contained in Enclosure No. 1 to this letter. In addition, we have concluded that the procedural and administrate measures you have already instituted will help prevent future pressure transients and that they should be continued even after the long-tenn hardware improvements have been implemented unless you can demonstrate that this would not be justified.
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Public Service Electric &
Gas Company FED 0 4 1977 On November 4, 1976, we held a meeting with licensees of Westinghouse PWR's to provide guidance on acceptable long-tenn overpressure protection methods.
A copy of the meeting stumnary is attached hereto.
We called the November meeting to inform the licensees of specific design cri~eria requirements.
Since Public Service Electric and Gas Company was represented at this meeting, we ass\\.lllle that your long-tenn protection methods will be based on the material presented at the meeting. The Division of Operating Reactors, to which Salem Unit No. 1 will be transferred in the near future for regulatory matters, has had additional correspondence with the licensees of PWR reactors under their purview.
As a result of this additional correspondence, licensees of Pt'R reactors have provided information regarding the proposed modifications and schedules of which you should be aware. Since Salem Unit No. 1 will be transferred to the Division of Operating Reactors in the near future, we are advising you of the findings by the Division of Operating Reactors.
regarding the proposed modifications by typical licensees.
In general, the responses provided by the licensees have been.based on their interpretation of the preliminary criteria contained in the August letters; however, they do not adequately address the final criteria as detailed in the November meeting. Since the overpressure protection system being proposed by licensees is based on their interpretation of the preliminary criteria, they may be sufficient as an interim remedy, but they will not be acceptable for the long term remedy until the final criteria are addressed and any deviation fully justified
- To insure that the Salem Unit 1 response meets the final criteria objectives and that your schedule* is acceptable, you should provide completely acceptable responses to the enclosed questions and select either of the following two implementation options:
- 1. Connnitment to a schedule for the installation of acceptable long-term improvements by December 31, 1977, which meet all the desigr. criteria discussed at the :~ovcmber 4, 1976 :meeting held in Bethesda, or
- 2. Connnitment to a schedule to achieve installation of interim hardware improvements by December 31, 1977, and installation of long term improvements that meet all the design criteria discussed at the November meeting during the first scheduled shutdown after December 31, 1977. The dual setpoint system of pressurizer relief valves described in the Westinghouse "Reference Mitigating System" design is an acceptable interim hardware improvement.
Public Service Electric &
Gas Company FEB 0 4 1977
'Ibe long tenn hardware improvements should meet the final design criteria discussed at the November 4, 1976 meeting. If deviations from the criteria are proposed, a detailed" justification should be provided.
Interim improvements need not meet all the design.criteria discussed at the November meeting but must represent good engineering practice and DlllSt not adversely affect plant safety or introduce potential common m:>de failures that could both cause the overpressure event and disable the protection system. For example, if air operated dual setpoint pressurizer relief valves are used as an interim improvement, air accUIJU.llators on t.~e actuation mechanism and alarms to indicate loss of instrument air should be provided to insure that protection is provicled and to alert the operator in the event t:.hat instrument air is lost. Whichever implementation option you select, your proposed design improvements JIUlst be submitted in sufficient time to allow our review and approval to meet the objective discussed above.
You are requested to identify to us the implementation option you have selected and to provide your response to the staff positions and requests for additional infonnation identified in the attached enclosure within 45 days of receipt of this letter.
Enclosures:
Sincerel~y,
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Karl Kniel, Chief Light Water Reactors Branch No. 2 Division of Project Management
- 1. Staff Positions and Infonnation Requests
- 2. Meeting Summary dated November 17, 1976 cc w/encls:
Richard Fryling, Jr., Esq.
Assistant General Counsel Public Service Electric & Gas Company 80 Park Place Newark, New Jersey 07101 Joseph B. Knotts, Jr ** Esq.
Conner & Knotts Suite 1050 1747 Pennsylvania Avcrrue, N. W
- Washington, D. C.
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ENCLOSURE l\\U. 1 ST:\\FF POSITIONS A~D INPOR\\ti\\TIO~ REQUESTS SALEM !\\'UCLEAR GENERl\\.TING STATIC:~, ~IT t.iO. 1 OOCKET NO. SO-t72
- 1. 111e staff considers it essential that all plant operators (e.g., reactor operators, equipment operators, instrument and control personnel) be made a\\~re of the details of the pressure transients which have taken place at Pl~'R facilities. POSITION:
You should conduct fonnal dis-cussions with the operators to review the causes of past pressure transients that have occurred at operating PWR facilities. The discussions should include the plant conditions at the time, the mitigating actions that could have been or were taken, and the preventive measures that could have been taken to prevent the event, inclu~ing the steps taken to prevent similar occurrences in the future. Plant similarities and distinctions should be identified, including their relation to plant startup, shutdown, and testing operations. With regard to this position, you are requested to provide the following infonnation:
- a. The date by which the fonnal discussions will be completed.
- b. A description of how the discussions will be conducted.
- c. Review of the past PWR Appendix G violations that have occurred and which are described in License Event Reports, and identify those which are not credible in your plant because of equipment differences. Provide a description of the distinctions.
- d.
Describe~ in detail, how you are reducing the likelihood of the other remaining credible events. Furnish schematic diagrams, P & I diagrams and procedural stunmaries necessary to support the effecti¥eness and reliability of these measures.
- 2. The majority pf reported pressure transient events have occurred while the plants \\\\'ere operating iii a water solid condition.
POSITION:
The staff requires that plant operation in a water solid condition be minimized, or if possible eliminated. Accordingly, those operations during which the plant is in a water solid condition should be identified and fully justified.. Towards this enc}, you are requested to provide the following information:
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- a. A description of the procedures and e\\*olutions of situation$
- that require the plant to be maintained in a water solid condition. Also justify why a nitrogen, air, or steam bubble cannot be maintained in these situ3tions.
- b. Sufficient background or supplementary infonnation such as system diagran:is, procedure Sl.UTllnClries and descriptions of
- equipment operation to justify your need for operating the plant in a water solid condition.
- 3. The inadvertent operation of SIS components during cold shutdown conditions has been responsible for several overpressure events.
POSITIQ~:
Based on the licensee submi ttals, the November iI'.eeting, *and discussions with NSSS vendors, the staff will require the deenergizing of SIS pumps and closure of SI header/
discharge valves during cold shutdown operations. Those.
situations during 'Which this position cannot be met must be fully described an~ justified. Accordingly, provide* the following information:
- a.
- b.
c..
- d.
A schematic diagram of the SIS showing the flowpaths into the RCS, including an identification of the pumps and valves to be closed cu,d deenergized.
The head-flow characteristics of each of the SIS ptmips~
Your tiJJte schedule for implementing the administrative and operating procedural changes required to meet this position.
An identificat1on of all ciretmlStances for which the SIS
- pumps and valves may not be isolated and deenergized,, and**
a description of the marmer by.which SIS injection would be prevented in such circumstances.
- e. The location of the breakers that will be opened, and the piaces from which they can be controlled.
f9 A description of the position indication and status signals which will be lost, if any, as a result of deenergizing these components.
- g. A detailed description of the administrative pn1cedurcs which will be used to assure proper equipment alignment and the supervisory personnel responsible fo.r maintain~ng control.
- h. A description of the overall impact on plant operations if the nitrogen pressure on the acctunulator is routinely lo\\,*ered while in cold shutdol*m *
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- 4. Several Appendix G violations have occurred during component or system tests while in cold shutclO\\m conditions; therefore, provide the following infonnation:
- a. Identify those components or systems that are routinely tested while in a cold shutdown condition that could cause overpressure transients.
- b. Describe the measures being taken to prevent over-pressurization during these tests.
S. The staff believes that a high pressure alann *used during low RSC temperature operations is an effective means to attract the operator's attention to a high pressure transient.
POSITION:
The staff requires the installation of an over-pressure alann, as delineated in Item E.l. of Attachment 1 to License DPR-70 (Amendment No. 3). Accordingly, provide the following infonnation:
- a. The alarm setpoint and reset pressure including the means to adjust the setpoint.
- b. The alann aruum.ciators and system sensors used.
- c. A discussion of the means being used to *assure the alann's availability and operability during all
.water solid operations. Describe the means being used to minimize the alann' s downtime during *all other cold shutdown conditions, including back-up alanns.
- 6. The RHR (or SCS) is nonnally connected to thr RCS and operating when the plant is in cold shutdm*m. The inadvertent isolation of the RHR system while water solid has caused 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 P\\\\'R overpressurizations:.Accordingly, pro\\*ide the following infonnation:
- a.
RHR (ol* SCS) design pressure, and a description of the system isolation valves and their arrangement (e.g.,
number and configuration of valves installed,*and whether they are pneumatic or motor operated).
- b. Interlocks, interlock setpoints, and alaJl!ls associated with each isolation valve.
- c. Nominal stroke time of isolation valves.
- l.
- d. The setpoint and capacity of Rf-IR (or SCS) relief and safety valves.
- e. All pressure alanns 1 setpoin_ts and associated annunciation for the system.
- 7. Several provi~ions are mentioned by other licensees in order to reduce the likelihood of an overpressure event due to inprope1* RCP operation while shutdo\\m, (e.g., Attachment 1 to the October 14, 1976 letter for the Wisconsin Michigan Power Company's et al, Point Beach Nuclear Plant Units No.
1 and 2, IX:icket Nos. 50-266 and 50-301). Please address the following with regard to the proposed measures:
- a. When the RCS is in a cold condition, what is the Appendix G pressure limit and the minimum pressure for RCP operation? Furnish any curves that help explain these limits.
- b. l\\'hen the RCS is water solid due to the addition of the RO> discharge pressure 1 what measures d6 you take to prevent overpressure events?
- c. Describe the techniques you use to reduce or to eliminate the RCS temperature imbalance should you have to start a RCP in a solid and id.le flow system.
- d.
~scribe the instrumentation used to detennine if the RCS is isothennal and the criteria for defining an isothermal system.
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A DOCKET NOS.:
UNITED ST ATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 November 17, 1976 50-344, 50-213, 50-315, 50-244, 50-247, 50-286, 50-305, 50-266, 50-301, 50-282, 50-306, 50-261, 50-295, 50-304, 50-206, 50-280, 50-281, 50-250, 50-251, and 50-334.
- LICENSEE/FACILITY:
ROCHESTER GAS & ELECTRIC CORPORATION (R. E. GlNNA)
CONSOLIDATED EDISON COMPANY (INDIM: POINT UNITS 2 & 3)
DUQUESNE LIGHT COMPAY (BEAVER VALLEY UNIT 1)
CONNECTICUT YAUKEE ATOMIC POWER COMPANY (HADDAM NECK)
WISCONSIN PUBLIC SERVICE CORPORATION (KEWAUNEE)
WISCONSIN ELECTRIC POWER COMPANY (POINT BEACH UNITS 1 & 2)
NORTHERN STATES POWER COMPANY (PRAIRIE ISLAND UNITS 1 & 2}
CAROLINA POl-IER LIGHT COMPANY (H.B.ROBINSON)
SOUTHERN CALIFORNIA EDISON COMPANY (SAN ONOFRE)
VIRGINIA ELECTRIC & POHER COMPANY (SURRY UNITS 1 & 2)
PORTLAND GENERAL ELECTRIC COMPANY (TROJAN)
FLORIDA POWER & LIGHT COMPANY ( TURKEY POINT UNITS 1 & 2)
COMMONWEALTH EDISON COMPANY (ZION UNITS 1 & 2)
INDIANA & MICHIGAN POWER COMPANY ( D.C. COOK, UNIT 1)
SUMMARY
OF MEETING HELD ON NOVEMBER 4, 1976, CONCERNING PROPOSED MEASURES TO PREVENT REACTOR VESSEL OVERPRESSURIZATION IN OPERATING WESTINGHOUSE
_(PHR) FACILITIES
- On November 4, 1976, we met with representatives of PWR licensees with Westinghouse designed plants to discuss measures being taken to prevent reactor vessel overpressurization.
A list of attendees is attached.
Significant discussions are sunmarized below.
We summarized the correspondesnce and discussions that have occurred with the Westinghouse licensees since our generic letter on reactor vessel overpressurization was issued in August 1976.
We then identified the criteria listed below as those that should be applied in the design of equipment which is to prevent pressure transients that would exceed the limits of Appendix G to 10 CFR §50.
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0 Sui111i1ary of 11/4/76 November 17, 1976
- 1.
Credit of Operator Action - No credit can be taken for operator action until 10 minutes after the operator is aware that a pressure transient is in progress.
- 2.
Single Failure Criteria - The pressure protection system should be designed to protect the vessel given a single failure that initiates the pressure transient. In this area, redundant or diverse pressure protection systems would be considered as meeting the single failure criteria.
- 3.
Testability - The equipment design should include some provision for testing on a schedule consistent with the frequency that the system is used for pressure protection.
- 4.
Seismic Design and IEEE 279 Criteria - Ideally, the pressure protection system should meet both seismic Category I and IEEE 279 criteria. The basic objective, however, is that the system should not be vulnerable to an event which both causes a pressure transient and causes a failure of equipment needed to tenninate the transient.
Representatives of the task group of Westinghouse utilities formed to evaluate the problem of reactor vessel overpressurization provided a description of the steps they have taken to respond to the requirements set forth in our August 1976 letter. A summary was given of the various types of Thermal and mass input transients being considered, and it was indicated that a "bounding" analysis is being perfonned to consider the worst case situation for all Westinghouse plants. The preliminary results of the mass input analysis show that the pressurizer power relief valves have both the capacity and time response characteristics to limit the resultant pressure surges. The task group, however, indicated that a more detailed analysis would be necessary in the case of the pump-start or thermal type of transient before any similar determination could be made.
The detailed plant specific analyses are not scheduled for completion for about six months. Since the power operated relief has evidently been selected by the licensees as the means to limit pressure transients. we urged that efforts be made to begin orderin~ the necessary equipment now rather than waiting 6 months for the plant specific analyses results.
We also urged that the licensees concurrently investigate other factors such that installation times can be minimized.
The licensees' task group agreed to look into these matters *
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.*.. November 17, 1976 We indicated that the "single failure criteria being assumed by the licensees was not consistent with the conventional single failure criteria required by the staff. The licensee agreed to examine this
- area further and to provide justification for any deviations from the
- conventional singl~ failure criteria. This infonnation as well as a discussion of the various administrative measures that the licensees intend to use to prevent pressure transients while shutdown are to be prepared with a target date of submittal to the staff by December 3,. 1976.
Attachment:
List of Attendees
~~~er Operating Reactors Branch #1 Division of Operating Reactors
i I
I
~
NRC STAFF MEETING WITH WESTINGHOUSE PHR LICENSEES NOVEMBER 4, 1976 NRC G.G.Zech R.L.Baer G.Lanik LB.Marsh J.E. Ouzts T.J.Carter I.Villalva W.A.Paulson S. lsrael J.Mazetis D.Tibbitts D. Neighbors W.T.Russell
.B.Fairtile
,J.Cook
.B.Erickson P.E.Harmon J.Stosnider R.M.Gamble R.Clememson M.Grotenhuis D.Hood W.Pike R.W.Klecker J.Wetmore M.Fietcher T.Wambach CAROL I NA Pm/ER & l l GHT ATIENDANCE LIST SOUTHERN CALIFORNIA EDISON W.C.Moody NUSCO B. Ilberman M. Kupinski P.F.Santoro CONNECTICUT YANKEE ATOMIC POWER CO.
J.Levine DUKE POWER CO.
R. W. Revels E. M. Geddie, Jr.
VIRGIN1A ELECTRIC & POWER CO.
D. *w. Speidel 1,.Jr.
A. L. Hogg, Jr.
ALABAMA POWER CO.
J. R. Campbell COMMONWEALTH EDISON CO.
co.
T.R.Trarran
~
E.E.O'Brien D,B.Waters W.A.Wogsland R.G.Black CONSOLIDATED EDISON CO.
C.W.Jackson P.M.Pivawer J.Makepeace POWER AUTHORITY OF THE STATE OF NEW YORK P.F.Altern J.M. Vargas WISCONSIN ELECTRIC POWER CO.
R. A. Newton WISCONSIN PUBLIC SERVICE CORF M.E.STERN ROCHESTER GAS & ELCTRIC R.W.Elias B.A.Snow AMERICAN ELECTRIC POWER P.W. Daley R.L. Shoberg J.G;~Dell Perico DQUESNE LIGHT COMPANY S. R. Porter J. J. Carey
.. - -. 1
\\.._....
/_-.:~--*
Attendance list FLORIDA POWER & LIGHT C.S.Pillar M.A.Schopmann WESTINGHOUSE - PWRSD W. G. Poulson F. Gilgliotti H. Gutzman A. M. Sklencar R. W. Fleming R. C. Jenkner NOTHERN STATES POWER CO *
- l. L. Tayl o~
PUBLIC SERVICE E&G CO.
L. K.Mil 1 er SOUTHERN COMPANY SERVICES,INC.
J. R. Lyons PORTLAND GENERAL ELECTRIC CO.
D. I. Herbon SNUPPS F. Schwoerer
- ~~------*-
I
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~ f. t I
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i t
APPENDIX B Design Drawings Unit #1 and #2
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LO I lt>G I~EOM ALL lTOM$ OELOW L0<.. 1ED IN AUX *
... ct--~---e-----,......---~
No. II PRESSURIZER POWER RELIEF VAL\\/.E IPRI
- I CONTR.OLC... 8.A~SE:t..ABl."t No.14 "iSA IPR/
LOSS OF AUX'. AIR SUPPLY'(<:.90PS:I)
O.H ALARM 8 (WINDOW K*Z4}
~AUTO BLOCK O.H* ALARM
. (WINDOW E*~Gi.)
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(2)
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No. II PRES:ioURIZER RELIE'F" STOP VAL.VI::
IPIC.G PC+S~E rnJ§)E~ERW, ~---------------------------------~~
O~ PZR. HI PRE:'5S.
IA 2sov. MCC DWG*20!.101-A* 8155
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- 24110\\0*B*'"I O.H. ALAR.M owi;;.* 24110~-~-ei'-G.I
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~~/X-CO""T"'C.TOR CLOS~O COIL AUX. R~LA.Y l'C45G.E.X* PZR..HI PRE.SS.AU,._ r::tELA"'1 (AR.440AR) 1:*c:457 E'X* PZll MIM. PRE.S-,. AUX.QELA"'I' (AA.440AR)
PD-985~ -
LOSS OF CONTROL. AIR PR:i'SS sw(CtOSI! AT <.90PSJ')
l'D*,862. - l OSSO~ AU.IC CO"'TROL. AIR PRESS ALAR:M SW.
C::::J *-LOCATION INDEX ABBREVIATIOH5 ~WIRl~GDIA<;RAMS:
v*- Cit.I RESPEC.TIVE:
VALVf'.. ____ OWCA*'2.0~4,._4-A* 1'1'!:11.
L. -
'llCIN ITY OF VALVE: ___ -
2.034-c;.4-A* i'Z.!>1.
R ZC.* PROC: G,R.?, '2 R.AClo( 2.0 _
2.20 I ~4* 5*°"54&
R"l:l-PROC G~P.I RACK'Z.1.:.___
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li131~** AUX. RELA."1' RACK se -
220234f..'!Gt*A*8"'4~
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2204-11-A* 1::.20
,4,~ANMUN* CAl!I. 117 (f"ROMl)_ _
U2 1ll<J*8-~5S4
-"! **ANNIJN CA8. l/5 (REAR)__
ZZ2764*8*j55f
_!l!'°J:ERENCE DRAWING~
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_OWC,*'2!.l!.SCO<;_S7-&-:Go1 ~'2'41.SOei*B-"'-lS
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PUBLIC SERVICE ELECTRIC AND GAS COMPANY E~GIHEERING DEPARTMENT NEWARK. N. J.
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