ML19351F525
| ML19351F525 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 01/02/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Abel J COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8101130426 | |
| Download: ML19351F525 (23) | |
Text
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UNifED STATES
[ } ',, fq 3 'j NUCLEAR REGULATORY COMMISSION E
WASHINcTON,0. C. 20555
.h JAN 2 1981 D'_%f
.2 Occket Nos: STN 50-454/455 ES STN 50-456/457
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Mr. J. S. Abel Director of Nuclear Licensing Q;
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Commonwealth Edison Company 7A zq
.P. O. Box 767 5
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Chicago, Illinois 60690 y
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Dear Mr. Abel:
Subject:
First Round Questicns on the Byron and Braidwood OL Application In our review of your application for operating licenses for the Byron Station, Units 1 and 2, and the Praidwood Station, Units 1 and 2, we have identified a need for additional information which we require to complete our review. The specific requests contained in the enclosure to this letter are the ninth set of our round one questions and cover some of the areas of our review performed by (1) the Instrumentation and Control Systems Branch and (2) the Mechanical Engineering Branch.
Please contact us if you desire any discussion or clarification of the enclosed requests.
Sincerely, l-t.
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- Rcbert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
As stated l
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I Mr. J. 5 Abel Director of Nuclear Licensing Commonwealth Edison Company Fost Office Box 767 Chicago, Illinois 60690 ccs:
Mr. William Kortier Mr. Edward R. Crass Atomic Power Distribution Nuclear Safeguards and Licensing Division Westinghouse Electric Corporation Sargent & Lundy Engineers P. O. Box 355 55 East Monroe Street Pittsburgh, Pennsylvania 15230 Chicago, Illinois 60603 Paul M. Murphy, Esq.
Nuclear Regulatory Commission, Region III Isham, Lincoln & Beale Office of Inspection and Enforcement One First National Plaza 799 Roosevelt Road 42nd Floor Glen Ellyn, Illinois 60137 Chicago, Illinois 60603 Myron Cherry, Esq.
Mrs. Phillip B. Johnson Cherry, Flynn and Kanter 1907 Stratford Lane 1 IBM Plaza, Suite 4501 Rockford, Illinois 61107 Chicago, Illinois 60611 Ms. Julianne Mahler Marshall E. Miller, Esq., Chairman Center for Governmental Studies Atomic Safety and Licensing Northern Illinois University Board Panel DeKalb, Illinois 60115 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 C. Allen Bock, Esq.
P. O. Box 342 Dr. A. Dixon Callihan Urbanan, Illinois 61820 Union Carbide Corporation P. O. Box Y Thomas J. Gordon, Esq.
Oak Ridge, Tennessee 37830 Waaler, Evans & Gordon 2503 S. Neil Dr. Richard F. Cole Champaign, Illinois 61826 Atomic Safety and Licensing Board Panel Ms. Bridget Little Rorem U. S. Nuclear Regulatory Commission Appleseed Coordinator Washington, D. C.
20555 117 North Linden Street Essex, Illinois 60935 Kenneth F. Levin, Esq.
Beatty, Levin, Holland, Basofin & Sarsany 11 South LaSalle Street i
Suite-2200 l
Chicago, Illinois 60603
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",5 ENCLOSURE 031!6 Perform an audit of Table 7.1-1 and i= prove the references listed under 031.19 "Conformance Discussed In".
(Refer to question 031.13 to avoid possible du-plication.) The following types of difficulties are noted:
A non-existent FSAR Section, 7.2.1.11, is cited for GDC-2.
a.
Section 7.1.2.2.3 is correct for GDC-3, however, Section 7.1.2.2.3 b.
erroneously refers to reference 3.
Many references are too broadly stated to be of value, e.g., 7.2, c.
7.3, Technical Specifications. Reference to individual paragraphs, using as many separcte references as needed.
Many control, safety-related, and safety-related support systems have 031.20 been left out of Chapter 7.0, as observed by comparison to similar Perform an audit of Chapter 7.and amend the FSAR to include all plants.
ISC systens applicable P. safety-related syste=s and their respective support systems. As. ainimum, list the missing systems and reference to appropriate FSAR discussions for those systems adequately treated else-where in the FSAR. This request for information is to be applied partic-ularly to Balance-of-Plant systems and to FSAR Sections 7.3 and 7.7.
(Refer to questions 030.11 and 030.16 to avoid possible duplication of effort).
i 031.21 An I&C LER for the Sequoyah station (LER No. 79-
/02L-0) refers to an undetectable failure in the P-4 interlock. Th'e LER description states that there is no testing procedure for this circuit. Yet FSAR Section 7.2.2.2.3(j) states that the "P" interlocks =eet the testing requirements of IEEE Standards 279-1971 and 338-1971.
If these interlocks are Verify that the "P" interlocks are tested.
not tested, provide a list of non-tested interlocks and related circuits
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031.7 i
and either justify the lack of test or define your plans to provide for tests in the future.
031.22 Safety-related support systems include HVAC syste=s.
Define the need for ventilation and cooling in terms of possible loss of availability of the supported safty-related system. That is, how long must diesel-gener-ators, RHR pu=ps, SI pumps, switchgear rooms, transformers, etc., func-tion on loss of cooling or ventilation in order to perform the required safety function. In which cases, if any, will loss of the HVAC function, as caused by I&C failure, result in loss of the required safety-related function.
031.23 Describe the arrangement of instrumentation and controls, with associated power sources, used to assure that the positions of BTP 10-1 (Auxiliary Systems Branch) will be satisfied in regard to auxiliary feedwater.
031.24 Explain why the PAMS variables listed in Table 7.5-1 were chosen to the exclusion of other possible variables, such as a.
Containment Sump Temperature e.
Containment Hy4rogen Cas Concentration b.
Pressurizer Water Temperature-f.
Control Room Radiation Level Essential Cooling Water Flow g.
Spent Fuel Pool Ventilation c.
Exhaust Radiation Level d.
Containment Radiation Level These variables have been chosen as PAMS at reactor stations similar to Byron and Braidwood.
031.25 The main steam isolation valves (MSIVs) have a two-train activation fea-If State whether any additional systems have two-train actuation.
ture.
Please include so, provide a list and a description of each such system.
schematic diagrams.
031.26 Provide an explanation or justification for the classification of the radiation detectors used for containment isolation. The FSAR, Figure 7.2-1, Sheet 8, indicates that these detectors are not safety-related.
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031.27 Explain how operation is intended to proceed on one unit while the other unit' remains under construction.
In your response, emphasize the instru-mentation and control of common systems, such as the control room, compo-nent cooling water, essential water, and possible e=ergency features, such as the DC bus switchover.
031.28 Describe hcw access to the remote shutdewn panels will be limited.
In this cennection, resolve the differences in layout between drawing M-9 and Fire Protection Report Figure 2.4-4A in the remote shutdown panel area.
Support your description with appropriate layout drawings.
t 031.29 Please locate more precisely in the FSAR the discussion of alignment functions referenced to Chapter 6.0 and Section 7.3, as stated on page 7.4-1.
Provide such additional discussion as may be needed for complete-ness.
031.30 Improve and insert the analyses defined by RG 1.70 for FSAR Section 7.4, especially in regard to loss of cooling water to vital equipment.
031.31 Provide explanations of how instrumentation is duplicated on remote shut-down panels and in other local panels so as to:
Maintain separation and isolation of redundant channels, a.
b.
Assure access to appropriate controls at either location in the event of emergencies.
c.
Prevent undue loss of reliability.
Supplement these explanations with appropriate schematic, layout, and viring diagrams.
031.32 Provide descriptions of how the Nuclear Rate trips are tested.
031.33 Provide descriptions of how the Reactor Coolant Pump undervoltage and ur. der frequency trips are tested.
031.9 031.34 Figure 7.7-11 indicates that Control Rod Bank D has 4 rods in group 1 and 4 rods in group 2, yet Figure 4.3-36 indicates 5 rods in bank D.
Please reconcile this difference.
031.35 FSAR Section 15.4.1.1 limits the accident study to the withdrawal of 2 banks of rods. However, the discussion in Section 7.7.1.2(c) implies that 4 control banks may be in si=ultaneous motion. Please resolve this possible contradiction.
031.36 In Section 7.2.1.1.2, it is stated that the manual trip switch re= oves the voltage frem the undervoltage trip coil and energizes the shunt trip coil. Distinguish between direct wiring to the respective trip coils and viring to trip logic, which in turn may activate or de-activate the trip coils.
Go on to explain how this design satisfies the single failure cri-terion and the separation requirements for redun' dant trains.
Provide appropriate sche =atic, viring and layout drawings to enable our verification of the above design.
031.37 Provide schematic drawings to show the interrelationship between ESF ac-t tivation of the emergency diesel generators and the loss-of power activa-
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tion.
l 031.38 Our position, taken during the Construction Permit review in regard to motor-operated valves, in the auxiliary feedwater system, was that these
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valves must have power removed during. operation,.while in the open posi-tion.
If your design has changed materially since issuance of the Construc-tion Permit, justify the use of these valves without power lockout.
Otherwise, identify these valves, add them to your response to qtection 040.12, and list them accordingly in the Technical Specifications.
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110.18 Mechanical _Engi_neering Branch 110.63 Due to a long history of problems dealing with inoperable and incorrectly installed snubbers, ard due to the potential safety significance of failed snubbers in safety relvted systems and components, it is requested that maintenance records for snubbers be documented as follows:
Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7 ab of Standard Technical Specifications 3/4.7.9 This exami-nation should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a mimimum verify the following:
(1) There are no visible signs of damage or impaired operability as a result of storage, handling, or installation.
(2) The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifictions.
(3) Snubbers are not seized, frozen or jsmmed.
(4) Adequate swing clearance is provided to allow snubber movement.
(5)
If applicable, fluid is to the recommended level and is not leaking from the snubber system.
(6) Structural connections such as pins, fasteners and other connecting hardware such as lock nuts, tabs, wire, cotter pins are installed correctly.
If the period between the-initial pre-service examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1,4, and 5 shall be performed. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must l
be repaired or replaced and re-examined in accordance with the above criteria.
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Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250* F should be verified as follows:
(a)
During initial system heatup and cooldown, at specified temperature l
l intervals for any system which attains operating temperature, verify j
the snubber expected thermal movement.
(b)
For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate l
the projected thermal movement.
(c) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
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110.19 i
The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test j
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j programs.
1 The pre-service inspection must be a prerequisite for the pre-cperational testing of snubber thermal motion. This test program should be specified in Chapter 14 of the FSAR.
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.110.64 10 CFR 50.55a has recently been revised with respect to pump and valve inservice testing requirements (See October 9,1979 Federal Register, pp. 57912-4)
Frovide a program for initial 120 cnth inservice testing of pumps and valves, as required by 10 CFR 50.55a(g)(4)(i). The applicable code for this inspection interval which would be required by 10 CFR 50.55a(g)(4)(1) is the Code endorsed by 10 CFR 50.55a(b)(2) 12 monthsprior to the date of issuance of your OL. Effective November 1,1979,10 CFR 50.55a(b)(2) endorsed the 1977 Edition with all agenda through Summer 1978. We therefor =
recomend that your progra'm be based on the 1977 Edition with all agenda through Sumer 1978. Your program should indicate which code requirements are idpractical to meet together with documentation for justification.why relief is necessary.
l The attached format should be used when submitting IST program.
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110.21 NRC STAFF COMMENTS ON INSERVICE PUMP AND VALVE TESTING PROGRAMS AND RELIEF REQUESTS The NRC staff, after reviewing a number of pump and valve testing programs, has determined that further guidance might be helpful to illustrate the type and extent of information we feel is necessary to expedite the We feel that the Licensee can, by incorporating review of these prograts.
these guidelines into each program submittal, reduce considerably the staff's review time and time spent by the Licensee in responding to NRC staff requests for additional informatiin.
The pump testing program should include all safety ralated* Class 1, 2, and 3 pumps whie.h are installed in water cooled nuclear pcwer plants and which are provided with an emergency power source.
The valve testing program should include all the safety related valves in the following systers excluding valves used for operating convenience only, such as manual vent, drain, instrument, and test valves, and valves used for maintenance only.
.P3R, a.
High Pressure Injection System b.
Low Pressure Injection System c.
Accumulator Systems d.
Containment Spray System
- Safety related - necessary to safely shut down the plant and mitigate the consequences of an accident.
7 110.22 e.
Primary and Secondary System Safety and Relief Valves f.
Auxiliary Feedwater Systems 9
Reactor Building Cooling System h.
Active Components in Service Water and Instrument Air Systems which are required to support safety system functions.
- i. Containmer. Isolation Valves required to change position to isolate containm2nt.
j.
Chemical & Volume Control System k.
Other key components in Auxiliary S stems which are required to directly support plant shutdown or safety system function.
1.
Residual Heat Removal System m.
High Pressure Core Injection System b.
Low Pressure Core Injection System c.
Residual Heat Removal System (Shutdown Cooling System) d.
Emergency Condenser System (Isolation Condenser System) e.
Low Pressure Core Spray System f.
Containment Spray System g.
Safety, Relief, and Safety / Relief Valves h.
RCIC (Reactor Core Isolation Cooling) System
- i. Containment Cooling System
- j. Containment isolation valves required to change position to isolate containment.
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110.23 Standby liquid control system (Boron System) k.
Autodatic Depressurization System (any pilot or control valves, associated 1.
hydraulic or pneumatic systems, etc.)
Control Rod Drive Hydraulic System (" Scram" function) m.
other key components in Auxiliary Systems which are required to directly n.
support plant shutdown or safety system function.
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Reactor Coolant System Inservice Pumo and Valve Testing Program Information required for NRC Staff Review of the Pump and Valve Testing I.
Program Three sets of P&ID's, which include all of the systems listed A.
above, with the code class and :ystem boundaries clearly marked.
The drawings should include all of the components present at the time of submittal and a legend of the P&I3 symbols.
Identification of the applicable ASME Code Edition ar.d Addenda B.
The period for which the program is applicable.
C.
D.
Identify the component code class.
E.
For Pump testing:
Identify Each pump required to be tested (name and number) 1.
2.
The test parameters to be measured 3.
The test frequency e
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110.24 F.
For valve testing:
Identify Each valve in ASME Section XI Categories A & B that will 1.
be exercised every three months during normal plant operation (indicate whether partial or full stroke exercise, and for power operated valves list the limiting value for stroke tine.)
Each valve in ASME Section XI Category A that will be leak 2.
tested during refueling outages (Indicate the leak test procedure you intend to usel Each valve in ASME Section XI Categories C, D, and E that 3.
will be tested, the type of test and the test frequency.
For check valves, identify those that will be exercised -
every 3 months and those that will only be exercised during cold shutdcwn or refueling outages.
Additional Information that will be Helpful in Speeding Up the Review II.
Process Include the valve location coordinates or other appropriate A.
location information which will expedite our locating the I
valves on the P& ids.
Provide P&ID drawings that are large and clear enough to be B.
read easily.
Identify valves tht are provided with an interlock to other I
C.
components and a brief description of that function.
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110.25 Relief Recuests frcm Section XI Recuirements The largest area of concern for the NRC staff, in the review of an inservice valve and pump testing progran, is in evaluating the basis for justifying relief from Section XI Requirements.
It has been our experience that many requeits for relief, submitted in these programs, do not provide adequate descriptive and detailed technical information. This explicit information is necessary to provide reasonable assurance that the burden imposed on the licensee in complying with tne code requirements is not justified by the increased level of safety obtained.
Relief requests which are submitted with a justification such as
" Impractical", " Inaccessible", or any other categorical basis, will require additional information, as illustrated in-the enclosed examples, to allow The intention our staff to make an evaluation of that relief request.
of this guidance is to illustrate the content and extent of information required by the NRC staff, in the request for relief, to make a proper evaluation and adequately document the basis for that relief in our safety The NRC staff feels that by receiving this information evaluatf or report.
in the program submittal, subsequent requests for additional information and delays in completing our review can be considerably reduced or eliminated.
I.
Information Recuired for NRC Review of Relief Recuests A.
Identify component for which relief is requested:
1.
Name and number as given in FSAR 2.
Function 3.
ASME Section III Code Class 4.
For valve testing, also specify the ASME Section XI valve category as defined in IWV-2000 n
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110.26' Specifically identify the ASME Code requirement that has been B.
determined to be impractical for each component.
Provide information to support the determination that the C.
requirement in (B) is impractical.; i.e., state and explain the basis for requesting relief.
Specify the inservice testing that will be performed in lieu D.
of the ASME Code Section XI requirements.
Provide the schedule for implementation of the procedure (s)
E.
in (D).
Examples to-Illustrate Several fossible-Areas '4here Relief May Be II.
Granted and the Extent and Content of Information Necessary to Make An Evaluation A.
Accessibility: The regulation specifically grants relief from the code requirement because of insufficient access pro-visions. However, a detailed discussion of actual physical arrangement of the component in question to illustrate the insufficiency of space for conducting the required test is necessary.
Discuss in detail the physical arrangement of the component in question to demonstrate that there is not sufficient space to perform the code required inservice testing.
110.27 What alternative surveillance means which will provide an acceptable level of safety have you considered and why are these means not feasible?
8.
Environmental Conditions (e.g., High radiation level, High temperature, High humidity, etc.)
Although it is prudent to maintain occupation radiation exposura for inspection personnel as low as practicable, the request for relief from the code requirenents cannot be granted solely on the basis of high radiation levels alone. A balanced judgment between the hardships and compensating increase in the level of safety should be caref211y established.--if-the health-and.
safety of the public dictates the necessity of inservice testing, alternative means or even decontamination of the plant if necessary should be provided or developed.
Provide additional information regarding the radiation levels l
at the required test location. What alternative testing techniques which will provide an acceptable level of assurance of the integrity of the component in question have you considered and why are these techniques determined to be impractical?
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110.28 C.
Instrumentation is not originally provided Provide infomation to justify that compliance with the code requirements would result in undue burden or hardships withuut What a compensating increase in the level of plant safety.
alternative testing methods which will provide. an acceptable level of safety have you considered and why are these methods detemined to be impractical?
Valve Cycling During Plant Operation Could Put the Plant in D.
an Unsafe Condition The licensee should explain in detail why exercising tests during plant operation could jeopardize the plant safety.
E.
Valve Testing at Cold Shutdown or Refueling Intervals in Lieu
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of the 3 Month Required Iriterval The licensee should explain in detail why each valve cannot be exercised during normal operation. Also, for the valves where a refueling interval is indicated, explain in detail why each valve cannot be exercised during cold shutdown intervals.
III. Acceotance Criteria for Relief Reauest The Licensee must sucessfully demonstrate that:
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Compliance with the code requirements would result in t
hardships or unusual difficulties without a compensating I
increase in the level of safety and noncompliance will provide an acceptable level of quality and safety, or 1
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110.29 2.
Proposed alternatives to the code requirements or portions thereof will provide an acceptable level of quality and
, safety.
Standard Format A standard fomat, for the valve portion of the pump and valve testing program and relief requests, is included as an attachment to this Guidance.
j The NRC staff believes that this standard fomat will reduce the time spent by both the staff in our review and by 'the licensee in their preparation of the pump and valve testing program and submittals. The standard format includes examples of relief requests which are intended to illustrate the application of the staadard fomat and are not necessarily a specific plant relief request.
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110.30 ATTAC! DENT STANDARD FORMAT VALVE INSERVICE TESTING PROGRAM SUBMITTAL t
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110.32 Legend for Valve Testing Exanole Format Q - Exercise valve (full strcke) for operability every (3) months LT - Valves are leak tested per Section XI Article IWV-3420 MT - Strcke time measurements are taken and cercared to the stroke time limiting value per Section XI Article IWV 310 CV - Exercise check valves to the position required to fulfill their function every (3).tonths SRV - Safety and relief valves are test'ed per Section XI Article IWV-3510 DT - Test category D valves per Section XI Article IWV-3600 ET - Verify and record valve position before operations are performed and after operations are completed, and verify that valve is locked or sealed.
CS - Exercise valve for operabilit)F every cold shutdown RR - Exercise valve for operability every reactor refueling 1
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110.33 Relief Recuest Basis System: Auxiliary Coolant System, Component Cooling 1.
Valve:
71 7 Category:
C Class:
3 Function:
pre ent backflow frc the reactor coolant pump coolinc coils Impractical test requirement: Exercise valve for operability every three months Basis for relief: To test this valve would require interruption of cooling water to the reactor ({inant pumps motor cooling coils. This action could result in damage to the reactor coolant pumps and thus place the plant.16 an unsafe mode of operation.
Alternative This valve will be exercised for operability Testing:
during cold shutdowns.
2.
Valve:
834 Category:
B-E Class:
3 Function:
Isolate the primary water from the component cooling surge tank during p1 nt opertion.
It is normally in the closed position, but routine operation of this valve will occur during refueling and cold shutdowns.
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Impractical Test Exercise valve (full stroke) for operability Requirement:
every three (3) months.
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110.34 Basis for Relief: This valve is not required to change position during plant operation to accomplish its safety function. Exercising this valve will increasa the possibility of surge tar.i line contamination.
Alternate Verify and re:ord valve positien before and Testing:
after each valve operation.
3.
Valve:
7448 Category:
A Class:
2 Function:
Isolate the residual heat exchangers from the cold leg R.C.S. backflow and accumulator backflow.
Test Requirements:
Seat leakage test This vaIve is located in a high radiation field Basis for Relief:
(2000 mr/hr) which would make the required seat leakage test hazardous to test personnel. We intend to seat leak test two other valves (875B and 876B) which are in series with this valve and will also prevent backflow. We feel that by complying with the seat leakage requirements we will not achieve a compensatory increase in the level of safety.
Alternative No alternative seat leak testing is proposed.
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