ML20041C429
| ML20041C429 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/09/1982 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Bauer E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| References | |
| REF-GTECI-A-01, REF-GTECI-A-11, REF-GTECI-A-17, REF-GTECI-A-40, REF-GTECI-A-43, REF-GTECI-A-44, REF-GTECI-A-45, REF-GTECI-A-46, REF-GTECI-A-47, REF-GTECI-A-48, REF-GTECI-CO, REF-GTECI-DC, REF-GTECI-EL, REF-GTECI-ES, REF-GTECI-PI, REF-GTECI-RV, REF-GTECI-SC, REF-GTECI-SY, TASK-A-01, TASK-A-1, TASK-A-11, TASK-A-17, TASK-A-40, TASK-A-43, TASK-A-44, TASK-A-45, TASK-A-46, TASK-A-47, TASK-A-48, TASK-OR NUDOCS 8203010372 | |
| Download: ML20041C429 (30) | |
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DISTRIBUTION:
Docket File bec: TERA FEB 9 1332 LB#2 File NSIC DEisenhut/RPurple NRC PDR RTedesco L PDR ASchwencer ACRS (16)
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y'2 OI&E "RO Region I Mr. Edward G. Bauer, Jr.
G Vice President & General CoJns.it:FEgl 219sas I Resident Inspector 9' >,c _g 7 SHanauer Philadelphia Electric Compan ze RMattson E"
2301 Parket Street
's HThompson Philadelphia, Pennsylvania 1910l[
RVollmer
Dear Mr. Bauer:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - LIMERICK:
UNRESOLVED SAFETY ISSUES The NRC staff is currently reviewing the portion of your application which addresses Unresolved Safety Issues. The review has indicated a need for certain additional infomation, as delineated in Enclosure 1.
To assist you in your response, we have also enclosed Enclosure 2, a section of the SER for the Clinton BWR Plant.
We would like to meet with you to discuss this information request some-time in early March at our Bethesda offices.
Any questions concerning the above should be directed to Dr. Harvey Abelson, Licensing Project Manager, at (301) 492-9774 Sincerely,-
A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing
Enclosures:
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Mr. Edward G. Bauer, Jr.
Vice President & General Counsel Philadel)hia Electric Company 2301 Mar cet Street Philadelphia, Pennsylvania 19101 cc: Troy B. Conner, Jr., Esquire Charles W. Elliott, Esquire Conner & Wetterhahn Thomas & Hair 1747 Pennsylvania Avenue, NW 123 North Fifth Street Washington, D.C.
20006 Allentown, PA.18102 Mr. Robert W. Adler Judith A. Dorsey, Esquire Assistant Counsel Limerick Ecology Action Commonwealth of Pennsylvania, DER 1315 Walnut Street, Suite 1632 505 Executive' House Philadelphia, PA 19107 P.O. Box 2357 Harrisburg, PA 17120 Mr. Karl Abraham Public Affairs Officer Honorable Lawrence Coughlin Region I House of Representatives U.S. Nuclear Regulatory Commission Congress of the United States 631 Park Avenue Washington, D.C.
20515 King of Prussia, PA 19806 Roger B. Reynolds, Jr., Esquire Mr. Jacque Durr 324 Swede Street Resident Inspector Norristown, PA 19401 U.S. Nuclear Regulatory Commission P.O. Box 47 Joseph A. Smyth Sanatoga, PA 19464 Assistant County. Solicitor County of Montgomery James M. Neill, Esquire Courthouse Associate Counsel for Del-Aware Norristown, PA 19404 Box 511 Dublin, PA 18917 Eugene J. Bradley Philadelphia Electric Company Joseph H. White III Associate General Counsel 11 South Merion Avenue 2301 Market Street Bryn Mawr, PA 19010 Philadelphia, PA 19101 Dr. Judith H. Johnsrud Mr. Vincent Boyer Co-Director Senior Vicc President Environmental Coalition on Nuclear Operations Nuclear Power Philadelphia Electric Company 433 Orlando Avenue 2301 Market Street State College, PA 16801 Philadelphia, PA 19101 Thomas Gerusky, Director Mr. Marvin I. Lewis Bureau of Radiation Protection 6504'Bradford TeFrace Dept. of Environmental Resources
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Philadelphia, PA 19149 5th Floor', Fulton Bank Bldg.
Third & Locust Streets Frank R. Romano, Chairman Harrisburg, PA 17120 Air A Water Pollution Patrol 61 Forest Avenue Director, Pennsylvania Emergency Ambler, PA 19002 Management Agency Basement, Transportation &
Safety Building Harrisburg,.PA 17120
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John Shniper Meeting House Law Bldg. & Gallery Lawrence Brenner, Esq., Chairman
- Mennonite Church Road -
Administrative Judge U.S. Nuclear Regulatory Commission Schuylkill Road (Rt. 724)
Washington, D.C.
20555 Spring' City, PA 19475 Dr. Richard F. Cole
- Robert L. Anthony Friends of the Earth of the Administrative Judge Delaware Valley U.S. Nuclear Regulatory Commission 103 Vernon Lane, Box 186 Washington, D.C.
20555 Moylan, PA 19065 Dr. Peter A. Morris
- Alan J. Nogee Administrative Judge The Keystone Alliance U.S. Nuclear Regulatory Commission 3700 Chestnut Street Washington, D.C.
20555 Philadelphia, PA 19104 W. Wilson Goode Managing Ofrector City of Philadelphia Philadelphia, PA 19107 William A. Lochstet 119 E. Aaron Drive State College, PA 16801 Walter W. Cohen Consumer Advocate Office of Attorney General 1425 Strawberry Square Harrisburg, PA 17120 Steven P. Hershey, Esquire Consumers' Eoucation & Protective Association Sylvania House Juniper & Locust Streets Philadelphia, PA 19107 Sugarman & Denworth Suite 510 North American Building 121 South. Broad Street.
Philadelphia, PA 19107
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. Donald S. Sronstein', Esquire The National Lawyers Guild I
' Third Floor l
1425 Walnut Street Philadel phia, PA 19102 l
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REQUEST FOR INFORMATION The Atanic' Safety and. Licensing Appeal Board in ALAB-444 determined.that.
the Safety Evaluation Report for each plant should contain an assessment of each signif.icant unresolved generic saf.ety question.
It is the staff's view that the generic issues identified as " Unresolved Safety Issues" (NOREG-0606) are the substantive safety issues referred to by the Appeal Board.
Accordingly, we are requesting that you provide your justification for pennitting plant operation pending resolution of these s
issues. This should include a description of any interim. measures in tenns of design or operating procedures or investigative programs that are.being pursued to address these concerns.
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The justification should provide an overall summary of your position on each issue rather than a reference to various sections of the SAR where related information is presented.
There are currently a total of 26 Unresolved Safety Issues.. We do not require infomation fran you at this time for a number of the issues Since a number of the issues do not apply to your type of reactor, or because a generic resolution has been issued.
information noted above for each of the issues listed below:However, we do request the 1.
Waterhammer (A-1) 2.
Reactor Vessel Materials Toughness (A-11) 3.
Systems Interaction in Nuclear Power Plants (A-17) 4.
Seismic Design Criteria (A-40)
'C'ontainment Emergency Sump Reliability (A-43) 5.
6.
St.ation Blackout (A-44) 7.
Shutdown Decay Heat Removal Requirements (A-45) 8.
Seismic Qualification of Equipment in Operatin Safety Implications of Control Systems (A-47) g Plants (A-46) 9.
10.
Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (A-48) e m
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p CLINTON APPENDIX C U. S. NUCI. EAR REGULATORY CO?NISSION (NRC)
UNRESOLVED SAFETY ISSUES C.1 Introduction The Nuclear Regulatory Commission (NRC) staff continuously evaluates the safety requirements used in its reviews against new information as it becomes available.
Information related to the safety of nuclear 1: Twer plants comes from a variety of sources including experience from operating reactors; research results; NRC staff and Advisory Committee on Reactor Safeguards (ACRS) safety reviews; and vendor, architect / engineer, and utility design reviews.
Each time a new concern or safety issue is identified from one'or more of these sources, the need for immediate action to ensure safe operation is assessed.
This assessment includes considfration of the generic implications of the issue.
In some cases, immediate action is taken to ensure safety, for example, the derating of boiling water reactors as a result of the channel box wear problems in 1975.
In other cases, interim measures, such 'as modifications to operating procedures, may be sufficient to allow further study of the issue before making -
licensing decisions.
In most cases, however, the initial assessment indicates that immediate licensing actions or changes in licensing criteria are not neces-sary.
In any event, further study may be deemed appropriate to make judgments as to whether existing NRC staff requirements should b'e modified to address the issue for new plants or if backfitting is appropriate for the long-term opera-tion of plants already under construction or in operation.
l These issues are sometimes called generic safety issues because they are related to a particular class or type of nuclear facility rather than to a specific plant.
Certain of these issues have been designated as unresolved safety issues in NUREG-0410, "NRC Program for the Resolution of Generic Issues Related to 01/21/82 C-1 CLINTON SER APP C
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l Nuclear Power Plants," dated January 1, 1978.
However, as discussed above,
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such issues are considered on a generic basis only after the staff has made an initial determination that the safety significance of the issue does not prohibit l
continued operation or require licensing actions while the longer term generic review ~is under way.
i C.2 ALAB-444 Requirements These longer term generic studies were the subject of a decision by the Atomic Safety and Licensing Appeal Board of the Nuclear Regulatory Commission.
The decision (ALAB-444) was issued on November 23, 1977 in connection with the Appeaf Board's consideration of the Gulf States Utility Company application for the River Bend Station, Unit Nos. 1 and 2.
In short, the board (and the public as well) should be in a position to ascertain from the SER f tself--without the need to resort to extrinsic documents--the staff's perception of the nature and extent of the relationship between each significant unresolved generic safety question and the eventual operation of the reactor under scrutiny.
Once again, this assessment might well have a direct bearing upon the ability of the licensing board to make the safety findings required of it on the construction permit level even though the generic answer i
to the question remains in the offing.
Among other things, the furnished information would likely shed light on such alternatively important considerations as whether:
(1) the problem has already been resolved for the reactor under study; (2) there is a reasonable basis for concluding that a satisfactory solution will be obtained before the reactor is put in operation; or (3) the problem would have no safety implications until after several years of reactor operation and, should it not be resolved by then, alternative means will be available to insure thct continued operation (if pernitted at all) would not pose an undue risk to the public.
This appendix is specifically included to respond to the' decision of the Atomic Safety and Licensing Appeal Board as enunciated in ALAS-444, and as applied to an operating license proceeding Virginia Electric and Power Comoany (North'knna Nuclear Power Station, Unit Nos. 1 and 2), ALAB-491, NRC 245 (1978).
C. 3 Unresolved Safety Issues l
In a related matter, as a result of Congressional action on the Nuclear Regula-tory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of I
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c 1974 was amended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:
UNRESOLVED SAFETY ISSUES PLAN SECTION 210. The Commission shall develop a plan providing for specification and analysis of unresolved safety issues relating to nuclear reactors and shall take such action as may be necessary to implement corrective measures with respect to such issues.
Such plan shall be submitted to the Congress on or before January 1,1978 and progress reports shall be included in the annual report of the Com-mission thereafter.
The Joint Explanatory Statement of the House-Senate Conference Committee for the Fiscal Year 1978 Appropriations Bill (Bill 5.1131) provided the following additional information regarding the Committee's deliberations on this portion of the bill:
SECTION 3 - UNRESOLVED SAFETY ISSUES The House amendment required development of a plan to resolve generic safety issues.
The conferees agreed to a requirement that the plan be submitted to the Congress on or before January 1,1978.
The con-ferees also expressed the intent that this plan should identify and describe those safety issues, relating to nuclear power reactors, which are unresolved on the date of enactment.
It should set forth:
(1) Commission action taken directly or indirectly to develop and implement corrective measu'res; (2) further actions planned concerning such measures;'and (3) timetables and cost estimates of such actions.
The Commission should indicate the priority it has assigned to each issue, and the basis on which priorities have been assigned.
In response to the reporting' requirements of the new Section 210, the NRC staff submitted to Congress on January 1,1978, a report, NUREG-0410, entitled "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants,"
describing the NRC generic issues program.
The NRC program was already in place when PL 95-209 was enacted and is of considerably broader scope than the Unresolved Safety Issues Plan required by Section 210.
In the letter transmit-ting NUREG-0410 to the Congress on December 30, 1977, the Commission indicated that "the progress reports, which are required by Section 210 to be included in i
future NRC annual reports, may be more useful to Congress if they focus on the specific Section 210 safety items."
It is the NRC's view that.the intent of Section 210 was to ensure that plans were developed and implemented on issues with potentially significant public 01/17/82 C-3 CLINTON SER APP C
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safety implications.
In 1978, the NRC undertook a review of over 130 generic issues addressed in the NRC program to determine which issues fit this descrip-tion and qualify as unresolved safety issues for reporting to the Congress.
The NRC review included the development of proposals by the NRC staff and l
review and final approval by ihe NRC Commissioners.
f The review is described in a report, NUREG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants - A Report to Congress," dated January 1979. The report provides the following definition of an unresolved safety issue:
An Unresolved Safety Issue is a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been developed and that involves conditions not likely to be acceptable over the lifetime of the plants it affects.
Furthermore, the report. indicates that in applying this definition, matters that pose "important questions concerning the adequacy of existing safety requirements" were judged to be those for which resolution is necessary to (1) ccmpensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety.
Quite simply, an unresolved safety issue is potentially significant from a public safety standpoint and its resolu-tion is likely to result in N'RC action on the affected plants.
All of the issues addressed in the NRC program were systematically evaluated against this definition as described in NUREG-0510.
As a result, 17 unresolved The safety issues addressed by 22 tasks in the NRC program were identified.
issues and applicable task numbers are listed below.
Progress on these issues was first discussed in the 1978 NRC Annual Report.
The numbar(s) of the generic task (s) (for example, A-1) in the NRC program addressing each issue is indicated l
in parentheses following the title.
(1) Waterhammer (A-1)
(2) Asymmetric Blowdown Loads on the Reactor Coolant System (A-2)
C-4 CLINTON SER APP C 01/17/82
j (3)
Pressurized Water Reactor Steam Generator Tube Integrity (A-3, A-4, A-5)
(4)
BWR Mark I and Mark II Pressure Suppression Containae. ts (A-6, A-7, A-8, A-39)
(5) Anticipated Transients Without Scram (A-9)
(6) BWR Nozzle Cracking (A-10)
(7) Reactor Vessel Materials Toughness (A-11)
(8) Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports (A-12)
(9) Systems Interaction in Nuclear Power Plants (A-17)
(10) Environmental Qualification of Safety-Related Electrical Equipment (A-24)
(11) Reactor Vessel Pressure Transient Protection (A-26)
(12) Resicual Heat Removal Requirements (A-31)
(13) Control of Heavy Loads Near Spent Fuel (A-36)
(14) Seismic Design Criteria (A-40)
(15) Pipe Cracks at Boiling Water Reactors (A-42)
(16) Containment Emergency Sump Reliability (A-43)
(17) Station Blackout (A-44)
In the view of the staff, the unresolved safety issues listed above are the substantive safety-issues referred to by the Appeal Board in ALAB-444 when it spoke of "...those generic problems under continuing study which have...
potentially significant public safety implications." Six of the twenty-two tasks identified with the unresolved safety issues are not applicable to Clinton because they apply to pressurized water reactors o.11y.
These tasks are A-2, A-3, A-4, A-5, A-12, and A-26.
Also, Tasks A-6, A-7, and A-8 only apply to Mark I or Mark II boiling water reactor containments. With regard to the remaining 13 tasks that are applicable to Clinton, the NRC staff has issued,
NUREG reports 'providing its resolution of seven of the issues.
These issues are listed below.
01/17/82 C-5 CLINTON SER APP C
Task Number NUREG Report No. and Title SER Section A-9 NUREG-0460, Vol. 4, " Anticipated 7, 15.6 Transients Without Scram for Light Water Reactors" A-10 NUREG-0619, "BWR Feedwater Nozzle 4.6 and Control Rod Drive Return Line Nozzle Cracking"
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A-24 NUREG-0588, Revision 1, " Interim 3.10, 3.11 Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" A-31 NUREG-0800, SRP 5.4.7 and BTP 5-1, 5.4.2
" Residual Heat Removal Systems" (incorporate requirements of USI A-31)
A NUREG-0612, " Control of Heavy 9.1. 5 Loads at Nuclear Power Plants" A-39 NUREG-0802, " Safety / Relief Valve--
6.2.1.8.2 Quencher Loads:
Evaluation for BWR Mark II and III Containments"*
A-42 NUREG-0313, Revisicn'1, " Technical 5.2.3 Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping"
- This report is scheduled for publication early in 1982.
The report documents NRC acceptance of the safety relief valve loads proposed by General Electric in Report #22A7000, Revision 1, Appendix 38 for GESSAR-II 238 Nuclear Island dated Novemcer 25, 1980.
01/21/82 C-6 CLINTON SER APP C
i The remaining issues applicable to Clinton Unit 1 are (1) Waterhammer (A-1)
(2) Reactor Vessel Materials Toughness (A-11)
(3) Systems Interaction in Nuclear Power Plants (A-17)
- (4) Seismic Desicn Criteria (A-40)
(5) Containment Emergency Sump Reliability (A-43)
(6) Station Blackout (A-44)
With the exception of Tasks A-43 and A-44, task action plans for the generic tasks above are included in NUREG-0649, " Task Action Plans for Unresolved Safety Issues Related to Nuclear Power Plants." The Task Action Plan for Task A-43 was issued in January 1981, and the Task Action Plan for A-44 was issued in July 1980. The info:mation provided in NUREG-0649 meets most of the informational requireaients of ALA8-444.
Each task action plan provides a description of the problem; the staff's approaches to its resolution; a general discussion of the bases upon which continued plant licensing or operation can proceed pending con.pletion of the task; the technical organizations involved in the task and estimates of 'the manpower required; a description of the inter-actions with other NRC offic's, ACRS, and outside organizations; estimates of e
funding required for contractor-supplied 'echnical assistance; prospective dates for completing the tasks; and a description of potential problems that could alter the planned approach or schedule.
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In addition to the task action plans, the staff issues the " Aqua Book" (NUREG-0606) on a quarterly basis.
This book entitled, " Unresolved Safety Issues Summary, Aqua Book," provides current schedule information for each of the unresolved safety issues.
It also includes information ielative to the implementation status of each unresolved safety issue.for which technical l
resolution is complete.
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The staff has reviewed the six unresolved safety issues listed above and the four new unresolved safety issues discussed iri Section C,4 as they relate to Clinton Unit 1.
Discussion of each of these issues including references to related discussions in the SER is provided in Section C.5.
The 3taff has 01/17/82 C-7 CLINTON SER APP C
a satisfactprily concluded its review for all but Tasks A-46,." Seismic Qualifica-tion of Equipment ~in Operating Plants," and A-48, " Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment." It will discuss resolution, c
of these issues in a' supplement to this SER.
Based on its review, the staff ha's conclud'ed for the reasons set forth in Section C.5 that, with the exception of Tasks A-46 and A-48, there is reasonable assurance that Clinton Unit 1 can be operated before the ultimate resolution of these generic issues without endangering the health and safety of the public.
C. 4 New Unresolved Safety Issues The staff has performed an in-depth and' systematic review of generic safety concerns identified since January 1979 to determine if any of these issues should be designated as new unresolved safety issues.
The candidate issues originated from concerns identified in NUREG-0660, "NRC Action Plan as a Result of the TMI-2 Ac:ident," ACRS recommendations, abnormal occurrence reports, and other operating experience.
The staff's proposed list was reviewed and commented' on by the ACRS, the-Office of Analysis and Evaluation of Operational Data (AE00),
and the Office of Policy Evaluation.
The ACRS and AE00 also propossd that several additional. unresolved safety issues be considered by the Commission.
The Commis-sion considered the above information and approved the fol. lowing four new unresolved safety issues:
4 (1) Shutdown Decay Heat Removal Requirements (A-45)
L (2) Seismic Qualification of Equipment in Operating Plants (A-46)
(3) Safety Implication of Control Systems (A-47)
(4) Hydrogen Control Measures and Nffects of Hydrogen Burns on Safety Equipment l
l (A-48) i A description of the above process together with a list of the issues considered is presented in NUREG-0705, " Identification of New Unresolved Safety Issues f
Relating to Nuclear Power Plants, Special Report to Congress," dated March 1981.
An expanded discussion of each of the new unresolved safety issues is also contained in NUREG-0705.
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The applicability and bases for licensing before ultimate resolution of the four new unresolved safety issues for Clinton Unit 1 is discussed in Section C.5.
C. 5 Discussions of Tasks as They Relate to Clinton Unit 1 This section provides the NRC staff's evaluation of Clinton Unit 1 for each of the applicable unresolved safety issues.
This includes the staff's bases for licensing before ultimate resolution of these issues.
NRC conclusions are based in part on information provided by the applicant in his letter of October 23, 1981, f om G. E. Wuller, Illinois Power Company, to R. L. Tedesco, NRC.
A-1 Waterhamer Waterhamer events are intense pressure pulses in fluid systems caused by any one of a number of mechanisms and system conditions such as rapid condensation of steam pockets, steam-driven slugs of water, pump startup with partially empty lines, and rapid valve motion.
Since 1971, over 200 incidents involving waterhamer in pressurized and boiling water reactors have been reported.
The waterhamers (or steam hamers) have involved steam generator feedrings and piping; the residual heat removal systems; emargency core cooling systems; and containment spray, service water, feedwater and steam lines.
Most of the damage reported has been relatively minor, involving pipe hangers and restraints; however, sev'eral waterhammer incidents have resulted in piping and valve damage.
The most serious waterhammer events have occurred in the steam generator feedrings of pressurized water reactors.
In no case has any waterhamer incident resulted in 'the release of radioactive material.
Under generic. Task A-1, the potential for waterhamer in various systems is being evaluated and appropriate requirements and systematic review procedures are being developed 'to ensure that waterhammer is given appropriate considera-tion in all areas of licensing review.
A technical report, NUREG-0582, " Water-hamer in Nuclear Power Plants" (July 1979), providing the results of an NRC t
staff review of waterhammer events in nuclear power plant.s and stating current staff licensing positions, completes a major subtask of generic Task A-1.
01/17/82 C-9 CLINTON SER APP C i
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Although waterhammer can occur in any_ light water reactor and approximately 118 actual and probable events have been repcrted in boiling water reactors as of September 1979, none has caused major pipe failures in a boiling water l
reactor such as Clinton and none has resulted in the offsite release of radio-activity.
As noted above, the most severe waterhammers observed to date have been in steam generators. Because the boiling water reactor does not use c steam generator, these worst cases are eliminated.
Furthermore, any waterhammer that may occur in feedwater or main steam piping will not impair the emergency core cooling system because all emergency core cooling system waf.er enters the reactor vessel through five separate reactor vessel nozzles independent of the feedwater and main steam piping.
To protect the Clinton emergency core cooling system (ECCS) from the effects of waterhammer, the ECCSs are equipped with jockey pumps that provide a contin-uous supply of water to the ECCS discharge piping.
These jockey pumps keep the ECCS lines water filled so that the ECCS pumps will not start pumping into voided lines and steam will not collect in the ECCS piping.
To ensure that the ECCS lines remain water filled, vents'have been installed and a Technical Specification requirement to periodically vent air from the lines has been imposed.
Further assurance for filled discharge piping is provided by pressure instrumentation at the piping high point.
An alarm sounds.in the main control room if the pressure falls below a predetermined setpoint indicating difficulty maintaining a filled discharge line. Should this occur, or if an instrument becomes inoperable, the required action is identified in the Technical Specifications.
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With regard to additional protection currently provided in plants against potential waterhammer events, piping design codes require consideration of impact loads.
Approaches used at the design stage include (1) increasing valve closure times, (2) piping layout to preclude water slugs in steam lines and vapor formation in water lines, (3) use of snubbers and pipe hangers, and (4) use of vents and drains.
In addition, the applicant will conduct a preoperational vibration and dynamic effects test program in accordance with Standard OM-3 of the American Society of Mechanical Engineers for all Class 1, Class 2, Class 3, and other piping systems 01/17/82 C-10 CLINTON SER APP C
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and piping restraints during startup and initial operation.
These tests will provide adequate assurance that the piping and piping restraints nave been designed to withstand dynamic effects caused by valve closures, pump trips, and other operating modes.
i Nonetheless, in the unlikely event that a large pipe break did result from a severe waterhammer' event, core cooling is ensured by the ECCS and protec-tion against the dynamic effects of such pipe breaks inside and outside of containment is provided.
In the event that Task A-1 identifies potentially significant waterhammer scenarios that have not been accounted for explicitly in the design and opera-tion of Clinton, corrective measures will be implemented at that time. The i
i task has not identified the need for measures beyond those already implemented.
Based on the foregoing, the staff concludes that Clinton can be operated before ultimate resolution of this generic issue without undue risk to the healtn and safety of the public.
A-11 Reactor Vessel Materials Touchness l
Resistance to brittle fracture is described quantitatively by a material property I
generally denoted as fracture toughness.
Fracture toughness has different values and characteristics depending on the material being considered.
For steels used in a nuclear reactor pressure vessel, three considerations are important:
l (1) Fracture toughness increases with increasing temperature.
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(2) Fracture toughness decreases with increasing load. rates.
.(3) Fracture toughness decreases with neutron irradiation.
1 In recognition of these considerations, power reactors are operated within restrictions imposed by the Technical Specifications on the pressure during heatup and cooldown operations.
These restrictions ensure that the reactor vessel will not be subjected to a combination of pressure and temperature that could cause brittle fracture of the vessel if there were significant flaws in the vessel material.
The effect of. neutron radiation on the fracture toughness t
01/17/82 C-11 CLINICN SER APP C
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of the vessel material over the life of the plant is accounted for in Technical I
Specificat. ion limitations.
The principal objective of Task A-11 is to develop safety criteria to allow a more precise assessment of safety margins during normal operation, transients, and accident conditions in older reactor vessels with marginal fracture toughness.
Based on its evaluation of the Clinton reactor vessel materials. toughness and g
subject to confirmation by review of d'ta to be supplied by the applicant in a
p g i. December, the staff concludes that adequate safety margins exist for brittle
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failure during operating, testing, maintenance, and anticipated transient
- g conditions. When Task Action Plan A-11 is completed and explicit fracture evaluation criteria for accident conditions are defined, all vessels will be reevaluated for acceptability over their design lives.
The materials of the Clinton reactor vessel meet the fracture toughness requirements of NB-2300 of the ASME Code.
Based on this fact and the fabrica-tion techniques employed on the Clinton vessel, it is estimated that the total fluence over th'e design life would result in a final fracture toughness value above the minimum Charpy impact requirement of 50 ft-lb.
In addition, the surveillance program required by Appendix H of 10 CFR 50'will afford an opportunity to reevaluate the fracture, toughness periodically during a minimum of the first half of the design life.
To ensure adequate safety margins, adjustment to the nil ductility transition (NDT) temperature and the development method for pressure / temperature curves are speci.fied in 10 CFR 50, Appendices G and H.
The amount of adjustment to the operating curves is a function of the fast neutron (greater than 1 Mev) fluence i
and copper and phosphorus content of the reactor pressure vessel (RPV) material.
For BWR-6s, the copper and phosphorus content of the material is closely controlled.
Furthermore, higa upper shelf toughness is specified and all values for core belt line material were in excess of 75 ft-lb.
The fast neutron fluence is low with respect to other reactor types because of the i
additional moderator (water) in the annulus between the core shroud and the RPV.
Therefore, the reactor pressure vessel material toughness (A-11) issue is of relatively low concern for SWR-6s.
I 01/17/82 C-12 CLINTON SER APP C
I t-In Clinton's case, the RPV limiting material in the core belt line contains 0.10 percent copper and 0.016 percent phosphorus.
The initial RT is -30 Add fs.
NDT Based on a predicted adjusted reference temperature (RT) as a function of cd M
is predicted fluence and copper and phosphorous content, the end-of-life RTNDT to be 41*F.
On this basis, the Clinton RPV has adequate safety margin with respect to the requirements of 10 CFR 50, Appendices G and H.
r Therefore, on the basis of the foregoing considerations, the staff concludes l
that Clinton can be operated before resolution of this generic issue without
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I undue risk to the health and safety of tne public.
t A-17 Systems Interaction in Nuclear Power Plants I
The staff's systems interaction program was initiated in May 1978 with the definition of Unresolved Safety Issue A-17 (Systems Interaction in Nuclear Power Plants) and was intensified by Task Action Plan (NUREG-0660) Item II.C.3 i
(Systems Interaction).
The concern arises because the design, analysis, and installation of systems are frequently the responsibility of teams cf engineers l
with functional specialties such as civil, electrical, mechanical, or nuclear.
[
Experience at operating plants has led to questicos of whether the work of these functional specialists is sufficiently integrated to enable them to minimize adverse interactions among systems.
Some adverse events that occurred i
in the past might have been prevented if the teams had ensured the necessary independence of safety systems under all conditions of operation.
The applicant has not described a comprehensive program that separately evalu-
[
ates all structures, systems, and components important to safety for the three l
categories of adverse systems interactions, that is, spatially coupled, func-tionally coupled, and humanly coupled.
However, there is assurance that Clinton Unit 1 c,an be operated without endangering the health and safety of the public.
The plant has been evaluated against current. licensing requirements that are founded on the principle of defense-in-depth.
Adherence to this principle results in requirements such as physical separation and indepenuance i
of redundant safety systems, and protection against hazards such as high energy line ruptures, missiles, high winds, flooding, seismic events, fires, human f
factors, and sabotage. These design provisions are subject to review against
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i 1
01/17/82 C-13 CLINTON SER APP C l
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review against the Standard Review Plan (NUREG-0800) which requires inter-disciplinary reviews of safety grade equipment and addresses different types of potential systems interactions. Also, the quality. assurance program that is followed during the design, construction, and operational phases for each plant contribute to the prevention of introducing adverse systems interactions.
Thus, the current licensing requirements and procedures provide an adequate degree of plant safety.
The NRC staff's current procedures assign primary responsibility.for review of various technical areas to specific organizational units and assign secondary responsibility to other units where triere is a functional interface.
Designers follow somewhat similar procedures and provide the analyses of systems and intei face reviews.
Task A-17 has been developing methods that could identify adverse systems interactions that were not considered by current review procedures. The first phase of this study began in May 1978 and was completed g!.7, in February 1980 by Sandia Laboratories under contract to the NRC staff.
The Phase I investigation was structured to identify areas where interactions are possible between systems and which have the potential of negating or serious 1y degrading the performance of safety functions.
The study concentrated on commonly caused failures among systems that would violate a safety function.
The investigation was to then identify those areas in which NRC review proce-dures may not have properly accounted for these interactions.
The Sandia Laboratories used fault-tree analysis on a selected light-water-reactor (LWR) design to identify component failure combinations (cut-sets) that could result in loss of a safety function.
The cut-sets were further reduced by incorpor.ating six linking systems failures into the analysis.
The results of the Sandia effort indicated a few potentially adverse systems interactions within the limited scope of the study.
The staff reviewed the interactions for safety significance and generic implications.
The staff concluded that no corrective measures needed to be implemented immediately except for the potential g
Interaction between the3wer-operated relief valve and its block valve.
This
['
interaction had been separately identified by the evaluations of the TMI-2 acci-y dent wnile Sandia was studying the selected LWR.
Because corrective measures J
01/21/82 C-14 CLINTON SER APP C l
p were already being implemented, no separate measures were needed under Unresolved Safety Issue A-17.
The applicant has taken some initial steps toward the performance of a separate evaluation of Clinton Unit 1 from a multidisciplinary point of view.
He has formed a Systems Interaction Survey Teum to analyze the "as-built" plant in areas where spatially coupled systems interactions c'ould adversely affect safety grade in'spection equipment.
NUREG-0660 provides for a systems interaction follow on study in Section II.C.3,
" Systems Interactions." Since. Apri'l 1980, the Office of Nuclear Reactor Regulation has intensified the effort both by broadening the st0dy of methods to identify potential systems interactions and by preparing guidance for audit reviews of selected plants for systems interactions.
The staff's recent experience provider, a basis from it is developing a more efficient review process for potential systems interactions. The process will provide for a resolution of l
Unresolved Safety Issue A-17, assimilate operating reactor experience, and rank identified systems interactions by their relative importance to safety.
~
It is expected that the development of systematic ways to identify, rank, and j
evaluate systems interactions will go further to reduce the likelihood of i
intersystem failures resulting in the loss of plant safety functions.
A comprehensive program is expected to use analytical methods, visual inspections, experience feedback, and' simulator dependencies experiments.
The LWR industry's current experience with systems interaction reviews is fragmented.
Experience like that gained by the Phase I study is an essential ingredient to the staff's l
considerations of a comprehensive systems interaction program.
After the resolu-
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tion of Unresolved Safety Issue A-17, the staff will determine whether the applicant must perform further evaluations for adverse systems interaction.
l
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A-40 Seismic Design Criteria - Short-Term Program i
l NRC regulations required that nuclear power plant structures, systems, and com-l l
ponents important to safety be designed to withstand the effects of natural
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phenomena such as earthquakes.
Detailed requirements and guidance regarding the seismic design of nuclear plants are provided in the NRC regulations and in 01/17/82 l
C-15 CLINTON SER APP C i
Regulatory Guides issued by the Commission.
However, there are a number of plants with construction permits and ' operating licenses issued before the NRC current regulations and regulatory guidance were in place.
For this reason, reviews of the seismic design of various plants are being undertaken again to ensure that these plants do not present an undue risk to the public.
Task A-40 is, in effect, a compendium of short-term efforts to support such reevaluation efforts of the NRC staff, especially those related to older operating plants.
In addition, some revisions to sections of the Standard Review Plan and regu tory guides to bring them more in line with the state of the art'will result The seismic design basis and seismic design of Clinton are still under review.
The staff's review of Clinton is discussed in Sections 2.5 and 3.7 o Should the resolution of Task A-40 indicate that a change is needed in these licensing requirements, all operating reactors including Clinton will be reevaluated on a case-by-case basis.
Accordingly, the staff has concluded that Clinton, subject to the resolution of the seismic issues still open, can be operated before ultimate resolution of this generic issue without endangering the health and safety of the public A-43 Containment Emeroency Sumo Reliability Following a postulated loss of-coolant accident, that is, a break in the reactor coolant system piping, the water flowing from the break would be collected in the suppression pool.
This water would be recirculated through the reactor system by the emergency core cooling pumps to maintain core coolin This water may also be circulated through the containment spray system to 4
remove heat.and fission products from the drywell and Wetwell atmosphere. Loss of the ability to draw water from the suppression pool could disable the emer-gency cooling and containment spray systems, i
The concern addressed by this tark action plan for boiling water reactors is i
limited to the potential for degraded emergency core cooling system (ECCS) performance as a result of thermal insulation debris that may be blown into the suppression pool during a loss of-coolant accident and. cause blockage of the 01/17/82 C-16 CLINTON SER APP C
1 l
P pump suction lines. A second concern, potential vortex formation, is not con-sidered a serious concern for Mark III containment because of the depth of the ECCS suction lines and the low approach velocities. The minimum ECCS suction
- submergence is 10 ft.
With regard to potential blockage of the intake lines, the likelihood of any insulation being drawn into an ECCS pump suction line is very small.
The potential debris in the drywell could only be swept into the suppression pool through the horizontal vents.
Any pieces reaching the pool would tend to settle on the bottom and would not be drawn into the pump suction because the suction centerline is 8 ft. above the pool bottom.
In additio.n, boiling water reactor designs employ strainers on the suction sized with flow areas 200 percent larger than the suction piping.
i Accordingly, the staff concludes that Clinton can be operated before ultimate resolution of this generic issue without endangering the health and safety of the public.
A-44 Station Blackout i
Electrical power for safety systems at nuclear power plants must'be supplied by at least two redundant and independent divisions.
The systems used to remove decay heat to cool the reactor core following a reactor shutdown are included among the safety-systems that must meet these requirements.
Each electrical division for safety systems includes two offsite ac power connections, a standby emergency diesel generator ac power supply, and dc sources.
Task A-44 involves a study of whether or not nuclear power plants should be designed to accommodate a complete loss of all ac power, that is, a loss of both the offs-ite and the emergency diesel generator ac power supplies.
Thi's issue arose because of operating experience regarding the reliability of ac power supplies.
A number of operating plants have experienced a total loss of l
offsite electrical power, and more occurrences are expected in the future.
During each of these loss-of offsite power events, the onsite emergency ac power supplies were available to supply the power needed by vital safety equipment.
However, in some instances, one of the redundant emergency power OU2U82 C-17 CLINTON SER APP C v
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m.
supplies has been unavailable.
In addition, there have been numerous reports of emergency diesel generators failing to start and run in operating plants during per. iodic surveillance tests.
A loss of all ac power was not a design-basis event for the Clinton facility.
Nonetheless, a combination of design, operating, and testing requirements has been imposed to ensure that this unit will have substantial resistance to a loss of all alternating current and that, even if a loss of all ac power should occur, there is reasonable assurance the core will be cooled.
These
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design, operating, and testing requirements are discussed below.
A loss of offsite ac power involves a loss of both the preferred and backup sources of.offsite power.
The staff's review and basis for acceptance of the design, inspection, and testing provisions for the offsite power system are described in Section 8.2 of the SER.
If offsite ac power is lost, three diesel generators and their associated distribution systems will deliver emergency power to safety related equipment.
The staff's review of the design, testing, surveillance, and maint'enance provisions for the onsite emergency diesels is described in Sections 8.3 and 9.6 of the SER.
Staff requirements include preoperational testing to ensure the reliability of the installed diesel generators.
If both offsite and onsite ac power are lost, boiling water reactors may use a combination of safety / relief valves and the reactor core isolation cooling system to remove care decay heat without reliance on ac power.
These systems assure that adequaw cooling can be maintained for at least 2 hr, which allows time for restoration of ac power from either offsite or onsite sources.
The issue of station blackout was considered by the Atomic Safety and Licensing Appeal Board (ALAB-603) for the St. Lucie No. 2 facility.
In addition, in view of the completion schedule for Task A-44 (March 1983), the Appeal Board recom-mended that the Commission take expeditious action to ensure that other plants and their operators are equipped to accommodate a station blackcut event.
The
. Commission has reviewed the Board's recommendations and determined that some interim measures should be taken at all facilities including Clinton while 01/17/82 C-18 CLINTON SER APP C
p Task A-44 is being conducted.
Consequently, interim emergency procedures and operator training for safe operation of the facility and restoration of ac power will be implemented.
This action will be completed by fuel load date.
i I
On the basis of the above considerations, the staff has concluded that there is reasonable assurance that Clinton can be operated before the ultimate resolution of this generic issue without endangering the health and safety of i
the public.
A-45 Shutdown Decay Heat Removal Requirements Following a reactor shutdown, the radioactive decay of fission products continues k
to produce heat (decay heat) that must be removed from the primary system. The principal means for removing this heat in a boiling water reactor while at high pressure is by means of the str.am lines to the turbine condenser. The condensate is normally returned to the reactor vessel by the feedwater system; however, the steam turbine-driven reactor core isolation cooling system is provided to maintain l
primary system inventory if ac power is not available.
When the system is at low pressure, the decay heat i,s removed by the residual heat removal systems.
This unres.olved safety issue will evaluate the benefit of providing alternate means of decay heat removal which could substantially increase the plant's i
capability to handle a broader spectrum of transients and accidents.
The study will consist of a generic system evaluation and will result in recommendations l
l regarding the desirability of and possible design requirements for improvements in existing systems or an alternative decay heat removal method if the improve-i ments or alternative can significantly reduce the overall risk to the public.
The Clinton reactor has various methods for the removal.of decay heat.
As discussed above, the decay heat is normally rejected to the turbine condenser and condensate is ieturned to the vessel by the feedwater system.
The reactor core isolation cooling system provides an alternate means of supplying makeup water to the vessel.
This turbine-driven pump takes suction from the reactor core isolation cooling storage tank and pumps to the vessel.
If the condenser is not available (for example, loss of offsite power), heat can be removed by means of the safety / relief valves to the suppression pool.
Also, the high 01/17/82 C-19 CLINTON SER APP C-
pressure core spray system is provided if the reactor core isolation cooling system is not available.
If the reactor core isolation cooling and high pressure core spray are unavailable, the reactor system pressure can be reduced by the automatic depres-surization system so that cooling by the residual heat removal system can be initiated.
When the condenser is not used, the heat rejected to tne suppres-sion pool is subsequently removed by the residual heat removal system.
The reactor core isolation cooling and high pressure core spray systems at Clinton have improvements over comparable systems at older boiling water reactors.
The reactor core isolation cooling system has been upgraded to safety grade quality (now required for all boiling water reactors), and the high pressute core spray is powered by its own dedicated diesel so it can operate with an assumed loss of all other sources of ac power.
Also, the residual heat removal system contains three pumps; the flow capacity of any single cump is sufficient to easily remove the decay heat.
Following the TMI accident, the industry performed and documented extensive analyses of feedwater transients and small-break. loss-of-coolant accidents to 3
support acceptability of current designs.
A report of these analyses was pro-vided to the NRC in NED0-24708A, Revision 1, dated December 1980.
The staff's cssessment of current designs related to, loss-of-feedwater transients and small
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loss of-coolant accidents is contained in NUREG-0626, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications."
I l
On the basi.s of.the above considerations, the staff has concluded that Clinton i
can be operated before the ultimate resolution of this. generic issue without endangering the health and safety of the public.
l 1
A-46 Seismic Oualification of Eouipment in Ooerating Plants The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant change j
during the course of the commercial nuclear power program.
Consequently, the 01/17/82 C-20 CLINTON SER APP C
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i margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably.The seismic qualification of the equipment in operating plants must, therefore, be reassessed to ensure the ability to bring the plant to a safe shutdown condi-tion when subject to a seismic event.
The objective of this unresolved safety issue is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants instead of attempting to backfit current design criteria for new plants.
This guidance will concern equipment required to safely shut down the plant, as well as equipment whose function is not required for safe shutdown, but whose failure could result in adverse conditions that.might impair shutdown functions.
Clinton was designed using current seismic design criteria, and commitments for seismic equipment qualification are to the latest codes and standards.
Require-ments for seismic equipment qualification in the Clinton FSAR include IEEE Std. 344-1975 and Regulatory Guides 1.92, " Combining Modal Response and Spatial Components in Seismic Response Analysis," and 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants." SRP Sections 3.9.2 and 3.10 have alto been considered in the qualification efforts.
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Since identification of hydrodynamic load effects on Clin' ton structures, an effort has been initiated to assess the effects of these loads (in combination with previously established seismic loads) on equipment required to safely shut down the plant.
This includes equipment not required for safe shutdown, but whose failure could adversely affect equipment required for safe shutdown.
This reassessment involves validation of equipment qualification by both 4A analytical methods and additional testing, when required.
It is anticipated that this effort wil.1 be completed by December 1981, and results will be j
9 reflected in an FSAR amendment to modify appropriate porti6ns of Sections 3.9 and 3.10.
The staff's evaluation of the applicants' seismic qualification pro-gram will be provided in a :upplement to this SER.
01/17/82 C-21 CLINTON SER APP C
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A-47 Safety Implicatfuns of Control Systems 1
This issue' concerns the potential for transients or accidents being mad l
severe as a result of control system failures or malfunctions.
These failures or malfunctions may occur independently or as a result of the accident or transient under consideration.
One concern is the potential for a single failure such as a loss of a power supply, short circuit, open circuit I
, or sensor failure to cause simultaneous malfunction of several con Such an occurrence could conceivably result in a transient more severe th those transients analyzed as anticipate'd operational occurrences A second concern is for a postulated accident to cause control system failures that l
would make the accident more severe than analyzed. Accidents could conceivably cause control system failures by creating a harsh environment in the area of the control equipment or,by physically damaging the control equipment Although it is generally believed that such control system failures would not lead to serious events or result in conditions that safety systems cannot safely ha in-depth studies have not been rigorously performed to verify this belief The potential for an accident that would affect a particular control system, and
~ ffects of the control system failures, may differ from plant to plant e
There-fore, it is not 'possible to develop generic answers to all of these concerns it is possible to develop generic criteria that can be used for future plant specific reviews.
The purpose of this unresolved safety issue is to verify the adequacy of existing criteria for contro.1 systems or propose additional ge criteria (if necessary) that will be used for plant specific review.
The Clinton safety systems have been designed with the goal of ensuring t 6
control system failures (either single or multiple) will not prevent automatic or manual i.nitiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe shutdown condition follow any anticipated operational occurrence or accident.
This has been accomplished by either providing independence between safety-and nonsafety grade system or providing isolating devices between safety and nonsafety grade systems.
i These devices preclude the propagation of nonsafety grade system equipment[
I faults so that operation of the safety grade system equipment is not impaired q
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01/17/82
~
C-22 CLINTON SER APP C
p A wide range of bounding transients and accidents is presently analyzed to ensure that' the postulated events,would be adequately mitigated by the safety systems.
In addition, systematic reviews of safety systems have been performed with the goal of ensuring that the control system failures (single or multiple) will not defeat safety system action.
Also, the applicant has been requested (NRC Information Notice 79-22, "Qualifica-tion of Control Systems," September 17, 1979) to review the possibility of conse-quential control system failures that exacerbate the effects of high-energy line breaks (HELBs) and adopt new operator procedures where needed, to ensure that the postulated events would be adequately mitigated.
As part of the review, the staff is also evaluating the qualification program to ensure that equipment that may potentially be exposed to HELB environments has been adequately qualified or an adequate basis has been provided for not qualifying the equipment to the limiting hostile environment.
The staff's evaluation of the applicant's response to Information Notice 79-22 and the adequacy of the qualification program will be reported in a supplement to the SER.
Further discussion of these issues can be found in Sections 7.7 and 3.11.
With the recent emphasis on the availability of postaccident instrumentation
~
(Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant Conditions 'uring and Following an Accident"), the staff D
reviews evaluate the designs to ensure that control system failures will not deprive the operator of information required to maintain the plant in a safe shutdown condition after any anticipated operational occurrence or accident.
The applicant was requested to evaluate its control systems and identify any cpntrol systems whose malfunction could impact plant safety.
The applicant is
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requested to document the degree of interdependence'of these identified control systems and identify the use (if any) of commo'1 power supplies and the use of common sensors or common sensor impulse lines whose failure could have potential safety significance. The status of these reviews and the staff's evaluation are documented in the Section 7.7 of the SER.
In addition, IE Bulletin 79-27 (" Loss of Non-Class 1E Instrumentation and Control Power System Bus During Operation," November 30, 1979) was issued to the applicant requesting that evaluations be performed to ensure the adequacy l
01/17/82 C-23 CLINTON SER APP C r
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1 of plant procedures for accomplishing ~ shutdown on loss of power to any electrical bus supplying power for instruments and contro:1 This item is
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coverect in the SER, Section 7.4.
However, as noted in Section 7.4 the review l
is not complete.
The subtask of this issue concerning the reactor overfill transient in boiling water reactors is currently under review by the BWR Owners' Group of which Clinton is a member.
Pending ultimate resolution of this item, the applicant I
has incorporated in the Clinton design., a commercial grade high' level trip (Level 3) of the reactor core isolation cooling, high pressure core spray, and feedwater systems to prevent the occurrence of overall transients.
l On the bas.is of the above considerations, and subject to the satisfactory resolution of the open items, the staff has concluded that there is reasonable assurance that Clinton can be constructed and operated before the ultimate resolution of this generic issue without endangering the health and safety of the public.
l A-48 Hydrogen. Control Measures and Effects of Hydrogen Burns on Safety
- Ecuipment i
Following a loss-of-coolant accident in an LWR plant, combustible gases, principally hydrogen, may accumulete inside the primary reactor containment as a result-of (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the water in the reactor core and the contain-ment sump; (3) the corrosion of certain construction materials by the spray solution; and (4) any synergistic chemical, thermal, and radiolytic effects of postaccident environmental conditions on containment protective coating systems and electric cable insulation.
Because of the potential for significant hydrogen generation as the result of l
an accident, 10 CFR 50.44, " Standards for Combustible Gas Control System in l
Light Water Cooled Power Reactors," and 'GDC 41, " Containment Atmosphere Cleanup,"
in Appendix A to 10 CFR 50, require that systems be provided to control hydrogen
. concentrations in the containment atmosphere following a postulated accident to ensure that containment integrity is maintained.
l 01/17/82 C-24 CLINTON SER APP C
P Regulation 10 CFR 50.44 requires that the combustible gas control system provided.be capable of handling the hydrogen generated as a result of degradation of the ECCS so that the hydrogen release is five times the amount calculated in demonstrating compliance with 10 CFR 50.46 or the amount corre-sponding to reaction of the cladding to a depth of C.00023 in., whichever amount is greater.
The accident at TMI-2 on March 28, 1979 resulted in hydrogen generation well in excess of the amounts specified in 10 CFR 50.44. As a result of this it i
becam. apparent to NRC that specific design measures are needed for handling.
larger hydrogen releases, particularly for small, low pressure containments.
As a result, the Commission determined that a rulemaking proceeding,should be
[
undertaken to define the manner and extent to which hydrogen evolution and other effects of a degraded core need'to be taken into account in plant design.
An l
t advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Register on October 2,1980.
Recognizing that a number of years may be required to complete this rulemaking proceeding, a set of short-term or interim actions relative to hydrogen control requirements was developed and implemented.
These interim measures were described l
.in a se~cond October 2, 1980 Federal Register notice.
I
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l For plants with Mark III containments such as Clinton, the proposed interim rule specified that either it must be demonstrated that the containment can
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withstand hydrogen burns or explosions or a detailed evaluation of possible
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hydrogen control measures must be performed and the selected measures installed.
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Clinton will comply with this interim rule through use of a hydrogen ignition system (HIS).
This system consists of glow plug igniters distributed throughout the containment and drywell.
This HIS is designed to ignite hydrogen at low
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concentrations thereby maintaining the concentration of hydrogen below its '
l detonable limit and preventing potential containment overpressure.
I To collectively evaluate the concerns associated with the hydrogen issue for l
Mark III containments, an owners' group has been formed.
This group is sponsoring I
l 01/17/82 C-25 CLINTON SER APP C j
)-
t analytical work with General Electric,- Offshore Power Systems,. and others.Cur-rent evaluations of this group indicate that containment pressures will remain well below the failure point as the result of the postulated hydrogen release and burn.
The staff is currently reviewing the HIS and the applicant's analysis of the ability of essential equipment to survive the hydrogen burn environment.
The results of this review will be provided in a supplement to this SER.
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i 01/17/82 C-26 CLINTON SER APP C
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