ML17309A167
| ML17309A167 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/22/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Maier J ROCHESTER GAS & ELECTRIC CORP. |
| References | |
| TASK-03-10.B, TASK-3-10.B, TASK-RR LSO5-81-06-105, LSO5-81-6-105, NUDOCS 8106290264 | |
| Download: ML17309A167 (49) | |
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Ihfo wang p~~, 'll CCAVAI5$gII lp Docket No. 50-244 LS05-81-06-105 Nr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corp.
89 East Avenue Rochester, New York 14649 UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, O. C. 20555 June 22, 1981
Dear Hr. Haier:
RE:
COMPLETION OF T R.
E.
GINNA NUC II-10.B -
P P
FLYWHEEL INTEGRITY Enclosed is a copy of our final evaluation of Systematic Evaluation Program Topic III-10.B.
This report has been revised to reflect the comments con-tained in your letter of August 29, 1979.
With th',s revised evaluation our review of SEP Topic III'-10.B is complete and will be a basic input to the integrated assessment of your facility.
The subject assessment compares your facility design with the criteria cur-rently used by the staff in licensing new facilities, This assessment may need to be re-examined if you modify your facility or if the criteria are changed before we complete our integrated assessment.
Si cereIy,
Enclosure:
As stated Dennis M. Crutchfield, ief Operating Reactors Bra h Ho.
5 Division of Licensing cc w/enclosure:
See next page f++g )IC(l
Mr; John E. Maier CC Harry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire
- Avenue, N.
W..
Suite 1100 Washington, D. C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen New York State F nergy Office Swan Street Building Core 1,
Second Floor Empire State Plaza
- Albany, New York 12223 Director, Technical Development.
Programs State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107-Ridge Road West
- Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York 14519 Director, Criteria and Standards Division Of'fice of Radiation Programs (ANR-460)
U. S. Environmental Protection Agency Washington, D. C.
20460 U. S. Envi ronmenta1 Protecti on Agency Region II Office ATTN:
E I S COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U.
S. Nuclear Regulatory Conmission Washington, D. C.
20555 Dr.
Emneth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Corrmission Washington, D. C.
20555 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.
1725 I Street, N.
W.
Suite 600 Washington, D. C.
20006 Ezra I. Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047
ROCHESTER GAS AND ELECTRIC CORPORATION R. E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-255 SAFETY EVALUATION REPORT MATERIALS ENGINEERING BRANCH DIVISION OF ENGINEERING
'SEP TOPIC III-10.B - PUMP FLYWHEEL INTEGRITY I.
INTRODUCTION The safety objective of this review is to assure that the integrity of the, primary reactor coolant pump flywheel is maintained to prevent failure at normal operating speeds and speeds that might be reached under accident conditions and thus preclude the generation of missiles.
II.
RE'/IEW CRITERIA The basis for review is outlined in Standard Review Plan (SRP),
Sec:ion 5.4.1.1 and the Regulatory Guide 1.14 (Revision 0), which describes and rec-ommends-a method'acceptable to the NRC staff in implzm'enting General Design.
Criterion 4, "Environmental and Missile Design Bases" of Appendix A of 10 CFR Part 50 with regard to minimizing the= potential for failures of the flywheels of the reacto~ coolant pumps.
I III. RELATED SAFETY TOPICS AND INTERFACES'nternally generated missiles protection is evaluated under SEP Topic III-4.C.
IV.
REVIEW GUIDELINES There are two parts to the recomnendation of Regulatory Guide 1.14.
The first part is related to the evaluation of the materials of cons ruction, design, fabrication, proof testing and pre-service inspection of the
flywheels.
The second part is concerned with the evaluation of the procedures used for the inservice inspection of flywheels.
V.
EVALUAT?ON a.
Haterial and Fabrication The flywheels are fabricated from rolled, vacuum-degassed, ASTH A-533 steel pl ates.
Flywheel bl anks are flarne-cut from the pl ate, with allowance for exclusion of flame affected ma;erials.
A minimum of three Charoy tests are made from each plate parallel and normal to the rolling direction to determine that each blank satisfies material toughness re-quirements.
The Nil-DuctilityTransition Temperature (NDTT) is less than +10'F.
The flywheel material has a minimum yield strength of 50,000 psi and tensile strength of 80,000 psi.
The flywheels are sub-jected to 1004 volumetric ultrasonic inspection.
The finished machined bores are also subjected to magnetic particle, or liquid penetrant exam-ination.
The pump flywheels are mounted on a shaft of radius 4.2 inches and consist of two large steel discs bolted together.
The discs are 75 and 65 inches in diameter.
Regulatory Guide 1.14 requires that the Nil-Ductility Transition Temperature (NDTT) for the flywheel material be no more than 10'F, that the Charpy V-notch (CVN) uppershelf energy should be at least 50 it-lb, and the minimum dynamic stress intensity. factor of the material should be 100 ksi
~in at the normal operating temperature.
This later requirement can be satisfied by demonstrating that the material has a
CVN energy level of 50 ft-lb at normal operating temperature.
Based on our review, we have determined that the NDT of the flywheels material at Ginna is less than 10'F.
Even though no Charpy test data or dynamic toughness values for flywheel A-533 material at the operating temperature of more than 100'F are given, that data for A-533 exist in
the literature because of its wide usage in the manufacture of the reactor pressure vessel.
The CVN values can be estimated above 100 ft-lbs and the stress intensity factor should also be above 100 ksi ~in at the operating temperature of 100'F.
Thus, all the design-fabrication requirements are satisfied.
The requirements of the Reg'ulatory Guide ensure that brittle fracture is unlikely and that a large Coleranc to flaw'-induced fracture exists.
This flaw tolerance meets the intent of the Regulatory Guide.
b.
Design The primary coolant pumps run at 1189 rpm, and may operate briefly at over speed of 109Ã (1295 rpm) during loss of outside load.
- However, the design speed is selec.ed as 125; of the operating speed.
At operating speed (1189 rom at Ginna) the bore stress due Co rotation is 1(000 psi.
The design specifications for the reactor coolant pumps include as a
design condition the stresses generated by a maximum hypothetical earthquake ground acceleration of 0.2g.
In no case does any bearing stress in the pump exceed or even approach a value which the bearing could not carry.
Each component of the primary'ump has been analyzed for missile generation.
Any fragments would be contained by the heavy stator.
The small fragments from the impeller would be contained by the heavy casing.
At the design overspeed of 1486 rpm, the maximum tangential stress reaches 21,500 psi, which is still less than 50% of the minimum yield strength at the operating temperature (100 to 150'F).
A bursting speed of 3900 rpm for the flywheels has been stated by the licensee.
The staff has determined the bursting speed for the flywheels to be 3400 rpm.
These results were obtained by calculating the tangential stress and'ssuming failure at flow stress.
Regulatory Guide 1.14 requires that the margin against ductile failure relative to the minimum specified yield strength be 3 and 1.5 at normal operating speed and design overspeed, respectively.
For the flywheels at
'L Ginna the margin against ductile rupture at normal operating speed is 3.7 and 2.33 at a design overspeed of 125>>.
The pump flywheels at Ginna have a wide margin of safety against ductile fracture and the requirements of Regulatory Guide 1.14 have been satisfied.
c.
Inservice Inspection Program Regulatory Guide 1.14 requires that an inservice inspection program for each flywheel should include the following:
a.
In-place ultrasonic volumetric examination of the areas of higher stress concentration at the bore and keyway at approximately 3 year intervals, during the refueling or maintenance shutdown ss required by the ASHE Code Sec.ion XI.
b.
A surface examination of all exposed surfaces and complete ultrasonic volumetric examination at approximately 10'year intervals, during plant shutdown coinciding with the inspec ion schedule as required by the ASNE Code Section Xi.
The present inser vice inspection program for Ginna consists of complete ultrasonic volumetric examination and surface examination of all exposed surfaces at approximately 10-year intervals, during plant shutdown coinciding with the inspection schedule, and inplace ultrasonic volumetric examination Z
of the areas of higher stress concentration at the bore and keyway at approximately 3 year intervals.
The Ginna pump flywheels do meet the requirements of the Regulatory Guide 1.14.
d.
Independent Staff Fracture Mechanics Evaluation The staff has performed an independent -fracture mechanics evaluation to determine the speed at which unstable crack propagation would occur for a 4 inch crack emanating from the key way.
The results of the fracture mechanics evaluation show that a 4 inch crack would remain stable at speeds upto 3000 rpm.
Based on the results of this analysis and other analyses performed at design overspeed for similiar plants, the staff concludes that a very large crack, on the order of 10 inches, would re-main stable at a design overspeed of 1486 rpm.
VI.
CONCLUSIONS Me have reviewed the material,.fabrication, design and inspection aspects of the pump flysheels at Ginna for compliance with the Regulatory Guide 1.14.
Me conclude that the requirements for fabrication and the margins against flaw induced fracture and yielding, required by the Regulatory Guide 1.14, have been satisfied for the flywheels.
The present inservice inspection program of the Ginna pump flywheels does meet the requirements of the Regulatory Guide 1.14.
Compliance with:the Guide provides a basis acceptable to the staff for satisfying, in part, the, requirements of General Oesign Criteria 4, "Environmental and Missile Oesign Bases".
TOPIC III-lO,C SEE TOPIC II 4 E
TOPIC III-ll III-12 SEE TOPIC II-2. 8 h
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50-10 50-237 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 AUG 17 g~
~wow QP.
ld CorJC4.6J~'ommonwealth Edison Company ATTt3:
Mr. Cordell Reed Assistant Yice President Post Office Box 767 Chicago, Illinois 60690 GentlemIen:
At our meeting with you on May 31,
- 1978, we indicated that our review of several SEP topics was essentially complete.
We also stated that completed topic assessments would be sent to you for information and review and would be placed in the Public Document Rooms.
Our initial evaluation of eight of these essentially complete topics is enclosed.
You are requested to carefully examine the facts upon-which the staff has'ased its evaluation arid respond either by confiming that the facts defining your plants are correct, or by identifying any errors.
If in error, please supply corrected information for the docket.
We encourage you to supply any other material for the docket related
'to these topics that you believe to be helpful.
At the May 31 meeting, the SEP Owners Group requested clarification of SEP documentation procedures and made several suggestions in that regard.
Enclosure 1 is our response to the request and suggestions.
It contains the documentation procedures to be used throughout the SEP program and discusses the bases for these procedures..
Our documentation of the eioht essentially complete topics in Attachment 1 illustrates the docurI>entation procedure to be used.
We woul.d appreciate any comments you may have to improve documentation of topic assessments.
V Er.cl osure:
Response
to the SEP Owners Group Suggestions SXJ Darrell G. Eisenhut, Assistant Director
~
~
for Systems 8 Projects Division of Operating Reactor s
Commonweal th Edison Company CC Hr. John M.
Rowe Isham, Lincoln 5 Beale Counselors at Law One First National
- Plaza, 42nd Floor Chicago, Illinois 60603 Hr.
B. B. Stephenson Plant Superintendent Dresden Nuclear Power Station Rural Route 81 Morris, Illinois 60450 Anthony Z. Roisman Natural Resources Defense Council 917 15th Street, N.
M.
Mashington, D.
C.
20005 I)orris Public Library 604 Liberty Street Horris, Illinois 60451
.TOPIC IV 1A - Operation with les's than all loops in service SEP Plants Affected - PMR's and BMR's OBEs Affected - Loss-of-Coolant Accident Discussion The majority of the presently operating BMRs and PMRs are designed to operate with less than full reactor coolant flow. If a PMR reactor coolant pump or a BMR recirculation pump becomes inoperative, the flow provided by the remaining loops is sufficient for steady state operation at a power level less than full povier.
Plants authorized for long term operation with one reactor coolant
')
pump out of service have submitted, and the staff has approved, the necessary ECCS, steady state, and transien calculations.
PKR and BMR licensees have Technical Specifications which
.reactor shutdown within a fairly short time if one of the loops becomes inoperable (with the exception of trio which below).
The remaining require a
'perating are discussed SEP APPLICABILITY The docketed material for the ll systematic evaluation program plants has been reviewed with respect to operation with less than all loops in service.
One licensee (Dresden
- 2) has requested authorization to operate with less than all loops in. service, the staff is reviewing the analyses submitted with the request and approval will be granted vihen the staf approves the analysis.
Five facilities (Yankee Rowe, Hillstone 1, Ginna, Palisades,and San Onofre) are not authori ed to operate with
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v r 1ess than all loops in service, Technical Specifications restrict this mode to. a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at which time the facility must have the idle loop'estored to: service or"shutdown.
Three facilities (Connecticut
- Yankee, Oyster Creek, and Dresden
- 1) have had an analysis reviewed and approved by the sta f which authorizes H-1 loop operation.
Two facilities LACBtiR and Big Rock Point) have had authorization to operate in the N<<l loop mode since they were licensed, however there is no supporting ECCS analysis to justify operation.
Conclusion This topic is complet for all the SEP facilities with the exception of LACBtlR and Big Rock Point, for the latter two if continued authorization is to be permit ed an analysis will have to be submitted which describ s
the thermal-hydraulic conditions of N-1 loop operation during ECCS, s.eady
- state, and transient conditions.
Until such an analysis s perfo...ed and approved.
Operation with less than all loops in service should be restricted to a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period at which time the plant should be shutdown unless the idle loop has been made operable.
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~O ky*y4 UNITEO STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 29, 1979 Docket No. 50-244 Mr. Leon D. White, Jr.
Yfce President Electric
& Steam Production Rochester Gas
& Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Mr. White:
RE:
TOPIC IV-1.A - R.
E.
GINNA NUCLEAR POWER PLANT Enclosed fs a copy of our. revised safety assessment of Topic IV-1.A, Operation With Less Than All Loops In Service.
This revision includes consideration of the comments received on the assessment issued by our letter dated February 6,
3979.
Your letter dated February 21,
- 1979, provided comments on the assessment.
This revision completes our assessment of Topic IY-1.A which will be used as input to the integrated review of the Ginna Plant.
If there are any errors in the facts of this revised assessment, please supply corrected information within 30 days of the date you receive this letter.
If no response fs received within that time, we will assume that you have no further comments or corrections.
Sincerely,
Enclosure:
Revised Assessment for Topic IV-1.A cc w/enclosure:
See next page Dennis L. Ziemann, Chief Operating Reactors Branch d2 Division of Operating Reactors
SYSTEMATIC EVALUATION PROGRAM TOPIC IY>>I-A:
Operation with less than all loops in service PLANT:
R.
E, Ginna Nuclear Power Plant Discussion The majority of the presently operating BNR's and PMR's are designed to permit operation with less than full reactor coolant flow, That is, if a PMR reactor coolant pump or a BMR recirculation pump becomes inoperative, the flow provided by the remaining loop or loops is sufficient for steady state operation at some definable power level, usually less than full power.
Plants authorized for long term operation with one reactor coolant pump out of service have submitted, and the staff has approved, the necessary ECCS, steady state, and transient analysis.
The remaining PMR and BMR licensees have Technical Specifications which require reactor shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one of the operating loops become inoperable and cannot be returned to operation within the time period, Evaluation The docketed material for the Ginna Nuclear Power Plant has been reviewed with respect to operation with less than all loops tn service.
The Ginna Technical Specifications [3,1,l.l.c,(i) and 3.l,l,l.c,(.$ $ )3.
permit operation with less than all loops in service up to 8,5 per cent (130Mwt) of full power provided that a predetermined shutdown margin ts maintainable.
The supporting analysis was provided by letter dated September 22,
- 1975, and included specific shutdown margin requirements depending upon whether one or both primary coolant loops were operable,
Mr. Leon D. Mhite, Jr.
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~ May 29, 1979 CC Lex K. Larson, Esquire
- LeBoeuf, Lamb, Leiby 5 MacRae 1757 H Street, N.
M.
Mashington, D.
C.
20036 Mr. Michael Slade 1250 Crown Point Drive
- Webster, Hew York 14580 Rochester Committee for Scientific Information Robert E. Lee, Ph.D.
P. 0.
Box 5236 River Campus Station Rochester, Hew York 14627 jeffrey Cohen Hew York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- Albany, New York 12223 Director, Technical Development Programs State of New York Energy Office Agency Building 2
'Empire State Plaza
In a safety evaluation supoorting'Amendment No.
10 to the Provisional Operating License, dated March 30,
- 1976, the staff approved the shutdown margin ana ys r in analysis and issued the proposed Technical Specifications.
Current staff criteria require that, unless an ECCS analysis is per-formed and approved, a reactor must be shut down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one of the operative loops becomes inoperable and cannot be returned to operation within that time period, The Ginna Technical Specifi-cations state that if the appropriate shutdown margin cannot be maintained the reactor will be brought to a hot shutdown condition until the proper shutdown margin can be established, There is no time limit on how long the plant can remain at 8,5 per cent power in the n-I loop configuration,
- However, based on the low power level, it is our judgment that the amount of stored energy in the fuel and the decay heat generated after shut down following a postulated loss of coolant accident (LOCA) will be sufficiently reduced to assure that peak clad temperatures are less than those calculated jn the ECCS performance analyses.
This is based on the fact that peak clad temper-atures are strongly affected by the stored energy of the fuel and the decay heat.
If the power level is reduced to less than 10 per cent of full power, the stored energy in the fuel is proportionaIly reduced resulting in peak clad temperatures significantly below those calculated for the accidents at full power, Fuel burnup and the cIad gap affect the relationship between power level and stored energy in the fuel but these are secondary effects so that reducing power leve y
1 b
a factor of 10 reduces the stored energy by approximately 10 with a substantial decrease in calculated peak clad temperature during the
- LOCA, Me find that although the restrictions at Ginna are different than those in our criteria, the difference is of no significant consequence.
Therefore, we conclude that continued operation in this mode is
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Wy*~4 Docket No. 50-244 LSOS05-045 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 May 26, 1981 Nr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Mr. Maier:
SUBJECT:
SEP TOPIC IV-2, REACTIVITY CONTROL SYSTEMS - R.
E.
GINNA NUCLEAR POWER PLANT Me have enclosed our final staff evaluation for SEP Topic IV-2.
The revised report now includes the references provided in your April 29, 1981 letter.
Sincerely,
Enclosure:
Topic IV-2 Final Report cc w/enclosure:
See next page Dennis M. Crutchfield, C
ef Operating Reactors Branch No.
5 Division of Licensing
John E. I'1aier y
H Voigt, Esquire
- euf, Lamb, Leiby and NacRae
~ New Hampshire Avenue, N.
W.
e 1100
- ington, O.
C.
20036 Oi 0=
l1ichael S lade railwood Circle
- ester, New York 14618 Bialik
~ant Attorney General E1vnmental Protection Bureau
~eiik State Department of Law 2 i>r Trade Center "en Y., New York 10047 jeffreCohen Ne!v Yo State Energy Office S<<an S'et Building Core l~econd Floor
+'<<:ate Plaza any (ew York 12223
'rector Technical Oevelopment Progra;s Sta.a"e of New York Energy Office "gency Bt.ilding 2 mPire State Plaza
" bany Mew York 12223 Rochester Public Library 115 South Avenue Rochester, Neiv York 14604 Supervisor of the Town of Ontario 107 Ridge Road West
- ontario, New York 14519 Resident Inspector R.
E. Ginna Plant c/o U.
S.
NRC 1503 Lake Road Ontario New York 14519
'Wa U.
Re AT 26 Neo Her Atc U-Was Or.
Ato U.
Was Or.
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Nati 1725 Suit Wash Ezra Assi Envi New 2
Wo New '
FINAL SAFETY EVALUATIOII REPORT SEP TOPIC IV-~2 RFACTIVITY COtfTROL SYSTEMS It~CLUDIfiG FUhCTIOHAL DESIGN AHD PROiECTIOH AGAINST SINGLE FAILURES R.E.
GItIHA NUCLEAR PO'iER PLANT DOCKET HO. 50-244 I.
I HTRODUCTION The purpose of this evaluation is to insure that the design basis for the Ginna reactivity control systems is consistent with analyses perfom d
to verify that the protection system meets General Design Criterion 25.
General Design Criterion 25 requires that the reactor protection system be designed to assure that specified acceptable fuel design limits are not exceeded for any single mal function of the reactivity control systems, such as accidental withdrawal of control rods.
Reactivity control systems need not be single failure proof.
However, the protection system must be capable of assuring that acceptable fuel design limits are not exceeded in the event of a single failure in the reactivity contro'l systems.
The re-view criterion, covered in this evaluation, is addressed in Section II.
Review areas that are not covered, but are related and essential to the completion of this topic, are 'covered by other SEP topics addressed in Section III. 'he scope of the SEP topics is defined in the "Report on the SystemIatic Evaluation of Operating Facilities" dated November 25, 1977.
This report is limited to the identification.and evaluation of inadvertent control rod withdrawals and mal positioning of control rods which.may occur as a result of single failures in the electrical circuits of the reactivity control systems.
II. REVIEM CRITERION The review criterion for his topic is based upon Section 7.7, Part II of the NRC Star: ard Review Plan.
In the specific case of the reactivity con-trol systems a single failure shall not cause plant conditions more severe than those Sr>>hich the reactor protection system is designed.
III.
RELATED SAFETY TOPICS The following listed review areas are not covered in this report, but are related and essential to the completion of this topic.
These review areas are covered by other SEP topics as, indicated below.
1.
Analyses of the consequences of control rod withdrawals and the malpositioning of control rods which may occur as a result of single failures in the electrical circuits of the reactivity control systems are covered by SEP Topic XY-8, "Control Rod
~
Yiisoperation (System Halfunction or Operator Error)"
2.
Analyses of reactivity insertions occurring as a result of inadvertent boron dilutions are covered in SEP Topic XY-10, "Chemical and Volume Control System Halfunction that Results in a Decrease in Boron Concentration in the Reactor. Coolant."
IV.
REVIEW GUIDELINES The purpose of this evaluation is to identify inadvertent control rod withdrawals and malpositioning of control rods which may occur as a
result of single failures in the electrical, circuits of the reactivity control systems for the R.E. Ginna Nuclear Power Plant.
j V.
EYALUATION Information was provided in Rochester Gas and Electric Corporation letter dated January 19, 1981, describing design features which limit control rod withdrawals and malpositioning of control rods caused by failures within the reactivity control systems at the R.E.
Ginna Nuclear
Power Plant.
Based upon the information provided by the licensee we conclude that the following may occur as a result of single failures:
1)
Two control rod banks may be simultaneously withdrawn.
2)
Two banks may overlap at other than the design value.
This conclusion is based upon the availability of alarm and interlock circuits associated with the rod control system such that certain consequential effects of single failures within the rod control system are precluded by the operability of these interlocks and alarms.
The basis for the assumption that these alarms and interlocks will be opera-ble is that a failure in the alarm and interlock circuits will be identified and corrected during routine maintenance or as a result of system fault investigation.
The effects of single failures occurring after an undetected failure has occurred in the alarm and interlock system are not included in the evaluation.
,This"is consistent with the basis used for plants currently under operating'icense review.
VI." CONCLUSION Each of the following two reactivity control system malfunctions have beenddressed as part of SEP Topic ZY-S, Control Rod Nisoperation, to verify that specified acceptable fuel design limits are not exceeded:
1)
Simultaneous withdrawal of two control rod banks.
2)
Overlap of two banks at other than the design value.
Fuel design limits are not exceeded for either of the above two mal-functions and thus, General Design Criterion 25 is met insofar as electrical failures within reactivity control systems are concerned.
VII.
REFERENCES 1)
Technical Supplement Accompanying Appli'cation to Increase
- Power, Section 14, February 1971.
.. 2)
MCAP-7778, Solid State Rod Control System - Full Length, by A. Blanchard and 0.
N. Katz.
3)
WCAP-8976, Failure Mode and Effects Analysis (FHEA) of the Solid State Full Length Rod Control
- Sequence, by M.
E.
Shopsky.
4)
Letter dated January 19, 1981 from J.
E. Naier (RG5E) to D. fI. Crutchfiold (ORB>5).
TOPIC IV-3 SEE TOPIC II-2.8
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,'Sg 9 gO LS05-81-11-069 UNITED STATES NUCLEAR REGULATORY COMMISSION IVASHINGTON,0 C. 20555 November 2i, 1981 LETTER TO ALL SEP LICENSEES Gentlemen:
SUBJECT:
TOPIC V-l, COMPLIANCE WITH CODES AND STANDARDS (10 CFR 50.55a}
This is to advise you that Topic V-l, Compliance with Codes and Standards (10 CFR 50.55a) has been deleted as a
SEP topic.
The purpose of Topic V-1 was to review the licensee's inservice inspection and testing programs.
for Class 1,
2 and 3, pressure
- vessels, piping, pumps and valves and other safety related components to assure compliance with ASME Code,Section III and XI as required by 10 CFR 50.55a, to assure that the integrity of components is maintained throughout service life.
Topic V-1 is a generic item being resolved under NRR Generic Items A-01, Inservice Inspection (ISI) and A-14, Inservice Testing (IST).
In the Fall of 1976 the NRC approved a
new regulation 10 CFR 50.55a, Inservice Inspection Requirements.
10 CFR 50,55a required that all plants prepare an ISI and IST program to the requirements of the 1974 Edition of Section XI ASME Boiler and Pressure YesseI Code.
The program was to be updated every 40 months for ISI and every 20 months for IST.
In the Fall of 1979, the regulation was changed to require that the ISI and IST program be updated every 120 months.
The basis for this change was that changes between the original and updated programs were found to be minor.
As of December 3,
1981 all SEP plants will have at least completed their first interval (120 months).
The ISI/IST programs are routinely reviewed by the NRC.
P.est review experience has shown that only minor hardware modifications, if any, result from these reviews.
Since the prnI.ram mainly results in changes to the technical specifications, and the'se will be implemented indepen-dently of the
- SEP, the program review need not be considered in the integrated assessment, Therefore, we have deleted the topic from the
- SEP, Sincerely, cc; See next page Dennis M. Crutchfield, Chief Operating Reactors Branch No.
5 Division of Licensing
Mr. David P. }{offman (Big Rock Point and Palisades)
CC Mr. Paul A. Perry, Secretary Consumers Power Company 212 West Michigan Avenue
- Jackson, Michigan 49201 Joseph Gallo, Esquire Isham, Lincoln 5 Scale 1120 Connecticut Avenue Room 325 Washington, D. C.
20036 Peter W. Steketee, Esquire.
505 Peoples Building Grand Rapids, Michigan 49503 Alan S. Rosenthal, Esq.,
Chairman Atomic Safety 4 Licensing Appeal Board U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 Mr. John O'eill, II Route 2, Box 44 Maple City, Michigan 49664 Charlevoix Public Library 107 Clinton Street Charlevoix, Michigan Chairman County Board of Supervisors Charlevoix County Charlevoix, Michigan 49720 Office of the Governor (2)
Room 1 - Capitol Building Lansing, Michigan 48913 Herbert Semmel Counsel for Christa Maria, et al.
Urban Law Institute Antioch School of Law 2633 16th Street, NW Washington, D.
C.
20460 Mr. David P.
Hoffman Nuclear Licensing Administrator Consumers Power Company 1945 W Parnall Road
- Jackson, Michigan 49201 U. S. Environmental Protection Agency Federal Activities Branch Region V Office "
ATTN:
Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Oscar H. Paris Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.. C.
20555 Mr. Frederick J.
Shon Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Big Rock Point Nuclear Power Plant ATTN:
Mr. C. J.
Hartman Plant Superintendent Charlevoix, Michigan 49720 Christa-Maria Route 2, Box 108C Charlevoix, Michigan 49720 William J.
- Scanlon, Esquire 2034 Pauline Boulevard Ann Arbor, Michigan 48103 Resident Inspector Big Rock Point Plant c/o U.S.
NRC RR g3, Box 600 Charlevoix, Michigan 49720 Mr. Jim E. Mills Route 2, Box 108C Charlevoix, Michigan 49720
Mr. David P. Hoffman CC Or. John H. Buck Atomic Safety and Licensing Appeal Board U., S. Nuclear Regulatory Coranission Mashington, 0.
C.
20555 Ms. JoAnn Bier 204 Clinton Street Charlevoix, Michigan 49720 Thomas S. Moore Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, 0. C.
20555
Mr. L. DelGeorge (Dresden 1
and 2)
CC Isham, Lincoln 8 Scale Counselors at Law One First National Plaza, 42nd Floor Chicago, Illinois 60603 Mr. Doug Scott Plant Superintendent Rural Route
<1 Morris, Illinois 60450 U. S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station RR 81 Morris, Illinois 60450 Mary Jo Murray Assistant Attorney General Environmental Control Division 188 W. Randolph Street Suite 2315 Chicago, Il1inoi s 60601 Morris Public Library 604 Liberty Street Morris, Illinois 60451 Chairman Board of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 Ill.inois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 lj. S. Environmental Protection Agency
'ederal Activities Branch Region V Office ATTN:
Regional Radiation Representative 230 South'Dearborn Street Chicago, Illinois 60604 The Honorable Tom Corcoran United States House of Representatives Washington, D. C.
20515 John H. Frye, III, Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington D.
C.
20555 Daniel Mintz Counsel for Petitioners
. (Citizens for a Better Environment)
Suite 1600, 59 E.
Van Buren Street Chicago, Illinois 60605 Mr. L. DelGeorge Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690-
Mr. John E. Maier CC Harry H. Yoigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N. M.
Suite 1100 Mashington, D. C.
20036 Nr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 Morld Trade Center New York, New York 10047 New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road Mest
- Ontario, New York 14519 Resident, Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York 14519 Nr. Thomas B. Cochran Natural Resources Defense Council, Inc.
1725 I Street, N. M.
Suite 600 Mashington, D. C.
20006
~ ~
U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Gr ossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington, D. C.
20555 Dr. Richard F, Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comnission Mashington,
- 0. C.
20555 Dr.
Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Conmission Mashington, D. C.
20555 Nr. John E. Maier, Vice President Electric and Steam Production Rochester Gas
& Electric Corporation 89 East Avenue Rochester, New York 14649
Mr. M. G. Counsil (Haddam Neck and Millstone 1) rtc William H. Cuday,
=squire Oay, Berry 8 Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Board of Selectmen Town Hall
- Haddam, Connecticut 06103 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant P. 0.
Box 128 Waterford, Connecticut 06385 Natural Resources Oefense Council 917 15th Street, N. M.
Washington, O. C.
20005 Resident Inspector c/o U. S.
NRC P. 0. Box Orawer KK Niantic, Connecticut 06357 Waterford Public Library Rope Ferry Road, Route 156 Waterford, Connecticut, 06385 First Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Materford, Connecticut 06385 John F. Opeka Systems Superintendent, Northeast Utilities Service Company P. 0.
Box 270 Hartford, Connecticut 06101 Mr. Richard T. Laudenat
- Manager, Generation Facilities Licensing Northeast Utilities Service Company P. 0.
Box 270 Hartfor d, Connecticut 06101-Connecticut 'Energy Agency ATTN:
Assistant Oirector Research and Policy Development Oepartment of Planning and Energy Policy 20 Grand Street Hartford, Connecticut 06106 Resident Inspector Haddam Neck Nuclear Power Station c/o U. S.
NRC East Haddam Post Office East Haddam, Connecticut 06423 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203 Superintendent Haddam Neck Plant RFD fl Post Office Box 127E East Hampton, Connecticut 06424 Mr.
M.
G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Connecticut Yankee Atomic Power Co.
Post Office Box 270 Hartford, Connecticut 061 01
Mr. Frank Linder CCFritz Schubert, Esquire Staff Attorney Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
- 0. S. Heistand, Jr., Esquire Morgan, Lewis 5 Bockius 1800 M Street, N.
W.
Washington, D. C.
20036 Mr. R. E. Shimshak La Crosse Boiling Water Reactor Dairyland Power Cooperative P. 0.
Box 135
- Genoa, Wisconsin 54632 Ms. Anne K. Morse Coulee Region nergy Coalition P. 0.
Box 1583 La Crosse, Wisconsin 54601 La Crosse Public Library 800 Main Street La Crosse, Wisconsin 54601 U. S. Nuclear Regulatory Comission Resident Inspectors Office Rural Route 01, Box 276
- Genoa, Wisconsin 54632 Town Chairman Town of Genoa Route 1
- Genoa, Wisconsin 54632
- Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 Alan S. Rosenthal, Esq.,
Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comaission Washington, D. C.. 20555 Mr. Frederick Milton Olsen, III 609 North 11th Street
- LaCrosse, Wisconsin 54601 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 Mr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Dr. Lawrence R. Quarles Kendal at Longwood, Apt.
51 Kenneth Square, Pennsylvania
'19348 Charles Bechhoefer, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Dr. George C. Anderson Department of Oceanography University of Washington
- Seattle, Washington 98195 Mr. Ralph S. Decker Route 4, Box 190D Cambri dge, Maryland 21613 Thomas S. Moore Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 Mr. George R. Nygaard Coulee Region Energy Coalition 2307 East Avenue
- LaCrosse, Wisconsin 54601 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South
- LaCrosse, Wisconsin 54601
Nr. I.
R. Finfrock CC G. F. Trowbridge, Esquire Shaw, Pittman, Potts and Trowbridge 1800 N Street, N.
M.
Mashington, D.
C.
20036 J.
B. Lieberman, Esquire Berlack, Israels 8 Lieberman 26 Broadway New York, New York 10004 Natural Resources Defense Council 917 15th Street, N.
M.
Washington, O.
C.
20006 J.
Knubel BMR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, Hew Jersey 07054 Deputy Attorney General State of Hew Jersey Department of Law and Public Safety 36 West State Street - CH 112
- Trenton, New Jersey 08625 Ocean County Library Brick Township Branch 401 Chambers Bridge Road Brick Town, New Jersey 08723 Mayor Lacey Township 818 Lacey Road Forked River, New Jersey 08731 Commissioner Department of Public Utilities State of New Jersey 101 Commerce Street
- Newark, New Jersey 07102 U.
S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, Hew York 10007 Gene Fisner Bureau Chief Bureau of Radiation Protection 380 Scotts Road
- Trenton, New Jersey 08628 Commi ssi oner New Jersey Department of Energy 101 Commerce Street
- Newark, New Jersey 07102 Licensing Supervisor Oyster Creek Nuclear Generating Station P. 0.
Box 388 Forked River, New Jersey 08731 Resident Inspector c/o U. S.
NRC P. 0.
Box 445 Forked River, New Jersey 08731 i4lr. I. R. Finfrock Vice President
- Jersey Central Power 5 Light Company Post Office Box 388 Forked River, New Jersey 08731
Mr. R. Dietch CC Charles R. Kocher, Assistant General Counsel James Beoletto, Esquire Southern California Edison Company Post Office Box 800
- Rosemead, California 91770 David R. Pigott Orrick, Herrington 8 Sutcliffe 600 Montgomery Street San Francisco, California 94111 Harry B. Stoehr San Diego Gas 4 Electric Company P. 0. Box 1831 San Diego, California 92112 Resident inspector/San Onofre NPS c/o U. S.
NRC P. 0. Box 4329 San Clemente, California 92672 Mission Viejo Branch Library 24851 Chrisanta Drive Mission Viejo, California 92676 Mayor City of San Clemente San Clemente, California 92672 Chairman Board of Supervisors County of, San Diego San Diego, California 92101 California Department of Health ATTN:
Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 U. S. Environmental Protection Agency Region IX Office ATTN:
Regional Radiation Representative 215 Freemont Street San Francisco, California 94111 Mr. R. Dietch, Vice
.=: sideni Nuclear Engineering and Operations Southern California Edison Company
.2244 Walnut Grove Avenue Post Office Box 800 Rosemead; California 91770
Mr. James A. Kay CC llr. J a...es
"=. Tribi1 e, P r s ident Yankee Atomic Electric Company 25 Research Drive Mestborough, Massachusetts 01581 Greenfield Commnity College 1 College Drive Greenfield, Massachusetts 01301 Chairman Board of Selectmen Town of Rowe Rowe, Massachusetts 01367 Energy Facilities Siting Council 14th Floor One Ashburton Place
'Boston, Massachusetts 02108 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building
- Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o U.S.
NRC Post Office Box 28 Monroe Bridge, Massachusetts 01350 Hr.
James A.
Kay Senior Engineer - Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Mass.
01701
TOPIC Y-2 SEE TOPIC II-4.E
TOPIC V-5 V-0 SEE TOPIC II-Z.B
~g REOy c~
~4 0
~y
~i 0
I O
.t ~
3;
~
f'~*~4 Docket No. 50-244 LS05-82 049 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 February 08, 1982 Hr. John E. Maier, Vice President Electric and Steam Production Rochester Gas
- 5. Electric Corporation 89 East Avenue Rochester, New York 14649
Dear Mr.. Haier:
SUBJECT:
SEP TOPIC V-5, REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION, R.
E.
GINNA NUCLEAR POWER PLANT Enclosed is a copy of our final evaluation of SEP Topic V-5 for the R.
E. Ginna Nuclear Power Plant.
This assessment compares this facility, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for licensing new facilities.
This revised evalua-tion factors in the information contained in your Harch 23, October 12 and October 20, 1981 letters on this subject, pertinent information from SEP Topic V-10.A and available 10 CFR 50, Appendix I submittals for R.
E, Ginna..
This evaluation concludes that the R.
E. Ginna reactor coolant pressure boundary leakage detection systems do not presently conform to all of the recommendations of Regulatory Guide 1.45 and presents the modifica-tions needed to establish compliance, The necessity for implementation of the modifications will be considered during the integrated assessment.
This assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed, Sincerely, Enclosure; As stated Dennis H, Crutchfield hief Operating Reactors Branch No.
5
~ Division of Licensing cc w/enclosure:
See next page
Mr. John E. Maier
~ '*
..Harry H. Yoigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae
'1333 New Hampshire Avenue, N.
W.
Suite 1100 Mashington, D. C; 20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau NNew York State Department of Law 2 Morld Trade Center New York, New York 10047 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York
]4519 Director, Bureau of Nuclear Operations State of Hew York Eneray Office Aaency Building 2 Empir e State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, Hew York 14604 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radi ation Representativ 26 Federal Plaza Hew York, Nhw York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Mashington, D. C.
20555 James P.. O'Reilly, Regional 4dministratr, Nuclear Regulatory Commission, Region Ii Office of Inspection and Enforcement 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Supervisor of the Town of Ontario 107 Ridge Road Mest
- Ontario, New York 14519 Or.
Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Or. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
20555
SYSTBQTIC EVALUATION PROGRAM TOPIC V-5 R.
E.
GINNA TOPIC V-5, Reactor Coolant Pressure Boundary Leakage Detection INTRODUCTION The safety objective of Topic V-5 is to determine the reliability and sensitivity of the leak detection systems which monitor the reactor coolant pressure boundary to identify primary system leaks at an early stage before failures occur.
<,)
REVIEW CRITERIA The acceptance criteria for the detection of leakage from the reactor coolant pressure boundary is stated in the General Design Criteria of Appendix A, 10 CFR Part 50.
Criterion 30, "guality of Reactor Coolant Pressure Boundary,"
requires that means shall be provided for detecting and, to the extent practi-cal, identifying the location of the source of leakage in the reactor coolant pressure boundary.
REVIEW GUIDELINES The acceptance criteria are described in the Nuclear Regulatory, Commission
. Standard. Review Plan Section 5.2.'5,'Reactor Coolant. Pressure Boundary Leak-.
age Detection."
The areas of the Safety Analysis Report and Technical Specifications are reviewed to establish that information submitted by the licensee is in compliance with Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"
- IV, EVALUATION Safety Topic V-5 was evaluated in this review for compliance of the infor-mation submitted by the licensee with Regulatory Guide 1,45, "Reactor Coolant Pressure Boundary Leakage Detection Systems."
The information in the Safety Analysis Report, Technical Specifications, the January 30, 1979 letter from RGSE to the NRC regarding SEP Topic V-10.A, the March 23, October 12 and October 20, 1981 letters from RG&E to the NRC regarding SEP Topic V-5, and the available 10 CFR 50, Appendix I review information for R, E. Ginna as well as the other information listed in the references was reviewed.
Regulatory Guide 1.45 recommends that at least three separate detection systems be installed in a nuclear power plant to detect an unidentified leakage from the reactor coolant pressure boundary to the primary containment of one gallon per minute within one hour.
Leakage from identified sources must be isolated so that the flow rates may be monitored separately from unidentified leakage.
The detection systems should be capable of performing their functions following certain seismic events and capable of. being checked in the control room.
Of the three separate leak detection methods required, two of the methods should be (1) sump level and flow monitoring and (2)
airborne particulate radioactivity monitoring.
The third method may be either monitoring of condensate flow rate from air coolers or monitor ing of airborne gaseous radioactivity.
Other detection
- methods, such as humidity, temperature and pressure, should be considered to be alarms of indirect indication of leakage to the containment.
In addition, provisions should be made to monitor systems interfacing with the reactor -coolant pressure boundary for signs of intersystem leakage through methods such as radioactivity and water level or flow monitors.
Plant incorporated systems and their corresponding features are tabulated in Enclosure 1.
Detailed guidance for the leakage detection system is contained in Regula-.
tory Guide 1.45.
Based upon our review of the referenced documents and the summaries presented in Enclosure 1,
we have determined:
1)
That the types of systems employed for the detection of leakage from the reactor coolant pressure boundary to the containment consist of the minimum three recommended by Regulatory Guide 1.45.
However, not one of the recommended systems have been sufficiently documented such that it meets all of the acceptance criteria of SRP 5.2.5 (See Table 1 for further details>.
2}
Table 1 lists leakage detection systems in addition to the minimum three systems recommended by Regulatory Guide 1.45.
If credit is to be taken for these systems in the evaluation of the adequacy of the entire leakage detection
- system, then these systems will have to be demonstrated to meet the criteria of the Guide.
3)
Provisions have been incorporated to monitor reactor coolant in-leakage to systemswhich connect to the reactor coolant pressure boundary.
However, from the review of the referenced information it is not clear that Table 2 includes all of the systems which connect to the reactor coolant pressure boundary.
Also, these systems have not been demonstrated to be able to withstand seismic events up to the OBE level.
4).
The Ginna Technical Specification
- 3. l. 5.3 does impose requirements concerning the operability of the leakage detection systems to monitor leakage to the primary containment, as recommended by Regulatory Guide 1.45.
However, the Technical Specification requirements do not conform to those given in current Standard Technical. Specification 3/4.4. 6.
In addition, corresponding surveillance requirements in the Standard Technical Specifications at e not contained in the Ginna Technical Specifications (Table 4.1-1).
5)
Information concerning the use of reactor coolant inventory balances, as indicated in Table 3, for detection of RCPB leakage is incomplete, therefore, its contribution to overall system effectiveness cannot be determined."
V.
CONCLUSIONS Our review indicated that the systems employed at R.
E. Ginna to measure reactor coolant pressure boundary leakage do not meet all of the recomnen-dations given in Regulatory Guide 1.45.
Specifically, our review concludes that:
1}
Although all of the recommended types of leakage detection systems for measurement of leakage from the reactor coolant pressure boundary to the containment have been incorporated in the facility, the systems do not meet all of the sensitivity, operability or surveillance criteria.
2)
Information concerning the leakage detection systems for the detection of inter-system reactor coolant pressure boundary leakage is incomplete.
Therefore, we cannot determine the extent to which Regulatory Guide 1.45 is met.
3}
Standard Technical Specification 3/4,4. 6 and the corresponding sur-veillance requirements concerning the operability of the reactor coolant pressure boundary to the containment leakage detection systems should be added to the R. E; Ginna Technical Specifications.
- Also, the current basis for Ginna Technical Specification
- 3. 1. 5. 3 and FSAR should be revised to state that the sensitivities of the reactor coolant pressure boundary to containment leakage detection systems.
4}
Information concerning the use of the primary coolant system inventory balance leak rate sensitivity and time required to achieve sensitivity is incomplete.
Therefore, we cannot determine the contribution of this technique to the overall leak detection sensitivity.
The necessity for any leakage detection system modifications will be considered during the integrated safety assessment.
1
~ I REFERENCES Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"
May 1973.
2.
R, E. Ginna FSAR, Sections 4.2 and 11.2.
3.
R.
E. Ginna "Supplement 1 to Technical Supplement Accompanying Applica-tion for a Full-Tenn Operating License, December 20, 1972, response to questions 7c and 10.
4.
Westinghouse Standard Technical Specifications, NUREG-0452, Revision 4.
5.
R.
E. Ginna Technical Specifications.
~ (
6.
7.
8.
9.
10.
12.
13.
Letter from John E. Maier, RG&E, to Boyce H, Grier, USNRC, Region 1, IE Bulletin 80-24, dated 12/31/80.
Letter from John E. Maier, RG&E, to Boyce H. Grier, USNRC, Region 1, LER 81-004, February 25, 1981.
Draft SEP Review of Topic IX-3, Station Service and.Cooling Water Systems, by letter from Crutchfield to Maier, June 23, 1981.
Letter from L. D. White, Jr.,
RG&E, to Dennis L. Ziemann,
- NRC, SEP Topic V-10.A, February 2,
- 1979, Letter from Dennis M. Crutchfield, NRC, to John E. Maier, RG&E, SEP Topic V-5, March 10, 1981.
Letter from John E, Maier, RG&E, to Dennis M, Crutchfield,
- NRC, SEP Topic V-.5, March 23, 1981.
Letter from Dennis M. Crutchfield, NRC, to John E. Maier, RG&E, SEP Topic V-5, July 22,
- 1981, Regulatory Guide 1.29, "Seismic Design Classification," Revision 3, September 1978.
.Table I:
REACTO COOLAHT RESSU E BOUHOARY LEAKAGE DETECTIOH SYSTEHS Re ulator Guide 1.45 Re ufrements P Iant'~~IIDM RCPB to Conta'ament System 1)
Sump Leva'onftorlng (Inventory) l 2)
Sump Pump Actgatlons Honftorfn
-- Time Haters 3)
Airborne ~articulate Radfoacti ft Honftorfn 4)
Alrborng "seous Radfoactf ft Ifonftorfn 5)
Condenst t 'low Rate from Afr coolers 6)
Contains>>
". Atmosphere Pressure..snftorfn 7)
Contaicf>>.. Atmosphere ltumldlt Honftorfn 8)
Contafn i< st Atmosphere
~ann eralnre aaanainrln g)
CVCS Incorporated Yes Yes Yes Yes Yes Yes Yes Yes Yes Leak Rate Sensitivity I gpm (2) 2-10 gpm 1-30 gpm HA 2-10 gpm 0.25 gpm me eq to Achieve Sensitivity ttA 1 hr.-,
I hr, 1 hr.
I hr.
ar qua e
or lthfch. Functf on Is Assured IIA SSE Control Room Indication For Alarms d" Indicators Yes Yes Yes Yes Yes Yes Ho Yes ocument-ation Ref-erence esa e
During H>r ~
mal Operst on Yes Yes Yes Yes Yes Yes Yes
~
~
IIA ~ Hot available; (1)'See References 2, 3, 5, and 7; (2) 0.013 gpm wfthfn twenty minutes assumfng the presence of corrosion products per Technical Specification 3.1.5.3.
I h
I.i 7
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REACTO COOLAHT RESSU E BOUIIOARY LEAKAGE OETECTIOM SYSTEHS Re ulator Guide 1.45 Re ulrements Plant:
~teters stat~assess e
Sys tous: Mhlch Interface 3r/ RCPB
) Secondary System
- 2) Secondary System
- 3) Component Cooling Mater System
- 4) Component Coo'ling Mater System Methods to lleasure RCPB In-Leaka e
on ensate r
Ejector Radla-BloMdown Monitor Surge Tank Level Ra lation lonltor Leak Rate Sensltivlty 0.0025 gpm~
0.16 gpm me eq to Achieve Sens1 tlvlty 1 hr.
art qua e
or Mhlch Function Is Assured RA Control Room 1ndlcatlon For Alarms A.lndlcators Yes Yes Yes cument-atlon Ref-erence Reference 2
Reference 8
References est 3
e Durl3g Mor-mal 3peratlon Yes Y s Yes
~ Total Leakage of 0.5 gallons necessary for lndlcatlon, a
r I el
~ n v,~ >u Iv e
'r,
.;.Table 3!
-,ed Rtd Inventor Retenen I
- . Leak Rate Sensltlvlty orrespond ng Time Required to
. Achieve Swsl tlvlty REACTO COOLANT RESSU E BODNDARY lEAKAGE DETECTION SYSTE Re nlntol Gntdn 1.45 Retntrenente P)ant:
. 'Normal 'Imentory Check L.. Instrument ation Required Nlth Corresponding Location:
Earthquake For Mhlch Instruaentatlon Oar<hare Functlonlng Is Assured:
Testable During Normal Operation:
Oocmnentatlon
Reference:
e'
~ e I!
Ie 4
v e