ML19308A266
| ML19308A266 | |
| Person / Time | |
|---|---|
| Site: | 05000580, 05000581 |
| Issue date: | 01/25/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19296A210 | List: |
| References | |
| NUREG-0423, NUREG-0423-S01, NUREG-423, NUREG-423-S1, NUDOCS 7902210114 | |
| Download: ML19308A266 (71) | |
Text
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j Safety 5%"
i Evaluation Itcpori neausator[c'c$Y0N05 related to construction of Officegf N c ar i Erie Nuclear Plant, y Units 1 and 2
"** " "; E" l":l""
on Ohio Edison Company, et al
,g,g Supplement No.1 ye
Available from National Technical Information Service Springfield, Virginia 22161 Price: Printed Copy $6.00 ; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.
T F
NUREG-0423 Supplement No. 1 January 25, 1979 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF OHIO EDIS0N COMPANY TULEDO EDISON CJMPANY CLEVELAND ELEC7RIC ILLUMINATING COMPANY DUQUESNE LIGHT COMPANY AND PENNSYLVANIA POWER COMPANY ERIE NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. STN 50-580 AND STN 50-501
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TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
AND GENERAL DISCUSSION......................
1-1 1.1 Introduction.......................................
1-1 1.11 Outstanding Items..................................
1-2
- 3. 0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS...
3-1 3.2 Classification of Structures, Systems and Components.........................................
3-1 3.2.2 System Qual ity Group Cl assi ficat ion........
3-1
- 3. 5 Missile Protection.................................
3-1 3.5.1 Missile Selection and Protection Criteria..
3-1 3.8 Desi gn of Se ismic Ca tegory I St ructures............
3-2 3.8.1 Concrete Containment.......................
3-2
- 5. 0 REACTOR COOLANT SYSTEM...................................
5-1
- 5. 4 Component and Subsystem Design.....................
5-1
- 5. 4. 3 D e cay He a t Remo va l Sy s t em..................
5-1
- 6. 0 ENGINEERED SAFETY FEATURES...............................
6-1 6.2 C o n t a i nme nt Sy s t em s................................
6-1
- 6. 2.
Contai nment Funct ional Des i gn..............
6-1 6.2.3 Co nta i nment Is ol at ion System...............
6-2 6.2.4 Combusti ble Gas Cont rol Syst em.............
6-2 6.2.5 Containment Leakage Testing Program........
6-3
- 6. 3 Emergency Co re Cool i ng Sys t em......................
6-5 6.3.5 Post Loss-of-Coolant Accident Baron Precipitation..............................
6-5 i
TABLE OF CONTENTS (Cont'd)
PAGE 6.5 Engineered Safety Feature Atmosphere Cleanup Systems.....................................
6-6 6.5.2 Control Room Habitability Provisions......
6-6 7.0 INSTRUMENTATION AND CONTR0LS............................
7-1 7.3 Engineered Safety Features Actuation System.......
7-1 7.3.1 Periodic Testing of Engineered Safety Feature Systems...........................
7-1 7.3.3 Feedwater Isolation System................
7-1 8.0 ELECTRIC POWER SYSTEMS..................................
8-1 8.2 Offsite Power Systems.............................
8-1 8.2.1 Offsite Routing and Distribution..........
8-1 8.2.4 Conformance With Acceptance Criteria......
8-2 9.0 AU X I L I AR Y S Y STE:iS.......................................
9-1 9.1 Fuel Storage and Handling.........................
9-1 9.1.3 Spent Fuel Pool Cooling and Cleanup System...................................
9-1 13.0 CONDUCT OF OPERATIONS...................................
13-1 13.1 Organizational Structure of Applicants............
13-1 15.0 ACCIDENT ANALYSES.......................................
15-1 15.2 Postulated Accidents..............................
15-1 15.E.1 Spectrum of Steam Piping Breaks Inside and Outside Containment.................
15-1 17.0 QUALITY ASSURANCE.......................................
17-1 17.2 Ohio Edison Company...............................
17-1 17.2.1 Organization..............................
17-1 ii t
TABLE OF CONTENTS (Cont'd)
PAGE 18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS..
18-1 20.0 FINANCIAL QUALIFICATIONS................................
20-1
21.0 CONCLUSION
S.............................................
21-1 APPENDICES APPENDIX A REPORT OF THE ADVISORY COMMITTEE ON REACTOR SA F E G U AR D S.....................................
A-1 APPENDIX B CONTINUATION OF THE CHRONOLOGY OF REGULATORY REVIEW.........................................
B-1 APPENDIX C NRC STAFF ACTIVITIES REGARDING GENERIC SAFETY ISSUES.........................................
C-1 APPENDIX D CHANGES AND ERRATA TO THE SAFETY EVALUATION REPORT ISSUED JULY 1978........................
D-1 LIST OF FIGURES PAGE FIGURE 17.1 OHIO EDISON COMP ANY ORGANI ZATION...............
17-3 LIST OF TABLES PAGE TABLE C.1 PRIORITY CATEGORY DEFINITIONS.....................
C-8 C.2 LIST OF TECHNICAL ACTIVITIES......................
C-9 iii
1.0 INTRODUCTION
AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Conmission's Safety Evaluation Report in the matter of the application by Ohio Edison Company, Toledo Edison Company, Cleveland Electric Illuminating Company, Duquesne Light Company and Pennsylvania Power Company (the apolicants) to construct and operate the proposed Erie Nuclear Plant, Units 1 and 2 (the facility) was issued in July 1978.
In that Safety Evaluation Report, we concluded that upon favorable resolution or the outstanding matters set forth in Section 1.11 of that report, we would be able to reach the conclusions required in acc.ordance with the provi-sions cf 10 CFR SJ.35(a). Section 1.11 of the Safety Evaluation Report identified 12 ou< standing items requiring additional staff evaluation.
Since the Safety Evaluation Report was issued:
(1) The Advisnry Committee on Reactor Safeguards completed its review of the Erie Nuclear Plant application on August 3,1978.
The Committee's report is included in this supplement as Appendix A.
Our response to the Committee's report is presented in Section 18.0 of this supplement.
(2)
The applicants have submitted Amendments 18,19 and 20 to the Erie Nuclear Plant Preliminary Scfety Analysis Report, responding to the items identified in Section 1.11 of the Safety Evaluation Report and this supplement.
(3)
The applicants submitted information on a change in the organizational structure of Ohio Edison Company, the applicant with the responsibility for the design, licensing and construction of the facility.
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(4)
The applicants submitted information concerning their financial qualifications to construct and operate the Erie Nuclear Plant.
However, the applicants announced that detailed studies were being performB'1 to determine if a delay for the Erie Nuclear Plant was necessary. We will complete our review of the applicants' financial qualifications when the applicants conclude these studies.
The purpose of this supplement is (1) to report our evaluation of the additional information submitted by the applicants since our issuance of the Safety Evaluation Report; (2) to report our evaluation of the outstanding items listed in the Safety Evaluation Report; (3) to address comments by the Advisory Com-mittee on Reactor Safeguards. Our conclusions regarding the individual out-standing items are found in the appropriate sections of this supplement. With the exception ;f the applicants' financial qualifications, we are able to reaffirm our conclusions stated in Section 21.0 of the Safety Evaluation Report.
Appendix B to this supplement is a continuation of the chronology of significant events of the licensing review of the Erie Nuclear Plant application. Appendix C discusses the Comission's ongoing activities on the generic safety issues appli-cable to the Erie Nuclear Plant. Appendix 0 is a listing of errata to the Safety Evaluation Report.
Each of the following sections in this supplement is numbered the same as the corresponding section of the Safety Evaluation Report. This supplement is an addition to and is not in lieu of the discussion in the Safety Evaluation Report.
1.11 Outstanding Items The Safety Evaluation Report identified 12 outstanding items which required further review in order to confirm that the proposed design would meet regulatory requirements. We have completed our review for each of those issues, and each has been acceptably resolved for the coastruction permit stage of review. The outstanding issues identified in the Safety Evaluation 1-2
Report are summarized below.
The resolution of each issue is discussed in this supplement under t n: designated sections.
(1)
In a letter dated June 9, 1978, the applicants modified their design to include two safety-grade isolation valves in each main feedwater line; this meets the BSAR-205 interface.
- However, the second isolation valve was classified as Quality Group C and we required that this valve be classified as Quality Group B (Section 3.2.2).
The applicants did not provide descriptive information on this design change; therefore, our review of the feedwater isolation system was incomplete (See Sections 7.3.3 and 15.2.1 ).
(2) We required that the main _ team lines or the borated water storage tank be protected from tornado missiles (Section 3.5.1).
(3)
For several seismic Category I structures, the applicants propose use of concrete which reaches its specified strength ninety days after placement. We required that when concrete cores are taken from these pours for testing, the cores should have an average strength equal to the specified strength for the pour to be considered adequate.
The applicants stated that an average strength of 85 percent of specified should be acceptable and provided justification for their position. Our review of this information was not complete (Section 3.8.1).
(4) We had not completed our review of the applicants' analysis of the forcing functions and transient moments acting on the reactor vessel and steam generator supports as a result of postulated pipe breaks in the respective containment sub-compartments in which each of the components is located (Section 6.2.1).
(5)
The radiation monitors associated with the continuous purge system which are intended to provide one of the diverse means 1-3
of closing the isolation valves in the continuous purge system were not specified as safety-grade. We required that they be safety-grade (Section 6.2.3).
(6) We required additional information on the applicarts' design in order to complete our analysis of hydrogen accumulation following a postulated loss-of-coolant accident (Section 6.2.4).
(7) The applicants had not provided justification for exempting certain containment isolation valves from local (Type C) leak tests. We required the applicants to provide this justification
'e to commit to locally leak test these isolation valves (Section 6.2.5).
(8) We required that the applicants' design provide the capability to monitor and indicate in the control room, the flow in the
" dump to sump" lines.
The applicants provided justification for not including these features; our evaluation of this infor.aation was not complete (Section 6.3.5).
(9) We had not completed our review of the applicants' probabilistic analysis of toxic chemical spills near the site (Section 6.5.2).
(10) We required aoditional information from the applicants on engineered safety feature systems that will not be tested during normal operation (Section 7.3.1).
(11) We required that the offsite preferred power system from the switchyard to the onsite Class IE power system, including the fast transfer from normal power sources to the preferred power sources, meet the requirements listed in Section 8.2.4, page 8-5 of the Safety Evaluation Report.
The applicants had submitted additional information on this matter; however, our review was not complete.
1-4
(12) We requested additional information on the financial qualifications of the applicants (Section 20.0).
With the exception of the applicants' financial qualifications, all technical issues previously identified in the Safety Evaluation Report have been resolved.
However, it should be noted that the ongoing NRC staff activities regarding generic safety tasks described in Appendix C of this supplement may result in future additional requirements on the applicants during the design, construc-tion and operation of this facility.
e e
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~ ammumum-3.0 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS 3.2 Classification of Structures, Systems and Components
' 2. 2 System Quality Group Classification The Erie Nuclear Plant design includes two isolation valves in series outside containment in each main feedwater line.
In the Safety Evaluation Report we stated that it was our position that both of the isolation valves in the main feedwater lines be classified as Quality Group 8.
The applicants position was that the first isolation valve (the valve closest to the containment) should be classified as Quality Group 8 and that the second valve could be Quality Group C.
The applicants provided additional information and further justification for the use of a Quality Group C valve in a meeting on August 15, 1978.
The applicants stated that the second vaive was not required for contain-ment integrity but was required for the termination of feedwater flow to the steam generators in the event of a main steam line break. Therefore i
for this second valve, operability is more important than provioing assurance of pressure boundary integrity via a higher quality group.
For the Erie design, the operability of the valve will be assured by a valve operator that has been qualified for the safe shutdown earthquake and by providing safety-grade actuation signals to close the valve in the event of a main steam line break accident.
As a result of our meeting with the applicants we have modified our position and we now conclude the classification of the second valve may be Quality Group C.
Therefore we find the applicants' design to be acceptable.
3.5 Missile Protection 3.5.1 Missile Selection and Protection Criter ia In the Safety Evaluation Report it was our position that either the borated water storage tank or the main steam piping outboard of the main steam 3-1
isolation valves must be protected from tornado missiles.
The bases for this position was that missiles generated by a tornado might initiate a main steam line break and also destroy the borated water storage tank, a portion of which is necessary to mitigate the consequences of a postulated n'ain steam line break accident.
In letters dated September 14, 1978 and January 8, 1979 the applicants provided analyses which demonstrated that the main steam lines can withstand the impact of each of the tornado missiles postulated for the site. The tornado n.issile spectrum for the site is discussed in Section 3.5.1 of the Safety Evaluation Report and was found to be acceptable. We reviewed the applicants' analyses and based on our review we conclude that a steam line will not rupture as a result of the impact of a tornado missile. Therefore, we find the applicants' design to be acceptable.
3.8 Seismic Design of Seismic Category I Structures 3.8.1 Concrete Containment The applicants propose the use of concrete (fly ash concrete) which will attain its specified compressive strength after 90 days of curing time.
The adequacy of a concrete pour is normally determined by testing of cylin-ders made at the time of concrete placement.
If this field testing indicates that the concrete strer.gth does not meet the specified strength, other non-destructive and/or destructive tests on the in-place concrete may be used. One such test is to core a sample from the concrete placement and corduct a destructive test to determine the core sample's strength. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, Division 2, " Concrete Reactor Vessels and Containments" states that if core samples are required to verify the adequacy of a concrete placement (1) the average strength of the core samples must be at least 85 percent of the specified strength, f'c and (2) that no single core sample can have a strength less than 75 percent of the specified strength.
It was our position that the average core sample strength meet or exceed 100 percent of the specified strength.
3-2
The applicants provided information showing that core samples generally have lower strength than the design strength due to the process of obtair.ing a core sample. We reviewed this information and met with the applicants to discuss the matter on August 15, 1978. As a result of our discussion with the applicants, we now require that the applicants inform us within 30 days of when the results of field tests indicate the need for subsequent con-firmatory tests of the in-place concrete. At thet. time, the applicants will provide us with the field tests results and the proposed method of evaluating in-place concrete strength. Upon evaluation of the applicants' submittal, we will determine if any further action is appropriate.
The applicants documented their commitment to provide the field test data within 30 days after the tests in a letter dated Scptember 8,1978. Based on thia cormiitment, we conclude that th's matter is resolved.
3-3
5.0 REACTOR COOLANT SYSTEM 5.4 Component and Subsystem Design 5.4.3 Decay Heat Removal System In the Safety Evaluation Report we stated that we require that the Erie Nuclear Plant have the capability to be taken to a cold shutdown condition in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> using only safety-grade equipment assuming a loss of offsite power or onsite power, and a single failure. The details of our requirements were (1) listed in Section 5.4.3 of Appendix A to the Safety Evaluation Report and (2) discussed in our letter to the applicants dated October 27, 1978.
In our letter, we required the applicants to commit to designing the total facility, both the balance of plant design and the BSAR-205 design, to meet our requirements for cold shutdown.
In a letter dated December 15, 1978, the applicants made this commitment. We find this commitment to be acceptable at the construction permit stage of review.
5-1
6.0 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.1 Containment Functional Design In the Safety Evaluation Report, we stated that the peak differential pressures calculated by the applicants for the structural design of the reactor cavity and the steam generator subcompartments were acceptable.
However, we stated that the applicants' analysis of the loads on the components supports as a result cf these pressures were not complete and that we would report on the resolution of this matter in a supplement to the Safety Evaluation Repnrt.
By letter dated September 14, 1978, tne applicants submitted additional information regarding the component supports analyses.
The applicants have calculated forces and moments acting across the reactor vessel, steam generator, pressurizer and reactor coolant pumps.
Nodalization sensitivity studies were performed to verif that the maximum forces and s
moments were being predicted.
The criterion for convergence used in the sensitivity study was defined to be less than a five percent difference in resultant horizontal force between successive models.
For the reactor cavity analysis, the applicants determined tht the largest forces and moments were predicted to occur for a one square foot cold leg break.
For the steam generator analysis, the applicants determined that the largest forces and moments were predicted to occur for a 5.585 square foot cold leg pump suction split.
A 40 percent margin was added to all calculated pressures in determining the calculated forces and moments acting on the components. We have found the applicants' sensitivity study and the margin added to the calculated pressures to be acceptable.
We have reviewed the applicants' analyses, and performed confirmatory analyses which showed good agreement with the peak loads and moments calculated by the applicants. We find that the applicants have developed acceptable models and, therefore, we find the subcompartment analysis acceptable.
6-1
We will further review the subcompartrent analysis at the opirating license review stage and will require the applicants to comply with any analytical or design requirements resulting from the resolution of Generic Task A-2,
" Asymmetric LOCA Loads."
6.2.3 Containment Isolation System In the Safety Evaluation Report we stated that the non-safety grade high radiation monitors that was to be used to isolate the continuous purge system was unacceptable. We requir:d that there be redundant radiation monitors for the continuous purge system and that the monitors must meet the requirements of Institute of Electrical and Electronics Engineers (IEEE) Standard 279-1971,
" Criteria for Protection Systems for Nuclear Power Generating Stations" and be qualified in accordance with the requirements of IEEE Std. 323-1974, " Institute of Electrical and Electronics Engineers Standard for ualifying Class IE Equipment for Nuclear Power Generating Stations."
The applicants have modified their design to include two radiation monitors for the continuous purge discharoe line. Each monitor will be powered from a separate Class IE power source and qualified in accordance with IEEE 323-1974.
The applicants' design for isolating the continuous purge system is now in conformance with IEEE 279-1971 and the electrical components will be qualified in accordance with IEEE 323-1974; theref're we find the applicants' design acceptable.
6.2.4 Combe;tible Gas Control System In the Safety Evaluation Report, we stated that we were not able to verify the applicants' aralysis of hydroger accumulation in the containment following a postulated loss-of-coolant accident and that we had requested additional information from the applicants.
6-2
Since the issuance of the Safety Evaluation Report, the applicants submitted a revised hydrogen accumulation analysis based on temperature dependent corrosion rates that we developed from experimental data.
The applicants' analysis, which used our recomc.anded corrosion rates, assumes that only one of the redundant hydrogen recombiners is in operation and that the recombiner begins operating 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the postulated loss-of-coolant cccident.
The applicants also stated that the capacity of a single recombiner wili be 100 standard cubic feet per minute. The applicants concluded that a single recombiner can limit the total accumulation of hydrogen in contain-ment to less than four volume percent.
Our confirmatory analysis, which is based on the applicants' input data, verifies the applicantr' results. Based on our review of the systems provided for combustible gas control following a postulated loss-of-coolant accident, we conclude that the design satisfies the provisions of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a loss-of-Coolant Accident," and General Design Criteria 41, 42 and 43 and, therefore, is acceptable.
6.2.5 Containment Leakage Testing Program In the Safety Evaluation Report we stated that the applicants' containment leak testing program was deficient in that justification had not been provided for exempting certain containment isolation valves from local (Type C) leak testing.
In a letter dated September 14, 1978, the applicants provided additional information on this matter.
With the exception of the lines that are connected to the secondary side of the steam generator the applicants have shown that all containment isolation valves either will be locally (Type C) leak tested or do not constitute containment atmosphere leak paths. The lines that are connected to the secondary side of the steam generator may become leakage paths to the environ-ment in the event of containment leakage through the steam generator tube bundles. The applicants provided, in response to our requests for information 6-3
an analysis of the containment atmosphere leakage through the steam generator tube bundles in lieu of a commitment to perform local leak tests on the isolation valves in these lines.
The applicants' analysis is based on the maximum allowable primary to secondary system leakage during normal operation, which will be limited to 1 gallon per minute. The applicants selected a correlation which, based on the one gallon per minute leak rate, maximizes the leak area that could exist in a steam generator tube bundle.
The applicants then maximized the flow of containment atmosphere through the leak area by assuming sonic flow.
The applicants included this calculated leakage in the radiological analysis of the loss-of-coolant accident.
We have reviewed the applicants' analysis and conclude that the applicants has conservatively calculated the amount of containment atmosphere which could pass through the steam generator tube bundles. Our independent dose calcula-tions show that the addition of the steam generator leakage does not increase the two hour exclusion area thyroid dose of 141 rem and increases the 30 day low population zone thyroid dose from 78 to 85 rem. These doses are well within the guidelines of 10 CFR Part 100. Therefore, we conclude that the applicants' approach for accounting for leakage through the steam generator tubes bundles is acceptable.
The following discussion relates to airlocks proposed for personnel entry into the containment.Section III.D.2 of Appendix J to 10 CFR 50 requires airlocks to be leak tested at six-month intervals, and after each opening during the intervals. Moreover,Section III.B.2 of Appendix J requires all penetrations to be leak tested at the calculated peak containment internal pressure corresponding to the design basis accident (Pa).
Based on plant operating experience, requiring an airlock to be leak tested after each opening is an impractical requirement when frequent airlock usage is necessary over a short period of time.
Testing an airlock for 6-4
i r
(5)
In the case of chlorine, an additional provision in the design of the Erie Nuclear Plant will be made with respect to the outside air intake rate for '.he control room.
If chlorine protection is required, the normal control room outside air intake rate will be limited to less than one air change per hour as described in Regulatory Guide 1.95,
" Protection of Nuclear Power Plant Control Room Operatces Against an Accidental Chlorine Release."
We will review this matter again during the operating license stage of review to determine which, if any, chemical monitors will be necessary.
However, based on the applicants' commitment to design the facility to include the provisions necessary for the protection of control room personnel, we conclude that this matter is resolved for the construction permit stage of review.
6-7
7.0 INSTRUMENTATION AND CONTROLS 7.3 Engineered Safety Features Actuation System 7.3.1 Periodic Testing of Engineered Safety Feature Systems In the Safety Evaluation Report, we stated that additional information had been requested from the applicants on testing of equipment in the essential service water system, the component cooling water system and the reactor building isolation system.
The applicants provided a list of the components that cannot be fully tested during normal operation and the justification for not testing during normal operation.
The applicants stated that these valves will be tested when a unit is shut down in accordance with item D.4.c in Regulatory Guide 1.22, " Periodic Testing of Protection System Actuation Functions." The details of the periodic tests will be addressed in the technical specifications which will be prepared at the operating license stage. We find this to be acceptable for the construction permit stage of review.
7.3.3 Feedwater Isolation System In the Safety Evaluation Report we stated that the feedwater isolation logic had not been changed to reflect a change in the feedwater isolation system and therefore our review was incomplete.
The revised isolation logic diagrams were submitted in Amendment No.18.
The proposed design now specifies that both the "A" and "B" feedwater isolation signals will be sent to both of the isolation valves in each feedwater line. Based on our review of the applicants' design, we conclude that the design is acceptable.
7-1
m.
8.0 ELECTRIC POWER SYSTEMS 8.2 Offsite Power Systems 8.2.1 Offsite Routing and Distribution Normal power to a unit's Class IE power system will be supplied from a 1
unit's main generator through the unit auxiliary transformers and non-j Class IE switchgear. An automatic transfer scheme will be provided to transfer the Class IE power system from the normal power source, the turbine generator via a startup transformer, to the preferred power source, the offsite power system via the startup transformers.
In the Safety Evaluation Reprt, we stated that the automatic transfer scheme should be designed suct that a single electrical failure would not cause the loss of both sources of preferred power from the offsite power system.
The applicants provided additional information on the automatic transfer in a letter dated September 5, 1978. The automatic transfer scheme will include primary and backup protective relays. The backup protective relays will use separate sensors and the switchgear for transferring each of the two preferred power sources will be electrically separate.
The power for autor'atic transfer will be provided by two non-Class IE 125 volt batteries. The batteries will be placed in separdte rooms and will be installed such that the loss of one battery will not preclude the automatic transfer of at least one preferred power Based on (1) our review of the preliminary design of the sou rce.
auto,aatic transfer provided by the applicants, and (2) the applicants' commitment to design the transfer such that a single electrical failure will not prevent the transfer of at least one prefer ed power source, we conclude that the applicants' preliminary design is acceptable.
8-1
8.2.4 Conformance With Acceptance Criteria In the Safety Evaluation Report, we identified several regulatory guides as acceptance criteria applicable to the port'on of offsite power system between the switchyard and the Class IE distribution system.
The applicants have responded to our position in a letter dated September 5, 1978. The applicants stated that the acceptance criteria ider.tified in the Safety Evaluation Report were not applicable because the offsite power system is not a Class IE system.
However, the applicants' pcoposed alternate measures for the (1) installation, inspection and testing control, (2) pre-operational and initial start-up testing, (3) periodic onsite testing, and (4) surveillance of system operability status ft the offsite power system between the switchyard and the Class IE power distribution system.
'N_
N We have reviewed the applicants' submittal and with the exceptions noted below we conclude that the applicants have proposed measures which will assure the availability of the offsite power system. We find tiie periodic surveillance tests for (1) the batteries in the switchyard ard (2) the bat-
/
teries which power the automatic transfer to be unacceptable. During the operating license review, we will prepare technical specifications which specify the frequency of surveillance and periodic testing of systems in the plant during operation.
Currently, it is our intent to require t. hat the surveillance and periodic testing for these batteries be the same as that required for Class IE batteries. The tests and surveillance of the batteries do not require special design provisions, therefore, we conclude that this matter can be left to the operating license stage of review.
We conclude that this matter is satisfactorily resolved for the construction permit stage of review.
8-2
9.0 AUXILIARY SYSTEMS 9.1 Fuel Storage and Handling 9.1.1 Soent Fuel Pool Cooling and Cleanup System In the Safety Evaluation Report, we stated that during periods of " abnormal" heat load conditions (17 normal refueling batches, plus one full core unload) three of the four spent fuel pool cooling pumps and heat exchangers would maintain the pool temperature below 170 degrees Fahrenheit.
We found this to be acceptable.
In Amendment No. 20 to the Preliminary Safety Analysis Report, the applicants modified their design such that the decay heat removal system, which will be connected to the spent fuel pool cooling system, will be used to handle the heat load in the spent fuel pool from the core unloaded during the " abnormal" heat load condition. Tne spent fuel pool cooling pumps and heat exchangers will be sized such that two of the four pumps and heat exchangers will handle the balance of the " abnormal" heat oad.
We find this to be acceptable based on the redundancy of components 'n the decay heat removal and spent fuel pool cooling systems.
9-1
13.0 CONDUCT OF OPERATIONS 13.1 Organizational Structure of the Applicants After the issuance of the Safety Evaluation Report, Ohio Edison Company announced a change in its organizational structure.
Ohio Edison Company is the applicant that will be responsible for the design, construction, quality assurance, testing, and operation of the Erie Nuclear Plant, Units 1 and 2.
The new organization is shown in Figure 17.1.
In the new organization the engineering and design for the project is the responsibility of Ohio Edison Company's Vice President (Generating Plant Engineering and Con-st ruction).
Reporting to him is the Manager, Erie Nuclear Plant Project.
The Manager, Erie Nuclear Plant Project is the overall coordinator of project activities.
The Senior Vice President (Generating Plant and Transmission System Operations) will have overall responsibility for the operation of the facility.
The changes in the quality assurance organi-zation are discussed in Section 17.0.
Based on our review of the changes, we conclude that the new organizational structure is an acceptable organization to implement their responsibilities relative to the Erie Nuclear Plant, Units 1 and 2.
13-1
15.0 ACCIDENT ANALYSES 15.2 Postulated Accidents 15.2.1
_rectrum of Steam Piping Breaks Inside and Outside of Containment In the Safety Evaluation Report we stated that the applicants had not submitted detailed information on a revision to the feedwater isolation system. Specifically, we requested the applicants to verify that the closure time for the added valves and volume of water between the second valve and the steam generator were consistent with the assumptions in the BSAR-205 accident analyses. The applicants submitted this informa-tion in Amendment No.19 and we conclude that the feedwater isolation time and the volume of water between the isolation valves and the steam generators is consistent with the assumptions in the BSAR-205 accident analyses and, therefore, th e refererce to the BSAR-205 analyses is acceptable.
15-1
17.0 QUALITY ASSURANCE 17.2 Ohia Edison Company 17.2.1 Organization After issuance of the Safety Evaluation Report, Ohio Edison Company announced an organization change which affected the Ohio Edison quality assurance program. The new organization was described in letters dated October 9, 1978, October 17, 1978 and December 1,1978.
The Preliminary Safety Analysis Report has not been updated to reflect this change, but the applicants have stated that this will be done in the future.
Figure 17-1 illustrates the new organization. The Manager, Quality Assurance has assumed all the duties of the Chicf Nuclear Quality Assurance Engineer in the previous organization discussed in the Safety Evaluation Report.
In addition he is also responsible for all non-nuclear quality assurance. The Manager, Quality Assurance reports directly to the Vice President (Generating Plant Engineerlag and Construction) as does the Manager, Erie Nuclear Plant Project.
Figure 17.1 depicts the independence of the Ohio Edison quality assurance personnel under the Manager, Quality Assurance from all other organizations connected with nuclear pr'je-ts.
With this corporate organization structure, we find that quality assura, has adequate independence and reports at a sufficiently high management level to accomplish its objectives without undue influence from cost and schedule.
One other aspect of the quality assurance program affected by the organization change was the Management Review Committee. The Management Review Committee annually assesses the quality assurance program to determine compliance with the program requirements. The Management Review Committee was previcusly chaired by the Vice President (Engineering); the committee will now be chaired by the Executive Vice President.
17-1
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t Based on our review of the infonnation submitted by the applicants we reaffirm our previous conclusion that the Ohio Edison Company organization complies with Appendix B to 10 CFR Part 50 and is acceptable.
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18.0 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFECUARDS i
At its 220th meeting on August 3-5, 1978, the Advisory Comaittee on Reactor Safeguards completed its review of the application by the applicants for permits to construct the Erie Nuclear Plant, Units 1 and 2.
A copy of the Committee's report on the f acility, dated August 9,1978, which contains certain comments and recommendations, is included as Appendix A to this report.
The actions we have taken or plan to take in response to the Committee's comments and recommendations dre described in the following paragraphs.
(1)
The Conmittee noted that our Safety Evaluation Report identified a number of outstanding safety items, and the Committee recommended that these matters should be resolved in a manner satisfactory to the staff.
Since the August 3,1978 meeting with the Conmittee, we have reviewed additional information and commitments submitted by the applicants concerning the outstanding issues identified by the staff.
Wiin the exception of the applicants' financial qualifications, we have resolved all these issues. Our evaluation and resolution of these issues are reported in appropriate sections of this supplement to our Safety Evaluation Report.
(2) The Connittee identified those generic issues relating to large water reactors which were discussed in the Committee's report to the Commission on n vember 15,1977 (Report No. 6), and which the Committee considered relevant to the facility.
The Committee noted that the issues shauld be dealt with by the staff and applicants as solutions are found.
We have transmitted the Committee's recommendations to the applicants for its consideration in proceeding with the Eric design. Appendix C to the Safety Evaluation Report discusses the disposition and status of the generic matters raised by the Advisory Committee on Reactor Safeguards.
Also, the items identified by the Committee (with the exception of Item Il-4, Instruments to Detect Fuel failures) are also covered by the NRC Generic Issues Program discussed in Appendix C to this supplement.
18-1
20.0 FINANCIAL QUALIFICATIONS The Cormi<:sion's regulations which relate to financial data and information re-quired to establish the financial qualifications of applicants for a f acility construction permit are Section 50.33(f) of 10 CFR Part 50 and Appendix C to 10 CR Part 59. On March 28, 1978 we requested that the applicants submit information on the projected cost of the Erie Nuclear Plant, Units 1 and 2 and their projected plans for financing the Erie Nuclear Plant. The appli-cants responded with submittals dated June 16,1978, July 28,1978, October 17, 1978 and October 28, 1978.
Shortly before we completed our review of this infernation, the applicants announced that they were reevaluating the construction schedule for several nuclear projects, including the Erie Nuclear Plant, and that the schedule for the Erie Nuclear Plant may lie delayed by several years. A letter dated December 21, 1978 from Ohio Ediscn Company, the applicant responsille for the design, construction, and operation of the Erie Nuclear Plant, cited increased construction costs, high environmental control costs on existing units, changing conditions in the capital markets and uncertainty over the need for generating capacity additions as the reasons for the reevaluation.
The applicants did not specify when their reevaluation would be completed.
Since our evaluation inherently depends upon the projected financing of the facility, which in turn is dependent upon its construction schedule, we will proceed in our review when the applicants have concluded their reevaluation.
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21.0 CONCLUSION
S i
In Section 21.0 of the Safety Evaluation Report, we stated, that upon favorable resolution of the outstanding matters set forth in Section 1.8 of the Safety Evaluation Report, we would be able to make certain conclusiens.
We have discussed these outstanding matters in this supplement. With the exception of the applicants' financial qualifications, we have indicated favorable resolution of the outstanding matters.
Accordingly, we conclude that, in accordance with the provisions of Sections 50.35(a) and 50.40 of 10 CFR Part 50:
(1) The applicants have described the proposed design of the facility, including but not limited to the principal architectural and engineering criteria for the design and has identified the major features or components incorporated therein for the protection of the health and safety of the public; (2) Such further technical or design information, as may be required to complete the safety analysis and that can be reasonably left for later consideration, will be supplied in the Final Jafety Analysis Report; (3 ) Safety features or components that require research and development have been described and identified by the applicants, and there will be conducted research and development programs reasonably designed to resolve safety questions associated with such features or components; (4 ) On the basis of the foregoing, there is reasonable assurance that (a) such safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of construction of the proposed facility, and (b) taking into consideration the site criteria contained in 10 CFR Part 100, the proposed facilities can be constructed and operated at the proposed location without undue risk to the health and safety of the public; 21-1
.i (5) Ohio Edison Company is technically qualified to design and construct the proposed facility; L
(6)
The issuance of a permit for construction of the facility will not be inimical to the common defense an' security or to the health and safety of the public.
e B
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APPENDIX A
[(] ') 5,( ' h UNITED STATES NUCLEAR REGULATORY COMMISSION
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- E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. D. C. 20655 August 9, 1978 Honorable Joseph M. Hendrie Chairman U. S. Nuclear Regulatory Camtission Washington, DC 20555
SUBJECT:
REFORT CN ERIE NUCLEAR PIRTT, UNITS 1 & 2
Dear Dr. Herdrie:
At its 220th Meeting, August 3-5, 1978, the Advisory Committee on Reactor Safeguards completed its review of the application of the Ohio Edison Company, Toledo Edison Cmpany, Cleveland Electric Illuminating Company, Duquesne Light Company, and Pennsylvania Power Company, (hereinafter referred to collectively as the Applicant) for a permit to construct the 3760 MWt Erie Nuclear Plant, Units 1 & 2.
A Subcommittee meeting was held in Elyria, Ohio on July 18, 1978 ard the plant site was visited by members of the Subcommittee on the preceding day.
Be Committee had the benefit of discussions with representatives and consultants of the Ap-plicant, the Babcock and Wilcox Car.pany, the Giltert/Camonwealth Can-panies, and the Nuclear Regulatory Commission (NRC) Staff.
We Committee also had the benefit of the documents listed.
The Erie Nuclear Plant, Units 1 & 2, incorporate the Babcock-205 Stardard Nuclear Steam System by reference, with certain minor modifications to the Babcock-205 design and interface criteria.
Se Canmittee reported on its review of the application of the Babcock & Wilcox Canpany for a Preliminary Design Approval for the Babcock-205 Standard Nuclear Steam System, in its letter of August 18, 1977. Se modifications to the Babceck-205 design include the sharing of several systems between Unit 1 and Unit 2, and a few changes brought about because they could not be accommodated in the Erie balance-of-plant design which was under concurrent review.
We proposed Erie Nuclear Plant will be located on a 1740 acre site in Erie County, Ohio within the Berlin and Vermilion Townships, 8.9 miles southeast of Sandusky, Ohio (the nearest population center,1970 popu-lation 32,700), and about 2.4 miles south of Iake Erie. S e exclusion area has been defined by two overlapping 2,600-foot radius circles each centered on one of the reactor buildings.
he radius of the low popu-lation zone is 2 miles.
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Honorable Joseph M. Hendrie August 9,1978 The Applicant and the Staff have agreed on a zero-period acceleration of 0.20g applied at the foundation level for the safe shutdown earthquake, and 0.07g for the operating basis earthquake.
'Ihe Committee considers these values acceptable for this plant.
The NRC Staff has identified a nurrber of safety items which will require resolution before issuance of a construction permit. 'Ibese matters should be resolved in a manner satisfactory to the NRC Staff.
With regard to the generic problems cited in the Committee's report,
" Status of Generic Items Relating to Light-Water Reactors: Report No.
6,"
dated November 15, 1977, items considered relevant to the Erie Nuclear Plant, Units 1 and 2 are:
II-3, 4, 5B, 6, 7, 9, 10; IIA-2, 3, 4; IIB-1, 2; IIC-1, 2, 3A, 3B, 4, 5, 6; IID-2; IIE-1.
'Ihese problems should be dealt with by the NRC Staff and the Applicants as solutions are found.
The ACRS believes that, if due consideration is given to the iters men-tiened above, the Erie Nuclear Plant, Units 1 & 2, can be constructed with reasonable assurance that they can be operated without undue risk to the health and safety of the public.
Sincerely yours, M
Stephen Lawroski Chairman
References:
1.
Erie Nuclear Plant, Units 1 & 2 Preliminary Safety Analysis Report with Amendments 1 through 18.
2.
Safety Evaluation Report by the Office of Nuclear Reactor Regu-lation, U.S. Nuclear Pegulatory Commission in the matter of Ohio Ediscn Company et al, Erie Nuclear Plant, Units 1 & 2, (NUREG-0423),
July 1978.
3.
Safety Evaluation Report related to the preliminary design of the BSAR-205 Standard Design, (NUREG-0433), May 1978, Office of Nuclear Peactor Regulation, U.S. Nuclear Regulatory Commission.
4.
Chio Edison Company letters:
a.
Lynn Firestone to Robert Baer, NRC dated June 26, 1978 regarding responses to NRC questions regarding fire protection.
A-2
Honorable Joseph M. Hendrie Augus t 9,1978
References:
(con' t) b.
Lynn Firestone (by Barry M. Miller) to Robert Baer, NRC, dated July 12, 1978 regarding transmittal of advance information regarding several outstanding items, c.
Lynn Firestone (by Barry M. Miller) to Robert Baer, NRC, dated July 14, 1978 regarding transmittal of advance information regarding the monitor to isolate the continuous purge system, post accident hydrogen accumulation, and post accident contair: rent temperatures.
A-3
APPENDIX B CONTINUATION OF CHRONOLOGY OF REGULATORY REVIEW Note: Documents referenced in tha chronology of the Safety Evaluation Report and in this Supplement to the tafety Evaluation Report are available for public inspection and copying for a fee at the NRC Public D cument Room,1717 H Street, N. W., Main Street, Berlin Heights, Ohio 44841.
July 3,1978 Issuance of Safety Evaluation Report.
July 12, 1978 Letter from the applicants transmitting information regarding outstanding items in the Safety Evaluation Report.
July 14, 1978 Letter from the applicants transmitting information regarding outstanding items in the Safety Evaluation Report.
July la,1978 Advisory Committee on Reactor Safeguards subcommittee meeting with staff and applicants.
July IP, 1978 Letter to the applicants transmitting " Barrier Penetration Databast NUREG/CR-0181.
July 28, 1978 Letter from the applicants transmitting proprietary financial infonnation.
August 2,1978 Representatives of the applicants and the Commission meet in Bethesda, Maryland to discuss several outstanding items in the review of the Preliminary Safety Analysis Report.
August 2, 1978 Letter to applicants transmitting " Nuclear Security Personnel for Power Plants, Content and Review Procedures for a Security Training and Qualification Program," NUREG-0219.
August 3,1978 Advisory Committee on Reactor Safeguards meeting with staff and applicants.
August 9,1978 Letter from the Advisory Committee on Reactor Safeguards.
August 11, 1978 Letter to the applicants regarding standard format for meteorological data on magnetic tape.
August 15, 1978 Representatives of the applicants and the Commission meet in Bethesda, Maryland to discuss applicants' appeal of three staff positions taken in the review of the Preliminary Safety Analysis Report.
August 15, 1978 Letter to the applicants advising them of pressurized water reactor steam generator workshop to be held September 7-8, 1978.
B -1
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I September 5, 1978 Letter from the applicants transmitting information which addresses two outstanding items in the Safety Evaluation Report concerning offsite power system.
o September 12, 1978 Amendment No. 19 to Preliminary Safety Analysis Report received along with a September 8,1978 letter from applicants containing information concerning commitments and resolutions.
September 14, 1978 Letter from the applicants transmitting additional information on outstanding issues.
October 9, 1979 Letter from the applicants transmitting information which addresses the Ohio Edison Company reorganization.
October 17, 1978 Letter from the applicants with information which address outstanding items.
October 27, 1978 Letter to the applicants issues sta'f position on requirements for cold shutdown.
November 6, 1978 Letter from the applicants which transmitting its cormitment regarding toxic chemical shipments past the site.
November 11, 1978 Amendment No. 20 to the Preliminary Safety Analysis Report received.
Dcoember 1,1978 Letter from the applicants transmitting additional clarification on two outstanding matters.
December 15, 1978 Letter from the applicants rroviding their commitment on the issue of cold shutdown.
December 21, 1978 Letter from the applicants discussing the uncertainty in their
~
schedule for the Erie Nuclear Plant.
January 8,1979 Letter from the applicants providing information on the response of a steam line to the impact of a tornado missile.
/
/
B-2 e
leakage within a limited time period following the initial opening is more practical, and still provides the desired confidence that the leak tightness of the airlock is within acceptable limits.
The applicants propose to leak test the airlock within three days of being opened at a pressure as stated in the plant technical specifications.
Furthermore, the six month total airlock leak test will be retained.
For several operating license applications, we have granted exemptions for applicants who propose this testing in lieu of the requirements of Appendix J.
We have found this testing to provide assurance of the con-tainment integrity equivalent to that required by the Appendix J testing requirements.
In conclusion, we find that the applicants' proposed leak testing of the airlocks warrants the granting of an exemption from the requirements of Appendix J to 10 CFR Part 50.
6.3 Emergency Core Cooling System 6.3.5 Post Loss-of-Coolant Accident Boron Precipitation In the Safety Evaluation Report, we stated that we requirca that the applicants' design include a flow monitor with indication in the control room in each of the two " dump to senp" lines.
The flow through the
" dump to sunip" lines would be used to prevent excessive boric acid con-centration in the reactor vessel following a postulated loss-of-coolant accident In Amendment No.19, the applicants modified their proposed design to include the monitors and indicators in the control room.
We therefore find the applicants' design to be acceptable.
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6.5 Engineered Safety Feature Atmosphere Cleanuo Systems 6.5.2 Control Room Habitability Provisions In the Safety Evaluation Report we stated that the applicants had submitted a probabilistic analysis of the hazard posed by the shipment of toxic chemicals past the site. We concluded that the analysis was unacceptable because (1) the applicants were unable to obtain detailed information on the number of ship-ments of toxic chemicals on the Norfolk and Western Railroad, the railroad closest to the site; and (2) our independent calculations indicated that the probability of an unacceptable accident involving chlorine, sulfur dioxide, or ammonia was greater than one chance in ten million per year.
However, the applicants have committed to the following:
(1) The necessary provisions will be made in the design of the Erie Nuclear Plant to accomodate toxic chemical monitors and adequate isolation capability for the toxic chemicals in question so that the monitors could be installed at any future date.
(2) Future attempts will be made to secure the necessary data from the Noriolk and Western Railroad.
(3) A hazards analysis will be submitted to the NRC for the toxic chemical shipped past the Erie site by rail, at a time after the issuance of a construction permit but prior to the issuance of an operating license.
(4)
Should the analysis show that the risk from certain toxic chemicals is unacceptable, the required monitors will be installed to detect the toxic limits of these chemicals and to automatically isolate the control room within an appropriate time period (e.g.,10 seconds for detection and isolation in the case of chlorine) from further intake of such chemicals.
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7 APPENDIX C NRC STAFF ACTIVITIES REGARDING GENERIC S/FETY ISSUES
Background
The NRC staff continuously evaluates che safety requirements used in its reviews against new information as it becomes available.
I,Jormation related to the safety of nuclear power plants comes from a variety oi sou ces.
Obvious sources of such information are experience from operating reactors, research results, NRC staff and Advisory Comaittee on R?o : tor safeguards 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 assure safe operation is assessed. This assessment includes consideration of the generic implications of the issue.
In some cases, immediate action is taken to assure safety, e.g., 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 prior to making licensing decisions. In most cases, howaver, the initial assessment indicates that immediate licensing actions or changes in licensing criteria are not necessary.
In any event, further study may be deemed appropriate to make judgments as to whether existing NRC staff requirements should be modified to address the issue for new plants or if backfitting is appropriate for the long term operation of plants already under construction or in operation.
These issues are sometimes called " generic safety issues" because they are related to a particular class or type of nuclear facility rather than a specific plant. These issues have also been referred to as " unresolved safety issues". However, as discussed above, such issues are considered on a generic basis only after the staff has made an initial assessment for individual plants and has made a determination that the safety significance of the issue does not prohibit continued operation or require licensing actions while the longer term generic review is underway.
These longer term generic studies were the subject of a decision by the Atomic Safety and Licensing Appeal Board of the Nuclear Regulatory Conmission.
The decision was issued on November 23,1977 ( ALAB-444) in connection vith the C-1
Appeal Board's consideration of the Gulf States Utility Company applica: ion for the River Bend Statior., Units Nos.1 and 2.
In the view of the Appeal Board, (pp. 25-29):
"The responsibilities of a licensing board in the radiological health and safety sphere are not confined to the consideration and disposition of those issues which may have been presented to it by a party or an " interested State" with the required degree of specificity. To the contrary, irrespective of what matters may or may not have been properly placed in controversy, prior to authorizing the issuance of a construction permit the board must make the finding, inter alia, that there is
" reasonable assurance" that "the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public." 10 CFR 50.35(a)
(
.... Of necessity, this determination will entail an inquiry f
into whether the staff review satisfactorily has come to grips with any unresolved generic safety problems which might have an impact upon operation of the nuclear facility under consideration."
"The SER is, of course, the principal document before the licensing board which refiac+s the content and outcome of the staff's safety review. The board s;.?"ld therefore be able to look to that document to ascertain the extent to which generic unresolved safety problems which have been previously identified in a TSAR item, a Task Action Plan, an ACRS report or elsewhere have been factored into the staff's analysis for the particular reactor -- and with what result.
To this end, in our view, each SER should contain a summary description of those generic problems under continuing study which have both relevance to facilities of the type under review and potentially significant public safety implications."
s "This summary description should include information of the kind now contained in most Task Action Plans. More specifically, there should be an indication of the investigative program which has been or will be undertaken with regard to the problem, the program's anticipated time-span, whether (and if so, what) interim measures have been devised for dealing with the problem pending the completion of the investigation, and what alternative courses of action might be available should the program not produce the envisaged result."
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"In short, the board (and the public as well) should be in a position to ascertain from the SER itself -- 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 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 i
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 availab'- to insure that continued operation (if permitted at all) would not pose an undue risk to the public."
In a related matter, as a resuit of Congressional action on the Nuclear Regulatory Commission budget for Fiscal Year 1978, the Energy Reorganization Act of 1974 was anended (PL 95-209) on December 13, 1977 to include, among other things, a new Section 210 as follows:
"UNRE50LVED SAFETY ISSUES PLAN" "SEC. 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 Commission thereafter."
The Joint Explanatory Statement of the House-Senate Conference Committee for the FY 1978 Appropriations Bill (Bill S.1131) provided the following additional information regarding tne Committee's deliberations on this portion of the bill:
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"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 conferees also expressed the inte.it 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 actions taken airectly or indirectly to develop and implement corrective measures; (2) further actions planned concerning such measures; and (3) time? '51es and cost estimates of such actions. The Commission sh.Jld 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 on the "NRC Program for the heso'." tion of Generic Issues Related to Nuclear Power Plants,"
(NUREG-0410).
The NRC program described in NUREG-0410 was already in place when PL 95-209 was enacted and is of considerably broader sccpe than the
" Unresolved Safety Issues Plan" required by Section 210. Although the NRC program does include plans for the resolution of generic technical issues of varying degrees of safety significance, it also includes generic tasks for the resolution of environmental issues, for the development of improvements in the reactor 'icensing proccss, and for consideration of less conservative design criterit or operating limitations in areas where over-conservatism may be urnecessarily restrictive.
The major elements of the NRC program are de' cribed below.
The NRC Generic Issues 'rogram A Technical Activitias Steering Committee was established to increase f
high level managem'.nt involvement in and to improve management oversight of generic technical activities.
The Steering Committee is chaired by l
the Deputy Direc.or, Office of Nuclear Reactor Regulation (NRR), and includes, as members, the four Division Directors in NRR.
The Committee's a
functions include assigning proposed generic tasks to priority categories, assigning lead responsibility to an NRR division for defining and executing each generic task, approving Task Action Plans, and regularly reviewing the progress of ongoing tasks.
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The Steering Committee's judgmental decisions regarding priorities and other matters, such as the assignment of an NRR division with lead responsibility and approval of the Task Action Plan for each task, are based upon recommendations resulting from an extensive internal review process.
This process begins in the NRR line organizations through their development, review, comment and concurrence on proposals for high priority tasks and Task Action Plans.
In addition, specific recommendations regarding these proposals are provided by the Steering Committee's Advisory Group following its detailed review.
The Advisory Group is made up of five senior technical staff representing each of the NRR divisions and the Director, NRR.
Implementation of the program began by making the judgments referred to above regarding the relative priority of a large number of ongoing, planned or suggested generic efforts.
The generic issues that were cen-sidered included those from the Advisory Committee on Reactor Safeguard's listing, those listed in NRR's former Technical Safety Activities Report, the 27 issues discussed in " Staff Discussion of Fifteen Technical Issues Listed in Attachment to November 3,1976 Memorandum from Director, NRR to NRR Staff "(NUREG-0138) and " Staff Discussion of Twevive Additional Technical Issues Raised by Responses to November 3,1976 Memorandum from Director, NRR to NRR Staff "(NUREG-0153),1/ and a number of other generic issues that were identified from a variety of sources as described earlier in this appendix (page C-1).
The Steering Committee adopted four priority category definitions as descriptive of the various g'neric technical issues. These definitions are presented in Table C.1.
As indicated by these definitions, issues were assigned to the various priority categories based on their judged safety, environmental or safeguards importance or their potential for improving the efficiency or effectiveness of the licensing process.
Initially, each of the four NRR divisions described and proposed to the Technical Activities Steering Committee those generic issues it considered to warrant the highest priority effort (Category A and Category B tasks).
Proposals were received for over 130 Category A tasks and cver 225 Category C tasks in April and May 1977, respectively. Many of these proposals were duplicates or could be readily combined as a part of another proposed generic
-1/ NUREG-0138 and NUREG-0153, published in November and December 1976 respectively, provided the staff's discussion of 27 technical issues identified by one or more members of the NRR staff as problems whose priority, progress or resolution was, in their opinion, unsati sf acto ry.
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issue. These proposals were reviewed in detail by the Steering Committee's Advisory Group. Following its review, the Advisory Group made recommendations to the Steering Committee for each task regarding the Priority Category to which it should be assigned and the NRR division that should be assigned lead responsibility.
The Steering Committee reviewed the division proposals and the recommendations of its Advisory Group, assigned each task to a Priority Category and designated an NRR division with lead responsibility (Leau Division) for each task.
Implementation of the NRC program has been a major effort that has required the participation of virtually every working and management level in NRR.
Decisions regarding the relative priorities of the hundreds of generic issues that have been suggested, although based on agreed upon criteria (Table C.1), in the final analysis are the product of the collective judgments of the individuals participating in the decision making process, in this case the preposing line organizations, the Steering Committee's Advisory Group, and the Steering Committee itself.
As indicated by the Priority Category definitions in Table C.1, those issues assigned to Category A are the most important generic tasks, including those most important in terms of safety significance.
The term " lesser... safety significance" in the Category B definition has not been defined. However, the report of the NRR Task Force (see NUREG-0410, Section 1.0) that originally developed the Priority Category definitions does offer some further insight.
In addition to developing the Priority Category definitions, the Task Force developed a set of criteria to be used to test each identified activity for assignment to the proper category.
The intent was that an activity, meeting one or more of the test criteria for a given category, would be assigned to that category.
The tests for Category A are:
1.
Resolution could remedy significant deficiencies in facility design or operation.
2.
Early resolution of issue could significantly improve the existing regulatory process.
3.
Other activities that are judged to require high level management attention and oversight.
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The tests for Category B are:
1.
Issue is important to safety, safeguards or environmental protection, but of smaller scope that does not require NRR wide coordination to obtain timely resolution.
2.
Resolution needed to confirm adequacy of previous staff judgments.
3.
Issue has potential of becoming a Category A issue.
From these tests it is important to note that any issue whose resolution is needed to "reax dy significant deficiencies in facility design or operation" would t e assigned to Category A.
Although some issues "important to safety" could be assigned to Category B, the intent was clearly not to assign issues that met the first Category A test to Category B.
Issues that are judged to meet the first Category A test are those that have the "potentially significant public safety implication (s)" referred to in River Bend.
Since such issues are not assigned to Category B, it is not necessary to meet all of the informational requirements of the River Bend decision for Category B, or lower category generic issues.
The third test for Category B indicates that issues assigned to Category B could be elevated to Category A if new information indicates that the issue is of greater importance to safety than originally judged.
None of the issues originally assigned to Category B, C or D had been elevated to Category A until recently.
Recent staff efforts have resulted in the elevation of several of these lower priority tasks as discussed later in this appendix.
Table C.2 provides a listing of the issues assigned to Priority Categories A, B, C and D by the Steering Committee from among those issues originally proposed as Category A and B tasks. A Task Action Plan has been approved by the Technical Activities Steering Committee for each of the Category A Tasks listed in Table C.2.
Category A Tasks Copies of Task Action Plans for the Category A tasks listed in Table C.2 are contained in " Task Action Plans for Generic Activities, Category A",
(NUREG-0371), published in November 1978. This is a revised version of the previous NUREG-0371 published in December 1977.
It is the staff's intention to maintain this document up to date by issuing revisions to the task action plans as they become available and adding new task action plans as new Category A issues are identified.
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s
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9 TABLE C.1 I
PRIORITY CATEGORY DEFINITIONS I
Category A:
Those generic technical activities judged by the staf f to warrant priority attention in tenns of manpower and/or funds to attain early resolution. These matters include those the resolution of which could (1) provide a significant increase in assurance of the health and safety of the public, or (2) have a significant impact upon the reactor licensing process.
Category B:
Those generic technical activities judged by the staff to be important in assuring the continued health and safety of the public but for which early resolution is not required or for which the staff perceives a lesser safety, safeguards or environmental significance than Category A matters.
Category C:
Those generic technical activities judged by the staff to have little direct or immediate safety, safeguards or environmental significance, but which could lead to improved staff understanding of particular technical issues or refinements in the licensing process.
Category D:
Those proposed generic technical activities judged by the staff not to warrant the expenditure of manpower or funds because little or no importance to the safety, environmental or safeguards aspects of nuclear reactors or to improving the licensing process can be attributed to the activity.
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TABLE C.2 LIST OF TECHNICAL ACTIVITIES Task No.
Title Category A Tasks A-1 Water Hammer A-2 Asyametric Blowdown Loads on PWR Primary Coolant Systems A-3 Westinghouse Steam Generator Tube Integrity A-4 Combustion Engineering Steam Generator Tube Integrity A-5 Babcock & Wilcox Steam Generator Tube Integrity A-6 Mark I Short Term Program A-7 Mark I Long Term Program A-8 Mark II Containment Pool Dynamic Loads A-9 ATWS A-10 BWR Nozzle Cracking A-ll Reactor Vessel Materials Toughness A-12 Fracture Toughness of Steam Generator and Reactor Coolant Puap Supports A-13 Snubber Operability Assurance A-14 Flaw Detection A-15 Primary Coolant System Decontamination and Steam Generator Chemical Cleaning A-16 Steam Effects on BWR Core Spray Distribution A-17 Systems Interaction in Nuclear Power Plants A-18 Pipe Rupture Design Criteria A-19 Digital Computer Protection Systems A-20 Impacts of the Coal Fuel Cycle A-21 Main Steam Line Break Inside Containment - Evaluation of Environmental Conditions for Equipment Qualification A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Response A-23 Containment Leak Testing A-24 Qualification of Class IE Safety-Related Equipment A-25 Nonsafety Loads on Class IE Power Sources A-26 Reactor Vessel Pressure Transient Protection (Overpressure Protection)
A-27 Reload Applications A-28 Increase in Spent Fuel Pool Storage Capacity A-29 Nuclear Power Plant Design for the Reduction of Vulnerability to Industrial Sabotage C-9
Table C.2 (Cont'd) l Task No.
Title A-30 Adequacy of Safety-Related DC Power Supplies A-31 RHR Shutdown Requirements A-32 Missile Effects A-33 NEPA Reviews of Accident Risks A-34 Instruments for Monitoring Radiation and Process Variables During Accidents A-35 Adequacy of Offsite Power Systems A-36 Control of Heavy Loads Near Spent Fuel A-37 Turbine Missiles A-38 Tornado Missiles A-39 Determination of Safety Relief Valve (SRV) Pool Dynamic Loads and Temperature Limits for BWR Containments A-40 Seismic Design Criteria - Short Term Program Category B Tasks B-1 Environmental Technical Specifications B-2 Forecasting Electricity Demand By State in the United States on an Annual Basis B-3 Event Categorization B-4 ECCS Reliability B-5 Ductility of Two-Way Slabs and Shells and Buckling Behavior of Steel Containment B-6 Loads, Load Combinations, Stress Limits B-7 5econdary Accident Consequence Modeling B-8 Locking Out of ECCS Power Operated Valves B-9 Electrical Cable Penetrations of Containment B-10 Behavior of BWR Mark III Containment B-ll Subcompartment Standard Problems B-12 Containment Cooling Requirements (Non-LOCA)
B-13 Marviken Test Data Evaluations B-14 Study of Hydrogen Mixing Capability in Containment Post-LOCA B-15 CONTEMPT Computer Code Maintenance B-16 Protection Against Postulated Piping Failures in Fluid Systems Outside Containment B-17 Criteria for Safety-Related Operator Actions B-18 Vortex Suppression Requ' 1ments for Containment Sumps B-19 Thermal-Hydraulic Stability C-10
Table C.2 (Cont'd)
Task No.
Title B-20 Standard Problem Analysis B-21 Core Physics B-22 LWR Fuel B-23 LMFBR Fuel B-24 Seismic Qualification of Electrical nf Mechanical Components B-25 Piping Benchrerk Problems B-26 Structural Integrity of Containment Penetrations B-27 Implementation and Use of Subsection NF B-28 Radionuclide/ Sediment Transport Program B-29 Effectiveness of Ultimate Heat Sinks B-30 Design Basis Floods and Probability B-31 Dam Failure Model B-32 Ice Effects on Safety-Related Water Supplies B-33 Dose Assessaent Methodclogy B-34 Occupational Radiation Exposure Reduction B-35 Confirmation of Appendix I Models for " Calculations of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Light-Water-Cooled Power Reactors" B-36 Develop Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units for Engineered Safety Feature Systems and for Normal Ventilation Systems B-37 Chemical Discharges to Receiving Waters B-38 Reconnaissance Level Investigations B-39 Transmission Lines B-40 Effects of Power Plant Entrainment on Plankton B-41 Impacts on Fisheries B-42 Socioeconomic Environmental Impacts B-43 Value of Aerial Photographs for Site Evaluation B-44 Forecasts of Generating Costs of Coal and Nuclear
'lants B-45 Need for Power - Energy Conservation B-46 Costs of Alternatives in Environmental Design B-47 Inservice Inspection Criteria for Supports and Bolting of Class 1, 2, 3 and MC Components B-48 BWR CRD Mechanical Failure (Collet Housing)
B-49 Inservice Incpection Criteria for Containment C-ll
Task No.
Title B-50 Requirements for Post-0BE Inspection B-51 Assessnent of Inelastic Analysis Techniques B-52 Fuel Assembly Seismic and LOCA Responses B-53 Load Break Switch B-54 Ice Condenser Containments B-55 Improved Reliability of Target-Rock Safety-Relief Valves B-56 Diesel Reliability B-57 Station Blackout B-58 Passive Mechanical Failures B-59 Review of (N-1) Loop Operation in BWRs and PWRs B-60 Loose Parts Monitoring Systems B-61 Allowable ECCS Equipment Outage Periods B-62 Reexamination of Technical Bases for Establishing SLs, LSSSs, etc.
B-63 Isolation of Low Pressure Systems Connected to RCPB B-64 Decommissiong of Reactors B-65 lodine Spiking B-66 Control Room Infiltration Measurements B-67 Effluent and Process Monitoring Instrumentation B-68 Pump Overspeed During a LOCA B-69 ECCS Leakage Ex-containment B-70 Power Grid Frequency Degradation and Effect on Primary Coolant Pumps B-71 Incident Response B-72 Development of Models for Assessing Risk of Health Effects and Life Shortening from Uranium and Coal Fuel Cycles B-73 Monitoring for Excessive Vibration Inside the Reactor Pressure Vessel Category C Tasks C-1 Assurance of Continuous Long-Term Integrity of Seals on Instrumentation and Electric Equipment C-2 Study of Containment Depressurization by Inadvertent Spray Operation to Determine Adequacy of Containment External Design Pressure C-3 Insulation Usage Within Containment C-4 Statistical Methods for ECCS Analysis C-5 Decay Heat Update C-6 LOCA Heat :ources C-12
Table C.2 (Cont'd)
Task No.
Title C-7 PWR System Piping C-8 Main Steam Line Leakage Control System C-9 RHR Heat Exchanger Tube Failure C-10 Effective Operation of Containment Sprays in a L0rA C-ll Assessment of Failure and Reliability of Pumps and Valves C-12 Primary System Vibration Assessnent C-13 Non-Random Failures C-14 Storm Surge Model for Coastal Sites C-15 NUREG Report for Liquid Tank Failure Analysis C-16 Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents for Radioactive Solid Wastes Category D Tasks D-1 Advisability of a Seismic Scram D-2 Emergency Core Cooling System Capaoility for Future Plants D-3 Control Rod Drop Accident (BWRs)
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NUREG-0371 meets most of the informational requirements of ALAB-444 for the Category A tasks applicable and relevant to the Erie proceeding. The Task Action Plans in NUREG-0371 provide a description of the problems; the staff's approaches to resolution; general discussions of the bases upon which continued plant licensing or operation can proceed pending completion of the task; technical organizations involved in the review and estinutes of the manpower required; descriptions of the interactions with other NRC offices, the Advisory Committee on Reactor Safeguards and outside organizations; estimates of any funding required for contractor supplied technical assistance; prospective dates for completing the tasks; and descriptions of any potential problems that could impact the plans.
Although 133 Category A, B, C and D generic tasks were identified by the staff and approved by the Steering Committee as listed in Table C.2, not all are applicable to each type or vintage of plant.
For example, in the case of Erie Nuclear Plant Units 1 and 2, the following Category A task are not applicable because they are peculiar to boiling water reactors:
A-6 Mark I Containment Short Term Program A-7 Mark I Containment Long Term Program A-8 Mark Il Containment Program A-10 BWR Nozzle Cracking A-16 Steam Effects on BWR Core Spray Distribution A-39 Determination of Safety Relief Valve (SRV)
Pool Dynamic Loads and Temperature Limits for BWR Containments The following tasks are not applicable because they are peculiar to pressurized water reactors supplied by other NSSS vendors:
A-3 Westinghouse Steam Generator Tube Integrity A-4 Combustion Engineering Steam Generator Tube Integrity Other generic tasks are not relevant to the licensing actions for any particular facility because they deal with improving the efficiency and/Cr effectiveness of the licensing process rather than plant safety.
These types of tasks include (1) efforts to improve guidance to applicants, C-14
licensees or staff reviewers, and (2) efforts to consider the relaxaticn of certain staff requirements that may be overly conservative. 2/ Category A tasks that fall into the first group are:
A-15 Primary Coolant System Decontamination and Steam Generator Chemical Cleaning A-19 Digital Computer Pro ection Systems A-27 Reload Application Guide A-28 Increase in Spent Fuel Storage Capacity A-34 Instruments for Monitoring Radiation and Process Variables During Accidents A-37 Turbine Missiles Category A tasks that fall into the second group are:
A-25 Nonsafety Loads on Class IE Power Sources A-38 Tornado Missiles A-40 Seismic Design Criteria - Short Term Program Other Category A tasks listed in Table C.2 are not relevant to the informational requirenents of ALAB-444 because they deal with environmental issues rather than safety issues. These include:
A-20 Impacts of Coal Fuel Cycle A-33 NEPA Reviews of Accident Risks The remaining 21 Category A generic tasks are, to varying degrees, related to plant safety and are applicable to the Erie Nuclear Plant. A discussion of 2/ Because the staff's view is that these types of generic tasks are not relevant to the licensing actions for any particular facility, they have not been addressed specifically for the Erie Nuclear Plant in this appendix. However, the Task Action Plans for these tasks are included in NUREG-0371.
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each of the issues addressed by these 21 plans as they relate to the Erie Nuclear Plant is provided in a subsequent section of this appendix.
Even these 21 tasks can be subdivided to better characterize the type of issue or the degree of safety significance of each.
For example, the following Category A tasks are best characterized as confirmatory, i.e.,
our present safety requirements are judged to be adequate; however, further study is believed to be prudent to confirm this judgment, perhaps on a more quantitative basis.
A-17 Systems Interaction in Nuclear Power Plants A-18 Pipe Rupture Design Criteria A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Building Response A-30 Adequacy of Safety-Related DC Power Supplies A-32 Missile Effects Category B, C and D Tasks Initially, the Category B tasks listed in Table C.2 were judged by the Steering Ccmmittee to be of lesser safety, safeguards or environnental significance than Category A tasks. Category C tasks listed in Table C.2 were judged to have little direct or immediate safety, safeguards or environmental significance and Category D tasks listed in Table C.2 were judged to have little or no importance to the safety, safeguards or environmental aspects of nuclear reactors or to improving the licensing process.
The staff has compiled a brief description of the Category B, C and D tasks listed in Table C.2 in " Generic Task Problem Descriptions, Categcry B, C and D Tasks", (NUREG-0471) published in June 1978. tio Task Action Plans have as yet been approved by the Technical Activities Steering Committee for these tasks.
Subsequently, the NRC staff conducted a preliminary evaluation of the relative significance of each of the Category A, B, C and D tasks from the standpoint of risk.
A draft report of this preliminary evaluation currently is undergoing peer and management review. The preliminary results of the risk-based evaluation indicate that the safety significance of several of the lower priority (Category B and C) tasks could be greater than indicated by the initial category assignments of the Steering Committee. With regard to the preliminary evaluation, bounding calculations of the risk C-16
(probability tines consequences) associated with accident sequences related to issues in the NRR generic issues - ogram were performed, where possible, using fault tree / event tree analyses.
The preliminary results of this effort and the initial results of the peer and management review indicate that the following seven Category B tasks and one Category C task have greater potential risk significance than originally judged:
B-30 Design Basis Floods and Probability B-34 Occupational Radiation Exposure Reduction B-55 Improved Reliability of Target Rock Safety Relief Valves B-57 Station Blackout B-63 Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary B-64 Decommissioning of Reactors C-3 Insulation Usage Within Containment B-18 Vortex Suppression Requirements for Containment Sumps In light of these prelimincry results and other considerations, the Technical Activities Steering Committee has recently acted to elevate Tasks B-18, B-57 and C-3 to Category A.
Tasks B-30, B-34, B-55, B-63 and B-64 were reconsidered by the Steering Committee, but were not elevated. Nonetheless, individual discussions of the safety significance of all these tasks (except B-55 which applies to boiling water reactors) as they relate to the Erie Nuclear Plant are provided below.
In the discussions, as in Table C.2, the original designations for Tasks B-18, B-57 and C-3 are retained, although they are being elevated to Category A.
The remaining Category B, C and D tasks listed in Table C.2 are not all applicable to all facilities and include tasks for improving regulatory guidance, for considering the relaxation of current requirenents and for confirming the adequacy of current requirements.
In addition, there are a nunter of tasks on environmental issues and a number directed at maintaining or improving staff capabilities to perform independent or audit calculations.
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Of those remaining Category B, C and D tasks that are related to plant safety and are applicable to the Erie Nuclear Plant, we have not identified any that could not be resolved either by system alterations using available techniques and equipment or by operational modifications in the event that our review of the issue revealed that current requirements required upgrading during construction or operation. On this basis and the Steering Committee's judgment that the Category B, C and D issues are of lesser safety significance than Category A issues, detailed information on these tasks is not necessary and we have not included any such informatien in this appendix for the remaining B, C and D tasks.
Discussion of Tasks As They Relate to the Erie Nuclear Plant A-1 Water Hanner As indicated in NUREG-0371, November 1978, since 1971 there have been about 100 incidents involving water hammer in pressurized water reactors and boiling water reactors. The water hammers have involved steam generator feedrings and piping, the decay heat removal system, emergency core cooling system, containment spray lines, service water lines, feedwater lines and steam lines. The systems most frequently affected by water hammer effects are the feedwater systems of pressurized water reactors. However, pressurized water reactors with Babcock and Wilcox once-through steam generators (steam generators similar to those to be installed at the Erie Nuclear Plant) have not experienced damaging water hammer incidents. While requirements have been defined for pressurized water reactors with steam generators with top feedrings, no requirements for modi-fications to the feedring in the once-through steam generator have been defined.
ndicated in Section 3.0 of the Task Action Plan for Task A-1, adequate pro-tection is currently provided against potential water hammer in other plant systems. We consider that the applicants have fulfilled the preliminary design requirements necessary at the construction permit stage of review.
Accordingly, our previous conclusions in the Erie Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-2 Asynretric Blowdown Loads on PWR Primary Coolant Systems The applicants have complied with all current staff requirements regarding this safety issue (Safety Evaluation Report, page 3-11). Analytical methods for assessing asymmetric blowdown loads for Babcock & Wilcox designs are being reviewed by the staff. A final, plant-specific analysis for the Erie Nuclear Plant will be completed during the operating license review for this f acility.
Accordingly, our previous conclusions in the Erie Nuclear Plant Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
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A-5 Babcock & Wilcox Steam Generator Tube Integrity The applicants have complied with current staff requirements for the con-struction permit stage of review.
Specific measures that will be taken by the applicants to ensure that the tubes will not be subjected to conditions that will cause deleterious wastage or cracking are described in Appendix A to the Erie Safety Evaluation Report on page 5-17 ard 5-18.
Based on our review of these measures, we concluded that construction permits for the Erie Nuclear Plant, Units 1 and 2 can be issued with reasonable assurance that there will be no undue risk to the health and safety of the public.
The efforts under Task A-5 regarding steam generator tube integrity may result in improved criteria that provide further assurance in this regard.
However, such improvet.ents are likely to be procedural rather than system modificatior and their application to the Erie Nuclear Plant is a matter that c reasonably be left to the operating license stage of review. Accoi.ngly, our previous conclusions in tia Erie Nuclear Plant Safety Eva'.uation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
_A-9 Anticipated Transients Without Scram Work on Revision 1 to the Task Action Plan for this generic issue is still underway.
The applicability of this issue to the Erie Nuclear Plant has been discussed in the Safety Evaluation Report for the Erie Nuclear Plant (pages 15-2 and Appendix A, page 15-14).
In the Safety Evaluation Report we concluded that any modifications necessary for the BSAR-205 class of plants can be considered in our review of the final design of the Erie Nuclear Plant at the operating license stage of review or in our review of the BSAR-205 Final Design Application. The following provides our basis, meeting the ALAB-444 criteria, in support of a decision on issuance of construction permits for the Erie Nuclear Plant, Units 1 and 2.
Because of the perceived potential for serious consequences resulting from anticipated transients without scram, a number of studies have been under-taken to assess the probabilities and consequencues of such events.
These studies have been performed by vendors, utility groups (i.e., Electric Power Research Institute), and by the NRC staff.
The latest NRC staff assess-ment is described in Volure 3 of NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors", dated December 1978.
In that document the staff has reaffirned that the present likelihood of severe consequences arising from an anticipated transient without scram event is acceptably small and presently there is no undue risk to the public from anticipated transients without scram.
This conclusion is based on engineerirg judgment in view of:
(1) the estimated arrival rate of anticipated transients with C-19
potentially severe consequences in the event of scram failure; (2) the favorable operating experience with current scram systems; and (3) the limited nunber of operating reactors.
Simply stated it is the staff's judgment that the individual and societal risk from anticipated transients without ; cram have been, and are today, acceptably snell.
The document also states that the staff has maintained since 1973 and reaffirms today that the future likelihood of severe consequences arising from an anticipated transient without scram event could become unacceptably large and measures should be taken to diminish it.
This conclusion is based on engineering judgment in view of:
(1) the potential severity of dntiLipated transients without scram; (2) the increasing number of nuclear power plants; ano (3) the need to maintain and improve further the safety margins provided to protect the public.
Simply stated, it is the staff's judgment that actions should be taken to require specific hardware changes that diminish the future contribution of anticipated transients without scram to the overall societal and individual risk arising from nuclear power plants.
In Volume 3 of NUREG-0460, the recommended hardware changes for rule making considerations are identified.
The recommended chances for a unit depend on (1) the unit's status (operating or under construction) and for plants now under construction, the construction permit issuance date; (2) reactor vendor (e.g., Babcock & Wilcox or Westinghouse); and (3) whether the appli-cation (a) utilizes a standardized nuclear steam supply system or balance-of-plant, and/or (b) replicates or duplicates a previous desigr.
For the Erie Nuclear Plant, Units 1 and 2, the report recommends that the design include diverse actuation circuitry for mitigation systems and increased relief capacity through the addition of pressurizer safety valves.
The report also recommends the demonstration, under the conditions calculated for specified anticipated transients without scram events, of the opera-bility for the valves needed for long-term cooling.
The report states that these modifications can begin as lat^ as four years before completion of construction without delaying a unit's startup.
Volume 3 cf NUREG-0460, which describes the rationale for specifying these modifications, is currently being reviewed by the Advisory Committee on Reactor Safeguards. The Regulatory Requirements Review Committee has completed its review and concurred with the staff approach described in Volume 3 of NUREG-0460 with regards to the Erie Nuclear Plant, Units 1 and 2.
We plan to issue requests to industry to supply generic analyses of anticipated transients without scram mitigation capability and anti-cipate presenting to the Conmission in May 1979 our recommendations for its actions to resolve the anticipated transients without scram issue.
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The above modifications are feasible for the units proposed for the Erie Nuclear Plants, Units 1 and 2.
As part of our ongoing review of this issue we have requested that the applicants provide a commit, rent that the Erie Nuclear Plant, Units 1 und 2 will be designed such that implementation of the above potential requirements will not be precludeded by design and con-struction prior to a final decision on the anticipated transients without scram issue.
In summary we conclude that, with the above commitment, the applicants for the Erie Nuclear Plant, Units 1 and 2 will have fulfilled the preliminary design requirements necessary at the construction permit stage of review and that a satisfactory solution to this generic task will be obtained before the Erie Nuclear Plant Units 1 and 2 are put in operation.
Therefore, there is reasonable assurance that the proposed Erie Nuclear Plant, Units 1 and 2 facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public.
A-ll Reactor Vessel Materials Toughness Section 3 of the Task Action Plan for Task A-ll indicates that current criteria related to fracture toughness together with the materials currently employed for reactor vessel fabrication are adequate to ensure suitable safety margins for the reactor vessels throughout their design lives. As indicated in the BSAR-205 Safety Evaluation Report (Appendix A to the Erie Safety Evaluation Report) on pages 5-12 and 5-13, the reactor vessels for the Erie Nuclear Plant will be designed, fabricated, tested and operated in conformance with current criteria and materials.
Our conclusion on page 5-13 of Appendix A to the Safety Evaluation Report indicates that conformance with such criteria will ensure adequate safety margins during operating, testing, maintenance and postulated accident conditions. This conclusion regarding postulated accident conditions was a judgment made by the staff based principally on empirical results from NRC experiments at the Oak Ridge National L aboratory. Further testing is currently being considered.
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As part of Task A-ll, we are evaluating reactor vessel material toughness under postulated accident conditions.
The results of this evaluation may indicate that some operating plants have to take measures late in life to assure adequate fracture toughness for postulated accidents.
A preliminary review indicates that adequate safety margins can be maintained even for these older plants for postulated accidents for up to approximately 20 years of neutron irradiation. Although this part of Task A-ll could conceivably provide information relevant to current
/intage reactor vessels, we believe that these vessels will have ddequate safety margins related to fracture toughness for postulated accidents throughout their design life. Accordingly, we anticipate that the task results will confirm our previous conclusions presented in the Erie Nuclear Plant Safety Evaluation Report.
A-12 Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports The applicants have committed to comply with current staff requirements.
Sectica 3 of the Task Action Plan for Task A-12 provides the basis for continued licensing pending completion of this generic task. On the basis of this discussion, we have concluded that our previous conclusions in the Erie Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic effort.
A-13 Snubber Operability Assurance The applicants will be required to describe the program for snubber operability assurance at the operating license stage of review. We anticipate that prior to the operating license review, our generic efforts under Task A-13 will provide comprehensive requirements for snubber operability assurance for use in the staff's review at that tin:e.
- Further, we expect that any future requirements that may be required for the Erie Nuclear Plant can be satisfactorily implemented at the operating license stage. Accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
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A-14 Flaw Detection As described in the Task Action Plan for Task A-14, our generic efforts, in conjunction with those being undertaken by industry organizations, may result in improved inspection techniques for the detection of defects in the reactor coolant pressure boundary. However, as described in Section 3 of the Task Action Plan, such improvements, although desirable, are not necessary to maintain adequate margins of safety. Accordingly, our previous conclusions in the Erie Nuclear Plant Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-17 Systems Interactions in Nuclear Power Plants The licensing requirements and procedures used in our review address many different types of systems interactions.
Task A-17 has been developed to confirm that our current review process encomp3sses all potentially adverse systems interactions.
As indicated in Section 3 of the Task Action Plan for Task A-17, we anticipate that this task will confirm that curi ant licensing requirements and procedures are adequate, although some modifications for improvement in the review procedures and licensing requirements may be made. On this basis, we believe that our previous conclusions in the Erie Nuclear Plant Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-18 Pipe Rupture Design Criteria The applicants have complied with all current staff safety requirments regarding pipe rupture design as discussed in Sections ?-
of the Safety Evaluation Report and in Section 3.6 of the BSAR-205 Sai"" Evaluation Report '
(included as Appendix A to the Erie Nuclear Plant Safety Lvaluation Report).
As indicated in Section 3 of the Task Action Plan for A-18, the task may result in adjustments to the current criteria to achieve a better balance between design of piping systems for normal operation and design to assure adequate protection against postulated pipe rupture. Such adjustments are desirable, but not necessary, to assure that plants such as the Erie C-23
.g
Nuclear Plant that meet current requirements, have adequate protection against pipe breaks. Accordingly, our previous conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-21 Main Steam Line Break Inside Containment As indicated in the Task Action Plan for Task A-21, the generic task will establish the acceptability of steam generator blowdown and containment analysis models to be used to calculate the wor st case main steam line break for equipnent qualification. The results of the generic task are expected to be available for use in the operating licensa review for the Erie Nuclear Plant. As discussed in the Safety Evaluation Report on page 6-3, the applicants have committed to use the peak temperature and pressure results of the worst case main steam line break analysis for Erie Nuclear Plant for the qualification of safety-related electrical equipment located inside containment.
We have concluded that this commitment can be carried out during construction and is acceptable at the construction permit stage of review. Accordingly, our previous conclusions in the Erie Nuclear Plant Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-22 PWR Main Steam Line Break - Core, Reactor Vessel and Containment Building Response This generic task is expected to confirm our prior determination that the postulated main steam line break accident has been conservatively evaluated using current staff licensing requirements.
The generic study, when completed, will include an evaluation of the reliability of non-safety grade equipcent and an evaluation of the adequacy of certain safety systems and operator actions necessary to mitigate the consequences of the main steam line break.
Further discussion of this matter is provided in the Task Action Plan for Task A-22.
No increases in requirements are foreseen that would require design changes or other modifications to the Erie Nuclear Plant as currently proposed by the applicants. How-ever, if design modifications are warranted, we will require that such changes are implemented. Accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this generic task.
A-23 Containment Leak Testing As reported in the Safety Evaluation Report at page 6-12, we have reviewed the applicants' containment leak testing program description and have C-24
concluded that it is acceptable for the construction permit stage of review.
Details of the applicants' containment leak testing program will be reviewed for conformance to Appendix J to 10 CFR Part 50 at the operating license stage of review. As indicated in Section 3 of the Tcrk Action Plan for Task A-23, the generic effort is to develop proposed cnanges to Appendix J to clarify its appl cation and to resoive any conflicting or impractical requirements. Acccrdingly, our previous conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-24 Qualification of Class IE Safety-Reisted Equipment The applicants have complied with all staff requirements at the construction permit stage of review, as reported in the Safety Evaluation Report on page 3-24.
In meeting these requirements, the applicants stated that Babcock & Wilcox would provide a topical report that is applicable to Erie Nuclear Plant that will describe a qualification program in conformance with IEEE 323-1974 for Clas; IE equipment supplied by Babcock & Wilcox.
We concluded that this commitment was acceptable for the construction permit stage of review, but that we will require the review and acceptance of this topical report to be completed prior to the operating license stage of our review. Task A-24 was developed to provide a mech.inism for conducting the generic review of equipment qualification progran; of the major nuclear steam supply system vendors and balance-of-plant equipnent suppliers.
The review of the Babcock & Wilcox qualification program discussed above is encompassed by this task and is a convenient and efficient means of meeting the Safety Evaluation Report requirement to complete this review by the operating license stage for the Erie Nuclear Plant. Accordingly, our previous conclusions in the Safety Evaluation Report regarding issuance of a construction permit are unaffected by this ongoing generic task.
A-26 Reactor Vessel Pressure Transient Protection (Overpressure)
The applicants' commitment to provide an overpressure protection system, and the staff approval of that commitment, is documented in the Safety Evaluation Report on page 5-4 of Appendix A.
Generic Task A-26 to define the criteria for overpressure protection system design and operations is essentially complete and the applicants previous commitments are consistent with the criteria resulting from the task. As stated in the Safety C-25
~.
Evaluation Report, we will review the details of the design and procedural provisions during the operating license stage of review. Accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this generic task.
A-29 Nuclear Power Plant Design for the Reduction of Vulnerabili.y to Industrial Sabotage The applicants have met all current requirements for a construction permit application, as discussed in the Safety Evaluation Report at page 13-7.
At the operating license stage of review we will require that the applicants demonstrate compliance with 10 CFR Part 73.55. As indicated in Section 3 of the Task Action Plan for Task A-29, the implemertation of 10 CFR Part 73.55 provides high assurance of protection of the health and safety of the public and accordingly, is an adequate basis on which to license the Erie Nuclear Plant. Although Task A-29 may identify design concepts that could provide alternative or more effective means of achieving protection against industrial sabotage, the implementation of such design improvements are not necessary to provide adequate protection of the Erie Nuclear Plant.
Accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-30 Adequacy of Safety-Related DC Power Supplies The staff's Mfort on this generic task is expected to confirm that the simultaneous and independent failure of redundant direct current power supplies is so unlikely as to be incredible, and that their failure from a common event is judged to have a low enough probability that adequate protection presently exists. This is the current staff view as discussed in Section 3 of the Task Action Plan for Task A-30.
Therefore, although the generic study will provide a quantitative assessment of reliabilities of direct current power supplies, particularly with respect to common mode failures, we have concluded that continued licensing and operation of nuclear power plants with the direct current power supply system designs now in use and proposed does not present an undue risk to the health and safety of the public. Accordingly, our previous conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-31 RHR Shutdown Requirements This generic task has been completed, and revised requirements have been developed and approved. These requirements were embodied in Regulatory Guide 1.139 " Guidance for Residual Heat Removal" issued for comment in C-26
May 1978. As indicated in the implementation section of the guide, implementation of these requirements will be reviewed on a case-by-case basis for all olants for which construction permit or Preliminary Design Approval applications were docketed before January 1,1978. The applicants have committed to the revised requirements.
A-32 Missile Effects As indicated in Section 3 of the Task Action Plan for Task A-32, the results of this generic task are expected to confirm the adequacy and conservatism of the current NRC licensing criteria for protection from the effects of impact from missiles caused by accide,its, tornados, or failure of the main turbine.
Based on this and the other information provided in Section 3 of the Task Action Plan, we have concluded that our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
A-35 Adequacy of Offsite Pnwer Systems The applicants have complied with all current staff requirements :s discussed in the Safety Evaluation Report on page 8-1.
As indicated in Section 3 of the Task Action Plan for Task A-35, if the task identifies areas where current criteria should be modified to increase safety margins, such modifications are not expected to be extensive. Further, for current construction permit applications such as the Erie Nuclear Plant, any forthcoming requirements are expected to be available for consideration for these applications well in advance of a decision on issuance of operating licenses. Accordingly, our previous conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by this ongoing generic task.
A-36 Control of Heavy Loads Near Spant Fuel The applicants have complied with our current criteria for fuel handling systems, as discussed on page 9-4 of the Safety Evaluation Report. As indicated in Section 3 of the Task Action Plan for Task A-36, any revisions to current requirements that might evolve from this generic study will be available well in advance of the planned operation of the Erie Nuclear Plant.
Further, the types of changes in currer.t requirements, if any, that could result are expected to be procedural and therefore, their implementation could reasonably be left to the operating license stage of review.
Accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this ongoing generic task.
C-27 s
e
C-3 Insulation Usage Within Containment The purpose of this task, as noted in NUREG-0471, is to gain a better understanding of how insulation might behave under pipe break accident conditions. Of principal concern is that under pipe break accident conditions, insulation might break loose from the piping resulting in the potential for impairing the effectiveness of the containment emergency sumps.
The preliminary results of the risk-based evaluation noted that blockage of the containment sump by pieces of insulation following a loss-of-coolant accident could lead to a serious accident, and for this reason, Task C-3 was determined to be potentially risk si t.iticant.
The preliminary evaluation notes, however, that proper design of the sumps and insulation should assure that the probability of such an event sequence is so small as to make the risk negligible.
Each unit of the Erie Nuclear Plant is designed with two functionally and physically independent containment sumps which are in full conformance with the criteria of Regulatory Guide 1.82, " Sumps For Emergency Core Cooling and Containment Spray Systems", (see Erie Preliminary Safety Analysis Report (PSAR), PSAR Section 6.2.2, page 6-2-3).
Conformance with this Regulatory Guide provides a high degree of assurance that the sumps are not susceptible to blockage by debris.
In addition, as stated in the Erie PSAR (Section 6.2.2, page 6.2-34) piping and equipnent insulation inside the reactor building will use metal reflective insulation which will not break up into small pieces. Such insulation as might be torn off by a postulated pipe rupture would sink to the bottom of the containment and would not contain floating particles that could block the pump strainers or cause a pump failure.
We conclude that the Erie Nuclear Plant design adequately considers the potential for sump blockage from insulation within containment and accordingly, our previous conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this generic task.
B-18 Vortex Suppression Requirements for Containment Sumps As described in NUREG-0471, problems with vortex formation have been noted during tests of containment sumps. This task is designed to develop criteria for sump design so as to avoid vortex formation and to define criteria for sump testing.
C-28
This task is related to Task C-3 in that the concern of both tasks is the reliability of operation of containment sumps.
The intent is to provide a high degree of assurance that there will be an ample supply of water for long-term cooling in the recirculation moae following a postulated loss-of-coolant accident. This matter is addressed in Regulatory Guide 1.79, "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors" which calls for testing of con-tainment sumps to verify vortex control and acceptable pressure drops across screening and suction lines and valves.
The preliminary risk-based evaluation identified Task B-18.s having negligible risk potential. However, during the peer and management review of this risk-based evaluation, it was determined that the overall concern regarding sump operability is sufficient to warrent elevating Task B-18 to a Category A task. Accordingly, Tasks B-18 and C-3, as that task relates to sump blockage, are being combined as a new Category A task to address containment sump reliability.
For the Erie Nuclear Plant, the applicants plan comprehensive tests to demonstrate the operability of the containment sumps in conformance with Regulatory Guide 1.79.
The Erie Safety Evaluation Report in Section 6.3.7 (page 6-17) notes the applicants' commitment to perform preoperational tests for the Erie emergency core cooling systems, including a test to demonstrate no adverse vortex formation during recirculation ecoling.
Thus, this issue is resolved for the Erie Nuclear Plant and, accordingly, our conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by the results of this generic task.
B-30 Design Basis Floods and Probability The purpose of this task was to prepare a paper for presentation to the Advisory Committee on Reactor Safeguards (ACRS) detailing the basis for design basis flood events used by the NRR staff in case reviews.
Additionally, the task was to address the possible use of probability estinates for the principal flood producing events.
This task has been completed and a report to the ACRS was issued in July 1977. The report presents a discussion and definitions of the Probable Maximum Flood events which may be used as Design Basis Floods for the review of nuclear power plants.
It also presents arguments for continued use by the staff of a deterministic approach for identifying the Probable Maximum Flood events in preference to possible use of a probabilistic approach. As indicated in the report, the NRR staff does not feel that C-29
- p.,
s a probabilistic approach is ap3ropriate at the present tiae "because of the lack of confidence one las in estimates of extreme events using current techniques and statistics". Ongoing research efforts, being conducted by NRC's Office of Regulatory Research (RES) in response to a request by NRR, are aimed toward developing improved methodological techniques for the probabilistic analysis of flooding.
The preliminary resuits of the risk-based evaluation discussed on page C-16 indicate that the probability of a flood-induced core meltdown accident at most sites is very low. However, the study notes that detailed probabilstic estimates have not been performed. On this basis, this task was rated as potentially risk significant in the preliminary evaluation.
The NRR staff has concluded that its present, determinist.. approach to selecting the design basis flood is adequately conservative and is the preferred method, particularly because there is low confidence in the probabilistic approach. Using the determinist 1c approach, the evaluation of Erie site led to the conclusion that the plant is sJequately designed to accomodate the design basis flood, as noted in Acction 2.4.2 of the Erie Safety Evaluation Report. Thus, this issue has been resolveo 'or the Erie Nuclear Plant, and accordingly, j
our conclusions in the Safety Evaluation Report regarding issuance of construction permits are unaffected by the results of this generic task.
B-34 Occupational Radiation Exposure Reduction The concern of this task is the development of additional criteria and guidelines to provide an improved basis for the staff to review rector plant designs and projected operations to support full implementation dt light water reactors of our reculatory requirement that occupational radiation exposures should be maintained as low as is reasonably achiev-able. The best guidance currently available is provided to applicants in Regulatory Guide 8.8," Information Relevant to Ensuring That Occupa-tional Radiation Exposures At Nuclear Power Stations Will Be As Low As Is Reasonably Achievable."
The preliminary results of the risk-based evaluation of generic tasks rates this task as potentially risk significant because occupational radiation exposures at operating nuclear facilities are averaging roughly 500 man-rem per reactor year and have generally been increasing with time.
C-30
Reduction of occupational exposures can be very important to reducing the overall radiologically-associated risks associated with the nuclear reactor industry. The preliminary evaluation does not indicate what, if any, reduction in occupational exposures could be achieved by completing this generic task.
This assessment of the significance of occupational exposures in the preliminary risk-based evaluation is consistent vith the NRR staff's view of the importance of occupational radiation exr: are reduction, as evidenced by the requirement to maintain such exposures as low as is reasonably achievable.
The staff review process at present provides reasonable assurance that occupational radiation exposures will be as low as is reasonably achievable.
Task B-34 is ained at upgrading the staff review acceptance criteria and providing more comprehensive guidance to licensees and applicants to reduce exposures. The guidance in Regulatory Guide 8.8 has been used for a number of years.
This task will draw from that experience and, with the aid of the results from supplementary studies and from research projects by the Department of Energy and the Electric Power Research Institute, will de alop additional criteria regarding techniques and methods to maintain occupational radiation exposure as low as is reasonably achiev-i able.
In the interim, currant licensing reviews, including that for the Erie Nuclear Plant, are being taken on a caseby-case basis and rely on existing staff criteria and requirements which are believed adequate to assure public health and safety.
A summary of the staff's review of the Erie Nuclear Plant related to occupational exposures is provided in Section 12.1 of the Erie Safety Evaluation Report. We concluded there that the applicants intend to design, construct and operate the Erie Nuclear Plant in such a manner that occupational radiation exposures will be as low as is reasonably achievable, and in accordance with Regulatory Guide 8.8.
Thus, the applicants are meeting the current staff criteria to assure that radiation exposures will be as low as are reasonably achievable and, accordingly, our conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this generic task.
B-57 Station Blackout This task is concerned with the capability inherent in a nucler.r plant to mitigate the consequences of a total loss of alternating current power (station black-out), i.e., a failure of the offsite power supply coupled with a fat ore of the emergency diesel generators. Accomodation of station blackout is not now a design basis requireaent for light water reactors; however, current staff criteria do require diverse drives for auxiliary C-31 h
e 9
feedwater system pumps.
The purpose of Task B-57 is to review the adequacy of current licensing requirements. The preliminary results of the risk-based evaluation indicate that Task B-57 is potentially risk significant on the basis of a risk assessment that considered pressurized water reactor plants not specifically designed for station blackout coaditions; i.e., for those plants which either do not have a steam driven auxiliary feedwater system or which may have a high degree of dependency of alternating current power in order to effectively operate or initiate operation of such a system.
As noted in Section 10.4.4 (page 10-5) of the Erie Safety Evaluation Report the Erie Nuclear Plant is designed with redundant emergency feedwater systems which include steam driven pumps as one of the diverse subsystems to provide emergency feedwater in the event of a loss of onsite and offsite power. All valves and controls necessary for operation of this subsystem use direct corrent rather than alternating currcnt power. Thus, the design of the Erie Nuclear Plant meets current staff requirements with regard to station blackout des {7n and, accordingly, does not fall into that class of plants for which station blackout is potentially risk significant as determined by the preliminary risk-based evaluation.
Further, the results of Task B-57 are expected to confirm the present staff view that a loss of all alternating current power does not have a significant safety impact on plants such as Erie. Accordingly, our conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by this generic task.
B-63 Isolation of Low Pressure Systems Connected to the Reactor Coolant Pressure Boundary The preliminary results of the risk-based evaluation indicate that because the failure of two check valves in series in the decay heat removal system was c dominant contributor to risks associated with the pressurized water reactor analyzed in the Reactor Safety Study, improved procedures for examining interfacing system isolation devices have the potential for reducing the likelihood that such accidents will be dominant in other types of pressurized water reactors.
As indicated in the problem description of this task in NUREG-0471, current reviews of license applications for construction permits and operating licenses are based on guidelines set forth in the Standard Review Plan, specifically Standard Review Plan 5.4.7.
The staff positions in Standard Review Plan 5.4.7 acceptably resolve this concern. However, since these C-32
~
guidelines were not available during the reviews of plants which are currently operating, Task B-63 involves the review of representative operating plants to assess the isoletion capabilities of low pressure systems.
This task has, in fact, been completed with a conclusion that there is adequate high pressure-low pressure isolation protection in operating reactors.
Since the Erie Nuclear Plant was reviewed using current criteria (this subject is discussed in Section 5.4.3 of Appendix A of the Safety Evaluation Report and in Section 5.4.3 of the Safety Evaluation Report), this task is not applicable to the Erie Nuclear Plant. Accordingly, our conclusions in the Safety Evaluation Report regarding the issuance of construction permits are unaffected by the results of this generic task.
B-64 Decommissioning of Reactors As noted in h0 REG-0471, this task deals with the development of methods and procedures for accomplishing reactor decommissioning.
The task is not expected to result in changes to nuclear plant design features.
The preliminary results of the risk-based evaluation indicate that this task was designated as potentially risk significant because of the possi-bility of significant exposures to a large number of plant personnel if decommissioning activities are not properly carried out. We agree that such exposures are an important consideration of any decommissioning Accordingly, the studies and resultant safety acceptance criteria program.
and guidelines for decommissioning operations developed under this task specifically include consideration of occupational radiation safety.
These acceptance criteria and guidelines will be available long before we can envision the neea for any major decommissioning operations at the Erie Nuclear Plant Thus, this issue may be safely left for resolution after the plant has gone into operation and, accordingly, our previous cen-clusions in the Safety Evaluation Report regarding the issuance of con-struction permits are unaffected by this ongoing generic task.
C-33
APPENDIX D CHANGES AND ERRATA TO THE SAFETY EVALUATION REPORT ISSUED JULY 1978 Changes and errata effected by this appendix to the report do not alter the staff conclusions presented in the earlier Safety Evaluation Report.
Table of Contents Page vi, line 10 Change " Spent Fuel Cooling" to " Spent Fuel Pool Cooling" Section 1.0 Page 1-6, line 12 Change " design procurement" to " design, procurement" Page 1-10, line 28 Change "(Section 6.2-5," to "(Section 6.2.5,"
Page 1-14, line 4 Change "the did not specify" to "they did not specify" Section 2.0 Page 2-10, line 23-24 Change the second sentence of this paragraph to read
" United States Route 6 and Ohio State Route 2 are approximately one and two miles north of the site, respectively.
Ohio State Route 113 is c.8 miles south of the site."
Page 2-21, line 22 Change " normal water supply" to " normal circulating water supply" Page 2-22, line 5 Change "from Lake Erie" to "from Lake Erie water."
Page 2-26, line 6 Change "kilometere" to " kilometer" D-1
Section 3.0 P_ age 3-2, line 28 Change "the downstream site" to "the downstream side" Page 3-19, line 32 Change " equip-ment to " equipment" Page 3-23, line 17 Change "the applicant will purchase Class IE motors type-tested" to " testing of Class IE motors will be" Page 3-23, line 18 Change "334-74" to "334-75" Page 3-23, line 20 Change "and Class IE valve operators type-tested" to " testing of Class IE valve operatirs will be" Page 3-23, line 23 Change "as augmented by" to "and" Page 3-23, line 26 Change " wires and cables type-tested" to " testing of wires and cables will be" Section 5.0 Page 5-4, line 27 Delete "which is in accordance with Regulatory Guide 1.45," Reactor Coolant Pressure Boundary Leakage Detection Systems," and is acceptable" Section 6.0 Page 6-3, line 2 Change "te containment" to "the containment" Page 6-4, line 2 Change "claculated" to " calculated" D-2
Page 6-5, line 31 Delete " shutdown cooling heat exchanger" Page 6-8, line 5 Replace "Since the return lines have no post-accident safety function, they will remain isolated following a loss-of-coolant accident." with "Following a loss-of-coolant accident, the inner valves will be opened during the long-term cooling period to allow flow through the dump-to-sump lines. "
Page 6-11, line 28 Change "12" to "two" Page 6-18, line 14 Change "0.025" to "0.625" Pcge 6-18, line 19 Change "N5.12" to N512" Page 6-19, line 3 Change "N.512" to "N512" Page 6-20, line 3 Change "controi room habitability system" to " control room charcoal cleaning system" Page 20, line 13 Change "high efficiency particulate air filter" to "two high efficiency particulate air filters" Page 21, line 10 Change "high efficiency particulate air filter" to "two high efficiency particulate air filters" Page 6-21, line 11 Change " absorber" to "adsorber" D-3
Section 7.0 Page 7-3, line 18 Change "protions" to " portions" Page 7-2, line 25 Delete " individually" Page 7-3, line 37 Change " refueling water storage tank" to " borated water storage tank" Section 8.0 Page 8-5, line 1 Change "our" to "out"
_Page 8-5, line 17 Change "followed" to "following" Page 8-7, line 10 Change "be" to "have" Page 8-9, line 13 Delete "by type test" Section 9.0 Page 9-1, line 18 Change " reactor drain tank" to " reactor coolant drain tank" Page 9-3,1:ne 14 Change " Spent Fuel Cooling" to " Spent Fuel Pool Cooling" Page 9-3, line 15 Change " spent fuel cooling" to " spent fuel pool cooling" D-4
Page 9-12, line 16 Change " Redundant radiation detectors" to " Redundant, non-safety grade radiation detectors" Page 9-12, lines 37 and 38 Change the first sentence of this paragraph to read:
"The spent fuel pool area and the general areas of the fuel handling area will be ventilated by the fuel handling open area ventilation system. Air from these areas will be exhausted through the fuel handling area exhaust system."
Page 9-14, line 37 Delete " Appendix A to" Page 9-16, line 15 Change " oil transfer pumps" to " oil pumps" Page 18, line 20 Change "weater" to " weather" Section 10.0 Page 10-3, line 18 Change " condensate feedwater" to " condensate-feedwater" Page 10-3, line 19 Change " recirculation" to " circulation" D-5
0"#
U.S. NUCLEAR REGUL AloRY CoMMISSloN (7 77)
NUREG-Oh21 BIBLIOGRAPHIC DATA SHEET Supplement No.1
- 4. TITLE AN D SUBTITLE (Add Volume No., if approprosto)
- 2. (Leave blank)
Supplement No. 1 to the Safety Evaluation Report for Erie Nuclear Plant, Units 1 and 2 3 RE CIPIE N T'S ACCESSloN No,
- 7. AUTHoRIS)
- 5. DATE REPORT COMPLE TED l YEAR MnNTH January 1979
- 9. PE RF oRMING ORGANIZATION N AME AND MAILING ADDRESS //oclude Zsp Code /
DATE REPORT ISSUED l YEAR MONTH U.S. Nuclear Regulatory Commission January 1979 Office of Nuclear Reactor Regulation e
(t,,,, b/,n )
Washington, D.C.
20555
- 8. (Leave blank)
- 12. SPONSORING oRGANIZ ATioN N AME AND M AILING ADDRESS (/nclude Zip Codel p
- 11. CONT R ACT No.
13 TYPE OF REPORT FE RIOD COVE RE D (/nclureve d,tes)
Safety Evaluation Report
- 15. SUPPLEMENTARY NOTES
- 14. (Le.cc b/atk)
Docket Nos. STN 50-580 and STN 50-581
- 16. ABSTR ACT (200 words or less)
Supplement No. 1 to the Safety Evaluation Report regarding the application by Ohio Edison Company, et al for pemits to construct the Erie Nuclear Plant, Units 1 and 2 (Docket Nos. STN 50-580 and STN 50-581), located in Erie County, Ohio, has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission.
The purpose of this Supplement is to update the Safety Evaluation Report by providing the staff's evaluation of additional information sutmitted by the applicants since the issuance of the Safety Evaluation Report and to address concerns expressed by the Advisory Committee on Reactor Safe--
guards.
The staff has concluded that the plant can be constructed by the Ohio Edison Company without endangering the health and safety of the public.
- 17. KEY WoRDS AND DOCUMENT ANALYSIS 17a. DESCRIPToRS 17b. IDENTIFIE RS!oPEN ENDE D TERMS
- 18. AV AILABILITY STATEMENT
- 19. SECURITY CLASS (This report) 21 No. OF PAGES
- 20. SECURITY CLASS (This page)
- 22. P RICE NRC FORM 335 (7 77)
UNITE D ST ATE S NUCLfAR RfGULATORY COMMISSION W ASHING TON, D. C. 20SSS POST AGE AND F E E S P AID u a mucLa A n nrGuLAf ony OF F ICI AL BUSI N E SS CO""'U" Pf N ALT Y F OR PHIV ATE USE. $M)O U S MAJt_
k d
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