ML20126D507
| ML20126D507 | |
| Person / Time | |
|---|---|
| Site: | 05000363 |
| Issue date: | 12/14/1979 |
| From: | George Minor AFFILIATION NOT ASSIGNED |
| To: | |
| Shared Package | |
| ML20126D493 | List: |
| References | |
| NUDOCS 8005020149 | |
| Download: ML20126D507 (39) | |
Text
. . - -.
80050:3O}
L AFFIDAVIT of GREGORY C. MINOR In The Matter Of FORKED RIVER NUCLEAR PLANT Regarding i
NEED FOR REASSESSMENT OF PROPOSED PLANT DESIGN FOR CONCURRENCE WITH TECHNICAL REQUIREMENTS FOR CP
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I. -
INTRODUCTION l l
My name is Gregory C. Minor and I have 19 years of i i
experience in the design, development, research, field work, and i management of nuclear reactor sys tems. I have worked for 16 g 1
years with the General Electric Company and 3 years as a technical !
cons ultan t . Since 1976, I have been a partner in MHB Technical 1 Associates working on a variety of reactor studies and nuclear i
safety issues. I am presently a consultant on several plant cases 'concerning the adequacy of current designs to meet exis ting regulations. Also, I am currently participating on a Peer Review Group of the NRC/TMI Special Inquiry Group , under the direction of Mitchell Rogovin. My complete experience record is attached (Attachmenc 1). _.,
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II. PURPOSE AND STATEMENT The purpose of this affidavit is to review the licensing and regulatory requirements and changes in recent years to assess the extent to which the Forked River Nuclear Plant may require a [
t reas sessment of the design's ability to meet the current regula- [
I" tions.
The delays in this plant have been extraordinary. In
, essence, the Forked River Nuclear Plant is an example of 1960 's technology , designed in the 19 70 's , to be built in the 1980 's , : :
and coming on line in the 1990 's . During these several decades ,
many changes have occurred (and undoubtedly, will continue to occur) -
in the regulations which protect the health and safety of the public, i It is my conclusion, given the number and significance of I, i
technical and environmental issues which have required additions [
to the regulatory requirements in recent years , plus the uncer- f tainty of the present s tatus of the Forke'd River Plant design, -
that a complete re-review of the Forked River Project is necessary Il prior to recommencement of construction in order to assess it's I i
ability to meet the regulations in effect at that time . ,
l i
III, DISCUSSION hl r:
A. S tandard Review Plan The Applicant for the Forked River Plant applied for iI a construction permit in 1970 and was issued a C.P. in July,1973. [l l-Thus , the Preliminary Safety Analysis Report' (PS AR) was not i 2
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prepared with the benefit of the Standard Format and Content document 1I firs t issued by the AEC in 1972. This makes it a more time-consuming process to evaluate the design in terms of current regulations. As time goes on and more regulations are revised and added, the gap between the Standard Review Plan E! ,
and Forked River PSAR will widen, making it more difficult to j assess the adequacy of the plant design. j B. Regulatorv Guides l The guides for minimum acceptable design are given in the NRC Regulatory Guides which of ten refer to industry standards or specific practices which are considered acceptable in complying with the requirements delineated in the Code of Federal Regulations. These Regulatory Guides are being revised ,
and added as needed to keep up with the current regulatory status.
It is reported that the Applicant at Forked River bas:
". . . . . upgraded the design as far is practicable .
to facilitate the process for obtaining an operating license. This includes reviews of the Regulatorv Guides through Regulatory Guide 1.96 and of che Standard Review Plan . . . . . . . .
[they also] have updated the Forked River design in some areas." 3/ (emphasis added)
~
1/ Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants , LRR Edition, Feb.1972, (Rev 1, Oct.
19 72, Rev. 2, 1975, Rev. 3,1978), now designated as Regulatory ,
Guide 1.70.
~
2/ Standard Review Plan, NUREG-75/087, Sep. 1975, describes the safety review conducted by NRC and is consis tent with recent revisions of Regulatory Guide 1. 70. -
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3/ Memo, R. A. Senedict to the Docket File 50-363, Jul. 24, 1979.
l This memo documents a meeting between NRC and the Applicant held on Jun. 20, 1979.
None of these stitements provide a clear description of the s tatus of the design and how well it has kept up with changing requirements over the 6 years since a C.P. was issued.
1
'l The Regulatory Guides alone have changed appreciably since 1973 and clearly go far beyond Regulatory Guide 1.96.
A cursory review of a recent listing shows that as of September 1979, there were an additional 48 issued Regulatory Guides (Division 1 Reg. Guides which apply to power reactors), 10
' Regulatory Guides in draft stage, 21 under development, 6 pro-posed for revision, and 23 being revised. bl Of these, many may have important significance to the design of basic s tructures, components, and systems of the reactor. These are a few examples:
Reg. Guide 1.97: POST ACCIDENT MONITORING Significant changes are being made. in the requirements of ins truments , control rooms, information and practices to better monitor l and understand an accident in pro'gress. This -
t ;
is largely the result of the experience at TMI-2.
- Reg. Guide 1.120: FIRE PROTECTION f I
Major changes in cabl.e spreading room design, cable construction, cable separation, and fire protection have been necessitated by the dis-covery of design in' adequacies at the Browns ~ '
Ferry reactor in the March 1975 fire. -
l
)
4/ U.S. NRC Regulatory Guide Series , Division 1, Powe*
Reactors , Table of Contents , Sep .1979. (Attachment 2) f
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[
l . - .- . . - . - .
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l Reg. Guide 1.122: SEISMIC DESIGN The seismic design practices have evolved as a result of i= proved understanding of soil-structure interaction and building response spectra pertinent to equipment design. Recently, !
, the Salem-1 reactor in New Jersey has undergone substantial changes to the seismic design of pipe supports in an effort to ' upgrade the design.
This has resulted in extensive outages for the plant.
Other Regulatory Guides have undergone significant revision 4
in recent years which could have substantial impact on the t
design of a plant being reviwed for a C.P. at this. time.
Examples are as follows:
Reg. Guide 1.7: CONTROL OF COMBUSTIBLE GASES FOLLOWING A LOCA ;
1 This was revised in 1978 and additional require- l ments are being considered as a result of the TMI-2 accident experience.
Reg. Guide 1.29: SEISMIC DESIGN CLASSIFICATION Three revisions since the Forked River C.P. was j issued.
Reg. Guide 1.75: PHYSICAL INDEPENDENCE OF ELECTRICAL SYSTEMS ;
Issued and revised twice since the C.P. datie for Forked River, this subject is being further evalu-ated as a result of the TMI-2 experience. l ASME SECTION III - DESIGN AND
~
Reg. Guide 1.84:
- FABRICATION
! This Reg. Guide has been revised 15 times since
- the C.P. date. However, the vessel for Forked River was designed to an older standard - ASME Section VIII- which will make comparison even more difficult. 5_/ .
5_/ Ibid 3. ,
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C. Systematic Evaluation Program (SRP)
The NRC has an entire program which is focused on.
the problem of how to evaluate the safety of older plants. It .
is called the Systematic Evaluation Program (SEP) mad is set up to review eleven older BWR and PWR nuclear plants to determine the degree to which they comply with the current NRC regulatory practices for licensing new plants (Regulations, Regulatory Guides ,
S tandard Review Plans, etc.) . SEP is intended to provide basic information regarding the areas of strength and weakness of ,
i I
older plant designs compared to current regulatory -practices.
Eventually, the SEP methodology is intended to provide the neces- E sary information to enable the NRC to decide if any deficiency (or the sum of several) has a sufficient impact on safety to require backficting of older operating plants. It may also be useful in re-evaluating plant designs where the C.P. period has ,
f been exceeded. -
f D. Unresolved Safety Issues In raviewing the application for a construction .'
permit for the River Bend reactor, The ASLAB held that a licensing {
bear.d, in order to discharge its statutory function, mus t determine whether the NRC Staff's review ". . . . . . satis factorily' has come to grips with any unresolved generic safety problems which might have an i= pact upon the operation of the nuclear facility under considera-tion." 1/ The ASLA3 detailed the level of r& view necessary in l
1/ ALA3-444, Nov. 23,19 77 " River Send" at page 27-28 ,
4 j
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order for a licensing board to properly discharge its functions:
. . . . .each SER should contain a summary descrip- ,
tion of those generic problems under continuing !
study which have both relevance to f acilities -
of the type under review and potentially signi-ficant public safety implications.
This summary description .should include information of the kind now contained in mos t .
Task Action Plans. More specifically, there should be an indication of the inves tigative 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 pe'nding the comple-tion of the investigation, and what alte rna-tive courses of action might be available should the program not produce the envisaged result.
In short, the board (and 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. ques tion and the eventual operation of the reactor under scrutiny. Once again this assessment might well have a dire,ct bearing upon the ability of the licensing board to I
make the safety findings required of it on the -
cons truction pe rmit level even though the i generic answer to the question remains in the -
offing." 2/
The ASLAS apparently intended that plants should not be licensed without Staff consideration of generic unresolva.d safety problems at the construction permit stage. It is difficult to see how a licensing board could make reasonable findings of safety absent such an analysis .
1/ Ibid 6 at p. 27-23 (foo: notes emitted) 7
The date of the Forked River C.P. obviously precedes the River Bend decision by several years. However, because plant construction is only 5 - 67. complete and the plan is to defer con-s truction for several years 8/ or even consider selling , ownership -
to ano the r p arty , 1/ their is jus tification for the NRC to re-review I b
the application for the impact of unresolved safety issues before i recommencing cons truction.
There is precedent for the suggested "re-review." A ,
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- recent example of a situation where .the question of backfitting ]
was considered after a construction permit was issued (before the [)
r -
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0.L. review) was the Midland Plant. Because of the extensive pl
=1 delays encountered by the Midland Plant Applicant (Consumers -~ l Power Company) , the Commission conducted a special review of the i applicability of a large number of Regulatory Guides and Staff g E
positions that had evolved since issuance of the Midland con-
[t-s t ruc tion p e rmi t . In the case of Forked River, the Midland action, E
. L plus the River Bend decision, would indicate that the design should [
?
be reviewed against such safety issues as the following:
{
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- 1. The 19 77 ACRS lis t of 28 unresolved issues . [
E' Taole II-l shows a recent lis ting, some dating back as f ar as 19 72 (e .g. , ATWS) .
- 2. The 27 safety issues identified by the NRC Staff.
8/ Letter, Ivan Finfrock (VP of JCP&L) to Robert Baer , U.S . NRC, Jul. 30, 19 79, subj ect: Forked River C.P. CPPR-96, page 2 j
e says cons truction probably will not occur for at least two years.
[r 9/ Tbid 8 at page ?. E l
2
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The NRC in 1974 tabulated the unresolved safety issues into an internal Technical Safety Activities Report (TS AR) . When a
released in 1975, the TSAR listed 223 , 3 items of concern to the NRC of which 173 -
E
- were categorized as having an important .
impact on the licensing review process. i< 4 The NRC apparently discontinued the TSAR l.. '
but in the fall of 1976, in response to i a memorandum by Rusche of the NRC, one ,
or more members of the NRC Staff identi- .
fied 27 technical issues as probless whose priority, progress , or resolution. was , in their opinion unsatisfactory. These'are ,
shown in Table III-2.
- 3. Safety issues described in NUREG-041010/
~
In April, 19 7 7 the NRC implemented a pro- g gram for planning the timely resolution -
of generic technical issues. Initially, each of the four NRC divisions reporting to the Office of Nuclear Reactor Regulation described and proposed those generic iteam it considered to warrant the highes t pri-ority. Proposals were received for 355 11/
tasks of which, following consolidation ,
and elimination, 133 tasks were eventually i selected for review. A s e t o f unifo rm [ '
criteria was applied to the generic tech-nical activities to establish their .,
priority for resolution. The 41 casks "
6 shown in Table III-3 were classified as Category A, warranting priority attention. [
The remaining issues were divided into 72 i -
Category B tasks , 17 Category C tasks , and !
3 Category D casks as shown in Table III-4. p
- 4. Safety issues resulting from the TMI-2 Accicent.
Several reports on the TMI-2 accident have made subs tantive recommendations for improve- .
ment of reactor safety; some of which would i
10/ NUREC-0410, NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants , Jan, 1978, U.S. NRC, I' l Wasning:en, DC, pages D-1 to D-7. (Note: NUREG-03 71, en:itled Approved Task Action Plans for Category A Generic h.
r Ac:i vi tie s , Volume 1, Revision 1, is contained as Appendix $i
F o f N U REG-0410. )
11/ Ibid 10, p . 7. l i
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,_,,i __q . _ , , ,9 _ , _ _ _. p%_ .,
' ACRS Ct;rER2C VSSUES RI90L1/r!CN PE53RNC PMORZTT TOR RETZVANT 70 RZ50LLT8CM ACRs CDrERIC I*rM PV!t B*.rP. ACRS NRC CRCUP TTi(Resolucion Pending Since December 18, 1972)
- 1. Turbine Missiles 1 I A A-37,A-32
- 2. Containment Sp~ rays x B C 10
- 3. Pressure Vessel Tailure By Ther=al Shock X A A-11
. Instruments to Detect (Severe) Tual C -
Tailure Z I SA. Excessive Vibration X X I I
- 33. Loose Parts Monitoring X X 5 5-60 I.
- 6. Non-Randos Multiple Tailures I I A C- 13 6A. Reactor Scram Systems Z X A A-9 A-35,3-36,
- 68. Alternacine Current Sources Onsica'& A B-57 offsite I I 6C. Direct Currenc Systems I I A A-30
- 7. Behavior of Reactor Tuels Under Abnormal I Condicions. I I A 5-22 y
- 9. Seismic Scram I I C D- 1
- 10. ECCS Capability for yucure Plancs I I A D-2 CRour II A: (Resolucion Pending since Tebruary 13, 1974) r
- 1. Ice Condenser Containments I B 5-$4
- 3. Steam Cenerator Tube Leakage I A A-3.A-4, A-5 ,
- 4. ACRS/NRC Periodic 10-year Reviev I Z C Policy CRCUP II la (Resolution Pending Since March 12, 1975) ,
- 1. Computer Reactor Protection Systes I B A-19 [.
- 2. Qualificacica of New Tuel Geomet y I I C 3 22
.. 3. SWR Mark III Containments I 8 A-39.5 10 4 Stress Corrosion Cracking in IVR Piping I B Policy ;
CRotfr II C: (Resolution Pending SLnce April 16, 1976) .
- 1. Locking out of ECCS Power Operated valves I I 5 5-8
- 2. Design Tescures to Control Sabotage % I A A-29 3A. Decontamination Z I 3 A-15
- 35. Decocumi s sioning I I E 3-64 i 4 Vessel Support Structures Z B A-2
{
- 5. Water Marinar I I A A1
- 6. Maintenance and Inspection 1 1* B 5-34 !,
- 7. SVR Mark I containeenes 1 A A-6.A-7, !
A-39 CROUP II Di (Resolution Pending since February 24, 1977).
- 1. Interfaces X X- A Policy.
A-17
- 2. Capability of Hermetic $eals % Z C C-1 7toVP !! t (Resolution Pending Since ?:ovember 15, 1977) ,
- 1. Se t'..s tr ac t are int e rs s tion - X -
1 C A 40.A-41
-la- r
- *'
- ruaunaAs sus
' ' REIIVANT TO PISCLUTICN PVR B'fR (NRC)
ISSUE
- 1. Treac=ent of Non-Safety Grade Equipment in Evaluations of Postulated Steam Line Break I A-22 Accidents
- 2. Lack of Independence of Interlocks on ECCS X Policy Valves I
- 3. Acceptability of Sving Bus Design of BWR-4 % Implamencation {
Plancs (a) l E
4 Loss of Off-Site Power Subsequent to Manual Safety Injection Reset Tollowing a LOCA I Implementation Analysis of Posculated Raactor Coolant Pump Imple=eneation ,
- 5. (2 loop pits).
Rocer Seisure Incidents I X [
E
- 6. Protection Against Single Tailures in I X 3-3 Raaccivi:y concrol
- 7. Passive Tailures Following a Loss of Coolant
. Accident I I 3-58.C-7 Probabilistic Assessment of Reliabilley (3reeder only) --
- 8. y X A-35,3-70 [
- 9. Frequency Decay E
- 10. Crid Stability I I A-35
- 11. Interpretation of GDC-19 Control Room I X --
X X 3-53
- 12. Load Break Switch /Cenerator Breaker
- 13. Instrument Trip Seepoines and Standard [
I X Implementation ,p Specificacion r 14 Computer Procaction System X A-19 E i
- 15. Overpressuri:ation I A-26 .
- 16. Auce=acic Rasetting of Reactor Trip System (RTS) Trip Biscable Relays ,X X --
f X X 3-58 E
- 17. Passive Mechanical valve Failures
- 5
- 18. Electrical C '>1e Penetracions of Rasecor I X B-9 Con tainment
~~
1a. Analys'is of the Interaction of Structures I X A-40,A-41 and the Supporting Soil f'
- 20. Comple'ceness of the Review of Plants Policy Referencing RES AR-3 %
- 21. Instruments for Monitoring 3o' ch Radiacion and Process Variables Durug Accidents % X A-34
- 22. Safety Implications of Control System X A-17 t
Tailures and Plant Dynamics %
l.
- 23. Effects of Steam Ceneracer Tube Degradation [
en che Consequences of a S:eam Line Break p' Accident % A-3.A*4.A-5 g f
24 *= proving the Availability of Offsi:e Power X X A-35 [
- 25. Environmental Qualificacien and/or Re-Quali- f fication of Safety-Related Iquipeene for a I A*21 S:ea= Line 3reak Accident ,
- 25. !=provement of 5'a'R Shu:down Reactivity X --
Perfor=ance (PRT) ,
- 27. Electrcmagnetic Pulse !!!ee:s of a High Altitude Explosion cf a l!ue' ear Weapon on i, l
Safety-Related Squipeent of Nuc*. ear Power X
-- 5 Plants I
,11 l
l TABLE III-3 ;
CATEGORY A TECHNICAL ACTIVITIES _ ,
RZLtVANT TO ,
TASK NO, TI?J PVR SVR .
X X s A-1 Vater Man =ner A-2 Asymetric Blowdown Loads on the fi Reactor Vessel Z I i
A-3 Westinghouse Steam Generator Tube Integrity Z p A-A Combustion Engineering Steam ;
Generator Tube Integrity I A-3 Babcoc.k & Vilcox Steam Generator X Tube Integrity A-6 Mark I shor1: Term Program X
. A-7 Mark I L ong Tern Frogram X A-8 Mark II Prograa I X A-9 AIVS X X ' f.
A-10 BVR Nozzle Cracking I [J Rasetor Vessel Materials Toughness X X A-11
- A-12 Tracture Toughness of Steam Gener-ator and Rasctor Coolant Fmp Supports I Z -
A-13 Snubbe rs X X A-16 Flav Detection I I A-15 Decontamination 1 1 .
A-16 Steam Effects on BVR Core Spray I -
Dis tribution A-17 Systems Interaction in Nuclear Power Plants X X A-18 Fipe Rupture Design criteria I I A-19 Digital Computer Protection Plants with digital computers.
Sys tees A-20, -Impacts of coal Tuel Cycle Environmental i A-21 Main Steam Line Break Inside [t' Containment I A-22 FVR Main Steam Line Break - Core and f Primary Coolant Boundary Response !!
(H5L3 Outside Containment) X . . .
A-23 Containment Leak Testing I I -
~
A-24 Qualification -of Class IE Safety-Related Equipment 1 I -
A="5 Nonsafety Leeds on Class IE Fower ,
sources 1 I A-26 Raactor Vessel Pressure Transient ~
- Protection (Overpressure) X E A-27 Reload Application Guide 1 1 '
- A-28 Increase in Spent Tuel Storage : '
Capacity
- X X A-29 Design Te atures to Control Sabotage Z I A-30 Adequacy of Safety-Related DC Power Supplies X I . i A-31 RJtR Shutdown Requirements X X
~
A-32 Evaluation of Overall Effects of .
X Missiles Z ,
A-33 NEFA Reviews of Accident Risks I,nvironees tal A-34 Instruments for Monitoring Radiation and Process variables During Accidents 1 I i A 35 Adequacy of Offisite Power Systems x x A-36 A-37 Control of Heavy Leads Sear Spent Twel ZI Turbine Missiles X X
["
A-38 Tornado Missiles K I ;
A-39 Determination of Safety Relief Yalve [
(5RV) Fool Dynamic I y A-40 Seismic Design Criteria Short Te7 Z Frogras A-41
- Seisate Design criteria - Long Teg g Progras l .
I i
12- [ ' ,
i k a l
l
- -- - - , , - m ---,c-- + ---w s--w-n. .-n.--+ w- - *w- - - -
- - ew a -- m ,
a ru2 La a s s - -v u e
- CA7!CORY F r!CE!!CAI ACTIVIT!!5 4
F.ELIVA:7 70:
BV2 PUR TA5% Mo.. ; 715:
Environmen:a1 Technical Specifications I I 3-1 I I 3-2 Torecasting Elec:Ts ci:y Se=and .
I I j 3- 3 Event Categorization :
I I !
3-4 ICCS Reliability 3-3 Due:ility of Two-Way Slabs and Shells and luckling i I I ;
3enavior of Sceal Containmenca I I 3-6 Leads, Loa'd Co=sinations. 53ress t'-4:s I
3-7 secondary Accident consequence Modeling I I 3- 8 Locking Out of ECCS Power Operated valves I I 3-9 Ilectrical Cable Penetrations of Containmene I
. 3-10 3ehavior of BWR Mark III Containment S ub comp ar==en t Scandard Problems ' I I 3-11 Concainment Cooling Requirements (Non-LCCA) I I 3-12 I I 3-13 Marviken Test Data Ivaluacien I
3 la Study of Hydrogen Mixing Capability is Containment Post-LOCA I CONTI)2T Computer Code Maintenance I I 3 13 3 16 Procaction Against Fosculated Finping Tailures La I I T1uid systems outside Con:ainment Cri:eria for Safery-Related Operator Actions I I 3-17 Vortex Suppression Requirements for Concainment Sumis I 3-18 Ther=al Hydraulic Sca'oility I I j 3-19 I I !
3-20 Scandard Probles Analysis
- I Core Physics
- I 3-21 .
I I 3-22 LER Tuel :
LMT3R Tuel LM73R.1 3-23 3-24 Seismic Qualificatien of Electrical and Mechanical I I ,;
,. Iquipment Piping Senchmark Probless I I 'h 3-25 Scrue: ural Integrity of Concainment Fenetraciens I I 3-26 Implementacion and Use of Subseccion N7 I I 3-27 Ra dionuclide / Sedimen t Transport Program I I 3-28 Effectiveness of Ulcicate Hear Sinks I 1 3-29 3-30 Design Basis T1oods and ?;obabill:y I I I I 3-31 Da:s Tv4ure Model 3-32 Ice ..fects on :stacy-Related '='ater Supplies Reectors in Northern United 5:a:es 3-23 Dese Assessment Methodology I I 1 36 cccusational Radiation Exsosure Reduction I I 3-35 Cenfir acien of Appendix ! Models for " Calculation of Re* eases of Radioac:ive Materials in Caseous and Liquid Iffluents fres Lignt Vater Cooled Power Reactors" I I .
1 36 Oevelos Desip. Testing and Maintenance Criceria for A:sosphere C.eanup Svste: Air Tiltracien and Adaeryclon Unit s f or Engineereo Saf ety Teature Systevu and f or j Normal Ventilation Systems I I i 3*;* Cheetcal Discharges :o Receiving *.*4cers I I l
3 *e Reconnais sance *.evel inves:154:icas I : -
i
.. l \
j I
TAB 12 III-4b 1 39 Transcissico Linas I I 5 40 Iffects of Power Plant Intrainment on Plankton X I 3 41 !=pae:s on Tisheries X *I 3-4 Socioecono=ic Invironeenea1 != pacts value of Aerial Photographs for Site Ivaluation Z I i 8-43 !
I I 3-44 Torecas:s of Generattng Costs of Coal and Nuclear Plants {
X I 5-45 Need f or Power - Inergy Corservation , (
I I 3 46 Costs of Alternatives in Inviron en:21 Design 5 47 Inse:vice Inspection of Supports class 1. 2. 3 and MC % I Cooonents 3-48 Sbt Control Rod Drive Mechanical Tailures 3-49 Inservice Inspection Criteria and Corrosion Prevention % I
(
Criteria f or Containments (
I I B-50 Pos: operating 3 asis Iarthquake Inspection l B-51 Assessment of Inelastic Analysis Technioues for X I i Iquipment and Coeponents
% 2 B-52 Tuel Assembly Seismic and LOCA Resoonses I 2 Lead Break switch 8-53 y
. I i.
B 54 Ice condenser containments 3
3-55 I::rproved Reliability of Target-Rock Saf ery-Reifef Valves I 1 3-56 Diesel Reliabili:7 Z I 3-57 S:ation Blackout f X X !
B-58 Passive Mechanical Tailures N-1 Loop Operation in 3'-Is and P7Rs X I, 5-59 I X 3-60 Loose Par:s Monitoring Systems ,
I 3-61 Analytically Derived Allowable ECC3 Iquipment Cutage Periods I 3 62 Re-Ixa=ination of Technical !ases f or Is:ablishing 5ts. J I *
,. L55s. and Reactor Protection System Trip Tunc=1oos .
5-63 Isolation of Low Pressure Systens Cennected to the Raactor 2 I
- Coolan: Pressure 3eundary I X f 5-64 Decocucissioning of Reactors I X 5-65 lodine Spiking Z I 3-66 Control Roos Infiltration Measurements 3-67 Effluent and Process Monitoring Instru=entation x I Pu=ro overspeed During a LOCA I I !
B-68 ECC5 Laakage Ix Containment I % ?
t B-69 S-70 Power Crid Trequency Degradation and Effect on Priman* f, Coo , an t P .:=p s I 5-71 *ncident Response B.*1 Nealth Iffects and Life Shortening from Uranium and Coal Tual cycles I I 5-73 Mona:oring for Iscessive Vibration inside the Reactor
- I Pressure '.'essel .,
i k
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TABLE III-4c CA ICCRT C TICc1 CAL AC TvT :rs I 1 i
RILIYMT TO:
- 3A F~.TR TA 51: ::0 1 - .-' ;
C-1 Assurance of Continuous *.cng-Ter: *ntegrity of Seals I I on Instrumentation and Ilectrical Iquipcent f
]
I C2 Study of Contain=ent Depressurization by Inadvertent Sorsy opera-ion to Deter _-ine Adecuac-r of Containment .I I Exts:.41 Design Pressure I I C-3 Insulation Usage Vithin Centainment
- I 'I C4 Statistical Methods for ICCS Analysis Z Z C5 Decay Heat Update X I !
- C-6 LOCA Heat Sources Z
.C-7 PVR System Piping C-8 Main Stes= Line Leakage Control Systems I C9 RRR Heat Exchanger Tube Tailures I I C 10 Iffective operation of Centainment Sprays in a LOCA I I C 11 Assessment
~
of Tallure and Reliability of Pu=:ss and Yalvo I I )
\
C-12 Primarv system Vibration Assessment I C-13 Non Random Tailures I X C 14 'Scorm Surge Model for Coastal Sites ,
I I C 13 NURIC Report for Liquid Tank Tailure Analysis I I on coasts C 16 Assessment of Agricultural Land in Relation to Power Plant Siting and Cooling Systen Selection X I farm lasd i C 17 Interim Acceptance Criteria for Solidification Agents !
for Radioactive Solid Wastes I 1 . I CATICCRY D ?!ca lCAL ACTIVITTIJ ,
I i
D-1 Advisability of a Seismic Scram X I D-2 Imergency Core Cooling 5ys tes capability for Tucu=e Flancs X I , ; j D3 control Rod Drop Accident X ,
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g be pertinent to a construction pemi review. The earliest report was NUREG-0560 on feedwater transients. 12/ The recommendations are shown in TaEle III-5.
The Lessons Learned Task Force also .
generated a list of 23 recommended areas of improvement described in Table III-6.
In addition, the Kemeny Commission made l'3/
more than 80 findings and 44 recommenda -
tions in the subjects shown in Table III-7.
Several of their findings and recommendations +
pertaining to the NRC licensing process and to technical issues have relevance to Forked River. Perhaps the most directly applicable recommendation is as follows:
- " Licensing procedures should fos ter early and meaningful resolution of safety issues before major financial i commitments in construction can occur." lj/ [. '
(emphasis added) F E. Documentation of Deviations f It has been recommended in the past that the NRC's I- 1 Safety Evaluation Report (SER) should include a summary which documents the pros and cons that have entered into the decision .
.t to exempt a plant from current NRC practices and s'afety require- [l ments. , k The NRC, in 1976, ins tituted a plan to document departures fl from the Standard Review Plan. EI While the NRC later drew back g
12/ NUREG.-0560, Staff Reoort' on the Generic Assessment of Feed-water Transients in Pressurized Water Reactors Designed by B&W Como any , U.S. NRC, May, 1979.
Report of the Presidents Commission on the Accident at Three M/ Mile Island - The Need for Change: The Legacy of TMI,, Oct.
19 79, Washington, DC. f
.t t
M/ Ibid 13, p. 65. r I-15/ NRC office letter No. 9,. Attachment F, Sep. 20, 1976, Mena- [:
~
randa, " Documentation of Deviations From the Standard Review' E
Plan."
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-TA312 III-5 NRR (NUREG-0560) RECOMMENDATIONS FOLLOWING TMI-2
- 1. TMI-2 changes per I&E bulletins.
- 2. Add instruments to detect subcooling of primary loop.
- 3. Study improved automation of some safety functions.
- 6. Improve ability to cope with TW transients.
- 7. Improve auxiliary FW reliability.
- 8. Refine NRC criteria and requirements for process control. .
- 9. Reevaluate control system , role and safety implica-Clons.
Provide a more direct reading of vessel water level.
10.
- 11. Decide proper balance between added automation and ,
improved operator response. p i
- 12. Set criteria for equipment important-to-safety but [ ;
not Class lE. .
t E
- 13. Better indication of POR valve position.
14 Revamp NRC reporting and data-assembly processes. i
- 15. Improve isolation of non-essential lines.
- 16. RHR design should be able to handle highly contami-nated coolant.
- 17. Training operator's actions to focus on core cooling.
- 18. Extend defense-in-depth concept to include operator. je
- 19. Consider augmenting operator with real-time, on-line- i analysis. [
- 20. Consider expanding training beyond single failure [
criteria. - !,
failures.
- 22. Expand simulator training. i!
.. 23. Increase refresh training rate for operators. l l E y v. procedures to include multiple failures 24 :
(e c. station blackout, reactivity anomalies, ATWS). #
- 25. Allow for operator improvisation.
- 26. Improve layout of control rooms. l
- 27. More emphasis on human factors. 1
- 29. Extend size range of small break LOCA analyses. ;
- 30. Extend time of events considered in code analysis. '
- 31. Improve NRC ability to analyze OTSG. ,
I
- 32. Revise standard review plan for transients.
- 33. Improve GDC regarding anticipated transients. p
- 34. Technical specification changes to improve surveillance. F c
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TA3LE III-6 RECOMMENDATIONS FOR SAFETY IMPROVEMEFT (SHORT-TERM)
BY THE LESSON-LEARNED TASK FORCE :
e
- 1. Emergency power supply requir'mentse for the pressurizer heaters, j power-operated relief valves and block valves and pressurizer [
level indicators in PWRs. [
- 3. Direct indication of power-operated relief valve and safety valve position for PWRs and BW3s. -
- 4. Instrumentation for detection of inadequate core cooling in PWRs
, and BWRs.
- 5. Diverse and more selective containment isolation provisions for [
- 6. Dedicated penetrations for external recombiners or post-accident -
purge systems. p
- 7. Inerting BWR containments.
- 8. Capability to install hydrogen recombiner at each light water f nuclear power plant. [
- 9. Integrity of systems outside can'tainment likely to contain radio-active materials (engineered safety systems and auxiliary systems) for PWRs and SWRs.
- 10. Design review of plant shielding of spaces for post-accident operations. .
g
- 11. Automatic initiation of the auxiliary feedwater system for PW3s. ..
- 13. Improved post-accident sampling capability. . [
- 14. Increased range of radiation monitors. ;
- 15. Improved in-plant iodine instrumentation. .
E Analysis of design and off-normal transients and accidents.
16.
- 17. Sh'if t supervisor's responsibilities. .
- 18. Shift technical advisor.
- 19. Shift and relief turnover procedures. [f
- 20. Control room access. ;-
E
- 21. Onsite technical support center. .
ll
- 22. Onsite operational support center.
{
- 23. Revised limiting conditiens for operation of nuclear plants j
based upon saiety system availability. !
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TABLE III-7 KEMENY COMMISSION RECOMMENDATIONS SUBJECT AREAS NO. OF RECOMMENDATIONS Nuclear Regulatory Commission 11 Utility and its .iuppliers 6 Training of operating personnel 4 Technical assessment 7
-Worker and public health & safety 5 ,
Emergency planning and response 6 -
Public's right to information 5 t
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~~~
from this position and failed to implement th es e pro-cedures, these directive by NRC management underscored the need 'I to have on record in the SER the factors entering the Staff's ,
analysis of facility safety. Absent such a lis tin 5 or "documen-tation of all deviations," it is nearly impossible for an inde-
.. {
pendent party to conduct an informed analysis and inquiry into any safety impacts these deviations may have. A complete listing and evaluation of potentially significant safety items which do not comply with present regulatory requirements is essential for a definitive assessment of plant safety on the Forksd ' River .
Nuclear Plant.
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IV.
SUMMARY
The design of the Forked River Nuclear Plant, being at least 10 years old at this time , and facing several years of delay [
plus many additional years of construction, is being outdistanced by a rapidly changing technk. cal and licensing environment. There-fore, it is important that the design be re-reviewed in such areas as the ability to meet current regulatory guides, the pressure
, vessel code criteria, the accident analyses in the PSAR, the l
resolution of safety issues applicable to the plant and implementa- I tion of the numerous recommendations resulting from TMI. The !
! l result should be a revised SER and a " documentation of deviations." ; j In keeping with the River Bend decision and because of the exten- l sive delays experienced and forecasted in the Forked River construc-
~ ~
tion, this re-review should be completed prior to recommencing construetien.
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.WM d i:
GREGORY C. MINOR -
Subscribed and sworn to before me this /D day of
. peacoceecemoseewe-occsc
/%6s , 1979, 3 orr.cLu, stti l-j CARLO F. cAEN.0 MonRY Pi;8CC CA'.!TCE?!!A g
3
'M
/ C ' gI d / j SANTA cura cou.'try ?
o My commiss6n Expires Se;,t. 18,19301 NOTARY PUBLIC l
I -
- ni inuu nmu . , '
RE S UME -
GREGORY C. MINOR MHB Technical Associates -
172 3 Hamil ton Avenue [
i Suite X San Jose, California 95125 [
(408) 265-2716 [
I-V EXPERIENCE 1976 - Pre s e n t:
Partner - MHB Technical Associates, San Jose, fe Caldei fo rni a .
ral and private l
Engineering and Energy consultant to state , Major activities include studies ,..
o rganizati ons and indi viduals . ;.
o f s a fe ty a n d ri s k involved in energy generation, p rovi ding tech- -
nical consulting to legislative, re g ul a to ry , public and private groups and expert witness in behalf Was co-editor of critique of'a state organizations of the Reactor and .
ll ci ti zens' groups. t Sa fe ty Study .(WASH-1400) for the Union of Concerned Scientis ts l' and co-author of a risk analysis of Swedish re a c to rs fo r the Swedish Energy Commission. Served on the peer Review Group of [
Actively L the NRC/TMI Special Inqui ry Group (Rogovin Committee). i involved in the Nuclear Power Plant Standa rds Commi ttee work fo r the Ins trument Society of America (IS A) . .f
~
1972 - 1976:
Manager - Advanced Control and Instrumenta tion Engineering, h:. ,
Gene ral Elec tric Company, Nucl ear Energy Division, San Jose, t Ca l i fo rnia .
[
Mana ged a 'desi gn and development group of thirty-four engineers and support personnel designing systems for use in the measurement, co n t ro l and ope ra tion of nuclear reactors. Involved coordination wi th other reactor design organiza tions, the Nuclea r Regula tory Commission, and cus tomers, bo th ove rseas. and domes tic.
i Responsibiliti.es included coordina ting and managing the designand new con troli and development of con trol sys tems , s a fe ty sys tems , The posi tion concepts for use on the next generation of reactors. j incl uded responsibili ty fo r s ta nda rds a ppli.ca bl e to control and i ns t rumen ta tion , as well as the desi gn o f short-term solutions to pJ-field problems. The disciplines involved included electrical and '
mechanical engineering, seismi c design and proces s computer con trol p ro g ra mmi n g. '
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1970 - 1972:
Manager - Reactor Control Sys tems Design - General El ec tri c '
Company, Nuclear Energy Division, San Jose, C al i f o rn i a .
Managed a group of seven engineers and two support personnel in the design and preparation of the detailed system drawings and con trol documents relating to safety and emergency systems for nuclear reactors. Responsibili ty requi red coordina tion with 7 j
other design organizations and interaction with the cus temer's '
' engineering personnel, as well a s regul a to ry pe rs onn el .
1963 - 1970:
Design Engi'neer - Gene ral Electric Company, Nuclear Energy Div.ision, San Jose, California.
Responsible for the design of specific Lead controlresponsibility design and instrumen for tation systems for nuclear reactors.
various subsystems of instrumentation used to measure neutron flux in the reactor during startup and intermediate power opera- i tion. Perfo rmed lead sys tem design function in the design of a major system for measuring the power generated in nuclear reactors.
Other responsibilities included on-site checkout and tes ting of a complete reactor control sys tem a t an experimental reactor in the Southwes t. keceived patent for Nuclear Power Moni to ring System.
1960 - 1963: - I Advanced Enginee ring Program - General Electric Company, Assign i
men ts in Wa shi ngton, Ccli fornia , and Ari zona. (
I Ro ta ti ng a s s i gnmen ts in a varie ty of disciplines: l
,, l i
- Enginee r - Rea c to r ma.i n te nance an d ins trumen t design, XE and D reactors, Hanford, Washington , Circuit design !
and equi pment maintenance coordina tion.
- Design En gi nee r - Microwave Depa rtment, Palo Al to, Cali-fornia. Work on design of cavity couplers for TWT's.
Phoenix, Ari zon a . l
- Desi gn Engi nee r - Compu te r Depa rtmen t, J Des i gn o f co re driving ci rcui try. ,
l
- Desi gn Engi nee r - Atomic powe r Equip, men t De p a r tme n t , San Jose, California. Ci rcui t desi gn and analysis. l
- Des i gn Engineer - Space Systems Department, San ta B a r b a r a ,,
Cali f o rn i a . Prepare con trol portion of satellite proposal .
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- Technical Staf f - Technical Military Planning PrepareOpera analysis tion. of (TEMPO), Santa Barbara, California. .
missile exchanges.
During this period, completed three-yea r General Electric program
.of extensive education in advanced engineering principles Also completed of highet courses in ma thema tics , p robabili ty a nd analysis. Management Training Program,[
Kepner-Tregoe, E f fecti ve Pres e nta tion , * !
and various technical seminars. -
I
[
EDUCATION
. Uni ve rsi ty lo f Cali fo rnia a t Be rkel ey , BSSE, 1960.
Advanced Course in Engineering year Curriculum, Gene ra l ;
Electric Company, 1963. ,
S tan ford Uni ve rsi ty , MSEE, 1966. .
h HONORS AND ASSOCIATIONS .
- Tau Beta Pi Engineering Hono ra ry Socie ty Co-holde r of U.S. Pa tent No. 3,565,760, " Nuclear Reactor i
Powe r Moni tori ng Sys tem," February 1971. 1
[
- Member: Ame ri ca n As s oci a ti on for Advance of Science. '
Nucl ea r Power Pl an t S tanda rds Commi ttee , Instru- !
Member:
ment Socie ty of America. [
F
?
PERSONAL DATA !
Born: June 7, 1937 Ma r ri e d , th re e children Residence: San Jose, California
,r---,y,-- , , -e-
~ .. - . . , , , . , , . , . , . ,
PUBLICATIONS AND TESTTMONY
- 1. G.C. Mi no r , S . E . Moo re , " Con trol Rod Signal Mul ti plFe exi nga ,"
b ru ry IEEE Transactions on Nuclear Science, Vol . NS-19, 1972.
- 2. G.C. Mi no r , W . G. Mil am, " An In tegra ted Control Room Sys tem for a Nuclear Power Plant," NED0-10658, presented at Inte rna tional Nuclea r Indus tries Fair and Te chnical Me'e tings October 1972, B asle , Swi tze rl a nd.
- 3. . The above article was also published in the German Technical
- Magazine, NT, March 1973.
- 4. Tes timony o f G. C. Mino r, D.G. B ri denba ugh , and R.B. Hubbard before the Joint Committee on Atomic Energy, Hearings held Fe b rua ry 18, 1976, and published by the Union of Concerned Scien ti s ts , Cambri dge , Mas s a chuse tts.
- 5. Tes timony o f G. C. Mi no r, D.G. B ridenba ugh , and R.B. Hubbard befo re the California State Assembly Commi ttee on Resources ,q Land Use, and Ene rgy , March 8, 1976. ,,
- 6. Tes timony of G.C. Minor and R.B. Hubbard before the Cali and fo rni Sta te Sena te Commi ttee on Public Utili ties , Transi t, Energy, March 23, 1976.
Tes timony o f G. C. Minor regarding the Grafenrheinfeld Nucle 7.
Plan t, March 16-17, 1977, Wurzburg, Germany.
- 8. Testimony of G.C. Minor before the Cluff Lake Board of Inqui!j Regina, Saskatchewan, Canada , Sep tembe r 21, 1977. ,
9'. The Risks of Nuclea r Powe r Reacto rs : A Review of the NRC -
KendaTT, Rea c to r Sa f e ty Study WISH_-1400 ( N U RE G - 7 5 / 014 )_ , H.
- et al, edi ted by G. C. Minor and R.B. H ubba rd for the Union f of Concerned Scientists, August 1977. "i
- 10. Swedish Reactor Safety Study: B a rs e b ~a'c(Published k Risk Assessment, by Swedi MHB Technical As s oci a tes , Ja nu a ry 1978.
Department of Industry as Document SdI 1978:1)
- 11. Tes timony by G.C. Minor before the Wisconsin Public Service Commission, Feb ruary 13, 1978, Loss of Coolant Accidents:
Thei r P robabil i ty an d Cons ec uen ce_.
- 12. Tes timony by G.C. Minor be fore the Cali fornia Legisla ture Assembly Commi ttee on Resources', Land Use and Energy, A8 310:
April 26, 1978, Sacramento, Cali fo rni a .
- 13. Presentation by G.C. Minor before the Fe de ral Minis try fe r Resea rch an d Technology (BMFT), Meetin g on Re a c to r Sa f e ty ,
Research, Man / Machine _ In terf ace in Nuclear Reac to rs , Augus t ji 31 and Sep temcer 1, 1978, Bonn, Germany, 1 1
1
. PUBLICATIONS AND TESTTMONY
- 14. Tes timony by G. C. Minor, D.G. B ridenba ugh , and R.B. Hubbard before the Atomic Safety and Licensing Board, September 25, 1978, In the Matte'r of the Black Fox Nuclear Power Station >
4 Cons truction Pe rmi t He a ri ngs , Tul s a , Okl a homa .
Mino r, D.G. B ri denba ugh, and R.B. Hubbard, Improvino th
- 15. G.C.
Sa f e ty o f LWR Powe r Pl an ts_, prepared for Sandia Labora tories by MHS Technical Associates, September 27, 1979.
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Septsmber 1979 i i
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U.S. NUCL EAR R EGULATORY COMMISSIGN REGULATORY GU!OE SER!ES OlVISION 1. POWER RE ACTORS f TABLE OF CONTENTS I
i i This table of contents Lists every revision of each regulatory guide in Division 1 with the date it was issue. E If the latest version of a guide (as of the date of this table of contents) was issued for early cocument, this is so noted.
Division 1, one of ten broad divisions in which regulatory guides are issued, contains those guides that were developed for power reactors. There may also be some guides issued in other divisions that would be of interest to those whose primary concern is in the area of power reactors. Accordingly, this issue of the table of contents includes, for the first tiene, a listing of regulatory guides issued in the other divisions that the NRC, ,
e staff has identified as possibly of interest to recipients of Division ! guides. This Usting will be updated from
- time to time, and suggestions for additions to it are encouraged. F.
Most regulatory guides contain a section headed " Implementation
- t. hat is Intended to provide information 'to appucants and licensees regarding the NRC staff's plans for using the guide. If a guide does not contain such ,
a section or if detailed information is needed on the staff's plans for using a regulatory guide with respect to a ?
specific permit or license or appUcation therefor, requests for such infomation should be addressed to the appro- ;
priate licensing project manager in the Office of Nuclear Reactor Regulation or Office of Nuclear 7taterial Safety and Safeguards. [c At an appropriate point in the development of a new regufatory guide or a proposed revision to an existing
- j[
guide, the guide and the associated value/ impact statement are f.ssued in draft form to involve the public in the early stages of the development of a regulatory position. These drafts have not received complete staff review I
and do not represent an official NRC staff position. They are temporarily identified by their task number and f issued to the s.une distribution list that is used for published guides in each division. Draft guides issued in h
this division are Listed in this issue of the table of contents.
All regulatory guides, including draf t guides, proposed revisions, and all published revisions, may be examined at the Commission's PubUe Document Room at 1717 H Street NW. , Washington, D.C. Regulatory guides [
E are not cepyrighted and Cemndsslor. approval is not required to reproduce them. Requests for single copies of draf t guides and proposed revisiuns should no made in writing to the (i.S. Nuclear Regulatory Cc.mmission. !
Wasnin gton , D.C. 20555, Attention: Director, Division.of Technical Infomation and Document Control. Copies !
.
- of active guides may be purchased at the current Covernment Printing Office (CPO) price. A subscript. ion serv-ice for future guides in specific divisions is available through the Government Printing Office. Information on i
) i subscription service and current CPO prices may be obtained by writing to the U.S. Nuclear Regulatory Coeunis-sion Washington, D.C. 20555, Attention: Publications Sales Manager.
Regulatory Guides issued Title Ret Year / Month Number Net Positive Suetion Head for Emergency Core Cooling and Containment 70A1 ,
1.1 (
Hest Remova.1 System IN:nps (Safety Guide 1)
-- 70M 2 1.2 Ther,nal Shocit to Reactor Pressure vessels (Safety Guide 2)
--- 70/11 .
Assumptions Used for Evaluating the Potential Radiological Consequences 1.3 1 73/06 E J of a '.oss of. Coolant Accident for Boiling Water Reactors 74/06 E 2 ,
r
--- 70/11 1.4 Aeseptions Used for Ivalusting the Potential Radiologies! Conmeces 1 73/06 of a Loss-of Coolantmecident for Pressurized Water Reactorf 2 74/06 I
i 1 ,
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, p.
- R ev.
Issued Year /t.1onth ,
[
N Number Title .
--- 71/03 1.5 Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors (Safety l Guide 5)
-- 71/03 1.6 Independence Between Redundant Standby (Onsite) Power Sources and Between Thetr Distribution Systems (Safety Guide 6)
Control of Combustible cas Concentratiens in Containment ToDowing ..- 71/03 1.7 1 76/09 a Loss of Coolant Accident 2 78/11
-- 71/03 1.8 Personnel Selection and Training 75/09 1
1-R 17/05
--- 71/03 1.9 Selection, Design, and Qualification of Diesel Cenerstor Units 78/11 Used as Onsite Electric Power Systems at Nuc! car Power Plants 1 (Tor Comment)
Mechanical (Cadweld) Splices in Reinforcing Bars of Category ! Con-
~ 71/03 1.10 1 73/01 crete Structures Instrument Lines Penetrating Primary Reactor Containment (Safety Guide 11) -~ 71/03 1.11 7&O2 Supplement to Safety Guide 11, Backfitting Considerations
-- 71/03 1.12 Instrumentation for l'arthqualtes 1' 74/04
-~ 71/03 1.13 Spent Tuel Storage Facility Design Basis (For Comment) 75/12 1
Reactor Coulant Nmp Flywheel Integrity (For Comment)
--- 71/10 1.14 1 75/C8
-~ 71/10 1.15 Testing of Retnforcing Bars for Category ! Cc.ncrete Structures 1 72/12 Deporting of Operating Infonnation--Appendix A Technical Sp eifications
--- 71/10 1.16 1 73/10 (Tor Comment) 2 74/09 3 75/01 4 75/06 Protection of Nuclear Power Plants Against Industrial Sat,otage -- 71/10 1.17 1 73/06 Structural Acceptance Test for Concrete Primary Resetor Containinents
--- 71/10 1.13 1 72/12 Non' destructive Exa.mination of Primary Contsinment Liner Welds (Safety --- 71/12 1.19 1 72/08 Cuide 19)
Comprehensive Vibration Assessment Program for Reactor Intemals Dudng
-- 71/12 1.20 1 75/06 Preoperational and initial Startup Testtng 2 76/05 Measuring. Evaluating, and Reporting Radioactivity [n Solid Wastes and --- 71/12 1.21 74/06 Releases of Radioactive Materials in Liquid and Cascous Effluents frena 1 Light Water-Cocied Nuclear Power Plants .
Psriodic Testing of Protection System Actuation Functions (54ety Citide 22) -~ 72/02 1.22 Onsite Meteorolugical Programs (Safety Guide 23)
~ 72/02 1.23 Assumptions Used for E ssluating the Pe:ential P.adiological Consequences --- 72/03 1.24 of a Pres: urised Water Rea tor RadioJClive Cas Sturage Tank 7 allure I (Safety Guide 24) i 2
- -- +
Issued Title Rev. Yes#.1onth Number Assumptions Used for Evaluating the Potential Radiological Consequences
--- 72/03 .
1.25 of a Tuel Handling Accident in the Fuel Handling and Storage Facihty for BoCing and Pressurised Water Reactors (Safety Cuide 25)
--- 12/03 1.26 QualJty Group Classifications and Standards for Water , Steam , and 7t/09 Radioactive-Waste-Containing Components of Nuc! car Power Plants 1 2 75/06 ;
(For Comment) 3 76/02 ;
- 72/03 1.27 Ultimate Heat Sink for Nuclear Power Plants (For Cc.mment) 1 74/03 2 76/01
--- 72/06 1.28 Quality Assurance Program Requirements (Design and Construction) "8/03 1
2 79/02
--- 72/06 1.29 Seismic Design Classification 73/08 i ,
1 2 "6/02 !
3 78/09 i l
.(
Quality Assurance Requirements for the Insta11stien, inspection, and Te, sting
-- 72/08 l 1.30 ' '
of Instrumentation and Electric Equipment (Safety Guide 30) ,
--- 72/08 Centrol of Temte Content in Stainless Steel Weld Metal 1.31 1 73/06 ! i 77/05 2
3 78/04 Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants --- 72/04 1.32 76/03 1
2 77/02 Quality Assurance Prngram Requiren.ents (Operation)
--- 72/11
- .33 1 77/02 2 "8/02 f i
Control of E1cetroslag Weld Properties ,--- 72/12 ;
1.34 ,
Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Contain- --- 73/02 (
1.35 1 74/06 4 ment Structures 2 76/01 I
-- 73/02 j
,. 1.36 Ncnmetallie Thermal Insulation for Austenitic Stainless Steel Quality Assurance Requirements for Cleaning of T1uid Systems and
-- 73/03 1.37 Associated Components of Water-Cooled. Nulcear Power Plants
--- 73/03 i 1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, 76/10 1
Storage, and Handling of items for Water-Cooled Nuclear Power Plants 2 77/05 Housekeeping Requiressents for Water Cooled Nuclear Power P! ants
.-~ 73/03 l 1,39 1 76/10 2 77/09 l
Qualification Tests of Continuous Duty Motors installed Inside the Contain-
--- 73/03 1.40 ment of Water Cooled Nuc! car Power Plants Prwperkt.ional Testing of Redundant on Site Elvetric Power Systems to
--- 73/03 l 1.41 .
i Ver"Jy Proper f.ead Group Assignments (Withdrumm Stre 41 FR 11891, 3/22/*6) l l 1.42 ! 1 Control of Stainless Steel Weld Cladding of I.4w Alloy Steel Components
--- 73/05 -
1.43 l,
--- 73/05 - l
! . 54 Control of the Use of Sensitiaed Stabless Steel I '
Reactor Coolant Pressure Boundary 1.cakage Detection Systems
--- 73/05 s 1 45 i ,
3 I
Issued R ev. Ye ar /t,*.on th Title Nurr.ber
--* 73/05 1.4C Protection Against Pipe Whip Inside Containment
-- 70/C5 1.47 Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Sy stems
--- 13/05 1.48 Design Limits and Loading Combinations for Seismic Category I fluid System Components
~- 73/05 1.49 Power Levels of Nuclear Power Plant.s 1 73/12
--- 73/05 1.50 Control of Preheat Temperature for Welding of Low- Alloy Steel 1.51 (Withdrawn ~See 40 TR 30510, 7/21/75)
--- 73/06 1.52 Design, Testing, and Maintenance Criteria for Post Accident Enr.neered- 1 ~6/07 Safety Feature At.mosphere Cleanup System Air Tihration and Adsorpt2on 2 18/03 Units of Light-Water Cooled Nulear Power Plants
-- 73/06 1.53 AppUcation of the Single Failure Criterion to Nuclear Power Plant Protection Systems
--- 73/06-2.54 Quauty Assurance Requirements for Protective Cnatings Applied to Water.
Cooled Nuclear Power P! ants
--- 73/06 1.55 Concrete Placement in Category 1 Str,uctures
-- 13/06 1.56 Maintenance of Water Purity in Boili.ng Water Henetors (For Co:nment) 1 13/07
~- 73/06 1.57 Dasign Limits and Leading Combinaliens for Metal Primary Reactor Contain-ment System Components
~~ 73/08 1.58 Qu.alifIcation of Nucinar Power Plant inspection, Exasninstion, and Testing Personnel
-- 73/03 1.59 Design Basis floods for Nuclear Power Plants 1 76/04 ,
2 77/08
-- 73/10 j 1 60 Design Response Spectra for Seismic Design of Nuc! car Pnwer Plants 1 13/12
.. ~- 73/10 1.61 Da.mping Values for Seismic Design of Nuclear Power Plsnts
~- 73/10 1.62 Xanual Initiation of Protective Actions
~- 73/10 1.63 Cactric Penetration AssembUes in Con'.ainment Structures for Ligh Water- 1 17/05 ,
Cooled Nuclear Power P! ants 2 7&/07 l
--- 73/10 1.64 QutUty Assurance Requirements for t.he Design of Nuclear Power Plants 1 75/02 2 16/06
-~ 7;fjo 1 65 Materials and Inspections for Re,ctor Nsel Closure Studs
-~ ---
1.66 (WithJeswn~See 42 FR 54473,10/06/77)
"* 73/10 1.67 Ins! Allation of CVerpressure Irotection Devices
~- 73/11 1.64 Initis! Test Programs for Water Cooled Reactor Power Plants 1 77/01 2 72/CS
- l
-~ 75/13 1.63.1 Precperationa und Initial Startup Testing of Feedwater and Conden- 1 77/01 sate Systems for soiting Water Reactor Power Piants 4
fuved Title R ev. Year / Month
. Number 1.68.2 Initial Startup Test Program to Demonstrate Remote Shutdown
--- 77/01 1 78/07 Capability for Water-Cooled Nuclear Power P! ants
--- 73/12 1 69 Concrete Radiation Shields for Nuclear Power Plants Standard For-Jat and Content of Safety Analysis Reports for Nuclear Power --- 12/02 1.70 72/10 I
Plants 2 75/09 3 75/11 Welder Qua'ification for Areas of Limited Accessil.ility --- "3/12 1.71 .
-- 73/12 1.72 Spray Pond Piping Made from Fiberglass-Reinforced Thermosetting Resin 78/01 1
2 78/11 Qualificatico Tests of Electric Valve Operators installed Inside the -- 74/01 1.73 Containment of Nuclear Power Plants
--- 74/C2 1.74 - Quality Assurance Terms and Deff.nf tions
--- 74/02 1.75 PhysicaJ Independence of Electric Systems 75/01 1
2 78/09 1,76 Design Basis Tornado for Nuclear Power Plants
-- 74/04 1.77 Assumptions Used for Evaluating a Control Rod Election Accident for --- 74/05 Pressurized Water Reactors 2.78 Assumptions for Evaluating thu Habitability of a Nuclear Power Plant --- 74/06
- Control Room During a Postulated Haardous Chemical Release 1.79 Preoperational Testing of Enu rgency Core Coo.ing Systec:s for Pressurized --- 74/06 Water Reactors 1 75/09 1.80 Preoperational Testing of Instrument Air Systems -- 74/06 1.81 Shared Emergency and Shutdown Electric Systems for Nulti itnit Nuclear --- 74/06 Power Plants 1 75/03 1.82 Sumps for Emergency Core Coeling and Containment Spray Systems -- 74/06 1.83 Inservice Inspection of Pressurized Water Reactor Steam Oenerator Tubes --- 74/06
~, 1 75/07 1.84 Design and Fabricativo Code Case Accep'tabilf tv ASME Section !!!, --- 74/06 Division 1 1 75/04 2 75/06 3 75/09
- 4 75/11
. 5 76/02 6 76/05 7 76/08 8 76/11 9 77/0?
10 77/08 11 77/11 12 78/03
, 13 74/07 14 78/11 15 79/05 1.85 Materials Code Case Accepta!nbty. ASME Section .II, Division 1 --- 74/06 1 75/04 2 75/06
- 3 75/09 5
Issued Rev. Yeard. ton th N umber Title -
4 75/11 5 76/02 6 76/05
- 7 76/C8 8 16/11 9 77/03 10 77/08 Il 77/11 12 75/03 13 ~8/07 14 73/11 15 79/05
-- 74/06 1.86 Tern:ination of Operating 1.icenses for Nuclear Re.ctors
-- 74/06 1 87 Cuidance for Construction of Class i domponents in Elevated-Tem;>crature 75/06 1
Reactors (Supplement to ASME Section !!! Code Cases 1592, 1593, 1594, 1195, sod 1596)
-- 74/0s 1.B8 CoUection Storage, and Maintena.nce of Nuclear Poscr Plant Quauty 1 .75/12 Assurance Records 2 76/10
-- 74/11 1.89 Qualification of Class IE Equipment for Nuclear Power P!asts
-- 74/11 1.90 Inservice !nspection of Prestressed Concrete Contain:nent Structures with 1 77/D8 Crouted Tendons
-- 75/0!.
1.9) Evaluations of Explosions Postulated to Occur on Tr:tnsportation Routes 3 78/02 Near Nuclear Power Plants (For Comment)
--- 74/12 1.92 Combining Modal Responses and Cpatial Components in Seismie Responsa 1 76/02 A nalysis {
1- 7(/12 1.93 AvnDability of Electric Power Sources
--- 75/04 1.94 Quality Assurance Requirements for InstaDation, Inspection, and Testing 76/04 -
of Structural Concrete and Structural Steel During the Construction 1 Phase of NucIvar Power Plants
- . ~ 75,*tt2 1.95 Protection of Nuclear Power Plant Control Hoom Operators Against no 77/01 1
Accidental Chlorine Release
.- --- 75/05 1.96 Ocs g n of Main Steam isolation Valve Leskage portrol Systems for Boiling 1 76/06 uni Rcactor Nuclear Power Planta Instrumentation for 1.ight Water-Cooled Nuclear Power Plants To Assess
--- 75/12 1.97 1 77/08 Plant Conditions During and FoDowing an Accident
--- 76/03 1 98 Assumptions Used for Evaluating the Potential Radiological Consequences of a Radioactive Offgas System Facurs in a Boiling water Reactor (For .
Comn..tn t)
--- 75/07 1.99
- Effects of Residual Elements on Predicted Radiation Daar. age to Resetor 1 77/04
. Vessel Materials
-~ 76/03 l
- 1.300 Seis: sic Qualification of Electric Equipment for Nucivar Power Plants 1 77/08
--- 75/11 1.101 Emergency Planning for Nutlear Puser P? ants 1 77/03
--- 75/10 1.10: Flusd Protection for Nuclear Power Plants 1 7C/09
--- 75/11 ,
1.103 Tost T6nsioned Pre'stresssg Systems for Concrete Reactor Vessels 1 *6/13 and Containments 3
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. 1 J fuuod Rev. Yeard.tonth Title Number .
1.104 (Withdrawn -See 44 FR 4931 S/ /79)
--- 75/11 1.105 Instrument Setp"oints 1 76/11
-~ 75/12 1.106 Thermal Overload Protection for Electric Motors on Motor Operated Valves 1 77/03 -
t
-- 75/11 1.107 Quahtications for Cement Crouting for Prestressing Tendons m Containment 1 7~/02 Structures
--- ~6/0S 1.108 Periodic Testing of Diese! Cenerator Units Used as Onsite Electric Power 1 71/08 Systems at Nuclear Power Plants
--- "6/03 1.109 Calculation of Annual Doses to Man from Routine Rc! esses of Reactor 1 77/10 Effluents for the Purpose of Evaluating Compliance m-th 10 CFR Part 50, Appendix 1.
-~ 76/03 1.110 Cost Benefit Analysis for Radwaste Systems for 1.ight Water-Cooled j.
Nuclear Power Reactors (For Comment) 4- 76/03 1.111 Methods for Estimating At:nospheric Transport and Dispersion of Caseous 1 77/07 -
Effluents in Routine Releases fre.ra Light Water. Cooled Reactors :
-~ 76/04 I 1.112 Calculation of Releases of Radioactive Materials in Caseous and Liquid O-R 71/05 Effluents from Light-Water Cooled Power Reactors .
s . --- 76/06 ,
1.113 Estimating Aquatic Dispersion of Effluents from Accidental and Routine 1 77/04 Reactor RrJeases for the Purpose of implementing Appendix !
-~ 76/02 1.114 Guidance on Deing Operster at the Controls of a Nuclear Power Plant 1 76/11 [ t
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--- 76/03 j g
1.115 Protection Aghinst Low-Trajectory Turbine Missucs 1 77/07 {
-- 76/06 i '
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- 1. 116 Qu611ty Assurance Requirements for Installation. Inspection,.and Testing O-R 77/05 ;
of Mechanical Equipment and Systems ,
r
--- 76/06 l 1.117 Tornado Design Classit' ration 78/04 1 -
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--- 76/06 :
! IIS Periodic Testing of Electric Power and Protection Systems 3 77/11 l 2 78/06 i l'
l.119 (Withdrumn- See 4: FR 33387. 6/30/77)
- --- 76/06 1.100 Tire Protection Cuidelines for Nuclear Power Plants (Tor Comment)
' 1 77/11
--- 76/08 1.121 Bases for Plugging Degraded PWR Stews Generator Tubes (Tor C<,n.:nent)
--- 7C/09 1.122 Development of Moor Design Response Spectra for Seismic Design of Moor- 1 78/02 Supported Equipment or Components
--- 76/10 ,
1.123 Quality Assurance Requirements for Control of Procurement of items and 1 77/07 {
Services for Nuclear Power Mants i
~ 76/11 1.124 Service Limits and Loading Cc:nbinations for Class 1 Linear-Type *Camponent 1 78/01 Supports
-- 77/03
- 1.105 Physical Models for Design and Operation of Hydraulic Structures and 1 75/10 Systems for Nuclear Power Plants
-- 77/03' 1.105 An Acceptacle Model and Related Statistical Methods for the Analysis of 1 78/03
! / Poel Densification r
\
i ese~m m --m , w - e- - v - .,. -,,----ry,s v ,,m -- v mm 4y,e ' - ~ - - nmvw-
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lssued Title Rev. Yaar/Montis Number * .
--- 77/04 1.127 Inspection of Water-Control Structures Associated with Nuclear Power Plants 1 78/03
--- 77/04 1 1:3 Instauation Design 'and Instaustien of Large Lead Storage Batteries for 1 78/10 Nuclear Power Plants
-- 77/04 1.139 M.intenance, Test ng, and Replacement of Large Lead Storage Batteries 78/0 2 1
for Nuclear Power Plants 1.130 Service Limts and Lnading Ccmbinations for Class 1 Plate-and- Shen Type -~ 77/07 1 78/10 i Component Supports
--- 77/08 .
1.131 Quahfication Tests of Electric Cables, Tield Splices, and Connections for
' Light Water Cooled Nuclear Power Plants (For Comment)
--- 77/09 1.132 Site Investigations for Foundations of Nuclear roaer Plants 1 79/03 1.133 Loosc Part Detection Program for the Primary System of Light Water-Cooled
--- 77/09 Reactors (Tor Comment) 1.134 Medical Evaluation of Nuclear Power Plant Personnel Requiring Operator
-- 77/09 1 73/03 Licen ses 1.135 Normal Water Level and Discharge at Nuclear Power Plants (For Conunent)
--- 77/ 09
-- 7'/11 1.136 Material for Concrete Containments 1 78/10 1.137 Fuel-OU Systems for Standby Diese! Cenerators (For Com' ment)
-- 77/01 1.138 Lauratory InvestiCations of Sous fur Engineeri.ng Analysis and Design of
-~ 78/04 Nuclear Pe=er Plants (For Comment) '
--- 78/05 1.139 Guidan,re for Residual 11 cat Removal (For Comment)
--- 78/03 p 1.140 Design Testing, and Maintenance Criteria for Normal VentDation Exhaust r System Air F0tratico and Adsorption Units of Light Water Cooled . [.
Nuclear Power Plants (For Comment) E 1.141 Containment 1 solation Provisions for Fluid Systems (For ("omment) -- ~5/04 E i-1.14: Safety-Related Concrete Structures for Nuclear Power Plants (other Than
-- 78/04 [
.. Reactor vessels and Containments) (For Con. ment) [ >
, E J.143 Design Guidance for Radioactive Waste Management Systems, Structures.
--- 78/07 E i
r and Caponents Instated in Light W6ter-Cooled Nuclear Power P! ants (For Comment)
-- 79/01 1.144 Auditing of Quality Assure ..e Programs fur Nuclear Power Plants {
(For Coinment) 1.145 Atmospheric Dispersion Models for Putential Accident Consequence
-- 79/08 Assensments at Nuclear Power Plants (For C<.mtnent) : 1 O
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-- __ ~ _
t Draft Regu! story Caldes . ,
1swed Task Year /Mone Number .
Title !
79/02 D1 805 5 Nuclear Analysis and Desi;n of Cwnerete Radiation Shielding for ,
Nuclear Power Plants Qualification of Quality Assurance Program Audit Personnel for 79/02 RS 810-5 Nucl ear Power Plants Functional Specification for Safety-Related Valve Assemblies in 79/02 SC 704 5 Nuclear Power Plants 79/04 SC 807 4 Determining Prestressing Forces for inspection of Prestressed (Proposed Concrete Containments R.C. 1.35.1)
Ultrasonic Testing of Reactor Vessel Welds During Inservice 19/05 SC "05-4 Examination l
Qualification Test for Cable repetration Fire Stops for Use in 79/07 RS 809-5 Nuclear Power Plants Inservice Inspection Code Case AcceptabQity--ASPIE Section XI, 79/08 SC 7:1-4 Division 1 e
Lightning Protection for Nuclear Power Plants 79/08 RS 705-4 Safety Related Permanent Dewatering Systems for Nuclear 19/09 FP 813-4 Pcwer Plants LWR Core Reloads; Cuidance on Applications for Amendments to 79/09 SC 521-4 Operating 1.icenses and on Bcfueling and Startup Tests i
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. _ _ . . . . . _ . . ~ . _ . . . . . _ . . _ . . . _ . _ _ . . _ _ _ . - .
s Propowd Revisions to Retslatory Guides a Task and R.G.
- Propoed lesved Nurntars Title Revision YsatAtcn2
[
RS 807-5 Personnel Selection and Training 2 79/02 [
1.8 t 1
Sr.810-4 :nservice Inspection of Ungrouted Tendens in Prestressed 3 79/04 I
. 1 35 Concrete Containments i
RS 901 5 Qualification of Nuclear Power Nant inspect. ion, Eaamination, 1 75/07 1.58 and Testing Personnel RS 050-2 Qualification T.s'. s ef !97tric Cables and Field Splices for 1 79/08 1.131 1.aght Water-Cooled Nursear Power Plants P.S 902 4 QuaUty Assurance Program Requir.rmen's (Operation) 3 19/08 1.33 f
RS 908-5 Quality Assurance Requirements for Instauation. Inspection, 2 79/09 !
1.94 and Testing of Structural Concrete, Structural Steel, {
Soils, and Foundations During the Construction Phase :
of Nuclear Power Plants -
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Am! story Guida Und-e Devetoo.wt
- Interhu Cuide on Tornado Missiles
- Single-Tallure Cr:teria for L!rht Water Reactor Plants TIuid Systems
- Eart.hquake Instrumentation Data ilanCin; for Nue! car Fo=cr Plants
- Pressurized Water Reactor and Bo(Jing water Reactor Containment Spray Design Criteria
- Criteria for Electrie, Instrumentation, and Control Portions of Safety Systems
- Design and Construction Deficiency Reporting Requirements
- Qual!!ication of Electric Modules for Nuclear Power Plants
- Quality Assurance P.equirements for Packagmg. Shipping, Receiving, Storage, and Handling of Ite os for Nuclear Power Plants
- Meteorological Extreme Air Temper 6tures for Design and Operation cf Nuclear Power Plants
- Extreme Windspeeds in Coastal Areas for Doign and Operation of Nuclear Power Plants
- Geochronologic Techniques Applaed to Nuclear Power Plant Siting
- Procedures and Criteria for Assessing Sou !.jquefaction Potentia) at Nuclear Thty Sites
- Poundation and Earthwork Construction for Nuclear Pnwer Plants
- Snow and lee Accumulations for the Desig s and Operation of NucJear Power Plants
- Geolngfral Mapping of Excavations for Nuch ar Power Plants
(
- Fracture Analysis of Plaws at Structural Discontinuities
- Inservice Monitoring of Core and Core Support Structure Motion t*ia Neutron-Flux Measurement .
- . Requirements for Qualifiention Tests and Production Tests for Piping and Equiprnent Snubbers
- Msthods of Analysis and Design of Reinforced Concrete Containrr.ent Structures
- Recomrnendations for Inservice Testing of Valves Required to Perfor-n a Safety F' unction in !.ight-Water Reactors 0
f 11 4
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(
Repu'atory Guides Seing Reve.ed
- Revision 1 to Regulatory Guide 1.9, "Se!<ction, Desip, and Quaufication of Diesel-Generator Unit.s Used as Standby (Cnsite) Electric Power Systems at Nuclear, Power Plants *
(
- Revision 2 to Regulatory Guide 1.12, " Instrumentation for Earthquakes" l.
t
- Revision 2 to Regulatory Guide ! 14. "Rtetor C.,clant Pump Tlywhee! Integrity"
- Revision 5 to Regulatery Guide 1.16. " Reporting of Operation Inforn.ation.. Appendix A Technical S t.ecifica tion s" .
Revision 1 to Regulatory Guide 1.05, " Assumptions Used for Evaluating and Pr essurized Water Reactors"
- Revision 1 to Regulatory Guide 1.50, " Control of Preheat Temperature for Welding # Low- AUoy Steel"
- R evision to Regulatory Guide 1.56, " Maintenance of Water Purity in Doiling Water Reactors"
- Update cf Revision 3 to Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis p Reports fer Nuclear Power P! ants" Revision 1 to Regulatory Guide 1.71, " Welder Qualarication for Areas of Limited Accessibility" f
- E
- Revision 1 to Regulatory Guide 1.80, "Precperational Testing of instrument and Control Air Systems"
- Revision 16 to Regulatory Cuide 1.84, " Design and Fabrication Code Case AcceptabtIity-- ASME I '
Section !!!, Division l'
- . Revision 16 to Regulatory Cuide 1.85. " Materials Code Case Acreptability-- AS*4E Section !!!, [
Division J" E for Nuclear Power P ants" j.
- Revision 1 to Regulatory Guide 1.89, "Qualificatic.n of IE Equipment f
- Revision 2 'o t Regulatory Guide 1.101. " Emergency Planning for Nuc!t.ar Power P1 wits"
. I
- Revision 1 to Regulatory Guide 1.133, " Loose Part Detection Program for the Primary System of }
Light-Water Cooled Nuclear Power Plants" Revision 'l to Regulatory Cuide 1.105, " Normal Water Level Disci..irge at N'uct:ar Power P! ants" f
- ip
- Revimn to Regulatory Calde 3.136, '.Material for Concrete Containments ( Article CC- COO of the [
'Cuie for Concrate Scactor Vessels and Containments')" t
- Rvvaian 1 to Regulatory Guide 1.137, " Fuel 00 Systems for Standby Diesel Cenerators' T
- P.rv'sion 1 to Regulatvry Guide 1.129 " Guidance for Residual Heat Removal" t
- Revieien 1 to Regulatory Guide !.140, " Design, Testing, and Maintenance Criteria for Normal VentDation Ethaunt Systern Air Filtration and Adsorption t* nits ef Light-Water-Cooled Nuclear Power Plants"
- Rcettion 1 to Begul.etory Guide 1.141, " Containment Isolation Provisions for Fluid Systems *
- Ru vissem I to Rirgulatory Culde 1.142, " Safety Related Concrete Structures for Nuclear Power P! ants (Other Than Reactor Vessels and Contair.ments)" i f
- Revision 1 to Regulstory Guide 1.143, " Design Gui&.nce for Raidwaste Nnager ent Structures, Systems, and Ccinponents instaued in Light. Water-Cooled %c1 car Power Plants" {.
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s Other Regulatory Guides of Possible Interest to Division 1 Recio.ents
. j fssued Titie R ev. Year / Month Numtwr .
1 7f,/04 4.1 Programs for Monstoring Radioactivity in the Environs of Nuclear Power Plants (Tor Co:vnent) 2 76/07 4.2 Preparation of Envire.,nmental Reports for Nuclear Power Stations
-- 74/05 4.4 Reporting Procedure for Mathematical Models Selected to Predict Hested Effluent Dispersion in Natural Water Bodies
-- 74/05 4.5 Measurements of Radionuclides in the Envi.renroent-Strontiura-89 and Strontium-90 Analyses 2 75/12 4.7 Ceneral Site Sultability Criteria for Nuclear Power Stations i Environmental Technical Specifications for Nuclear Power Plants
--- 75/12 ! i 4.8 *
(For Comment) a 1
Terrestrial Environmentaj Studies for Nuclear Power Stations 1 77/C8 4.11 Performance. Testing, and Procedural Specifiestions for Thermolumi- 1 77/07 4.23 nescence Desimetry: Environmental Applications Quality Assurance for Radinlogica) Monitoring Programs (Norinal
-- 77/12 4.15 Operations) Effluent Streams and the Environment (For Comment)
Serial Numbering of Tuel Assemblies for Light-Water Cooled Nuclear --- 12/12 5.1 Power Reactors Control of Personnel Access to Protected Areas Vital Areas, and -- 73/06 5.7 Material Access Areas }
5.12 . Central Use of locks in the Prutection and Control of TacDities and
--- 73/11 f Special Nuclear Wierials ,
i 75/06 i 5.29 Nuclear Material Control Syrtems for Nuclear Power Plants 1 ,
Plant Security Forre Duties
-- 75/01 5.43 Perimeter intrusion Alarm Systems ! 76/06 5.44
- Standard Format and Content of Safeguards Cuntingency Pians for --- 78/03 5 54 Nuclear Power Plants (For Comment) i Administrative Guide for Packaging and Transporting Radioactive Material -- 74/06 7.1 I
-- 74/06 I j 7.2 Packaging and Transportation of Radicactively Contaminated Biological Materials Procedures for Picking Up and Receiving Packages of Radioactive -- 75/05 7.3 Materia.1 (For Cearment)
Leakage Tests on Packages for Shipment of Radioactive Materials (For - - - - 75/06 7.4 Comme n t) ,
Ad uinistrative Cuide for CLtaining Exem;,tiwns Froen Certain NRC Require- C-R 77/05 7.5 monts Over Particactive %ts rial Shii.aients Ocsign Criteria for the Structural Analysis of Shipping Cask Containment ! 78/03 7.6 Vessels I
Ar.ministrative Guide fur Verifying Con.pliance h*ith Packaging Require- ~~ 17/08 7.7 a ments for %iisments of Rt.<hoactive Malarials (For Cominent) s 1.8 Lead Combinations for the Structural Analysis of Shipping Casks (Tur -~ 77/05
, s Commen t) i 1?-
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lssued /
Title Rev. Year / Month (
Nember ' *
--- 79/03 7.9 Standard Por:nat and Content of Part 71 Applications for Approvaj of Packar ng of Type B, Large Quanuty, and Pissue Radioacuve Mate-rial (For Commen,t)
--- 73/02 8.1 Radiation Symbol b
--- 73/02 f 8.2 Guide for Administrative Practices in Radiation Monitoring [-
-- 73/02 [.
P0m Badge Perfor:r.ance Criteria t 9.3
--- 73/02 f 8.4 Direct Reading and Indirect Reading Pocket Desiraeters
--- 73/05 8.6 Standard Test Procedure for Ceiger-Mu~Uer Counters
.--- 73/05 8.7 Occupatfonal Radiation Exposure Records Systems 3 78/06 8.8 Information Relevant to Ensudng That occupational Radiation Exposures .
- at Nuclear Power Stations Will Be As Low As is Reasonably Achievable
--- 73/09 8.9 Acceptable Concepts, Models, Equations, and Assuroptions for a Bionssay Program 1-R 77/05 c.10 Operating Phuosophy for Maintaining occupational Radiaticn Exposures As Low As !s Reasonably Achievable s
1 75/11 8.13 Instruction Concerning Prenatal Radiation Exposure
! 77/08 8.14 Personnel Neutron Desimeters
--- 76/10 8.15 Acceptable Programs for Respiratory Protection 1 19/06 8,19 Occupational Radiat' ion Dose Assessment in Light-Water Reactor Power .
Plants -Design Stage Man-Rem Estimates I 1 19/09 i
- 8. ::0 Applications of Bionssay for 1-135 and 1-131 L 3 77/05 l 10.1 CompCation of Reporting Reqtiirements (or Persons Subject to NRC ,
Regulations *
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/ o, UNITED STATES y ni g NUCLEAR REGULATORY COMMISSION 3 s , t -
WASHINGTON. o. C. 20SS5 WL t 4 373 r
I 00CXET NO. 50-363 !
AFDLICANT: JERSEY CENTRAL PCWER AND LIGHT COMFANY FACILITY: FCRKED RIVER NUCLEAR STATION
SUBJECT:
MSETING HELD ON JUNE 20, 1979 A' meeting with th'e applicant's representatives was held in Bethesda, Maryland on June 20, 1979. The purpose of the meeting was to discuss the applicant's i request for an extension to the Forked River construction pemit. The ::eeting participants are listed in the Enclosure.
.i.
The staff presented some of its thoughts concerning not only the construction permit extension but also the potential difficulties in conducting an operating f-j license review in the future considering the rather long elapsed time since issuance of the construction permit.
The censtruction permit application was filed in 1970 and the construction permit was issued in 1973. Plant construction has proceeded slowl,y and is now only about .
3". ccmpl ete. Howver, approximately 350 million dollars has been expended in [
design and equipment procurement. Assuning construction does not begin again for i tw or three years, the plant design would be about 15 years old when the RC i begins the operating license review. The design wuld probably deviate considem )
' ably from the then-current Standard Paview Plan's acceptance criteria, making the j staf f's review more difficult and more time-consuning. For example, the ASME Ccde i to which the reactor vessel was purchased was the 1971 edition. The containment !
vessel was designed in accordance with Section VIII of the ASME cede rather than i 2ection III, as would be the case for plants of more recent design. The seismic E cesign criteria are old, but the applicant believes that a re-analysis would show {
that the plant would meet current criteria. }
The NRC environmental review was completed in 1973. Since then, additional generic environmental concerns have arisen that will have to be addressed.
The reasons for needing an extension to the construction permit were,"as given in the applicant's August 31, 1978 letter, primarily financial in nature. In order to show good cause for the extension, the facters contributing to the .
delay should be beyond the applicant's centrol. As written in the August: 31
!Etter, it is noc clear to the staff that the factors cited were indeed beyond the applicant's control . Furthermore, the occurrence of the accident at Three
.* fie Island Unit 2 has uncouctedly exacerbated the applicant's financial con-cition, raising questions abcut the a:clicant's financial capability to complete {
the ccnstruction of Forked R:ver. [
g m d Up G / b ) h),
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.t. Arnold of General Public Utilities, the parent corporation of the applicant, presented a brief chronology of events since the issuance 'of the construction -
3 permit in July 1973. Construction was halted in 1974 due to financial difficul-ties, and was begun again in 1977. On August 31, 1973 the applicant filed a request to extend the latest date for ccmpletion of construction fran Cetober 4, 1978 to February 1,1985. On April 3,1979, construction was halted in order to conserve funds that might be needed, as a result of the accident at Three Mile -
Island Unit 2, for other purpcses. The substructure is about 60% complete, but construction has not reached above grade level. Most major equi;=ent items are 1.n storage at the site.
The applicant stated that it has upgraded the design as far as ' practicable to facilitate the process for obtaining an operating license. This included reviews of the Regulatory Guides through Regulatory Guide 1.95, and of the Stan- i da'rd Review Pl an. They have kept up-to-date on.the cperating license review of !
San Oncfre Lhits 2 and 3, which is similar in design to Forked River, and have updated the Forked Rivte design in some areas.
Mr. Arnold explained that ocean cooling, as an alternative to the ccoling tower, is more e.xpensive and is less desirable environmentally. He stated that the State of i4w Jersey may be willing to grant a variance on the salt deposition ;
1imit such that the cooling tower will be acceptable.
According to Mr. Arnold, Jersey Central will be short on its base load capability and on reserves by 1981, and GPU's comitment to the Pennsylvania-New Jersey-Maryland (PJM) pool will also be short. GPU will install gas turbines to try to ;
meet the shortage. Forked River construction will not begin again for at least ,
two years . Jersey Central will probably try to buy participation in other plants I to carry them into the 1980's, and may 1.cok for participation in the Forked River
. plant by other utilities. -
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l As a result of the discussiens at this meeting, the applicar.t agreed to send us ,
a letter that (1) states that construction work has been halted, (2) requests that we not take action at this time regarding their August 31, 1978 request for CP extension, and (3) states that they will notify us when a decision has been j made to re-start construction work. In. addition, they said they will provide '
detailed information supporting a showing of good cause for construction permit extension and a detailed description of plant design changes intended to meet new safety and environmental standards.
- R. A. Benedict Light Water Reactors Branch fio. 2 Divisien of Project Management
Enclosure:
Attendance List ces w/ enclosure:
See next page {
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- J;rssy Carstral Pow:r & Light Comptriy J
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(') Macisen Ave'tue at Punch ScM Acto Mornstewri. New Jersey 07960 (20t) 45542M file: 2415.6 2L21.1.1 July 30, 1979 i Robert Baer , Branch Chief
'Lig ht Wa te r Re ac to rs Branc h No . /*
Nuclear Regu.lato ry Commission Was hing to n , DC 20555
SUBJECT:
Forked River Nucien: S ta tion Docks t Number 50-363 Construc tion Fermit No. CPPR-96 i
Dear Mr. Baer:
l This vill confirm our June 20, 197 9 =eeting to discuss the sta tus of the :
construction program for Jersey Central Power & Light Company's (JCP&L) !
Torked River Nuclear Sta tion. Specifically, we discussed the sta tus of the construc tion program for the Station immediately prior to the TMI-2 accident I and the effects of the accident on the cash resources available to General !
Public Utilities Corporation (GPU) and its subsidiaries, including JCP&L, and their relationship to the Forked River construction program. As a result, we requested that the NRC defer temporarily any action on our August 31, 1978 request for an extension of the completion dates for the Station.
i As we had discussed, a construc tion permit was issued for the Forked I River Station on July 10, 1973 and construction commenced immediately. !)
However, shortly thereaf ter construction work at the site was slot-ed down l l
until early 1977. The reasons for such delay and the rationale for re- - !
scheduling the in-service date to December,1983, were described in my August 31, 1918 letter to Mr. S. A. Varga, as part of a request fo r an l extension of the earliest and latest completion dates for the Station -l
( Ar.tachment 1, hereto.)
}
, As of the date of the TMI-2 accident (March 23,197 9), engineering and construction work for the Torked River Station was proceeding on a schedule ' j consistent with the December,1983, in-service date. Construction ass 5.6% l cemplete, engineering was 47.27% ccmplete and the aajor equipment :onnected i vich the nuclear steam supply system had been delivered to the site. JCP&L's total investment in the Station was approximately $350 million.
Following the TMI-2 accident, and because of the enor=ous cost s imposed oti che companies in the GPU system, on April 3,1979, GPU announced the i suspension of work on major construction projects as well as other =easures to l teduce cash ouciays. As a result, nest engineering and all construction work l on the Forked River Station has been deferred and all on-site equipment placed I in storage. '
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f?A Mr. Robert Bacr Page 2 As we discussed at our June 20, 197 9 meeting, JCPSL needs additional generating capacity presently to meet its base load requirements; by 196L JCP5L will need additional capacity to meet its peak lead requirenants.
Nevertheless, it is clear that the GPU and JCP&L financial situation vill preclude recoc=encement of construction at the Forked River Station for some time . And, as we had explained during our =eeting, G7U and JCP&L have not had I sufficient time either to evaluate fully the financial Lap 11 cations of the New I.
Jersey Board of Public Utilities' recent rate decision, or to determine what further rate relief will be necessary to support the r esumptio n o f c ons t ruc-tien ac tivities at the Forked River sita. In any event, we did indicate that recom=encement of construction probably would not occur for at least two years .
In the interim, JC7&L will explore o ther alternatives for meeting its
,near-term capacity requirements including the possiblity of purchasing additional capacity f rom sources located outside the CPU system. JCP&L also vill investigate the feasibility of sharing vich others the costs and owner-s hip o f t he Fo rked Riv er Sta tio n.
Importantly, we plan to continue the program Laplemented during the period of the initial delay in commencement of construction; that is , JCPSL vill evaluate new NRC regulatory criteria and apply such criteria, where E practicable mad applicable, to the design of the Forked River Station.
Additionally, we plan to advise the NRC of any decision to recocmence en-gineering and construction activities a t the Focked River site and vill at t ha t time propose any new completion dates and provide supplemental justifi-cation for extension of such dates.
In view of the foregoiIng, we believe that it would be appropriate for the NRC to defer at this time any action on our August 31, 1978, request for an extension of the ccepletion dates for the Station. ,
{ l Very truly yours, [
Ivan R. Fin ' ek, r. [
Vice President '
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