ML20136F380
| ML20136F380 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/24/1985 |
| From: | Phillips L, George Thomas Office of Nuclear Reactor Regulation |
| To: | NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP) |
| Shared Package | |
| ML20136F370 | List: |
| References | |
| NUDOCS 8601070365 | |
| Download: ML20136F380 (33) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
)
)
CLEVELAND ELECTRIC ILLUMINATING
)
Docket No. 50-440 OL COMPANY, ET AL.
)
50-441 OL
)
(Perry Nuclear Power Plant,
)
Units I and 2)
)
JOINT AFFIDAVIT OF LAURENCE E. PHILLIPS AND GEORGE THOMAS CONCERNING SINGLE LOOP OPERATION CONTENTIONS RAISED BY "0CRE" I, Laurence Phillips, being duly sworn do depose and state as follows:
I am employed as a Section leader in Reactor Systems Branch, Division of.BWR Licensing in the Office of Nuclear Reactor Regulation. A statement of my professional qualifications is attached.
(Attachment 1).
I,~ George Thomas, being duly sworn do depose and state as follows:
I am employed as a Nuclear Engineer in Reactor Systems Branch, Division of BWR Licensing in the office Nuclear Reactor Regulation. A statement of my professional qualifications is attached.
(Attachrnent 2).
The purpose of our joint affidavit is to respond to the (0CRE) contentions regarding, Single Loop Operation (SLO).
~
1.
The applicant, The Cleveland Electric Illuminating Company (CEI),
informed the staff in their letter PY-CEI/NRR-0369L dated October 28, 1985 that appropriate changes to the technical specifications will be submitted after the recirculation flow limit has been detennined during start-up testing. Even though the applicants submitted SLO analysis in Amendment 22 to their FSAR, the staff has not approved Perry for SLO. The staff will 8601070365 860103
{DR ADOCK 05000440 PDR
2 evaluate SLO when CEI submits the technical specification changes required for SLO. At present SLO is not approved for Perry.
The staff response to specific OCRE contentions are as follows:
2.
Contention B-1.
" Applicants should analyze the progression and consequences of an anticipated transient without scram ("ATWS") initiated by the inadvertent startup of the idle recirculation loop when operating at 70% of rated thermal power with single loop operation. The analysis should demonstrate that this event will meet the safety criteria outlined in Section 15C.3 of the FSAR."
3.
The ATWS rule 10 CFR 650.62 deals directly with the ATWS events and identifies the equipment needed to mitigate an ATWS. The rule does not require the analysis of ATWS events. The analyses considered in developing the rule included the most severe ATWS events and thus bound any other ATWS events, including any SLO ATWS, since they would occur at reduced power.
4 The applicant provided the ATWS analysis in Appendix C of the FSAR.
The Perry ATWS analysis assumed an initial power level of 100% for the the limiting transients. This initial condition bounds the maximum power level expected in SLO.
For these reasons, the Staff does not consider this contention as a significant safety issue.
5.
Contention B-2.
" Applicants have not demonstrated that the seizure of the operating recirculation pump when operating up to 70% of rated thermal power with a single loop will not exceed fuel safety limits, assuming scram functions, and that ATWS initiated by this event will meet the safety criteria of FSAR Section 15C.3."
6.
Generic analyses were performed by General Electric for a large core BWR/a plant to determine the impact of recirculation pump seizure accident on one-recirculation pump operation. Thermal power of 75% and core flow of 58% were assumed. The Minimum Critical Power Ratio (MCPR) was determined to be greater than the fuel cladding integrity safety limit.
Therefore, steam blanketing of fuel rods would not occur and no fuel failures would occur as a result of this analyzed event. Although these
, results are for a large core BWR/4 plant, similar results are expected for Perry which is a BWR/6 plant.
7.
Board Notification BN-84-062 addresses a potential problem of sustained local power oscillations due to lack of local power monitoring in the scram circuitry. Technical Specifications for single loop operation will reouire surveillance of local power range monitors and operator actions to suppress oscillations as was done in the foreign reactors described in the Board Notification.
- 8..Because the recirculation pump rotor seizure is categorized as an " Accident Event" based on low probability of occurrence, a failure to scram need not be considered in addition to the recirculation pump rotor seizure. The Staff does not consider Contention B-2 as a significant safety issue.
9.
Contention B-3. " Applicants have not demonstrated that the traversing incore probe ("TIP") noise uncertainty values reported in FSAR Section 15.F.2.2 are applicable to single loop operation up to 70% of rated thermal power; consequently, the minimum critical power ratio
("MCPR") may not be determined in a conservative fashion."
- 10. The Safety Limit FCPR is determined using the General Electric Thermal Analysis Basis (GETAB) which statistically combines all of the uncertainties in operating parameters and procedures used to calculate 1
critical power. The uncertainties used in the statistical analysis are i
independent of single loop versus two loop operation except for measure-ment uncertainties relative to core total flow and TIP reading.
- 11. The "TIP uncertainty" consists of several components which are combined statistically (by a square root of the sum of squares). One of these components is the random noise factor, which is the variation in TIP output averaged over six inch axial segments of TIP travel determined by w--
. repeated traverses of the sama instrument tubes.
For analysis of normal reactor operation, this component has been determined to be a 1.2%
uncertainty at a Icr level (NED0 20340, Process Computer Performance Evaluation Report, June 1974). For first cycle operation, the combined value of the other components of TIP uncertainty is 6.2%, which results in a statistically combined total TIP uncertainty of 6.3%.
For reload cores, the TIP noise remains the same but other uncertainty components increase to give a total TIP uncertainty of 8.7%.
For Single Loop Operation (SLO),
all components of the TFP noise remain the same as for normal operation except for the TIP noise component, which increases to 2.8%.
The statis-tically combined total 'IP uncertainty for SLO is 6.8% for the first cycle and 9.1% for reloads.
- 12. The sensitivity of the safety limit MCPR to TIP measurement uncertainty is such that a ?% increase in TIP uncertainty (e.g., from 6%
to 8%) results in about 1% increase in MCPR. The increase in TIP uncer-tainty from 6.3% (8.7% reload) to 6.8% (9.1% reload) increases the MCPR Safety Limit by about 0.002, e.g., from 1.06 to 1.062. This increase is inciaded, along with the effect of an increase in flow measurement uncer-tainty from 2.5% for two loop operatinn to 5,0% for SLO, in the determin-ation of the increase of safety limit MCPR from 1.06 to 1.07 (see FSAR Section 15.F.2). When the Applicants submit techni al specifications for SLO, the higher value for MCPR must be included.
It is possible that measurements at other SLO power-flow conditions (e.g., up to 70% power) would produce a TIP noise uncertainty larger than 2.8%.
However, to produce an additional incremertal increase
, in the Safety Limit MCPR (e.g., from 1.07 to 1.08) would recuire an increase in TIP uncertainty of more than 1.0 percent, which is equivalent to a TIP noise value of about 4.8% (rather than 2.8%). TIP noise measure-ments made for two loop operation at various power-flow levels during beginning of cycle start-up tests show no significant sensitivity to power level (e.g., Washington Nuclear Plant Unit 2 Final Start-up Report, April 17, 1985, Table 3-16). On this basis we would not expect sufficient increase in noise level at the 70% power plateau for SLO to result in an incremental increase in Safety Limit MCPR.
- 14. The staff does not consider Contention B-3 as a significant safety issue.
- 15. Contention B-4.
" Applicants' Technical Specifications for single loop operation up to 70% of rated thermal power should include limits on the core plate pressure drop."
- 16. Technical specifications when approved for SLO will require surveillance and limits on core plate pressure drop. However, the Staff requires this to avoid possible excessive jet pump vibratfor and not for regulation of incore coolant finw. The Staff does not consider this contention to be a significant safety issue.
- 17. Contention B-5.
" Applicants have not demonstrated that single loop operation up te 70% of rated thermal power will not aggravate the strorp variability in flow rate along the fuel channel seen in fast BWR transients, or that this phenomenon has been conservatively accounted for in analyses of fast transients."
- 18. Contention B-5 refers to an iterative method for transient CPR evaluation which has not been used in licensing safety analyses and has not been reviewed by the staff. The methods used by the licensee for both two loop and single loop operation are the standard General Electric analyses which have been widely used and accepted by the staff for licensing evaluations.
t
- 19..As stated in FSAR Section 15.F.3.1, the consequences of abnormal transients initiated from single loop operation are considerably less severe than those initiated from the two loop operational mode. The maximum single loop power level of 70% results in substantial operating thermal margin in comparison to the full power operating conditions which were included in the two loop safety analyses.
In fact, FSAR results for the Generator Load Rejection With Bypass Failure transient initiated from single loop operating conditions show no significant reduction in thermal margin due to this transient.
20.
In the Staff's opininn, Contention B-5 does not raise a signifi-cant safety issue with respect to the safe operation of Perry.
We attest that the foregoing affidavit is true and correct to the best of our knowledge and belief.
O 1
J-(f _
.L h u. Y_ w r u d fe Laurence E. PhiT1 fps ~
/Geerge Troiiias S
Section Leader Nuclear Engineer Reactor Systems Branch Reactor Systems Branch Division of BWR Licensing Division of BWR Licensing Office of Nuclear Reactor Regulation Office of Nuclear Reactor Regulation and sworn to before me Subscrjbp~' day of M a t W, 1985 this[#
E cY3 h S k <c (
Notary Public My comission expires: Tt r
ATTACHMENT 1 PROFESSIONAL QUALIFICATIONS OF LAURENCE E. PHILLIPS REACTOR SYSTEMS BRANCH DIVISION OF B0ILING WATER REACTOR LICENSING (BWR) 0FFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COP 9tISSION I am employed as a Section Leader in the Reactor Systems Branch, Division of PWR Licensing in the Office of Nuclear Reactor Regulation.
I graduated from the University of Cincinnati with a Chemical Engineering degree in 1954. After serving two years as an officer in the the United States Army, I have been continuously employed in the nuclear engineering profession since January,1957.
I received a M.S. degree with nuclear physics ma;ior from Union College of Schenectady, N.Y., in 1961.
I am a registered Professional Engineer, Certificate #E-026547, in the state of Ohio.
In my present work assignment at the NRC, Division of Boiling Water Reactors (BWR), i have supervisory responsibility for the review and evaluation of Boiling Water Reactor systems, BWR core design and operating performance, dnd radiological Considerations associated with reactor accidents.
In addi-tion, my Section participates in the review of analytical models used in the licensing evaluation of the core thermal-hydraulic behavior under various operating and postulated accident transient conditions.
I have had similar supervisory responsibilities with NRC since 1976. Over a period of several years, I have participated in the study of thermal hydraulic stability in BWPs, and have supervised the recent technical resolution of Generic Issues B-19, " Thermal Hydraulic Stability," and B-59, "(N-1) Loop Operation in BWRs and PWRs.
Prior to joining the NRC staff in December, 1974, I was employed by NAI Corporation as a Senior Associate.
In this capacity, I was responsible for the development and application of computer codes for analysis of nuclear reactor cores.
I acted as a consultant to nuclear operating utilities in the use of these codes for analysis of their operation, and in the solution of general nuclear engineerina problems. My tenure at NAI was from 1967 through 1974.
From 1962 to 1967, I was employed by Allis Chalmers Mfg. Co. My assignments during that period included supervisory responsibility for the safety analyses and licensing of the Lacrosse Boiling Water Reactor.
From 1958 to 1962, I was employed by Alco Products where I was project manager for the design, development, and fabrication of heat exchange equipment for nuclear liquid metal projects. Prior to that I was with the Nuclear Division of the Martin Company.
ATTACHMENT 2 PROFESIONAL QUALIFICATIONS OF GEORGE THOMAS DIVISION OF BOILING WATER PEACTOR LICENSING (BWR) 0FFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY COMMISSION I have been employed as a Nuclear Engineer in the Reactor Systems Branch, Division of BWR Licensing, Office of Nuclear Reactor Regulation, since October 1980.
I serve as a reviewer in the area of reactor systems (BWRs). This involves performing reviews and evaluations of those portions of the applications for Operating Licenses and submittals regardino proposed modifications in licensed nuclear power plants for which the bra ich has responsibility to assure public health and safety.
Since 1981, I participated in the Perry review for the Reactor Systems Branch.
I received a Bachelor of Science degree in physics for Kerala (India)
University in 1963. Additional graduate and professional courses were taken in Nuclear Engineering, University of Pennsylvania and Engineers Club, Philadelphia, PA - 1975. Other educational background and training includes:
Power Plant Engineering - 1976 (diploma for International Correspondence Schools, Scranton, PA); PVR Technology Course - 1980 (NRC sponsored); BWR/6 Simulator Course - 1981 (NRC sponsored); Tarapur Atomic Power Station (India)
- Reactor Operators Training Program - 1969; GE BWR - Training at Tarapur by GE - 1967.
From 1967 to 1972 I served as a Reactor Operator on the Indian Atomic Energy Commission's first commercial nuclear power station, Tarapur 1 & 2 (a BWR built by Bechtel and GE). There I participated in construction tests, pre-operational tests, and nonnal operations of the station.
Frcm 1973 to 1975 I was employed by United Engineers and Constructors (UE&C),
Philadelphia, PA.
Initially I was a Test and Start-up Engineer in the Construction Division of UE8C.
In this capacity I wrote various procedures and systems descriptions for a BWP. Subsequently, I worked as a staff Nuclear Engineer on the Nuclear Technical Staff of UE8C.
I was engaged in providing technical expertise and consultation services to all nuclear projects of UE&C.
From 1975 to 1980 I was a Systems Engineer in the Power Division of Stone &
Webster Engineering Corporation.
I performed detailed engineering and design of reactor systems of a BWR. My duties included project interface and coordination work with the ASSS supplier (GE) and the client (utility).
\\
~
vue PART 50 o PROPOSED RULEMAKING
- 3..
47 F R 13369 important in providing reasonable resulted in protracted discussions during Pubhshed 3/30/82 assurance that the facility (would) be the liceruling process, and comment pe"od "P"es en182.
constructed and operated without undue misapplication and misinterpretauon of
[-
comments rec *'*d 8" 'h d'"
hazard to public health and safety "
requirements by plant operating staffe 3
enH be considered if it is practical Technical specifications that were after a license was issued.
to do so but assurance of.
formulated in SCCordance with this In recognition of these difficulues.b cons.deration cannot be given regulation, as it was then written, AEC. in 1972, instituted the Standard.
Y[
I (segpt as to comments genersIly Contained more detailed Technical Spedfications (STS) Pstgrum, l
,ece,,e4 on or detare this date.
design information than was considered Sets of S'IS were developed for reacter to be necessary to assure safe reactor types designad by each mactae operation.Dese technical manufacture: (Sde latest revisloes g(
-. l 10 CFR Part 50 specifications proved to be difficult to these documente 'see: NUREG.csag, Technical Speccadons W Wedea' organize. unduly restricted flexibility of Rev. 4. Fall 1981; NUREC-0123. Rev. g, J
Power Reactors reactor operation, and neceseitated the Fall 1980; NUREG-0212. Rev.2. Fall 1-processing of many changes that were 19e&. and NUREG-0103. Rev. 4. Fall lego caemer: Nudear Regulatory not significantly related to safety.
for Westinghouse. General Electric.
I Cosiminion.
In December toes the Atomic Energy Combustion Engineering, and Babek Acnosc Proposed rde.
Commission (AEC). predecenor of the and Wilcox. respectively). The STS S
NRC, amended its regulations in provide applicants with model sunsenAny:De Comrmasion is propesing iI 50.36 and 50.50 (33 FR 18812). Section specifications to be used in formulaun to change its regulations pertaining to 50.3e was amended to indude a more plant-specific technical specifications.g technical specifications for nudear precise definition of those categories of Hey have served to make technical power reactors. The proposed chanses technical specifications that must be specifications for facilities licensed would reduce the volume of technical induded in an application for an since 1974 more consistent with one specifications that are made part of an operating license. He amended another, and they have tended to redines cperating license. thereby reducing the regulation narrowed the scope of the the number of disagreementa between number of change requesis which material contained in technical applicants and the NRC staff regarding licensees wedd have to submit to the specifications by definmg five specific items to be induded as technical NRC. The proposed changes,if adopted, categories of technical specifications.
specifications.
h are espected to produce an ne five categories defined for nuclear Improvement in the safety of nuclear reactors are:(1) Safety limits and Current Problem
^8
[
power plants through more efficient use limiting safety system settings. (2)
Disagreements among parties to a cf NRC and licensee resources, hmiting conditions for operation. (3) recent NRC licensing proceeding (In the
$g caTt Comrnent period expiree lune 1.
surveillance requirements. (e) design Matter of Portland General Electric 1982. Comments received after this date features, and (5) administrative controls.
Company, et al. (Trojan Nuclear Plant),
i will be considered ifit is practical to do The design information that was AI.AB-631. 9 NRC 283 (1979]). have so. but asennee of canalderation required to be retained included only highhghted the need to estab!!sh spectSc
-aan be afven except as to comments those items which, if altered. would critaria in the regulations for deciding received on or before thee date.
have a significant effect on safety.
which items derived from the safety Acoarseese Interested persens are Amendments to l 250, among other analysis report must be included in the invited to ru$ smit written comments and things, darified requirements for technical specifications incorporated in 7
suggestkms to the Secretary of the keeping records of design changes and the license for a facility.
Comsdesian. UA Nudear Regu! story pefined more adequately the term In addition. the substantial growth in
'I Commission. Washington. DC 30655 unnviewed safety question." ne latter both the number of items and in the Artentiese Dodeting and Service change establiebed criteria for allowing detail of the requirements contained in Branch.
licensees to make certain kinds of technical specifications that has taken Poa rupmeen meronasAnces costrAcTt changes, tests, and experiments (Le, place since the STS were instituted Mr. D. Skovbolt. Offica of Nuclear those not involving an unreviewed indicates that more precise definitions Reactor Regulation. lJA Nudeer safety question or a change to technical of the existing categories of technical Regulatory Commission. Washington.
speciScations) without prior NRC specifications contained in 150.36 an
(
3 DC 20555 (301/492-M46).
approval. Dese amendments to iI 50.36 needed.ne Commission is concerned t
muv iseronesAnoot Each and 50.50 (1) eliminated detailed design that the increased volume of technical license for operation of a nuclear power information from technical specifications lessens the likelihood that 3
reactor keued by NRC contains specifications, which in turn reduced the licensees will focus attention on matters trchnical speci$ cations which set forth need for a large number dchange of more immediate importance to safe thi specific characteristics of the facility requests, and (2) resulted in a system of operation of the facility.
m cod the conditions for its operation that technical specifications and regulations While each of the requirements in
< b cn required to provide adequate that more effectively directed the todsy's technical specifications plays a protection to the health and safety of the s uention d both licensees
- management role in protecting public health and public. Technical specifications cannot and the NRC to matters important to safety, some requirements have greater j
be changed by licensees without prior
- safety, immediate importance than others in NRC approval As knowledge in the field of reactor that they relate more directly to facility Bedsmed safety increased, the level of complexity operation. %ese are the requirements and detail in technical specificauons that pertain to items which the facility A'
BeforeUs I 50.36." Technical also increased, and a divergence in operator must be aware of and must 3
Specifications.' of the Commission's content of technical specifloations from control to operate the facility in a safe M
regulation to CMt Part 50 requind one facility to another began to emerte.
manner. To a large extent, the relative Af technical specifications to indude In addition. an increasing diversity of f
"those significant design features.
opiafon between app!! cants and the Q.' cope
==r tw obiained from ilm DMadam af opereting proceduns, and operating NRC staff, as to what should be Twank=i tararmenas and Dommwat Ca**4 I,J[
7.
limitations which (were) considered included as technical specifications.
s= a.s=1.uny canmawwn. we
, f,j t
j June 29.1964 (reset)
'.m*A s
50.PP,-40
- v.
PART 50 o PROPOSED RULE MAKING I
importance of these requirements, as changes to the regulations.He major are ofloeg4erm "-- - ----- to
(
distinguished from those related to long-features of these changes and the De Bfth type defines the term effects or concerns, may have been principles upon which they are based daracteristice of the plant and ofte that~
I diminished by the incnase in the total are discussed below.
are not expected to change at au unless 3
volume of technical specification the licensee decides to alter to plant I
i requiressenta.
L Ceneral Priadples design in some way. Des.
ase met
~
Moreover, the increased volume and in reviewing the safety analysis report of concers to the de -to< lay t
detail of wh*=1 specincations and the for a fedlity, b staff reviewe the
?Pweden, but am lame 4mm t
ruultant inmm in the number of methods of analyses, the underlying 4
l
i' N six 4 Pe.e eve proposed change requests that must be assumptions, and the results and cutmia, can be evided into two F ~--
hm inmand the conclusions of the analyses to determine if the P ant has been designed so as not subtypes: m pennining 2 sW staEing Paperwork burden for both licensees l
and reponelbludm. &e oew and the NRC staff. Die is because to Present undue risk to the health and to management overview and control g saas requiru eat wehabt safety d es pubhc. Sm d he plant changes and operations &
sped 8 cations be included in each opereting license; thus, any proposed analyses are quandfative in natum, fwen is more imporunt h b change, regardlae of its importance to whue othm are are quahtadvs. but an immediate operation of the plant, while
~~
safety. must be processed se e license of the analyses rely on underlying
, the latter le more important over the amendment. For changes involving sumnpdons.While many of these long tena.
matters oflesser importance to safety.
mumptions are explidt, such as those Plant functions also can be segregated the processing of a license amendment pertaining to plant operating mode.
in a sindar way.Noe that are with the aseodated incnmd system lineups, or speciSc parametere, considered of immediate importance to paperwork has had no significant otliere an more impudt such as those safety are those aseodated with:
beneSt with regard to protecting the associated with the degradation of
- 1. Protecting the integrity of fission public health and safety.
equipment over the life of the plant or product barriere:
the management control over plant
- 2. Controlling reactivl Proposed Solution operation and maintenance.N
- 3. Cooling the fuel:
fundamental purpose of technical
- 4. Limiting the release of radioactive As a first step in attempting to resolve specifications is to define and preserve Seele products foDowing an acx$ dent.
the difBculdes associated with the those underlying assumpdoes that are
%ne functions med be mda the curmat syskm M kchnical expected to, or could, vary with time or constant cognizance and contro! of the P ant opmmr to mm safe plant l
speciScations for nuclear power drcumstances, throughout the life of the reactore, the Commiselon published an plant, and thus to preserve the validity operation. Other funcdons, such as Advance Notace of Proposed of the safety analysis.& speciBcatior.s em awociated wie se midgetion d Rulemaking (ANPR) on July a.1980(45 addre a vasty M d"' ' _
the effects of netural or man-made FR 45018), requesting commente on 'he phenomena (fires, floods, earthquakes, desirebility of changing its regulations assumptions but are of stm 3eamral etc.), serve to support these four 8
on technical specifications to: (1) types:
functions, bot are not1 subject to the level Establish a standard for deciding which
- 1. Velase Qwess variables that of operator control associated with the must be kept within certain bounds.
four function hated above.
iterns derived from the safety analysis eb 8
of Sca eh e
me,, e. -.
f I
ca a ad ty:(2) ta e,ekm dm,
,, am., for.-
..g d
.,_ e., m.t b. m_
'd technical specifications to focus em
- 4. r nn,lition (or quality) of equipment kh'usho Ia directly on the aspects of reactor and structures that must be maintained.
- th' I "*** O' *ff'""'
operation that are important to the
- 5. Physical characteristka of the plant I" '
P "* "** * *d ' " "
protection of the health and oefety of the and site that must remain fixad and vari us p ant functions.
public:(3) define a new category of
- 6. Adminletrative controls (e,s shift requiremer.ts that would be of leuer erafRng. myiew and audit) that must be IL Overview of Proposed New System immediate importance to safety than maintained.
of Speca6 cations technical specificatione. thereby De first three types or speci8 cations
%e proposed changes to 10 CFR Part
~
providing greater Dexibility to both the can be t of as definine the 50 would establish a new system of NRC and licensees in processing
" bounds'* normal plant operetion specifications divided into two general proposed changes: and (4) establish within which the conclusions of ee categories.De categories, which are appropriate conditions that must be met safety analysis mport am expected 2 discussed in further detailin succeeding by licensees to make changes to the remain vaud.nese first thme types meties. are:
mquiremente in the new category m!ak dimetly to se opere mde of
- 1. Technical Spectfications. and without prior NRC approval.
eyp an an
- 2. Supplemental Specifications.
be db
=
Proposed specification in both Comments received in response to the
,p g, g mg,gg,,g, p,,,
g ANPR wem strongly in favor of a rule categories would be included in an app ication for a license (in Chapter 16 l
I change to incorporate these concepts.
N foure type ochcerns es condidon of the FSAR). and would be reviewed Copies of the comments received from
,, q,,gggy og og,,,,,,,g,,ct,,,,
p and approved by the staff, but only individuals. and a summary of all the and is expect to change slowly if at
- I" O' **"87 N "*h"I**I comments received, am ave 11able for all, over en ext==,8==8 period of tiene.
specifications would be made daractly inspectico and copying at the Hus, generall. specifications of this i
Commiselon's Public Document Room.
are not oftamediate conosen to the Part d ee opunting license. As in se H Street NW., Washington, DC
.to-day opersuon of the plant. but P,,'k y
e hnical e ciBca no ne NRC staff has considered these assmo.Sve tryse a.,re ensemanBr es** 88tud l' Spedfication in the supplemental
'ne lhel wg g,, hahmi'ai category would be documented in the comments and has developed proposed spednesuams ear %sser pesar einem-FSAR and linked to the licensing I
50-Ph-41 June 29,1984 (reset)
PART 50 o PROPOSED RULE MAKING document through a new bcense and would consist of five subcategottes, condition can be met." For KOs in the cor stion in i 50.54. ne licensee would each of which is discussed separately current system that have a lower level be ellowed to make changes to the below.
nf importance to safety (i.e, am not supplemental specificatjons within (i) Safety hmits-Safety limits would essential to the functions listed above),
cert:in bounds and under prescribed be defined (see propsoed
" remedial action" has often been only a conditions, without obtaining prior NRC l 50.36(d)(1)(i)); the same way they are reportmg requirement. With the approval. As further discussed in currently defined in the present system proposed system only tarkniemi section III.B below, these changes would of technical specifications (tn specificaticas with hn==& ate be required to be reported along with a i 50.36(c)(1)); i.e, they would be limits importance to safety won!d be included safzty evaluation for each. ne NRC on important process variables needed as O!Is: thus, specific ac* ion to chauge would review these changes and to protect the integrity of finalon product the facility operating mode willbe supporting documentation in the same barriers. Lf a safety limit is exceeded.
required when an 014 is not met.
manner as design changes, tests, and the licensee must; shut down the plant (iv) Check andlest regmrensants Gs.periments are currently reviewed nottfy NRC; review the matter to (CTRJ. Check and test requirements undir to CFR 50.50. Any change made deternune appropnste actions to would be a newly defined subcategory by a bcensee that was not adequately preclude recurrence: and restart the similar to the survait1==re justified could be quickly revoked upon plar t only aher authorization is received requirements m l 50.38(c)(3), but
- writtro notification by the apprvpriate from NRC.
would nodaclude requirements relating NRC Regional Administrator. A change (ii) Umitistg safety sietem settings to " ' ' ' calibration or inspection to to i 50.54 would be made to require that (LSSS/. ISSS would also be defined (see assure the necenary quality of systema supplemental specifications documented proposed i 50.3e(d)(1)(W)) the same way and components is maintained * * '"
in thi PSAR must be adhered to by the they am cumntly defined in hoe calibratme or inspection bcansee. Unde this system. the licensee 15ase(c)(1). that is, as settmas foe regarements would be included in the has some latitude to rearme the automatic protective devices associated definition of facility monitoring supplementa specifloations, whica beve with variables having signifiant safety provisions which are discussed below e traser degree of safety sigrufirmnce functions. %ese settings are to be under supplemental specifications.
than tedmical specifica tions. Prior NRC chosen so that the automatic protective Check and test requiremente would be cpproval will not be necessary, and the devices =ch will correct an ahmormal defined (eee praposed i 50.36(d)(1)(iv) reduced workload for the Ikeusee and otustion before a safety li=!t is as those penodic checks and tes's thz staff will permit grester exceeded, time pmeerving an ensantial needed to assure that operation will be concentration on more significant
- I*'7 within the safety limits and that the ISS mrttzrs.
.f an summade pmtsettve dmce does and OLCs are met. These are the checks nis new system of specifications o t hincdon pmperty, the bcensw most and tests that are generally performed taka would be put in effect for new operating ka by plant operators during. or just prior ah lic2nses issued too days after the to, operation to determme tf procesa effective date of the amended rule.
,g, variable are within acceptable bounds.
There would be no backfit requirement p
g, gp and if systems and equ;pment are an the for existing operating fadhties, though f
donollimier andconditions camct operating state or standby this cculd be done if a bcensee were to "9"**'*
(OLC/. Olfs would be a newly defined status. As with the current IfOs and subcategory, =f =itar to the existing surveillance mquimments the technical III. Major Features of Proposed Rule
" limiting conditions for operation" 8pecifications would be struct.tred to A. Changes to Definitions of Technical (IIOs) defined in i 50.3e(c)(2), with attain a one. tune correspondence Specification Cotesones some important differences. OLCe between these checks and tests and would be defined (see proposed each OLO, to assure that operators keep
- 1. The term technical specifications i 50.3e(d)(1)(iB)) as limits on important abreast of plant status with respect to would become a category of process variables and on conditions each OLC.
specifications and would cor.sist of relating to the operating state and (v) Operationalstaffing andreporting
. operational speciScstions and principal standby status c4 systems and requirements (OSR). OSR would be a design feature spectfications.
components nese limits and conditions newly defined subcategory which would
- e. OperutionolSpeerficutions.
are associated with the perfortnance of contain a subset of the requirments Operational specifications would be the functions of controlling reactivity, currently contained in Administrative d1 fined (see proposed i 50.3e(d)(1)) as coohng the fuel, and limiting the release Controls as defined in 150. 36(c)(5).
those specification imposed upon of redioactive fission products following OSR would be defined (see proposed ficiltiy operation that are necessary to an accident. De 01r eubcategory is
$50.30(d)(1)(v)) as those items relating to assure that the facility is operated narrower in scope than the axisting 140 shift cmw composition and within limits and under conditions that definition, since ILOe include items responsibility, as we'l as reportmg. that cre consistent with the assumptions to addressing virtually any equipment ir, are necessary to sosttre operation in a the sWty analyris report regardmg the the t " required for safe operation of safe manner.Rese items are valus of process variables and the the
'ty "
considered to be ofimmediate short-opersting state and standby status of When an 01E is not met, the 1"*"""
term importance to safety and thus are systems and components assodated must: Shut down the plant or take included in operadonal technical with the four safety functions identifled specified remedial action to place the in Section I above. His definition would facility in a safe condition rati! the 014 specif cations and would be part of the cet:blish a framework for decidag can be met; notify NRO Ld review the openting license.
which ite no derived from the safety matter to determine appropriate actions b.PrznefpalDesign Feature cnslysis report are to be incladed to the to preclude rec.arrence. He preeest 140 Spec $ codons. Prindpal de@ festum opera tional sp*cdf! cations. Operational definition includes plant shatdown as a sPocifications would conatst of those specifications must be written for all required action, but it also allows (for f,tems that am cumatly categorized as n:rmal modes of fadlity operstlos matters of lesser leportance)"any design features in the present system including shutdown and refueling ney remedial action permitted by the "II'4"l specifications, as defined in wruld be part of the opereting license tarharal sPedfications until the the existing 8 50.36(c)(4). Dey would be l
defined (see proposed i 50.30(d)(2)) in l
June 29.1934 fr" 50.PR-42
PART 50 o PROPOSED RULE MAKING essentially the same way; that is, those administrative in nature, such as instance, administrative agations.
, nysical characteristics of the facihty reporting to NRC instituting fire patrols, rather than 10 CFR 30.55 that I
wtuch. if altered. woud have a etc and review the matter to prMude proposed changes to tians have 1
sisruScant effect on safety and are not recurrence. Plant shutdown would not to be nuessed
. % ssefore.
)
included in cther categories of technical be required because items in this administrative spedScatione benema speciScations.' Principal design featum subcategory are not considered to be of esPeciaUy impor: ant if. as intended. the l
speciScations identify the physical irr. mediate importance to safety, pmposed rule shifts more controllrum characteristics of ti.a plant and site that However, documentation oflicensee NRC to licensees, eBowleg them to I
may not be changed without prior NRC review of these matters would be make ch before NRC esetow.he approvat hoe specifications are not required to be made available on a NRCinte to carefully moedler considered ofimmediate importance to regular basis for NRC audit.
changes to administrative spedScottons i
safety and therefore are not included in
- b. Monitoring pmvisions. Monitoring to ensure that essential adelaistrative the Appendix A operational technical Provisions would be defined (see controls are rigorously maintained.
l specifications;however,because of their proped i 50.3e(eX2)) as provisions a C%er to Wemerrsd great importance to assurance of the relating to monitoring. inspection.
Sp,,jffg, vslidity oi Le acddent analysis, they testing. and cabbration needed to
~
will be made part of the facility provide long-term assurance that th, ne licensee would be permitted to operstmg license.
necessary quality of systems, make changes to supplemental components, and structures important to specifications (see proposed
- 2. Supplemento/ Specific =rtions.
safety is maintained. He monitoring I 50.38(f)(1)) without prior NRC Supplemental specifications would provisions would assure that the FSAR approval. provided the changes do not consist of those items needed to assumptions regarding the condition of involve a conflict with the technical preserve safety analysis assumptions equipment and structures would remain specifications incorporated in the regarding important safety functions not valid over the hfe of the plant.%e license or do not result in a decease in included in operational technical monitoring provisions would contain their effectiveness as explicidy defined specifications, assure that the newesary those survetuance requirements that are for each type of provision feee proposed
... m b ts, and perfortned at relatively long intervals 6 50.36f f) (2). (3). and (4)). & rule
[n
.s jg -... '
. assure and are directed toward determtrung the contains tests to be applied for each erview and
' " "b8h state of quali condition of type of supplemental specification.
d fac s
n equipment. A portion of these hoe tests are consulered by the O y
- A8-y'pe jn would perfed by technidens rathee than by whether there is a deavase in the inspections, etc., are generally Commission to be appropriate to judge not made irec9y Pus ant is in a effectiveness of a revised specification.
operating license, but would be Indirectly linked to the licensing When the perfonnance of' inspections
"" "* I supP smental specification only after it docurnent and be enforceable bY or tests requimd by a moaltortug regulations prescribed in i 50.54 (see provision reveals a defect, the boensee
'"""""I' proposed i 50.54(x)). Enforceability is would be required to declare the system,
& liansee would be required to separately discussed in subsection C component. or structure to be inoperable Mview PmPowd change to below.ne licensee would be allowed and take the action appropriate for that supplemental spedfications in the same to make changes to supplemental system. component or structure as marm as is curmntly required fw specifications, withh certain bounds s'ipulated in the spedScation. his is no proposed design changes by to CFR and under prescribed conditions, different from what would be required 5m And as with I so.se design without prior NRC approval Dese by the current system of technical changes, the revised i SCL36 would bounds and conditions an discussed in specifications if the perfor==nar of a require (see proposed i 30.3e(f)(5)) the subsection B below. Supplemental surveillance requiNment revealed a licensee to maintain reoards of changes specifications would consist of three defect.
made to the supplemental specifications subcategories, each of which is
- c. Administmtive provisions.
which would be available for NRC discussed separat,ely below:
Admirustrative provisions would be audit. including a written safety
- s. Contro/pmvisions. Control defined (see proposed i 50Je(e)(3)) as evaluation which provides the basis for provisions would be defined (see provisions rotating to organisation, deter =Inine that the change does not pmPosed i 50.36(e)(1)) as provisions recordkeeping, review and audit, and involve a decrease in effectiveness of relating to the contml of variabin and reporting necessary to aneure effective the provision. and to report all these the operating state and standby status management overview and control of changes to NRC.his would be done on of systems and components aseodated facility changes and operaticas.no a prompt basis. NRC review of these with important safety functions not included in operational technical administrative brevisions would contain changes would be conducted in a those items in t current systeen manner simDar to that of reviews of to l
8Pecifications. Examples of these de ignat'ed as admmistrative controls.
CF1t 50.50 changea; i.e., NRC will, by functions are the mitigation of the as defined in i 50.3e(c)(S), that are not audit. ascertain that the licensee has
' e to at I r man-made included in the operational technical exercised responsible and pruder t pheno na (firn. Doods, earthquakes-spec 18ca6ans as opwenood staEing WM ud h uf4 sh etc4 e subcategory would include and reporting requirmnents.Deee items made by the licensee are -aiaraat with g
syYes
, Tecti and
(*,*a H
~
pPrwing a
the t
av are not euential to the four safetF control. but are not considered to be of In addition to the staF review of functions discussed earlier.& control immediate in to the safe provisions would include requirements opaea6on of reports to assess licensee performance.
for periodic checks and tests to assure Administrative =pacinaations are the monitoring of plant operations by l
the provisions are being met.
essential to the entire ocatrol scheme of the NftC resident laspector will include When a control provision le not met, changes to technical spedficanons, consideration of changa to j
the licensee must take appropriate becaum they govern how other supplemental specifications. Since the action. which generally would be spectAcanons can be changed. For licensee is rsquired to document its 50-P's C June 29,1984 (reset)
PART 50 o PROPOSED RULE MAKING basis for the change to the supplemental and (g)). Technical specifications laaued recesignated paragraph (g), and new I
specificadon prior to effecting it the before that date would not be requind paragraphs (d), (e), and (f) are added to I
documentadon is available for the to be changed; however, upon request read as follows:
re:idInt inspector's review whenever by a licensee to convert the existmg I so.3s specencetnene.
C '
de: ired.
(a) Each applicant for a license se pe te orma the NRC Changes to supplemental a&
ope de d a pduh w l
with the technical specificadons f. Pa# %
Ad StaM*
utilization facility shah include in Ita 9
specifications that involve a conflict t
application proposed technical incorporated in the license or a decrease As required by Pub. I,96-511. this specifications in accordance witii the in thstr effectf veness would require NRC proposed rule wiH be submitted to the requfrements of this section.He cpproval prior to their implementation OfEco of Management and Badget for techrdcal specifications must be derived (see proposed i 50.36(f)(6)).
clearanca of the reportmg and from the analyses and evaluation
"&WMM"
$d W h & Wh W C EnforcoobshtyofSpecifications and amendments thereto, submitted L'nder the existing system of technical V. Regulatory Flaxibdity Ad Statement under l 50.34. Rese specificadons are
,pecincations presenbed by ( 50.36. the In accordance with the Regulatory described in paragraphs (c). (d), and (e) enforceability of specifications is Flexibility Act of 19eo. 5 U.S.C. 604(b),
of this section. A summary statement of assured by making them a part of the the Commission hereby certifies that the bases or reasons for the hcense. Under the proposed system this rule wiu not,i! promulgated. have a specifications, other than those covering prescribed by the proposed changes to sigriificant economic impact on a administrative controls or provisions, or to CF7 Part 50. only technical substantial number of small entities.
operational staffing and reporting specifications, as more narrowly This proposed rule affecta only the requirements, must also be included in difined. would be incorporated directly licensing and operation of nuclear the application, but will not become part into the licensing document.
power plants. Re companies that own of the technical specifications.
Supplemental specifications would not these planta do not fall within the scope (b) Each license authorizing operation be pcrt of the four comers of the license of the definition of **small antities" set of a production or utilization facility of a as sah. However, in order to link the forth in the Regulatory Mexibility Act ce type descnbed in il 50.21 or 50.22 of hcense to the supplemental the Small Business Size Standards set this part willinclude technical specificstions. I 50.54. " Conditione of out in reguletions issued by the Small specifications. For a nuclear reactor Lacenses." would be modtfled (see Businese Administration at 13 CHL Part operating license issued before (180 proposed i 50.54(x)) to require licensees
- 21. Since these companies an dominant days after the effective date of this to a'nde by all specifications, including in their service areas. this proposed rule amendment) and for a fuel reprocessing the supplemental spectficabons does not fall within the purview of the plant, the license will inclu le technical documented m the FSAR. as amended Act.
specifications in the categones set forth by changes made, and recorded and Pursuant to the Atomic Energy Act of in paragraph (c) of this section. For a reported in accordance with the 1954, as amended, the Energy nuclear reactor operating license issued proposed i 50.36(f)(5). In addition, a Reorganization Act of1974 as amended, on or after (180 days after the effective prouston would be added to sive and section 553 of T1tle 5 of the United date of tMs amendment) the license will authonty to the appropnete NRC States Code, notice is hereby given that include technical specifications in the Regional Administrator to immediately adoption of the foHowing amendments categories set forth in paragraphs (d (1) revote any change made by a licensee to 10 CFR Part 60 le contemplated.
and (2) of this section. The Commission to supplemental specifications that is may include additional technical ludgid not to have adequate PART $N UCENSING OF specifications as it finds appropriate.
Justification.This delegation oflicensing PRODUCDON AND UTIUZADON (c) Technical specifications for a authonty to the Regional Admirustrator FACIUTIES nuclear reactor operating lic nse issued po ci e by the
- 1. De authority citation for this part before (180 days after the effective date Office of Nuclear Reactor Regulation.
meds as foUows of &is ameneny and for a fud s
p n I clude items in Thr licensee could provide additional A mehortty: Seca.108,104.181.18:.183, tee, gg information if it desires to further justify es S'al saa,337, tea, es3. e64, ese, see, as a chinge but would be required to amended tu UAC 212a, 2134. 2ao1. 2222.
s btsin NRC spproval befor, 2213,2230k seas. 201,202,20s, se Stat 1243, (d) Technical specifications for a tenstituting the change. Additionally, 1244.1246 (42 U.Sf. 5841. 5842,5844L unlana nucleat reactor operating license issued othemse noted. Section Safe also Isaaed chInges to supplemental spedfications on or after (180 days after the effective will be subject to possible enforcement undu sec 122. 88 9st 939 (42 UAC 2152) date of this amendment) willinclude acti:n in accordance with NRC amen $
items in the following categories:
St t.
C 2234)
Enforcement Policy. Chan8es which ar*
Secnons 50.100-50.1021seved under sec. ist, (1) Operationo/ specifications.
Inad quately reviewed, supported or es Stat. sea (4a UAC 22 sol.
Operational specifications are justified or incorrectly implemented may For 'he purpoem of sec. 223. 88 Stat ase, as specifications impond upon facility result in a violation.These violationa amended (42 UAC az73k li 30.10 (el. ibl, operation that are necessary to assure will be evaluated in the same way and (c), so 44, snee, ease, east, and 5asota) that the facility is opereted within limits violations of to CFR 80Je are evehieted are issued under sec.1sth, es Sist sea, as and under conditions that are consistent
~)
la accordance with NRC Enforcement amended (42 UAC 2act(b)k ll ento (bland with h anumpdons in the safety Policy.
(c) and 5o.54 are tsausd ander sec. telt es anal @ W "8ardmg b values of Stat Dee, as==Aad (42 Uit 2201(ill; and D ApplicebilityofProposedule Il so&(el, sase(bl. saro, soft, sa72, and process variables and the operating so rs are toened ander see teto, es Stat. seo, state and standby status of systems and 3
De Proposed rule,if adopted, would as amended (42 UAC 2stn(o)).
components that are associated with the
?
8Pply to nuc. ear plants receiving an performance of the fonctions of
- Pereting lleense on or after a date 180 2.la i saas, the utie is avised.
controlling reactivity, cooling the feel.
[
days after the effecties date of the final Paragraphs (a) and (b) am remed, the protecting the integrity of finalon prodmel -
2 p
'*Ie (*se the Proposed i 50.36 (b). (d),
intadoctory text of pareyspb (c) is bamers, and limiting the release of revised paragraph (d)is revised and y,
l
% 24,1984 truet) 50.PR 44 e
y.
l1
PART 50 o PROPOSED RULE MAKING radioactwe fluion products following (ivl CAsch anddest segeuensenta important to safety is maintained. When en acciderst. Operational specifications Check and test requiremenas are the results of a monitoring provision are to be imposed on sil normal asodes sequaements nieting to periodic chek = activity indicate that the necessary of facihty operation includmg shutdown and tasas to ensure that fassaity quality is lacknas, the beensee shall and refuehns and are to consist of items operation will be within the safety limits declare the system, ea=pa===8 se of the following types:
and that the limiting anfety systems structure to be inoperebie and take (i) Sofety Fatifs. Safety limits are settings and operationallimits and appropriate action as permitted by the lunats upon isoportant proomas variables conditions are ast.
speci6mtions.
which are found to be necusary to (v) Opemtsanalste/fug androportarts (8) Administratrve Proviasons.
reasonably prosect the integrity of requirements. Operational stafing and Administrative provimiens ase certain of the physical barriers which reporting requinments are requirements provisions relating to orgmassation.
guard abainst the uncontrolled release of nlating to shet crew composition and quali6 cations of personnel gm.m.
radioactivity. If any safety limrt to responsibility and reporting that are recordkeeping, review and audit. and exceeded, the reactor must be shut neoemsary to assare operation in e safe reporting necessary to assure effective dewn.h licensee shat! notify the manner management overview and omntrol of Commieston, review the matter, and (2) Principoldesign frenrre facihty changes and operations.
~
record the results of the review, spec 6Teotiortst Princtpal design feeture (f) Changee so supplerienaal meloding the cause of the condition and specifications are sped 6catione relating specifications. (1) A licensee may make the basis for corrective action taken to to those feetures of the fecility, such as changes to supplemental specincetions preclude recurrence. Operation may not meterials of constrodion and geometric without prior Comadesion opproval.
be resumed until authorized by the arrangements, whi&. If aftered or unless the dange involves:(i) A conflict Commission.
modified, would have a signfficant effect with the tedinical spec 16 cations (ii) Limiting safety system settings.
on safety and are not covered by incorporeted in the licever, or (ii) a Limiting safety system settings are technical spedfications required by decrease in the e6ectiveness of the settings for automatic protective devices pareyeph (d)(1) of this section.
provisions of the s+ph.=al related to those variables having (e) Supplemensat specifications. For specification.
signincant safety funcnons. Where a nuclear reactors liwnsed to operate in (2) A change to a control provision is limiting safety system sertme is accordance with techmcal specifications deemed to involve a decease in the specified for a variable on widch a of the tpe desafbed in peregraph (d) of effectiveness of the provie6an:(i)If the safety limit has been placed. the setting this section the final safety analysis controls on variables or on performance must be choecn so that automatic report must also include supplemental levels that define the required opereting protective action will merect the spedfications. S-T -
state or standby status of system and
~'
abnormal situation before a safety limit specifications are specifications relating components are relaxed or (ii)if the is exceeded. If, during operation, the to maartares, oestrol. and frequency of the periodic check or test is autometic safety system does not administration necessary to assure that decreased more than is justified by the function as required. the licensed shall the quality of equipment, the proper history of test results: or (iii)if the take action es stipulated in the operating state and standby status of required action, in the event the specification, whid may indude important support systems, and afectswe provision le not met,is relaxed.
shutting down the reactor: motify the management overview and control of (3) A change to a monitoring provision Comnussion: review the metter, and facility changes and operetions are is deemed to involve a decrease in the record the results of the review, maintained.Sepplemsstal-S" "- 1 effectiveness of the provision:(i)If the including the cause of the osaditica and are to consist ofitems of the following frequency of the monitanns, inspection.
basis for corrective action taken to types:
testing or calibration is deceased preclude recurrenoe (1) Contm/ Provisions. Control without a compensating change in the (in) Opcietionallunits and condrtions, provisions are provisions relating to the soceptance criterion or an increase in Operationallimits and cendrtaons are control of variables and the spesettes the sensitivity or accuracy of the method limits en important process eariables state end steney semens of syneens and used, unless the cumulative history of and conditions relating to the operatag components==aadated with imiportant test results clearly sup a reduction state and standby status of systems and safety functaons not desenhed in in frequency. or (if)if sensitivity or components that are===adated with the paragraph (d)(1) of tble sectnan, such as accuracy of the method used to perform performarww of the functicas of the mitigation of the effects of natural or the monitoring. '--*= testing. or controlling reactivity. moling the fuel.
man.made phenomena. End control calibration is duasseed without a protectag the integnty of fission product provision must inchade penadic decia compensating change in the acceptance barriers, and limiting the release of or tests to assure that the provimien is criterion or increase in the frequency of radioactive fiaaion products following being met. When a control provision is the monitonng. inspection, testing or en accident. When an operational lianit not met, the Boensee shall take calibretion or(iii)if the acceptance or conditinn of a nuclear reactor is not appropriate action as permitted by the criterios for the monitoring. ! ;- -
-O -
met, the bcensee shall shut down the specification. The licensee shad review testing, or calibration is relaxed without reactor or Igliow spedfied remedaal the matter and record the teaults of the a compensating inmense ha he action, as stipalated by the review, including the emuse of the frequency. eensitivity, er somsrecy of the specifications, to place the facility in a condition and basis for corvootive action method used.
safe conditica entil the operetional knit taken to priaeluda reassumon.
(4) A ch7e to an administrative or modstpoo can be met. The licensee (2) Moriiterity P)oensioma Monstering provision is deemed to imselee a shah notfy the f'a==uaalan review the proviasons are, _ _ - __ relatug to deaease in the effecovemass of te matter, and record the resalts of the monitorieg. '-, - -t tending, and
-. "- (i) N the level af sammagement review lacluding the cease of the calibretion needed to provide overview or control is deoensed; or (ii) condition and the basis for conectrve assurgaos that the aseeseary of if the assarsaca of tb t gamhey of action taken to preclude recurrence.
systemus, ar==pa===sa and sensoeuses operations or of persumed in decreased.
50.PR-45 June 29,1984 (reset) l
PART 50 o PROPOSED RULE MAKING entire safety smalysts report es technical on that report and other questions Igg) g b usefulnese of the n!ating to property insurance for wasedkeeping in usessing matters specifications.
nuclear utilities, (2) At the initiethre of the Commluton important to safety la decreased, or(iv) or the licensee. any license may be DAfts: Comment period expires g the method or timeliness of amended to inchsde technirmt September 22.1982. Comments recalved management review of changes to specifications of the scope and content after the expiration date will be
,pedficagone is changed, which would be required if a new cousidered if it is practical to do ao but (5) %e licee889 shall maintain records of changes in the supplemental license were being leeued.
assurance of consideration cannot be spec 1Scations made under this section.
- 3. In l 50.54, a new paragraph (x) is, given except as to comments filed as Rese records must include a written added to road as follows.
before that date.
safety evaluation which provides the i 50.54 m eHoonosa, an-sane. Written comments should a
i be submitted to the Office of the i
buis for the determination that the change does not involve a decease in (x)%e licanaes aball maintain and Secetary of the Commission, U.S.
I the effectivenen of the provisions.ne operite the facility in accordance with Nudear Regulatory Commisalon.
j.
records must also include an indication the spedfications provided in i 50.36 of Washington. DC 20555. Copies of the
{
cf review and approval by the licensee's this part. Changes to the specificationa comments may be examined at the NRC oosite safety review eqanization.%ese may be made only with price Public Document Room. mr H Street.
ncords must be available for inspection Co-tantos approval or as prescnbed NW., Washington DC 20655.
ct the facility before implementation of in i 50e(flof tais part. A change made Copies of the report entitled " Nuclear the change. Within three days of by the hcensee under i 50.3e(f) of this Property Insurance: Status and cpproval of a change to a supplemental part must be revoked immediately upon Outlook." NUREC-4891. may be specification. the lionesee shall furnish written notification by the appropriate ottained ander the NRC/GPO Sales ta the apt.opriate NRC Regional NRC RegionalMminiawator that the p,,.pam at a cost of $6 00 by writing to Administrator shown in Appendix D of Part 20 of this chapter (with a copy to justificatioaaprovided for the change is the Director. Division of Technical the Director of Nudear Reactor inadequate. When this notification is Information and Document Control. U.S' Regulation.U.S. Nudear Regulatory recetved by the licensee, the change Nuclear Regulatory Commission Commission. Washington.DC 20655) a must not be implemented.ce must be hhington.DC 20555.Ris report is report containing a brief description of revoked if already implemented, and abo asa:!able for inspection and
~
each change. including a copy of the may not be reinstated without prior copying at the NRC Public Document safety evaluation. Any report submitted
%Manton approval Founi.1717 H Street.NW Washington.
by a licensee under this paragraph will Dated at Washingtan. D.C this 2sth day of DC.
be made part of the public record. %e Mard lear-Fon rummen mronesation contact records of chanan made to the For the Nudeer Ragulatory Comniassion.
Mr. Robert S. Wood. Office of State supplemental specificabons must be at il GE.
Programs. U.S. Nudear Regulatory maintained for a period of at least 5 3,,,,,p %. ies Commission. Washington. DC 20555.
se
( A licensee who wants a change to 47 FR 27371 "N'
the supplemental specifications that Pubt.shed 6/24/s2 the NRC published a report." Nuclear invohes (i) a confhet with the technical property insurance: Status and Outlook" specifkations incorporated in the comment period exp res 9/22/s2.
INUREG-0891). by Dr. lohn D. Iong.
Lcense: or (U) a decease in the comments received after the Chairperson and Professor ofInsurance effectiveness of the provisions of the expirat.on date wist be cons.dered at Indi na Unnersity. This report was specification, shall submit a proposed if it is practicas to do ao bu, written as an outgrowth of the Hrve change, along with the basis and ensurance of consideration cannot Mdc Island-2 accident after it became justification for the proposed change, for be given except es to comments apparent that nuclear utilities may need fd*d on or before that dare.
approval by the Commission prior to rnore property insurance than has been implementing the proposed change.
avsblAe e NRC staH asked Mang (7) A peoposed change to the 10 CFR Part 50 to w nte the report. in part. to answer six supplemental specifications that 4""" *"" "' I* U * * ';
involves (1) a conflict with the technical ggendet Pr h
- 1. What has been the development of specifications incorporated in the y,
each principal source of nuclear license, or (ii) a decrease in the Aenect: Nuclear Regulatory property insurance used as of early 1982 effectiveness of the provisions of the Commission.
by nuclear utilities in the United States?
specification, or (iii) an unreviewed
- 2. What are some of the disHap%
safety question. shall ge treated as a Actiose Advance notice of proposed features of nuclear property lasurance proposed change in the facihty or rulemaking.
as offered by the principal sources?
procedures, as described in the safety aussesamv: On March 31,1982 the 3 How much nudear property analysis report, requiring an amendment Nudear Regulatory Commission (NRC) insurance was offered by each of these to the license. A licensee who desires published in the Tederal Register (47 FR sources as of lanuary 1.19827 such a che shall submit an 13750) an laterim final rule mquiring
- 4. Assuming that present plans come opplication amendment to its license utility licensees to purchase on-site to fruition how much nuclear property pursuant to l SILee, PfDPerty insurance to be used for insurance is likely to be offered by and (g)(1) Die section does not modtfy the decontamination expenses arising from of these sources as of January 1.19B3.
j technical speciScotions included in any an socident. He NRC subsequently
- 5. What, if any, principal sources of effective date of this amendment). A Published a report on property insurance nudear property insurance are Mkely to
- .l ticense issued before (180 days aftes the license which does not contain technk al PrePand by Dr. John D. Long. Professor emerge in the pnvste sector by January i
- -
ofInsurance at Indiana Univeralty L 1983?
6 What problems serious enough to specifications la deemed to include the (NUREG-Oert).nia advance notice of wareant action of the U.S. Nuclear J-y*..
proposed rulemaking requests comments
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June 29.1984 (reset) 50-PR-46 g
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5 September 17, 1985
(
,/
SECY-85-306 RULEMAKING ISSUE (Notation Vote)
For:
The Comissioners From:
William J. Dircks Executive Director for Operations
Subject:
STAFF RECOMMENDATIONS REGARDING THE IMPLEMENTATION OF APPENDIX R TO 10 CFR 50
Purpose:
To provide the Comission with recomendations for expediting and improving the implementation of Appendix R (Fire Protec-tion), and to provide a discussion of related issues raised by a member of OGC and by several fire protection staff members.
I The initial inspections for compliance with the safe shutdown
Background:
provisions of Appendix R to 10 CFR Part 50, at facilities where compliance with Appendix R was believed to be complete by the licensee, revealed a number of instances of noncom-pliance and enforcement actions were being proposed.
In late l
l 1983, the ED0 initiated a consultation between headquarters and the regions which concluded that uncertainty still existed over the staff requirements and suggested that a series of regional workshops be organized to discuss fire protection and answer questions from the licensees. The Nuclear Utility Fire Protection Group (NUFPG) organized a meeting to precede these workshops to solicit industry experience in fire protec-tion programs and compile for the NRC questions which they felt needed answering.
The NUFPG also initiated discussions with the staff on whether a licensee may propose an alternative means of complying with l
certain Appendix R issues covered by Generic Letter 83-33, if backed by a technical justification.
In such cases, an j
exemotion would not be needed if an alternative means of
Contact:
W. Johnston Ext. 27331 SECY NOTE:
This subject is scheduled for discussion at an Open Meeting at 10:00 a.m.,
Thursday, October 3, 1985.
An advanced copy of this paper was provided Commissioner Offices on September 17.
l The Comissioners compliance is provided in areas where Appendix R is non-specific. The result of these discussions was a document called " Interpretations of Appendix R."*
The workshops were held in the spring of 1984. At these workshops a package of NRC guidance was distributed to each attendee which included NRC staff responses to industry questions and the " Interpretations" document. The cover memo for the package explained that it was a draft package which would be issued in final fom via a generic letter following the workshops.
During the same timeframe, the staff began to develop enforce-ment guidance on Appendix R.
Draft guidance was developed in the spring of 1984 and it was detemined that, until the guid-ance and the interpretations of the rule were completed, enforcement actions based upon Appendix R should be placed on hold.
Four cases arising in late 1983, early 1984 and two more rec,ent cases were and still are in this status.
On May 30, 1984, the staff briefed the Commission on the status of the implementation of Appendix R and discussed the issues raised by several NRC fire protection engineers in a Differing Professional Opinion (DPO) and by a member of OGC. At this meeting, the Comission requested that the implementation guidance be submitted for Comission approval prior to its issuance to industry.
Following the Comission meeting, NRR was requested by the EDO to form a Steering Committee on Fire Protection Policy to examine all current licensing, inspection and technical issues, and to develop policy recomendations aimed at expediting Appendix R compliance for pre-1979 plants. The Steering Comittee drew its membership from three of the Regions, IE, and ELD.
Richard Vollmer, then of NRR, was Chairman. Follow-ing a series of meetings, which included discussions with all j
agency fire protection engineers, the Steering Comittee issued a report to the EDO (Enclosure 1) in October 1984 containing a number of recomendations for improving the consistency of the levels of fire protection safety and for enhancing the implementation of Appendix R.
Following coment by the Program Offices and the regions, the EDO directed in December 1984 that the Steering Comittee Report be issued for public comment.
Coments were received from the NUFPG, utilities, and consultants. On May 3, 1985, the Fire Protection Steering Comittee issued updated recommendations which reflect resolution of these coments (Enclosure 2).
- This document was later disputed by a member of OGC, and several staff members initiated a differing professional
I 5
. The Comissioners Current Status Although Appendix R became an effective rule in 1981, approximately half of the operating reactors still ties are not currently scheduled for compliance until the This situation, as discussed with the Commis-late 1980s.
sion in May 1984, has arisen partially because of the com-plexity of achieving compliance with the Rule for because industry did not manage its approac Comission-required backfitting of certain portions of Appendix R.
guidance, which has been generated in respon Some The backlog of licensing review extends well into 1986.
of this could be given back for licensee action, b guidance.
work may also be generated.
As discussed previously, the development of this guidance an regional meetings in 1984 resulted from early inspectionsThis sit showing noncompliance with Appendix R. Based on a larger sample of changed somewhat for the better.
ten pre-1979 and twenty post-1979 safe sh of the analyses needed to establish required post-fire shutdown capabilities.
a significant reduction in the number of deficiencies observ Some of this is attributable to a better understanding of the requirements as discussed in the regional workshops.
In the past year, we have been inspecting NTOL plants prior to issuing an operating license specifically for the safe workshops, the applicant usually makes items of Appendix R.
l The staff then uses the time before or during the inspection to come to resolution on these deviati the staff guidelines.
The increased understanding since the workshops has reduce significantly the open items to be resolved at t j
licensing.
substantial compliance with regulatory requirements.
3 The staff proposes to adopt, in large part, the positions recomended by the Fire Protection Policy Steering Comittee Discussion:
for the guidance of the licensees in the implementation of Appendix R.
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The Commissioners The Steering Committee's principal recommendations are as follows:
1.
Promptly issue a generic letter (Enclosure 3) infonaing all licensees that:
a.
Extensions to 50.48c schedules will no longer be granted except where a licensee has an approved "living schedule."
~
b.
Inspections will include plants with uncompleted modifications as well as plants with completed modifications.
In addition, licensees may request an inspection to help avoid costly design and implementa-I tion decisions with which the staff may disagree.
c.
Documentation of valid analysis supporting fire protection features must be available for inspection.
d.
Quality arsurance measures should be in effect to assure that fire protection systems will function as intended.
e.
Licensees are required to notify the NRC when deficiencies are discovered pursuant to 10 CFR 50.72 and 50.73, f.
A standard fire protection license condition is available which will clarify the licensee's options in making modifications to the fire protection program.
i 2.
Accompany the generic letter with the following guidance i
documents:
o Interpretations of Appendix R (Enclosure 4).*
o Appendix R Questions and Answers (Enclosure 5).
o Fire Protection License Condition (Enclosure 6).
3.
Adoption of proposed Guidance For Enforcement Actions Concerning Fire Protection Requirements (Enclosure 7).
Further consideration of the positions taken by the Steering Committee regarding extensions to 50.48(c) schedules and a proposed standard fire protection license condition have raised policy and legal issues which are worthy of the Commission's attention.
- These interpretations represent staff positions, and should not be considered as official agency interpretations issued by the General Counsel.
See 10 CFR 1.32; 10 CFR Part 8.
{.
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'd The Connissioners Schedular Exemptions The Steering Connittee's position is that the Appendix R I
implementation schedule was established by the Connission in 10 CFR 50.48(c), promulgated together with Appendix R in November of 1980. Allowing time to evaluate the need for alternative or dedicated shutdown systems, which require prior NRC approval before installation, and time for design of and NRC review of.such systems, the Commission envisioned that implementation of Appendix R would be complete in four to five years, or approximately by the end of 1985. The schedule for many fire protection modifications was initially tolled by the filing of exemption requests under 50.48(c)(6).
This delay was anticipated by the rule.
In addition, the staff has issued numerous exemptions to 50.48(c) deadlines pursuant to 10 CFR 50.12. The latest completion date approved so far is in 1987 Some licensees have proceeded expedi-tiously to implement Appendix R and are now finished or nearly finished with that effort. Others have engaged in lengthy negotiations with the staff while continuing to file requests for schedula extensions, and thereby have barely begun Appendix R modifications needed to comply with Sections III.G and III.L. Schedule extension requests have been received seeking implementation dates of 1990 and beyond.
As the 50.48(c) schedule was intended to be a one-time schedule in the 1985 time frame and ending(particularly connencing in the 1980-1982 time frame, extensions well beyond this schedule where major modifications remain to be completed) underinine the purpose of the schedule which was to achieve expeditious compliance with NRC fire protection requirements. The staff believes that further wholesale extension of fire protection schedules via exemption from 50.48(c) requirements is legally unsound and could expose the Connission to a court challenge.
On the other hand, there are a number of considerations favor-ing granting additional schedular exemptions where a genuine i
hardship can be established. Recent PRA's have shown that thecongributiogoffirestocoremeltprobabilitiesisin the 10- to 10- range for reactors with no alternate shut-down capability, and about a factor of 20 improvement if fully protected. Recent inspections are showing greatly improved compliance, a large percentage of the necessary changes have already been completed, and defense in depth provides knowledge that individual deficiencies do not cause significant safety effects. As a consequence of these con-siderations, the generic letter has been modified to state that additional schedular exemptions may be requested under 10 CFR 50.12, but such requests will be granted sparingly based on an evaluation using the following criteria:
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The Commissioners 6-1.
The utility has, since the promulgation of Appendix R in 1980, proceeded expeditiously to meet the Connis-sion's requirements.
2.
The delay is caused by circumstances beyond the utility's control.
3.
The proposed schedule for complet. ion represents a best effort under the circumstances.
4.
Adequate interim compensatory measures will be taken until compliance is achieved.
License Conditions Most licenses contain a fire protection license condition.
License conditions for plants licensed prior to January 1, 1979, contain a condition requiring implementation of modifications connitted to by the licensee as a result of the reviews against the Branch Technical Position. These license conditions were added by amendments issued between 1977 and February 17, 1981, the effective date of 10 CFR 50.48 and Appendix R.
In April 1981, i
the Connission decided that, until a fire protection rule for future plants was approved, "new licenses (those issued after January 1, 1979) should contain a condition requiring compli-ance with connitments made by an applicant and agreed to by the staff."
Because of the evolving nature of the approach to fire pro-tection, license conditions for plants licensed after Janaury 1, 1979, vary in scope and content.
Some only list open items that must be resolved by a specified date or event, such as exceeding five percenc power or the first refueling outage.
Some reference a connitment to meet Appendix R; some reference the Final Safety Analysis Report (FSAR) and/or the NRC staff's SER. These variations have created problems for licensees and for NRC inspectors in identifying the operative and enforceable fire protection requirements at each facility.
These license conditions also create difficulties because they do not specify when a licensee may make changes to the approved program without requesting a license amendment.
If the fire protection program connitted to by the licensee is required by a specific license condition, the provisions of 10 CFR 50.59 may not be applied to make changes without prior NRC approval.
Thus licensees may be required to submit amendment requests even for relatively minor changes to the fire protection program. One means of reducing this diversity and to increase
.,-.-.___.__y.__
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O 1
The Comissioners unifomity is to propose a uniform license condition which clearly identifies the approved fire protection program for
,l the facility and specifies the conditions under which it may be modified.
The Steering Comittee has proposed that this be recomended as an option to the licensees and remain a requirement for new plants.
Some industry opposition has been expressed and the CRGR expressed the view that it is doubtful that many oparating plants will voluntarily exercise this option.
At the time that the Commission required the license condition on new plants it was expected to be an interim measure until a 1
fire protection rule was developed.
Since that time, the staff has recommended and the Comission has agreed that a rule for new plants was not needed. An alternative approach, recoanended by CRGR, which also has the virtue of clearly identifying the approved fire protection program and specifying the conditions under which it may be modified, but without elevating fire pro-tection to a status above that of other safety related plant features, is to incorporate the fire protection program for all plants into the FSAR for the facility and remove the requirement for a license condition'.
In this manner, the fire protection program, including the. systems, the administrative controle, the organization, and other plant features described in the FSAR will be on a consistent status with other plant features described in the FSAR.
The provisions of 10 CFR 50.59 would then apply directly for changes the licensee desires to make in the program.
Some questions of enforceability have been raised with regard to this approach.
The generic letter has been modified to state that licensees should include the fire protection program in the next update of their FSAR with the intent that thereafter modifications to the program could be made in accord with 10 CFR 50.59.
One i
proviso would be necessary, that the " accident previously con-sidered" in the context of the unreviewed safety question i
detemination under 10 CFR 50.59 be the postulated fire in the i
fire hazard analysis for the fire area affected by the change.
The staff would then entertain application for amendment to remove the fire protection license condition. The staff is I
continuing to evaluate various other approaches to the elimi-l nation of the license conditions, including the need for a rule which would require compliance with the fire protection plan described in the FSAR.
Such a rule could be a means for elimi-nation of present license conditions and technical specifications.
For the plants remaining to be licensed the staff proposes to require that the fire protection program be submitted as part of the FSAR.
Changes under 10 CFR 50.59 would then be allowed under the groundrules discussed above.
No license condition would then be required.
r 3s The Commissioners 8-Interpretations Document The Steering Committee carefully reviewed the " Interpretations" and concluded that they will provide an improved means of achieving compliance with Appendix R.
Concurrence with this position was received from the Regional Administrators and there were no substantive comments on this document received from the public. The Appendix R Questions and Answers include the questions posed by the industry and are supplemented with additional questions identified at the regional workshops.
These responses may be used as guidance for design, revfew, and inspection activities. Minor revisions have been made in consideration of public comments and for consistency with the
" Interpretations" document.
Trubatch Comments The Commission also requested that the staff comment on the OGC memorandum and the differing professional opinions.
Mr. Trubatch raises objections to five parts of the Inter-pretations decument:
fire boundaries, cold shutdown, automatic detection and suppression, fire damage, and alternate shutdown. A response to Mr. Trubatch's legal concerns has been prepared by OELD (Enclosure 8).
Two comments from Mr. Trubatch relate to the staff's handling of fire damage and alternate or dedicated shutdown. The staff has not changed its position regarding the need for licensees to consider the effect of suppressants when assuring that "the structure, system, or component under consideration is capable of performing its intended function during and after the postulated fire, as needed." The purpose of this inter-pretation of fire damage is to clarify that cosmetic damage, which does not interfere with the above capabilities, would not be a basis for rejection of an exemption request.
l The position of the staff has also not changed in regard to the l
need for adequate separation of alternative shutdown equipment from normal shutdown equipment. However, Appendix R, in mentioning " room, or zone" in addition to " area," clearly implies that compliance may be achieved by other than separation by fire areas.
The staff has recognized this and has granted exemptions in the past. What is changing is the recognition that compliance with Appendix R can be established with adequate justification and exemptions need not be requested. The " room" concept must be justified by preparation of a detailed fire hazards analysis that demon-strates that a single fire will not disable both normal shutdown equipment and the alternate shutdown capability.
c I
d The Comissioners S_taff DP0 The staff DP0 is primarily concerned with Section 4 Fire Area Boundaries, and Section 5, Automatic Detection and Suppression, of the Interpretations of Appendix R.
The Interpretations allow penetrations in fire area boundaries and partial area coverage for fire suppression and detection if it can be established by analysis that adequate protection is provided Such analyses for the fire hazards associated with the area.
have previously been approved by the staff as exemptions.
By contrast, the " Interpretations" do not require submission of fire hazards analyses for staff review or the filing of exemption requests prior to fire protection inspections; but The DP0 states the analysis must be available for NRC audit.
that not requiring exemption requests is in conflict with Generic Letter 83-33 which was issued previously to clarify staff positions on these and other issues, and which is deemed preferable.
The Steering Comittee has revieived the positions defined in the " Interpretations" and has concluded that they are techni-cally and legally consistent with the requirements of Appendix R.
It is expected that the DP0 will be resolved accordingly.
However, formal resolution of the DP0 is being deferred pend-ing Comission action on the recomendations of the Steering Comittee.
The Steering Comittee on Fire Protection Policy established by the EDO has made a number of recomendations for improving Sumary and the consistency of the levels of fire protection safety and
==
Conclusions:==
These recom-for enhancing the implementation of Appendix R.
mendations, having received public coment, have been incorporated r
in a generic letter with attachments which the staff proposes The staff believes that the recom-to send to the licensees.
mendations, with changes discussed above, will provide the most consistent and workable means of enhancing the implementation of Appendix R and intends to proceed expeditiously with their issuance.
That the Comission direct the staff to:
Recomendations_:
Issue the generic letter with attachments (Enclosures 4, 5 1.
and 7 to this paper).
Conduct fire protection inspections in accordance 2.
with currently allocated fire protection resources and as described in Section B of the generic letter.
i
- The Comissioners 3.
Utilize the guidance in Enclosure 7 as the criteria for enforcement of Appendix R requirements and pro-ceed to evaluate and issue pending enforcement cases in accordance with this guidance.
4.
Remove the fire protection license condition from future licenses if the fire protection plan is incorporated in the FSAR.
5.
Evaluate the appropriate approach and need for fire protection Technical Specifications in conjunction with the ongoing effort on Technical Specification improvements.
f' L ".~ k 7.,I
/
I l'.'(i Wfiliam J. Dircks Executive Director for Operations
Enclosures:
1.
Memo dtd 10/26/84 to Ofrcks fm Fire Protection Policy Steering Comittee 2.
Memo dtd 5/3/85 to Denton fm Vollmer 3.
Generic Ltr on Fire Protection 4.
Interpretations of Appendix R 5.
Appendix R Questions & Answers 6.
Fire Protection License Condition 7.
Guidance for Enforcement Actions Concerning Fire Protection Requirements 8.
Memo dtd 6/27/84 to Vollmer fm Olmstead 9.
Steering Comittee Resolution of Public Coments f
.-._-,r.
IN RESPONSE, PLEASE REFER TO:
M851003A
,,,h UNITED STATES y
),
NUCLEAR REGULATORY COMMISSION
- ^*"*"*C"'
ACTION - Taylor /Denton l
Y8' November 6, 1985 o
Rehm gg, omes or THE stensTAny Arfngham
/
1 William J. Dircks, Executive Director MEMORANDUM FOR:
for Operations t
HerzelH.E.Plaine,GenergCounsel Samuel J. Chilk, Secreta f
FROM:
STAFF REQUIREMENTS - STA'I'J!j OF INTERPRETATION
SUBJECT:
OF APPENDIX R - FIRE PROTtOCION, 2:00 P.M.,
THURSDAY, OCTOBER 3, 1985, COMMISSIONERS' CONFERENCE ROOM, D.C. OFFICE (OPEN TO PUBLIC ATTENDANCE)
The Commission met with acaff to discuss the status of the implementation of fire protection requirements, the fire protection rule contained in 10 CFR 50.48 and Appendix R contained in SECY-85-306.
The Commission requested a report from the staff on current fire protection enforcement actions as well as a suggested schedule for affected licensees to comply with Appendix R requirements.
(SECY Suspense:
11/29/85)
(IE/NRR)
Chairman Palladino urged the Commissioners to vote promptly on Additionally, he would like to see the staff i
incorporate a license condition in each facility license to This should ensure enforceability of fire protection programs.
be done in addition to incorporating fire protection programs The license condition into the next edition of the FSARs.
should allow flexibility which permits modifications of fire
)
protection programs in accordance with 10 CFR 50.59.
Commissioner Asselstine requested that OGC provide its comments on staff's reply to the Trubatch memorandum on Appendix R requirements.
(OGC) 1985 the General Counsel provided (Subsequently, on October 24, a response to the Commission.)
R ec'd u ft. E n Deta..)I.
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~~
~
2-Commissioner Asselstine asked the staff to send to him the cost-benefit analysis required by the backfit rule of the new or dif ferent interpretations of Appendix R.
If such cost-benefit analysis does not now exist, he asked that the required backfit analysis be performed and sent to him before any new or modified interpretations are transmitted to licensees.
(NRR) *
(SECY Suspense:
11/29/85)
- Note: Proposed response received ED010/18/85 cc:
Chairman Palladino
~
Commissioner Roberts Commissioner Asselstine Commissioner Bernthal 1
Commissioner Zech Commission Staff Offices PDR - Advance DCS - 016 Phillips e
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Classes of Reactor Accidents e
31 f
C by a 9fu _ _ jMM
,.s.~1 while sam-. -
Econtrol rods can adversely affec l
fissioning in the core, causing exon
?
I spots). Another possible cause for f
o f
,N of coolant flow to a fuel bundle in g$ 4W6Dttypre. Such a weta boiling of coolant on the surfaces h
l p
1 M stearn sve ed
[
(
thin steam layer or " steam blanke sulate the rods from the coolant Motten fuei j
NN o, the heat removal from the fuel ro i<
explains why beads of water roll I'
Reactor iron. instead of boiling oII. The be j
the hot surface by a thin stean d'
F beads, so that they evaporate ver f
h","
7 ",
could then heat up and melt or i p
4 outpouring of molten or crumbi q
esame fuel rods, possibly causing them i
ing steam blanketing there or b E
Could Conceivably be a self prop
[
Fgre 9 Rornan candle" effect in SWR senare velve closure accident ing.18 The concern a that 1
e sot, rneco,e,,co,npo,d ofa' eros oe open4nse rnetar minute or so to the point where bones packed together inside each bon, w ~W." is a fuel od bun Process would be initiated, eve die The boxes in which the fuel rnetong % mes could constrain the
[
l fuei to discherge out the enos. n shown and the fissioning were stoppe y
steam explosions or fuel distorti prevent a SCRAM. Then, if a la rfl
'effect could conceivably be to i I-by compressing core steam bub L
present in the BWR core-that 18. eliminating the steam valve Jejecting) several contml m s
(
closure accident would not eliminate this kind of autocatalysis.
F It should be noted that the only theoretical analysis of the
- worsen the accident.
I PEA potential of present-day. Large reactors (water-cooled, UOr Therefore, PCMAs could rt sel with an instantaneous re es f
fueled) is contained in an " internal report" (PTR-738) of the AEC's Perature coolant and cause a p p,
National Reactor Testing Station. which was prepared in 1964 missiles and pressure waves
(
This report ca'culates a " catastrophic" explosion potential and is v; hole core meltdown and, fin.
f fairly thorough in examining the uncertainties. It also recommends lease directly into the atmosp a fundamentally important set of esperiments, including full-scale
' PEA, would be a more severe E
E reacto: destructive tests to investigate the PEA potential, which without effective ECCS.
T we shall discuss later.
Other PCMA possibilities b
. control rod at the normal mote I
Power-Cooling Mismatch Accidents a power excursion but which r
derin addition to increasmg 5
The most worrisome PCMA involves eteessive fissioning or un-namely, the intensity of loca i
E I
dercooling in a small region of the core while the reactor is oper.
yould be ineffective in this sit ated at full pg. Such a situatiun could be caused, for examp!,.
m.
6
l!
l!
T i
i r
g j
- ~
s c
t r
[
31 Classes of Reactor Accidents by a misarrangement of the ori@'M ta M [i(, tin'g h es enMent r~h, while maintaining steadyMuff nowAMispositioning
, j q
of control rods can adversely affect the spatial distribution of the l'
j i
fissioning in the core, causing excessive nonuniformity (local hot L
spots). Another possible cause for a PCMA would be a blockage of coolant flow to a fuel bundle in a BWR by sonneleiQM at caught up under the core. Such mishaps could lead to excessive c
boiling of coolant on the surfaces of the affected fuel rods until a thin steam layer or " steam blanket" forms on the surfaces to in-5 l
sulate the rods from the coolari, which will drastically reduce the heat removal from the fuel rods. (This steam blanketing effect I'
explains why beads of water roll around on a hot skillet or flat j
l' iron, instead of boiling off. The beads of water are insulated from d
the hot surface by a thin steam layer which forms under the
[
beads, so that they evaporate very slowly.) The affected fuel rods g
could then heat up and mmt or crumble in several seconds. The outpouring of molten or crumbled fuel would contact adjacent y
fuel rods, possibly causing them to similarly overheat by trigger-
[
ing steam blanketing there or blocking coolant th w. The result could conceivably be a self-propagating cascade of fuel rod melt-i ing? The concern is that the core could deteriorate within a
[
l minute or so to t' point where an uncontrollable core meltdown process would be initiated, even if the situation were detected l
and the fissioning were stopped by a SCRAM. Indeed, initial
(
f steam explosions or fuel distortion might jam the control rods and
[;
prevent a SCRAM Then, if a large steam explosion occurred, the
(
effect could conceivably be to produce a severe power excursion by compressing core steam bubbles in a BWR or by blowing off f
[
~
{
(ejecting) several control rods in a PWR, which would greatly worsen the accident.
F Therefore. PCMAs could rupture or explode the reactor ves-
{
set with an mstantaneous release of the high pressure, high-tem-
{
g g
perature coolant and cause a prompt containment rupture due to
(
missiles and pressure waves of steam explosions, followed by i
whole core meltdown and, finally, by a large fission product re-l t
lease directly into the atmosphere. This accident, like a severe
/
g PEA, would be a more severe chain of events than a DB-LOCA l
[
without effective ECCS.
Other PCMA possibilities include (a) a withdrawal of a single l
[
control rod at the normal motor speed. which is too slow to cause jT a power excursion but which can cause the core power level to rise in addition to increasing the nonuniformity of the fissioning.
7
(
namely. the intensity of local hot spots in the core (a SCRAM would be meffective in this >!tuation); (b) a withdrewel of a group i
E t
=
SS Classes of Reactor Acd 32 Chapter Three t
which " cooling water" (for e e
of rods without SCRAM in a PWR; and (c) a fuel bundle mispo.
The reactor heat energy, thes sitioned in the core (PWR). For these and the previous examples, form of heated lake water as i
the core power level would not be reduced, which maximizes the is the same, except that there danger of cascading fuel meltdown when steam blanketing occurs.
Hence, there is a chain o There are other PCMAs which involve a reactor coolant sys-stay intact 'i the coolant is e i
tem fault that would reduce the heat removal from the core but, chain is broken-for example through negative reactivity feedback, would reduce the core a failure of the outside cooli i
power level as well. Yet, steam blanketing could still occur. An the turbine steam valve-th example is an instantaneous coolant pump stoppage (seizure) that overpressurize. If the pressuri would reduce the coolant flow. The coolant in the core would ant pressure relief valves (i heat up so as to reduce the reactivity and thus reduce the reactor heated reactor coolant into th power; but the reduced flow would cause steam blanketing be.
but in such an event the i 5
fore the core heat (power) output could decline and thus would cooled, and pumped back int threaten fuel rod breakup, even if a SCRAM occurred to shut off a loss-of-coolant situation in i
the fissioning altogether. If a SCRAM failed to occur, this pump-core meltdown. (The ECCS c i
seizure accident would be more severe, since the core power level the reactor.) The heat of the i
~
would remain substantial. In all, there are a great many PCMA by the containment cooling s-possibilities. As with PEAS, full-scale tests would be necessary containn.ent, for a containm a
to verify the theory for those PCMAs in which the reactivity of the coolant and thereby le r
could conceivably rise.
This relief valve-containme The PCMAs discussed so far involve a failure of the coolant stitutes anotiser vital chain to remove the heat from the fuel There is also a class of PCMAs the chains must work to avt i
not leading immadiately to fuel overheating but leading instead the reactor is shut down, the to coolant overpressure, which would threaten to cause a catas.
must also be removed and tri 1
trophic reactor vessel rupture or a severe LOCA. These PCMAs by a heat-exchange chain.
involve a failure to remove heat from the coolant.
The coolant's function, besides slowing down neutrons, is to Serious accidents result heat-exchange chains will be absorb the fission-and af ter-heat of the reactor core. This heat b
dents " to distinguish them must then be removed from the coolant before it recirculates cussed earlier. Specific exan through the reactor, or else it will overheat and over pressurize.
. as follows:
Normally, the coolant in a PWR gives up its heat by passing Turbine valve closing w through the steam generator. In the process, the boiler water is stops the flow of steam (hes heated into steam, which is then piped to the turbine. That is, the the core remains at a high po reactor core heat, which was absorbed by the coolant, is trans.
ferred to the boiler water, where it ends up in the steam. The l overheats and pressurizes u resultant loss of boiler water must be continually replenished, l
two minutes,8' which violate r
PsL8 Since the reactor desig however, and this is done by pumping "feedwater" into the steam f
, sel rupture might be assuma generator. To conserve water, the steam, after passing through the Z
turbine, is recycled by condensing it back into water to become margin to withstand the exti
> a BWR causes a power exct F
the feedwater. But for this to be possible, the heat of the steam team bubbles in the core must be removed, which is done by the turbine, in generating electricity, and by the " condenser " The condenser cools and con.
pwer excursion problem d since no steam bubbles are r
=
denses the used steam by means of an array of cold tubes through 2
4 UNITED STATES OF AMERICA NUCLEAR REGULATORY C0pmISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
)
)
CLEVELAND ELECTRIC ILLUMINATING
)
Docket No. 50-440 OL COMPANY, ET AL.
)
50-441 OL
)
~ (Perry Nuclear Power Plant,
)
Units 1 and 2)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO MOTION TO REOPEN THE RECORD FILED BY OHIO CITIZENS FOR RESPONSIBLE ENERGY" in the above captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, by deposit in the Nuclear Regulatory Commission's internal mail system, this
- 2nd day of January, 1986:
- Dr. Jerry R. Kline
- James P. Gleason, Chairman Administrative Judge Administrative Judge Atomic Safety and Licensing Board 513 Gilmoure Drive U.S. Nuclear Regulatory Commission Silver Spring, MD 20901
' Washington, DC 20555
- Mr.-Glenn 0. Bright Donald T. Ezzone, Esq.
Administrative Judge Assistant Prosecuting Attorney Atomic Safety and Licensing Board 105 Main Street U.S. Nuclear Regulatory Connission Lake County Administration Center Washington, DC 20555 Painesville, OH 44077 Jay Silberg, Esq.
Susan Hiatt Shaw, Pittman, Potts and Trewbridge 8275 Munson Road 1800 M Street, NW Mentor, OH 44060 Washington, DC 20036
- Atomic Safety and Licensing Board Terry J. Lodge, Esq.
U.S. Nuclear Regulatory Commission 618 N. Michigan Street, Suite 105 Washington, DC 20555 Toledo, OH 43624 John G. Cardinal, Esq.
Janine Migden, Esq.
Prosecuting Attorney Ohio Office of Consumers Counsel Ashbabula County Courthouse 137 E. State Street Jefferson, OH 44047 Columbus, OH 43215
- Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commissien Washinoton, DC 20555
- Docketing & Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 s
Colleen P. Woodhead Counsel for NRC Staff e
9