ML19256A993
| ML19256A993 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/08/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Bilby C CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| FOIA-80-162, RTR-NUREG-0460, RTR-NUREG-460, TASK-15-22, TASK-RR NUDOCS 7901170371 | |
| Download: ML19256A993 (9) | |
Text
Tis ea g
UNITED S TATES NUCLEAR REGULATORY COMMISSION y
,4 3. I. ' D I $
WASHINGTON, D. C. 20555 b%, ' '"..f E
.../
JANUARY 8 5 79 Docket No. 50-155 Mr. C. R. Bilby Vice President Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Bilby:
Attached for your infonnation is a copy of NUREG 0460, Volume 3 which details our current view related to ATWS.
In this supplement a variety of options are considered regarding ATWS. We intend to select one of the ATWS options in the near future and to pursue it to adoption.
However, it is important to note that all of the options under serious consideration by the NRC staff (options #2, 3, and 4 in Volume 3 of NUREG 0460) regarding resolution of the ATWS issue for BWRs require installation of an RPT. While you have committed to install a RPT on your facility, Big Rock Point Plant, you have not yet begun to take steps toward such installation, on the grounds that you were awaiting firmer requirements by NRC. The NRC staff now has a firm position that RPT is required for your facility. Therefore, we see no bases for any further delay in implementing an RPT for your facility. The RPT designs discussed in this letter are compatible with ATWS requirements.
To expedite your installation of an approved RPT, the staff is providing a modified description (Appendix A, attached) of design requirements which provide some additional flexibility over those previously provided (May, 1978), but which the staff has found acceptable for RPT systems to be installed in the near future.
For all operating plants, the Monticello RPT design described in NED0 25016 and summarized in Appendix B has been accepted by the staff as meeting the Appendix A criteria.
Sections of NEDO 25016 related to ARI should be ignored as that system is not addressed by this letter.
Some operating plants have already installed the "BWR/4" or " Hatch" RPT, and the staff also accepts that design as meeting the Appendix A criteria provided the changes specified in Appendix B, or equivalent changes, are incorporated.
?S011?Q3y i
d Mr. C. R. Bilby ~
Both the Monticello design and the modified "BWR/4" or " Hatch" design utilize generator field breakers which have been modified so that they are provided with two trip coils. One coil for each breaker is actuated only by reactor pressure and water level sensors in RPT division A, and the other coil is actuated by pressure and level sensors in RPT division B, thereby providing redundancy of power supplies available to the overall system and increasing trip reliability.
Either the Monticello or nedifiad "BWR/4" or " Hatch" design, would be an acceptable RPT design provided diverse final trip relays of a different type are used, or obtained from a different manufacturer than the primary scram relays used in the RPS.
The staff has not reviewed the specific design of the time delay circuitry recently proposed for the Monticello RPT design for low-level initiated pump trips. We agree that time delays on the order of 10 seconds are desirable to avoid making the consequences of a postulated LOCA more severe, and we agree that such delays of around 10 seconds have insignifi-cant effect on ATWS consequences (for low-level initiated ATWS pump trips only). Therefore, we find incorporation of such circuitry on either RPT design discussed above to be acceptable, provided:
1.
The time delay is realized only for low-level initiated pump trips;
- and, 4
2.
The circuitry is incorporated in such a way that it does not signifi-cantly affect the overall reliability of the RPT; that is, that no single failure in the timing circuit (s) can cause failure of the pump trip to occur. This could be accomplished, for example, by use of a separate, independent timing (delay) circuit with each low-level sensor, or equivalent.
Implementation as soon as possible of an RPT in accordance with the attached design criteria will provide an increased level of safety over the lifetime of the plant and should be installed as promptly as is reasonable.
Mr. C. R. Bilby The staff has given careful consideration to the concern expressed by some licensees that RPT design requirements may change in the future.
We have concluded that the design criteria outlined in this letter (Appendix A) are, for operating plants, equivalent to those enclosed with the May, 1978 letters to all BWR licensees, and we intend to effect no changes to those criteria in the future.
We believe that RPT design, procurement, and installation can be accomplished within a two year period without requiring additional outage time beyond refueling outages.
We have given consideration to steps that can be taken at present, in order to reduce the risk from ATWS events during the interim period before recirculation pump trip circuitry and any other necessary plant modifications are completed. We have determined that many of the following steps are practicable and appropriate for your facility for this interim period. We therefore, request that you inform us within 90 days that you have done the following:
1.
Developed emergency procedures to enable operators to recognize an ATWS event, including consideration of scram indicators, rod position indicators, flux monitors, vessel level and pressure indicators, relief valve and isolation valve indicators, and containment temperature, pressure, and radiation indicators.
2.
Train operators to take actions in the event of an ATWS including consideration of manually tripping the recirculation pumps and scramming the reactor by using the manual scram buttons, changing individual rod scram switches to the scram position, stripping the feeder breakers on the reactor protection system power distribution buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the RHR in the pool cooling mode to reduce the severity of the containment conditions.
Early operator action as described above would provide significant protection from those ATWS events which occur at low power levels where the rise in the vessel pressure and the containment temperature is limited to acceptable values by manual recirculation pump trip and actuation of the existing standby liquid control system.
If the operator were to promptly (in a few seconds) trip the recirculation pumps to assure that the short term rise in vessel pressure is not excessive, protection will also be provided for those ATWS events where the common mode failure occurs in either the electrical portion of the scram system or in some portions of the drive system.
Mr. C. R. Bilby Within 90 days inform us of your schedule for implementation of your
~
commitment to install an RPT system for your plant. Such system should conform to the acceptable systems described in this letter and your schedule should be consistent with the staff's overall objective of assuring that an acceptable RPT system is installed at your facility within two years.
Sincerely, Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
NUREG 0460, Volume 3 2.
Appendices A and B cc w/ enclosure No. 2:
see next page
-c
~
' Consuners Power Canpany cc Mr. Paul A. Perry, Secretary Consumers Power Canpany 212 West Michigan Avenue Jackson, Michigan 49201 Judd L. Bacon, Esquire Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Hunton & Williams George C. Freeman, Jr., Esquire P. O. Box 1535 Richmond, Virginia 23212 Peter W. Steketee, Esquire 505 Feoples Building Grand Rapids, Michigan 49503 Charlevoix Public Library 1 07 Clinton Street C h3: 'evoix, Michigan 49720
.c E
APPENDIX A CRITERIA FOR HIGH PRESSURE-LOW LEVEL INITIATED RECIRCULATION PUMP TRIP RPT TO BE INSTALLED IN OPERATING BWRs BEFORE NOVEMBER 1, 1979*
A.
General Functional Raouirement The RPT system shall automatically initiate the appropriate action whenever the conditions monitored by the system reach a preset level.
B.
Independence and Integrity The RPT system and components shall be independent and separate from components and/or systems that initiate anticipated transient (s) being analyzed and diverse from the normal scram system to minimize the probability of disabling the operation of the mitigating system.
Diversity can be achieved by incorporating as many of the following methods as is practicable:
1.
Use of'RPT final trip relays from different manufacturers (required).
2.
Use of energized versus de-energized trip status.
3.
Use of AC versus DC power sources.
It shall be demonstrated that the function of the RPT system and components will not be disabled as a consequence of events being analyzed.
Diversity of the RPT pressure and level sensing devices (including relays used in such sensing devices) from similar or identical devices used on the RPS is not required, since failure of those devices on both the RPT and the RPS is not likely to cause an ATWS due to the presence of other diverse trips on the RPS (high flux, valve position, etc.).
- The NRC staff has reviewed the Monticello RPT design and the " Hatch" RPT design, and finds that they meet these criteria (provided the changes specified in the cover letter are made to the " Hatch" design). Plant specific reviews will be conducted only as necessary to ascertain that the plant design is the same as, or equivalent to, one of the approved designs.
I
, C.
Equipment Qualification The RPT system equipment and components shall be tested to verify that the system will provide, on a continuing basis, its functional capability under conditions relevant to postulated ATWS events, in-cluding extremes of conditions (as applicable) relating to environ-ment, which are expected to occur in the lifetime of a plant.
D.
Periodic Surveillance and Preventative Maintenance Testing and Calibration Periodic surveillance and preventative maintenance tests and calibra-tion requirements shall be identified to provide continuing assurance that the RPT system, including sensors and actuated equipment, is capable of functioning as designed and that system accuracy and per-formance have not deteriorated with time and usage. These requirements shall be particularly directed toward the detection of those failures or degradation of accuracy and performance which would not otherwise be likely to be detected during the course of normal operations.
Integrated system testing shall also be performed to verify overall system performance.
E.
Quality Assurance A quality assurance program in conformance with the requirements of 10 CFR 50 Appendix B shall be applied to the RPT system design and equipment.
F.
Administrative Controls
+
Administrative controls shall be established to control the access to all set point adjustments, calibration and test points.
G.
Information Readout The RPT system shall be designed to provide the operator with accurate, complete and timely information regarding its status. For those functions, including operations, test or maintenance, and calibration, which require direct operator interaction, human engineering factors such as information displays (e.g., display formats, layout and con-trols) and functional controls (e.g., methods, location and identiff-cation) shall be included in the design.
. H.
Maintainability The design shall include measures which enhance maintainability to reduce mean-time-to-repair and to assure the continued availability and reliability of the system for the life of the plant. The system design shall include features which facilitate the recognition, loca-tion, replacement, repair and/or adjustment of malfunctioning equipment and components or modules.
.y E
Appendix B Acceptable RPT Designs Monticello RPT Design The Monticello design simultaneously trips both MG. sets "A" and "B" generator field breakers upon receipt of either reactor high pressure or low-low water level control logic input signals. The logic to each breaker is two-out-of-two (pressure) or two-out-of-two (level)
(2/2 or 2/2),
i.e., contacts "A" and "C" or contacts "B" and "D" must close to trip the breaker. The Monticello design employs diversity, testability, separation and redundancy.
Modified BWR/4 or Hatch RPT Design The modified "BWR/4" or " Hatch" design results in the independent (separate) trip of each of the two recirculation pumps upon receipt of either one reactor high pressure signal or one low-low water level signal. The logic to each MG set "A" and "B" generator field breaker is one-aut-of-two (level) or one-out-of-two (pressure) (1/2 or 1/2).
The modified "BWR/4" or " Hatch" design employs diversity, testability, separation, and redundancy.
The modification to the existing " Hatch" design which makes it acceptable is accomplished as follows:
- 1) Add a second trip coil to each recirculation loop's M-G set generator field breaker, as per the identical modification made to Monticello.
- 2) Connect one of the pressure sensors and one of the low level sensors in RPT train A to the old (existing) trip coil in the recirculation loop A M-G set generator field breaker. Connect one of the pressure sensors and one of the low level sensors in RPT train B to the new trip coil in the recirculation loop A M-G set generator field breaker.
- 3) Connect the other pressure sensor and. the other low level sensor in RPT train A to the new trip coil in the recirculation loop B M-G set generator field breaker. Connect the other pressure sensor and the other low level sensor in RPT train B to the old (existing) trip coil in the recirculation loop B M-G set generator field breaker.
y z
A NRC PUI L1 occtniMNT RCOM g
p@/pd }e f
r
,7 e
UNITED STATES OF AMERICA f
NUCLEAR REGULATORY COMMISSION g
BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of Portland
)
Docket No. 50-344 (Proposed General Electric Company
)
Amendment to Facility Oper-et al.,
(Trojan Nuclear
)
ating License NPF-1 to Permit Plant)
)
Storage Pool Modification)
)
INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION 790117C377 (er
e Table of Contents Page I.
Introduction.
1 II.
Oregon Exceptions No. 1-6 and 8 2
A.
Summary.
2 B.
The ASLB's failure to impose Oregon's requested.
2 technical specifications is contrary to the NRC's obligation to protect the public health and safety.
C.
Oregon's proposed technical specifications 5
are necessary in order to protect the public health and safety.
D.
The ASLB's " burden on the operator" and.
6
" rigidity" standards are not appropriate and in any event were applied inconsistently.
E.
Specific " Technical Specification" Exceptions.
8 1.
Water Chemistry Controls.
8 2.
Corrosion Coupons 9
3.
Use of Alternate Storage Cavities 10 4.
SFP Water Temperature 10 5.
2,000 ppm Boron.
11 6.
Full Core Reserve 12 F.
Conclusion Re Exceptions No. 1-6 and 8 13 III. Oregon's Exception No. 7.
13 A.
The ASLB's ruling reflects a misunderstanding.
13 of Oregon's argument.
B.
Statement of Facts 15 1.
Consideration of cumulative impacts 16 2.
Consequences of deferral.
17 3.
Failure to weigh and balance the five 18 factors C.
Argument.
19 1.
Since it cannot be concluded that deferring.19 authorization to use the modified SFP to store more than 4/3 cores would result in substantial harm to the public interest, authorization should be withheld pending completion of the GEIS 2.
The NRC staff failed to weigh and balance.
20 the five factors contained in the NRC's 9/75 Notice D.
Conclusion Re Exception No.
7.
21 IV.
Request for Oral Argument.
22 FOOTNOTES.
23
t Table of Authorities Page Cases Natural Resources Defense Council v. NRC,.
20 539 F2d 824 (2d Cir 1976), cert. granted, 430 US 944 (1977), vacated and remanded 434 US 1030 (1978) (memorandum opinion)
Power Reactor Co. v. Electricians, 4
367 US 396, 406 (1961)
NRC Decisions Boston Edison Company (Pilgrim Nuclear.
21 Generating Station, Unit No. 2), ALAB-479, 5/25/78 (7 NRC
)
Niagra Mohawk Powar Corp. (Nine Mile 2
Point Nuclear Station, Unit 2), ALAB-264, 1 NRC 347, 357 (1975)
Northern States Power Company (Prairie 3
Island Nuclear Generating Plant, Units 1 and 2), ALAB 455, 7 NRC 41, (1978)
Statutes Atomic Energy Act of 1954 as amended:
Section 103 (42 USC 2133)
.3 Section 182 (42 USC 2232) and (42 USC 2232a).
.3 Section 186 (42 USC 2235) 5 Chapter 18 (42 USC 2271-82).
5 National Environmental Policy Act:
42 USC 4321
.20 NRC Regulations 10 CFR Part 2, Subpart B.
6 10 CFR 50.36(b) 4 10 CFR 50.40 3
NRC Policy Statement
" Notice of Intent to Prepare Generic.
14 Environmeittal Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel" 40 FR 42801 (9/16/75)
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of Portland
)
Docket No. 50-344 (Proposed General Electric Company,
)
Amendment to Facility Oper-et al.,
(Trojan Nuclear
)
ating License NPF-l to Permit Plant)
)
Storage Pool Modification)
)
INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION I
INTRODUCTION Oregon does not oppose immediate modification of the Trojan spent fuel pool ("SFP") to accomodate additional storage.
Its exceptions to the ASLB's initial decision serve two purposes:
(1) to argue that the NRC is proceeding unlawfully by authorizing at this time use of the modified SFP beyond that authorized in the original Trojan operating license; and (2) to insist that certain technical specifications be imposed to insure operation of the modified SFP in compliance with NRC standards.
The first point is made in connection with Exception No. 7 and will not be summarized here.
As to the second purpose, Oregon disputes the method of regulating the operation of the modified SFP approved by the ASLB.
It is Oregon's position that regulation of the safety aspects of the operation of a nuclear power plant should not be left to a series of mere " gentlemen's agree-1 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
s ments" made by the operator to the NRC staff in a license amendment application.
Even where, as here, no question has been raised as to the good faith of the applicant, a clear basis should be established for regulating the safe daily operation of the Troj an SFP.
The technical specifications suggested by Oregon, which to a large extent are derived from representations made by PGE in its safety analysis, would provide such a clear basis for maintaining the safety of the SFP.
Plainly put, by means of Exceptions No. 1-6 and 8, Oregon is asking this panel to substitute its judgment for that of the ASLB, in order to reach a result which Oregon submits is pref-erable from a public policy standpoint.
Niagra Mohawk Power Corp (Nine Mile Point Nuclear Station, Unit 2) ALAB 264, INRC 347, 357 (1975)
II OREGON'S EXCEPTIONS NO. 1-6 AND 8 A.
Summary Oregon Exceptions No. 1-6 and 8 share three basic argu-ments.
7 hose arguments are set forth below preceding a dis-cussion of the supporting facts in the record.
B.
The ASLB's failure to impose Oregon's requested technical specifications is contrary to the NRC's obligation to protect the public health and safety.
Technical specifications suggested by Oregon relating to the safe operation of the Trojan SFP are designed to ensure a 2 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
clear, enforceable basis for regulation in order to protect the public health and safety.
The ASLB declined, without adequate reason, to adopt most of the specifications suggested by Oregon.
(Oregon Exception 8).
The Appeal Board's decision in the Prairie Island SFP case sets the standards by which this proposed SFP expansion should have been considered by the ASLB:
.in essence, section 50.40 [of 10 CFR] requires that the Commission be persuaded, inter alia, that the applicant will comply with all ap-plicable regulations, that the health and safety of the public will not be endangered, that the issuance of the amendment will not be inimical to the health and safety of the public, and that any applicable requirements of 10 CFR Part 51 (governing environmental protection) have been satisfied. "
Northern States Power Company (Prairie Island Nuclear Generating Plant, Units 1 and 2) ALAB 455, 7 NRC 41, 2 CCH Nuclear Regulation Reporter 1 30,267.02 (1978).
This standard is, of course, consistent with section 103 of the Atomic Energy Act itself (42 USC 2133; See also 10 CFR 50.40 (common standards) and 10 CFR 50.57(a) (issuance of operating license)).
Section 182(a) of the Act (42 USC 2232) expressly contemplates the use of technical specifica-tions to enable the NRC to find that the public health and safety will be protected:
.In connection with applications for licenses to operate production or utili-zation facilities, the applicant shall state such technical specifications,.
3 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued. '.' 42 USC 2232a.
The NRC may require technical specifications not orig-inally offered by the applicant, based on safety analysis and evaluations.
Thus, technical specifications intended to insure safe operation are directly related to the NRC's key finding, in any license proceeding, that the public health and safety will be protected.
Power Reactor Co. v. Electricians 367 US 396, 406 (1961)
The technical specifications proposed by Oregon are entirely consistent with the directions of the Act and the NRC's regulations.
These specifications -- water chemistry controls (Exception No. 1); a corrosion coupon program (Exception No. 2); use of alternate storage cavities (Exception No. 3); SFP water temperature maximum (Exception No. 4); maintenance of 2,000 ppm of boron (Exception No. 5);
and maintenance of full core reserve (Exception No. 6) --
are interrelated, and collectively provide protection to the public health and safety by ensuring that the SFP liner and equipment will not be subject to corrosion, and will be 4 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
easily repairable; that criticality will be prevented; and that the consequences of accidents will be minimized.
C.
Oregon's proposed technical specifications are necessary in order to protect the public health and safety.
The ASLB's fundamental decision that expansion of the Trojan SFP will not be inimical to the public health and safety rests upon a series of conclusions or findings that certain events, such as corrosion or criticality, will not occur.
Each of these con-clusions is in turn based on a series of assumptions regarding factual circumstances.
The technical specifications suggested by Oregon are intended to ensure that the ASLB's assumptions relating to spent fuel storage circumstances are well-founded as a matter of operating reality.
There is little or no factual difference between the ASLB and Oregon as to the importance to safe Trojan SFP operation of water chemistry controls, water temperature ceilings, cou-pons (or other corrosion detection devices), maintenance of 2,000 ppm of boron, or utilization of alternate storage cavi-ties for freshly discharged fuel.
Each of these matters may be found as an assumption or circumstance accepted as fact in ASLB's initial decision.
The difference lies in the extent to which these underlying facts are to be made binding on PGE.
Oregon proposes to have these key assumptions made directly en-forceable by means of imposing technical specifications on the licensee / operator of Trojan.
See Section 186, (Revocation) and Ch 18 of the Atomic Energy Act of 1954 (42 USC 2235, 42 USC S - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
2271 - 42 USC 2282) and 10 CFR Part 2 Subpart B (Procedure for Revocation of License) 10 CFR 50.100, 50.110.
The ASLB chose not to do so.
Oregon maintains that its position represents a more sound regulatory policy.
D.
The ASLB's " burden on the Operator" and " rigidity" stand-ards are not appropriate, and in any event were applied inconsist-ently.
The rationale advanced by the ASLB for refusing to impose Oregon's proposed technical specifications is not consistent with the NRC's statutory obligation to protect the public health and safety.
For example, the ASLB appar-ently rejected as "too rigid" a technical specification relating to use of alternate storage cavities for freshly discharged fuel.
It rejected a technical specification relating to maintaining water temperature at 140 F. on S.e same grounds.
The only other test announced by the ASLB related to the use of technical specifications is whether or not a " burden would be imposed on the plant operator.
ASLB Finding 25, p. 8 Je also, ASLB Finding 53, P.
34 (Oregon Exception 8).
There is no support in the Act or in the NRC's regulations for either test.
Moreover, assuming arguendo that the ASLB's " rigidity" and " burden on the operator" standards were lawful, as to all of Oregon's Exceptions, the applicant is apparently wil-ling or able to follow Oregon's proposals, and voiced no real concern regarding " rigidity" or " degree of burden".
6 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
To illustrate, PGE represented that it will implement water chemistry controls (PGE Ex 2, Tables 8-6 and 3-7),
that it will use alternate storage cavities for freshly discharged fuel, (PGE Ex 2 p. 4-4), and that 140 F will be the SFP water temperature maximum (PGE Ex 2 p. 3-16).
PGE has stated that maintenance of 2,000 ppm boron in the SFP is not an onerous burden (Tr 6126-6127).
PGE also asserted that maintenance of a full core reserve was a basis for its choice of new rack capacity (PGE Ex 2 p. 6-1 and 3).
Finally, PGE agreed that it is capable of running a valid corrosion coupon program (Tr 3079, 4581).
Thus, by its own standards, the ASLB should have applied Oregon's proposed technical specifications.
Furthermore, the ASLB applied its standards in an incon-sistent manner.
The 2,000 ppm boron technical specification imposed by ASLB relating to installation of new racks (ASLB
- p. 32-33, Finding 52) is to prevent criticality because of an accident having an unknown probability.
Oregen identi-fied a probability for projectiles striking stored spent fuel (Oregon Ex 1 p. 30).
The ASLB apparently determined that such a scenario was incredible.
(See ASLB p. 33, Finding 51 and p. 35 Finding 55).
However, it is not apparent what test the ASLB used to classify events as " credible" or " incredible".
It is very difficult to rationalize the ASLB's distinction be-tween the risk of criticality occuring during reracking opera-tions, necessitating a boron technical specification, and 7 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
criticality occuring due to accidents occuring at other times but involving the same SFP.
E.
Specific " Technical Specifications" Exceptions As to each of Oregon's Exceptions No. 1-6 and 8, there is evidence in the record upon which this Appeal Board could base the requested technical specifications.
In its Ex-ceptions, Oregon raised specific alleged errors under the
" umbrellas" of six general subjects.
Oregon intends the following discussion to be applicable to those general items of concern.
Each specific subpart of each general exception will not necessarily be discussed.
This should not in any way be construed as an abandonment of any of the general exceptions.
1.
Water Chemistry Controls (Oregon Exception 1) -
The ASLB found that corrosion will not affect spent fuel racks, SFP liners or stored fuel assemblies (ASLB p. 14, Finding 16; p. 15, Finding 17; and p. 20, Finding 31).
It then concluded that corrosion would not cause significant offsite radiation releases and on-site occupational ex-posures.
(ASLB p. 45, Finding 73).
All ( f these findings depend to a large extent on an assumption regarding main-tenance of PGE's proposed SFP water chemistry.
(See ASLB
- p. 13, Findings 14 and 15).
The expert testimony supporting the ASLB's ultimate find-ing on corrosion is highly contingent upon maintenance of PGE's represented water quality (NRC witness Weeks and PGE witness
.8 - INTERVENOR STATE OF OREGON'S BRIEF 1N SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
Johnson, Tr 2829, 3092, 4588, 4600-02).
In the absence of water chemistry control, corrosion could result in releases of radioactivity, causing a threat to the public health and safety.
PGE indicated an intention to control water chemistry in the SFP at the quality relied upon by Drs. Week's and Johnson.
(PGE Ex. 2, Tables 3-6, 3-7).
2.
Corrosion Coupons (Oregon Exception 2) - The ASLB misinterpreted Oregon's position on the use of coupons (Com-pare ASLB p. 9, Finding 10 with Tr. 3442-43 and 7477-78).
Oregon believes coupons are useful in discovering abnormal con-ditions, not for measuring corrosion under normal conditions.
PGE has no plans for evaluating the impact of out-of-speci-fication SFP water chemistry on racks and liners (Tr 3096).
Such chemistry would result whenever makeup water was provided to the SFP directly from the Columbia River (Tr. 3070-72)
Use of corrosion coupons would be useful for evaluating the conse-quences of out of specification chemistry (Tr. 2852, 4581, 4609-4610).
Visual examination would detect only gross damage (Tr. 2944) and only on the outer rows of the racks (Tr. 2945).
Thus the ASLB's findings that SFP leaks can be repaired, (ASLB p. 15, Finding 18), that the modifications will facilitate repairs (ASLB p. 23, Finding 34), that shipping casks for re-moval of fuel from the SFP are not required (ASLB p. 25, Finding 35), that conditions for SFP repairs are acceptable (ASLB p. 24, Finding 36), and that SFP components or fuel assemblies will not be subject to adverse corrosion (ASLB p. 14, Finding 16) are all 9 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
dependent, to some extent, on an ability to assess the impacts of off-specification SFP water chemistry.
PGE is capable of admini-stering a valid corrosion coupon program (Tr. 3077) and such a program would have utility at Trojan (Tr. 4609).
3.
Use of Alternate Storage Cavities (Oregon Exception
- 3) - The new storage racks will permit closer storage of freshly discharged spent fuel (13.3 inches v. 23 inches - PGE Ex. 2,
- p. 2-1 and 4-5).
Therefore, an object dropped in the SFP would potentially cause greater consequences since more fuel would be affected.
PGE committed to use every other storage cavity for freshly discharged fuel, thereby providing spacing (26.3 inches) greater than what now exists (PGE Ex 2, p. 4-4).
This insures that the consequences of a dropped object or projectile will not be increased by the modifications as older fuel, placed between freshly discharged fuel, will have a negilgible contri-bution to accident consequences (PGE Ex 2, p. 4-4)
Thus, an assumption regarding the use of alternate storage cavities for freshly discharged spent fuel is implicit in the ASLB's finding that the consequences of projectile impacts are acceptable to the public health and safety (ASLB p. 51, Finding 47).
4.
SFP Water temperature (Oregon Exception 4) - The record shows that a maximum SFP operating temperature of 140 F was used as a basis for many safety conclusions.
An underlying assumption regarding maximum SFP water temperature is implicit in finding that little corrosion will occur at 10 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
the temperature of the SFP water (ASLB p. 12, Finding 13),
that SFP components or fuel assemblies will not be subject to corrosion (ASLB p. 14, Finding 16), that no SFP liner corrosion will occur (ASLB p. 20, Finding 31), that the weld heat sensitized zones in the racks and liner would not be subject to corrosion (ASLB p.
15, Finding 17), that a small increase in temperature will not be detrimental to SFP equipment (ASLB p. 37, Finding 61) that SFP cooling equipment will not be burdened (ASLB p. 31, Finding 62), and that temperature would not affect off-site releases of radioactivity and occupational exposures (p. 45, Finding 73).
The consequences of routine operation at temperatures in excess of 140 F were not evaluated.
(Tr. 4852) 5.
2,000 ppm Boron (Oregon Exception 5) - If 2,000 ppm boron were retained in the SFP criticality would not be possible, even with fresh fuel.
(Tr 5169 and Oregon Ex 1, p. 20)
Absent any prohibition, fresh fuel may be stored in the SFP.
In fact, PGE could envision operating circumstances which would result in storage of fresh fuel in the SFP (See Tr 5188)
There fore, the ASLB could not properly conclude that only depleted or spent fuel will exist in the SFP (See ASLB p. 30, Finding 46).
Main-tenance of 2,000 ppm boron would allow storage of fresh fuel while preventing criticality.
The proposed modification will increase the probability of criticality.
The ASLB recognized that closer spacing of fuel increases the probability of criticality (ASLB p. 30, 11 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
Finding 46).
The ASLB states incorrectly that this is offset by the fact that fewer projectiles can wedge themselves between fuel assemblies (Op cit).
However, the " wedge" mechanism relied upon is not operative, since the tops of the storage racks are welded together (PGE Ex 2, Figure 3-4).
Projectiles may crush the fuel and racks without wedging between fuel assemblies.
Without a technical specification requiring 2,000 ppm, boron in any amount (including none at all), is allowed by the Trojan license.
Absent a technical specification requiring 2000 ppm boron in the SFP, the ASLB cannot conclude that under actual conditions substantial amounts of boron will remain in the SFP after refueling (ASLB p. 30, Finding 46 and p. 33, Finding 50).
Since criticality is not precluded, the ASLB cannot find that releases of radioac tivity and occupational exposures are insignificant (ASLB p. 83, Finding 73).
Therefore, it cannot conclude that the modification presents an acceptable risk to the public health and safety.
6.
Full Core Reserve (Oregon Exception 6) - SFP liners and racks serve the basic function of safely storing spent fuel.
In order to continue these functions over the life of the SFP Trojan operating license it is necessary to be able to perform repairs if needed.
Repairs may be needed because of corrosion to liners and racks.
The use of Columbia River makeup water could result in chloride concentrations in excess of proposed technical speci-12 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
fication limits (Tr. 3072 and 3099).
Such chloride concentra-tions could affect weld heat affected areas in the stainless steel for the liner and racks.
(ASLB p. 15, Finding 17).
The ASLB noted that full core reserve is needed to accomplish repairs to some parts of the liner (ASLB p. 22, Finding 34 and PGE p. 38 follows TR 4180).
F.
CONCLUSION RE EXCEPTIONS 1-6 AND 8 It is clear from the record that the technical specifica-tions proposed by Oregon are directly related to the safe operation of an expanded SFP at Trojan.
These specifications impose no undue burden on the licensee -- PGE presented no substantial evidence to that effect.
Their imposition would ensure that the assumptions underlying the ASLB's fundamental safety findings will be grounded in fact throughout the oper-ating years of the Trojan facility.
The public health and safety would be better served if Oregon's proposed technical specifications are adopted by this panel.
Such a result would be consistent with the Atomic Energy Act, the NRC's regula-tions and sound regulatory policy.
III OREGON'S EXCEPTION NO. 7 A.
The ASLB Ruling It is not possible to state neatly the ASLB's ruling on Oregon's contentions regarding the permissible extent of utilization of the modified SFP at Trojan.
A reading of Finding No. 88 (p. 57-58) indicates that the ASLB mis-13 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
construed Oregon's argument.
Thus it is necessary to set that position forth here in summary fashion.
First, Oregon contends, as a general proposition, that by reason of the Nuclear Regulatory Commission's " Notice of Intent to Prepare Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel" 40 Fed Reg 42801, September 16, 1975 ( the "9/75 Notice")
license amendments authorizing increased on-site storage of spent fuel cannot be issued prior to completion of the gen-eric environmental impact statement ("GEIS") described in such notice, unless deferral of an individual licensing action would result in substantial harm to the public interest.
The Board's initial decision does not address Oregon's legal premise.
In Finding Number 86 (at page 56) the Board concurred
. with the staff's conclusion in the EIA that deferral or substantial restriction of the action here proposed would result in substantial harm to the public interest."
As to this Finding, Oregon contends that the Board erred.
The second aspect of Oregon's exception assumes that that the 9/75 Notice is to be taken other than literally.
In Finding Number 87 (at page 57) the ASLB concluded that the NRC staff had " adequately applied, weighed and balanced the five factors set forth in the [9/75 Notice]. This in turn is based upon the ASLB's conclusions in Findings Number 85 and 86 (at pages 54-56).
Oregon contends that the ASLB erred in three respects:
14 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
9 (1)
The evidence establishes that neither the staff nor the applicants considered cumu-lative impacts thoroughly; (2)
The evidence does not establish that de-ferring the authority to store more than 4/3 cores in the SFP would result in sub-stantial harm to the public interest; (3)
The record establishes that the staff did not weigh and balance the five factors.
B.
Statement of Facts In its 9/75 Notice the NRC announced that, because of long-range policy considerations, it would prepare a GEIS on the handling and storage of spent light water power reactor fuel.
40 Fed Reg at 42802.
At the same time, the NRC refused to impose a " moratorium" on all licensing actions intended to ameliorate possible shortages of spent fuel storage capacity pending completion of the GEIS.
The NRC stated that the five factors which it analyzed in refusing to impose a moratorium would be " applied, weighed and balanced" in any environmental impact statement or environmental impact appraisal prepared in "any licensing action intended to ameliorate a possible shortage of spent fuel storage capacity during this interim period.
(The " interim period" refer-enced is that between September 16, 1975 and completion of the GEIS.)
The NRC stated in the 9/75 Notice, as one of its objectives, that the GEIS should not ".
. serve as a justification for 15 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
a fait accompli."
40 Fed Reg at 42802.
The record reflects that the NRC had authorized increases in on-site spent fuel storage in approximately 24 instances.
(Tr 5698).
Approximate-ly 20 more similar requests were pending before the NRC.
(Tr 5701).
If 44 requests are granted, this would represent increasing storage capacity at two-thirds of the nation's com-mercial nuclear reactors.
(Tr 5701).
The record reflects that in only three of these situations was reactor shutdown imminent.
(Tr 5698-5702).
Evidence was submitted in this proceeding regarding the five factors.
As to the two factors in controversy, and as to the balancing process, the evidence can be summarized as follows:
(1)
Consideration of Cumulative Impacts PGE, in Ex. 2, and even more particularly the NRC in Exhibit 1-B, evaluated the environmental impacts of the mod-ification as they relate to the locale of the Trojan Plant.
However, cumulative environmental impacts were in fact overlooked.
NRC Exhibit 1-B itself is silent regarding cumulative impacts.
(See p. 25 thereof).
Staff witness Donohew testified (follows Tr 5558) that, to the extent there are a number of licensing actions similar to that proposed for Trojan, there will be impacts experienced throughout the nation which would not otherwise be exper-ienced.
(Tr 5561).
Cumulative impacts, taken in this sense, were not explored by the NRC staff.
16 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECIS'.ON
O e
(2)
Consequences of Deferral The Board based its ultimate conclusion upon the fact that full core discharge capability ("FCDC") might be lost in the spring of 1979, creating a situation whereby the Trojan Plant might have to shut down if ad hoc necessary inspections and maintenance required discharge of a full core from the reactor.
The Board accepted that there was a 50% probabil-ity that Trojan might have to discharge a full core between the spring of 1979 and 1982.
PGE was unwilling to testify that deferral would result in substantial harm to the public interest.
Through witness Frewing's (follows Tr 5638), PGE stated that deferral could result in substantial harm to the public interest.
Mr. Frewing testified that he used the word "could" intentionElly.
(Tr 5654).
Trojan's first refueling occurred in April 1978--
one-third of a core of spent fuel was placed into the original SFP racks.
PGE presently intends to follow an annual l
refueling cycle.
Assuming such a cycle, the existing SFP racks will be completely filled in the spring of 1981 and PGE would have to shut Trojan down in the spring of 1982 due to a lack of spent fuel storage capacity on site.
(Tr 5647).
With its existing racks, PGE will lose FCDC in the spring of 1979.
(Tr 5647).
PGE's proposed SFP modification would permit the on-site storage of three and one-third cores of spent fuel.
17 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
8 If FCDC capability is not preserved, this would result in a full spent fuel pool after the spring 1987 refueling.
If FCDC is to be maintained, the modofied SFP would be " full" after the spring 1984 refueling.
There was testimony by the applicant to the effect that between spring, 1979 (when FCDC would be lost if the existing racks remained in place) and 1982 (when the existing racks would be filled, assuming an annual re-fueling schedule), there would be a greater than 0.50 chance that Trojan would have to discharge a full core.
(Owens, Tr 5644, 5649-51).
The probability assigned by PGE in any one year was 0.24.
As noted above, the NRC's GEIS on the handling and storage of light water power reactor spent fuel was scheduled to be completed in late summer or early fall of 1978, 3-1/2 years prior to the time at which PGE must shut down Trojan due to a filled SFP, and six months prior to the date by which FCDC will be lost.2 (3)
Failure to Weigh and Balance the Five Factors A review of PGE 1013 (PGE Ex. 2) and the staff's EIA (Ex. 1-B) demonstrates that neither weighed and/or bal-anced the five factors.
It is not possible to discern how either PGE or the staff did anything more than provide evidence directed to each factor individually.
- Moreover, the staff, in its analysis, tended to simply accept con-clusory statements of the applicant, and performed only 18 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT
^" ITS EXCEPTIONS TO THE INITIAL DECISION
a limited inuependent substantive analysis.
- See, e.g.,
Tr. 2224-25, 5735-37, 5776-5788.
C.
Argument 1.
Since it cannot be concluded that deferring auth-orization to use the modified SFP to store more than 4/3 cores would result in substantial harm to the public interest, authorization should be withheld pending completion of the GEIS.
Our first argument rests on one fundamental premise:
that the NRC meant what it said in its 9/75 notice.
Oregon continues to read that notice (despite contrary evidence in the form of numerous operating license amendments author-izing increased on-site spent fuel storage capacity) as constituting a commitment by the NRC to proceed circum-spectly with individual licensing actions pending comple-tion of its GEIS.
The language in the 9/75 Notice expressing an intention that the GEIS would not serve to justify a fait accompli, coupled with the language therein regarding the consequences of deferring individual actions creates, in our opinion, the following presumption:
In the interim between September 1975 and completion of the GEIS, individual licen-sing actions to increase on-site storage of spent fuel should be denied, unless such denial would result in substantial harm to the public interest by causing reactor shutdown.
19 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
Oregon relies upon Natural Resources Defense Council v.
NRC, 539 F2d 824 (2d Cir 1976), cert. granted 430 US 944 (1977), vacated and remanded 434 US 1030 (1978) (memorandum opinion) as justification for holding the NRC to a literal read-ing of the 9/75 Notice.
Even if not necessarily mandated by the National Environmental Policy Act (42 USC S 4321 et. seq.),
the NRC's decision to review expanded on-site storage of spent fuel on a generic basis represents a commitment to complete the generic review prior to implementing individual solutions.
By proceeding with what has turned out to be wide-scale " interim licensing", the NRC in effect has breached its commitment.
Under the facts of this specific proceeding, however, the commitment can and should be honored.
It is literally impossible to conclude from the evidence in this record that substantial harm would result if action upon PGE's application were deferred until the GEIS was completed.
PGE itself was not even willing to make such a claim.
Given the inability to legitimately conclude that deferral would result in substantial harm to the public in-terest, the ASLB could not properly find that the license amendment should be granted at this time.
2.
The NRC staff failed to weigh and balance the five factors contained in the NRC's 9/75 Notice.
Oregon's second argument assumes that the 9/75 Notice is to be read other than literally.
That Notice requires that the five factors be " weighed and balanced", as well as
" applied".
One factor -- cumulative im' pacts -- was not con-20 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
9 sidered thoroughly.
As to another factor -- the consequences of deferral -- the ASLB decision is not supported by the record.
A review of PGE's application (S,ee PGE Ex.
2, Ch. 6 and Frewing Test. pp 5-8 following Tr. 5638) and of the NRC staff's environ-mental impact appraisal (See NRC Ex.
1-B, pp 23-26) demands a conclusion that the factors were never weighed and balanced.
The staff's failure to weigh and balance the five factors is compounded by the fact that it did virtually no independent analysis with respect to the fifth factor.
This indicates that in performing its duties under 10 CFR 51.7(b), the staff did not meet the standards of the NRC's Pilgrim case.
Boston Edison Company (Pilgrim Nuclear Generating Station, Unit No. 2),
ALAB-479, May 25, 1978 (7 NRC
).
Given the context of individual licensing actions afforded by the 9/75 Notice, the NRC staff should be expected to provide a "hard look" at whether deferral would harm the public interest.
The staff did not do so here.
D.
Conclusion Re Exception No. 7 Whether the 9/75 is to be taken literally or not, author-ization to use the modified Trojan SFP for storage of more than 4/3 cores of spent fuel should not be granted at this time.
There is no evidence to justify a conclusion that deferring such authorization pending completion of the GEIS would result in substantial harm to the public interest.
To the contrary, such a deferral would constitute the honoring by the NRC of a commitment accepted at face value by Oregon.
21 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
IV REQUEST FOR ORAL ARGUMENT Pursuant to 10 CFR 2.763, Oregon requests that the Appeal Board panel hold oral argument on its exceptions.
Respectfully submitted, bl -
<c uw RICHARD M.
SANDVIK Assistant Attorney General Ac ~ h J.
AJoh hf, [
r FRANK W.
OSTRANDER, di'.
/ A/M Assistant Attorney General Of Counsel to the Oregcn Energy Facility Siting Council and Oregon Department of Energy 22 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
FOOTNOTES 1
Oregon believes that under 10 CFR 2.743(i) the Appeal Board can take notice of the fact that Trojan has remained shutdown since the March 1978 refueling, during the pendency of a proceeding concerning the seismic adequacy of the walls of the plant's control building.
In the matter of Portland General Electric Company et al., (Trojan Nuclear Plant) -- Docket No. 50-344 (Control Building).
It is anticipated that an interim ASLB order authorizing opera-tion of Trojan will not be issued until December 20-22, 1978, at the earliest.
Thus, a spring 1979 refueling is somewhat unlikely.
2 To the best of our knowledge,the final GEIS has not yet been issued.
23 - INTERVENOR STATE OF OREGON'S BRIEF IN SUPPORT OF ITS EXCEPTIONS TO THE INITIAL DECISION
CERTIFICATE OF SERVICE I hereby certify that on December 13, 1978, copies of the foregoing "Intervenor State of Oregon's Brief in Support of Exceptions to the Initial Decision" were served upon the parties of record listed below, by then depositing in the United States Post Office at Portland, Oregon, full, true and correct copies thereof, in sealed envelopes with postage pre-paid, addressed to the said parties of record listed below:
Sheldon J. Wolfe, Esq., Chairman Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Frederick P. Cowan, Member Apt. B-125 6152 N. Verde Trail Boca Raton, Florida 33433 Mr. Frederick J.
Shon, Member Atomic Safety and Licensing Board U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Joseph R. Gray, Esq.
Counsel for NRC Staff U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Columbia County Courthouse Law Library, Circuit Court Room St. Helens, Oregon 97051 Atomic Safety and Licensing Appeal Board Panel U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Page 1 - CERTIFICATE OF SERVICE
hs, Docketing and Service Section Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Susan M. Garrett 7325 S.E. Steele Street Portland, Oregon 97206 Mr. David B. McCoy 348 Hussey Lane Grants Pass, Oregon 97526 Sharon S. McKeel P.O. Box 8786 Portland, Oregon 97208 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. Warren Hastings Attorney at Law Portland General Electric Company 121 S.W.
Salmon Street Portland, Oregon 97204 John J.
Haugh Attorney at Law 555 Benjamin Franklin Plaza One S.W. Columbia Portland, Oregon 97258 Alan Rosenthal, Chairman Atomic Safety & Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 L d E kt. D l RICHARD M. SANDVIK Assistant Attorney General S
%& jf FRANK W.
OSTRANDER, J R.'
/ /W 'S Assistant Attorney General Of Attorneys for Intervenor State of Oregon Page 2 - CERTIFICATE OF MAILINC