ML20054H602
| ML20054H602 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 06/17/1982 |
| From: | Curran D, Jordan W HARMON & WEISS, NEW ENGLAND COALITION ON NUCLEAR POLLUTION |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML20054H599 | List: |
| References | |
| NUDOCS 8206240252 | |
| Download: ML20054H602 (39) | |
Text
'
I UNITED STATES OF AfiERICA ~ qi
" i:42 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ..
)
In the Matter of )
)
PUBLIC SERVICE COMPANY OF NEW ) Docket Nos. 50-443 HAMPSHIRE, et al., ) 50-444
)
(Seabrook Station, Units 1 and 2) )
)
NECNP'S REPLY TO THE RESPONSES BY THE APPLICANT AND THE NRC STAFF TO NECNP'S CONTENTIONS Introduction At the outset, NECNP wishes to make clear its position on the status of NRC Regulatory Guides, which have figured prominently in the objections by-ths Applicant and tne NRC Staff to NECNP's contentions. As we stated in the introduction to our contentions, we do not rely upon Regulatory Guides as legal requirementc. Rather, we rely on the failure to comply with a Regulatory Guido ac sufficient factual basis for the admission of a contention. Although Regulatory' Guides are not enforceable as regulations, they
. represent the Staff's position on acceptable methods of implementing NRC regulations. They thereby provide a factual basis for a contention that the threshold level of cafety that they represent must be met in some way in y -4' 8206240252 820617 PDR ADOCK 05000 4 0
2 order to achieve compliance with the underlying regulation.
Alternative methods of satisfying the regulations are acceptable only "if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission." Introduction to U.S. N.R.C.
Regulatory Guides. Consequently, when the Applicant does not meet the requirements of a Regulatory Guide and alternative measures are either not presented or insufficiently documented, basis for a contention on the subject exists. We have reworded the relevant contentions to make it clear that we do not rely on Regulatory Guides as legal requirements.
I.A.1 Environmental Qualification Both the Applicant and the Staff would restrict NECNP's environmental qualification contention to a claim of non-compliance with GDC 4 as implemented by CLI-80-21, 11 NRC 707 (1980). However, as noted by the Commission in that decision, CLI-80-21 does not incorporate the lessons learned at Three Mile Island, where equipment previously thought to be environmentally qualified did not survive the accident environment. In addition to the existence of a harsher environment than the equipment was qualified for, the accident lasted much longer than the matter of minutes the equipment was qualified to function. Further, as a result of the unexpected extended duration, the equipment was required to withstand environmental conditions such as
i 3 submergence in water for which the equipment had not been qualified. Thece developments constitute a factual basis for NECNP's contention that GDC-4 itself requires that the Applicant not only satisfy CLI-80-21 but also show that equipment can withstand the conditions and lengthy duration of an accident such as occurred at Three Mile Island. We note that data on the qualified life and duration of functioning under accident conditions are still missing from the Applicant's FSAR.
The Staff also argues that NECNP must specify which particular pieces of equipment are not qualified. NECNP responds that all of Applicant's equipment is unqualified insofar as it has concededly been qualified to an outdated standard. FSAR at 1.8-33. In addition, the burden is on the Applicant to show which equipment must be and has been qualified. Since the Applicant has not provided cuch information, as would be required by the proposed rule on environmental qualification, 47 FR 2876, NECNP cannot be required to identify particular nonconforming equipment.
I.A. 2 Environmental Qualification of Electric Valve Operators Ac in the casc of contention I.A.1, NECNP maintains that electric valve operators must be environmentally qualified to meet GDC 4 as implemented by CLI-80-21 and as may be further required to provide a reasonable assurance
i 4
that the equipment can survive an accident environment of the harshness and duration experienced at Three Mile Island Unit 2.
I.A.3. Environmental Qualification for Hydrogen Burns The Applicant and Staff object to NECNP's Contention I.A.3. on the ground that matters related to hydrogen control arc covered by 10 CFR 50.44. Their objection is misplaced.
This contention does not challenge the adequacy of hydrogen control at Seabrook, but asserts that a higher level of hydrogen release must be concidered for the purpoce of environmental qualification. The hydrogen control requirements of Section 50.44 may differ from hydrogen release assumptions for the purpose of environmental qualification, just as the 5% standard of 50.44 differs from the 17% assumption for the purpose of ECCS acceptance criteria under 10 CFR 50.46(b). The Commission has given no indication that those acsumption must be the same for other purpoces.
In addition, the Commission's recent proposed rule on hydrogen control, 46 FR 62281, makes it clear that the Commission views the two questions of hydrogen control and environmental qualification for hydrogen burn as ceparate and distinct. The very fact that the Commission proposes to require environmental qualification for hydrogen burn
i 5
establishes that the Commission is making provision for a hydrogen release greater than that for which control measures are required. NECNP simply asserts that the amount to be assumed must be equivalent to the amount released at Three Mile Island.
I.B.1 Environmental Qualification of Mcchanical Equipment In responding to the criticisms by the Applicant and the Staff concerning this contention, NECNP first would like to correct the typographical error in the first paragraph, which should have referred to "GDC 34" instead of "GDC 3". GDC 34 requires that a system to remove residual heat be provided. Residual heat removal systcms are considered to perform safety functions because they transfer decay heat from the reactor core; therefore they should be safety grade and environmentally qualified under GDC 4.
Applicant and Staff object to this contention on the ground that there is no regulation requiring environmental i
qualification of steam dump valves, turbine valves, and the steam dump control system. However, NRC regulations contain no list of any equipment which must be environmentally qualified--the requirement applies to " structures, systems 1
and components important to safety." GDC 4. If a system is 3
i considered important to safety, it must be designed and
. 6 qualified to perform its safety function and to accomodate the effectc of and to be compatible with the environmental conditions associated with the most adverse credibic accidents, including loss of coolant accidents.
NECNP maintains its position that the Three Mile Island accident showed that the steam generator heat dumping system ic "important to safety" because it may be relied upon for residual heat removal, and that therefore it must meet GDC 4 and GDC 34. For clarity's sake, NECNP would reword the contention as follows:
The Applicant has not satisfied the requirements of GDC 4 and GDC 34 in that all systems required for residual heat removal, such as steam dump valves, turbine valves, and the entire steam dumping system are not safety grade and environmentally
, qualified.
In cupport of this contention, NECNP also notes that the development of additional reliable heat removal capacity is a new Unresolved Safety Issue which must be resolved at the licensing stage. NUREG-0705, Task A-45.
I.B.2. Duration of Environmental Qualification The Applicant has mado no objection to this contention, but the Staff objects that GDC 4 does not require that time durations for environmental qualification be identified.
i
- . _ . .. -. - _ - . _ _ _ _ _ . _ _ . . _ . , _ , , . . _ _ _ _ . . . _ , . _ . _ _ _ ~ - - . ~ . _ _ _ . _ . - - . _ _ , _
i 7
NUREG-0588, made enforceable by CLI-80-21, specifies that: 1 Equipment that is used to perform a necessary cafety function must be capable of maintaining functional operability under all service conditions postulated to occur during the installed life for the time it is required to operate. (emphasic added)
Id. at 1. When the NRC issues a 40-year operating license, it must be able to provide reasonable assurance that the plant can be safely operated for the entire license term, not just a few months or a few years. The Commission has no means of providing that reasonable assurance if there is no evidence that the plant equipment can be expected to function safely for that period; or how long it can be expected to function and how it will be replaced or repaired. The Applicant has provided no information on the qualified life of environmentally qualified equipment at the Seabrook facility, depriving'the public of a reasonable assurance that the plant can be operated safely during its lifetime.
I.C. Environmental Qualification--Emergency Feeduater Pump-House HVAC The Applicant has responded to this contention by listing parts of the Emergency Feedwater Pumphousc HVAC which are environmentally qualified, and claiming that no regulation exists which requires environmental qualification of more equipment. As we stated in response to the objections
8 to' contention I.B.1, NRC regulations do not list specific components which must be qualified, but rather require environmental qualification of systems important to safety.
The HVAC system must be relicd upon to maintain a certain temperature in the event of an accident in ordcr to assure the safe operation of the emergency feedwater system.
It must therefore be environmentally qualified so that no single failure prevents it from performing its safety function. GDC 4. The Applicant has listed which components in the emergency feedwater pump house . IIVAC are environmentally qualified--but has failed to mention which components are not. For instance, although supports for cables are environmentally qualified, the Applicant does not indicate 4
whether the cables themselves have been environmentally qualified, although they are essential to the safety function. The apparent failuro to qualify the cabics constitutes a factual basis for the contention that equipment in the emergency feedwater pumphouse HVAC has not been enviornmentally qualified. At this point, it is not NECNP's responsibility to guess what other equipment that has not been environmentally qualified is part of the cmergency feedwater pumphouse HVAC. NECNP has sufficient basis in the fact that the Applicant is unable to vouch for the environmental qualification of the entire UVAC system.
1
[
.,. . . _ . _ . . _ . _ _ _ _ _ , _ . - - . . _ , . __.__._..._ .- _.,_,.__._ --__.___.~. . ~ _ . _ - _ _ _ _ . _
. 9 I.D.l. Testing of Reactor Vescel Welds In response to the objection by the Applicant and the Staff that NECNP improperly ceeks mandatory compliance with a Regulatory Guide, NECNP has reworded this contention as follows:
The Applicant has not complied with GDC 1 in that it has not fully implemented Regulatory Guide 1.150 with respect to ultraconic testing of reactor vecsel welds during preservice and inservice examination, nor does it indicate any alternative means by which a level of safety equivalent to that achieved by Reg. Guide 1.150 will bc met.
I.D.2 Testing of Protection System Actuation Functions Applicant and Staff object to this contention on the
~
ground that Rep.~Guidc^1.22~is~not a mandatory standard.
NECNP therefore has reworded this contention in that light.
In addition, we have altered the wording slightly to clarify l our understanding of Reg. Guide 1.22 that where testing at pcwer is not possible, the Applicant must test actuation devices by alternative means as described in the Reg. Guide i
and provide, " compelling justification" where actuated equipment l
l 1s not tested during operation. NECNP has reworded this l
contention, adding the support of NUREG-0737, as follows:
The Applicant has not satisfied GDC 21 by complying with Reg. Guide 1.22 regarding the testing of protection systems and actuation devices, or providing a suitable alternative means of achieving the same level of' safety. In particular, the Applicant does t
I
10 not provide for the testing at full power of twelve safety functions (see FSAR at 1.8-9),
justify that omission, or provide for other reliable means of testing them as required by Rog. Guide 1.22. This also violates Task II.D.1 of NUREG-0737, which requires testing of reactor coolant system relief and safety valves under operating conditions for design-basis transients-and accidents.
I.D.3 Testing of Leakage Detection System In response to the Applicant's and Staff's objection that NECNP sceks mandatory compliance with a Regulatory Guide, NECNP rewords this contention as follows:
The Applicant has not provided a reasonable assurance that the leakage detection system for the Scabrook reactor will operate when needed because not all of the system is to bc tested during plant operation, as required by GDC 21. The Applicant does not comply fully with Reg. Guide 1.22, and it has not presented any alternative means of satisfying GDC 21 which will achieve the same level of safety ac provided by Rog. Guide 1.22. Only the airborne radioactivity
- detector has the capacity to be tested during power operation. FSAR at 1.8-17. The Applicant thereby also fails to satisfy GDC 30, which requires the development of adcquate leakage detection i systems.
The Applicant also objects to-thic contention on the basis that NECNP has not identified all components of the leakage detection system which must be tected at full power in order to demonstrate compliance with GDC 21. NECNP bases this contention in part on the Applicant's acknowledgement that it has not complied with all the terms of Reg. Guide 1.22,
k 11 ;
in conjunction with its failure to identify other, equivalent means of satisfying GDC 21. NECNP's citation to Reg. Guide 1.22 provides sufficient specificity. The contention extends to all aspects of the leakage detection system that are covered by Reg. Guide 1.22. The Applicant, not NECNP, knows and can identify those components of tite leakage detection system which do not meet Reg. Guide 1.22.
I.D.4 Periodic Tesning of Electric Powcr and Protectic.i Systems In response to the Applicant's and the Staff's objection that NECNP may not seek mandatory compliance with a Regulatory Guido, NECNP rewords this contention as followc:
The Applicant has not complied with GDC 21 in that it has not complied with Reg. Guide 1.118 Rev. 2, nor has it identified any equivalent alternative means of catisfying GDC 21. In particular, the Applicant indicates compliance with an outdated 4 ctandard, IEEE 338-1975, which has been superseded 2
by IEEE 338-1977. Furthermore, the Applicant improperly assert: that it need not comply with IEEE 338-1975 whenever the standard staten that an
. action "chould" be taken or a requirement "should" be met. All the provisions of the IEEE standard should be treated as mandatory unless the Applicant can show an alternative means of achieving the same level of safety.
4 I.E. Reactor Coolant Pump Flywheel Integrity This contention does not, as claimed by the Applicant and the Staff, elevate a Regulatory Guide to a regulation.
The contention ascerts that an applicant must meet the l Regulatory Guides which implcment the General Design Criteria, or provide an equivalent alternativo.
, - - - - , - < - - .--,w, , .o- ,-v- r , w y r-y-,-. v-- -2 . a.- ,, - - < = w -
12 4
I.F. Diescl Generator Qualification In response to the objection of the Applicant and the Staff that NECNP secks mandatory compliance with a Regulatory Guide, NECNP rewords this contention as follows:
The Applicant has not mot the requirementc of GDC 17 or Criterion III, App. B in that it has not implemented Reg. Guide 1.9 or demonstrated any '
equivalent means of satisfying those requirements.
In particular, the Applicant does not indicate compliance with IEEE 323-1974 as required by Reg.
Guide 1.9, or indicate reasons for noncompliance and alternative means of achieving the same level of safety as provided for in the Reg. Guide, i It is NECNP's position that lack of compliance with a Regulatory Guide, absent identification of alternative means of catisfying the applicable regulatory requiremont, is sufficient basis fcr a contention charging noncompliance with the requirement. It is the Applicant's responsibility, not NECNP's, to identify discrepancics and explain them.
I.G. Pressure Instrument Reliability NECNP disagrees with the criticisms of both the Staff and Applicant with respect to this contention. In particular the Staff's objection is internally inc2nsistent in arguing that our conclusory language must be deleted, while appending similar conclusory language to the version proposed by the Applicant. Nonetheless, NECNP would accept the following revision, which is based on the versions of the Staff and Applicant:
NECNP contends that there is not reasonable assurance that the public health and safety will be protected in light of the RCS wide range .
- pressure instruments being utilized at ceabrook,
1 l
13 which cannot be relied upon to provide accurate information. Reliance upon the instruments could result in inappropriate operator actions or premature or late tripping of RCS pumpo during the course of a small break loss-of-coolant accident.
I.H. Decay Heat Removal Capacity The Applicant and the Staff. object that there is no regulatory requirement for larger heat exchanger capacity at Seabrook. The basis t'or this contention lies in Tack A-45, a new Unresoived Safety Issue described in NUREG-0705. A critical element of this unresolved safety issue is "the adcquace of existing shutdown decay heat removal requirements." NUREG-0705 at A-2. The Applicant has stated in its FSAR that its heat exchanger capacity is smaller than that of other comparable reactors, and that cooldown time will be longer. FSTR at Table 1.3-1, Sheet 2.
This statement indicates a conflict with the directive of Task A-45 which must be resolved before a license may issue.- Virginia Electric and Powcr Co. (North Anna Nuclear Power Station, Units 1 and 2), ALAB-491, 8 NRC 245 (1978).
NECNP has raised a viable inference that the issue of adequate decay heat removal has not been resolved by the Applicant at the licensing stage.
14 I.I Inadequate Provisions for Achieving Cold Shutdown The Applicant and Staff object to this contention on the ground that NECNP has not identified which systems, structure and components necessary to bring the plant to cold shutdown must meet the General Design Criteria pertinent to cold shutdown. In addition, the Applicant contends that NECNP has improperly elevated a Regulatory Guide to the status of a regulation.
IE Bulletin No. 79-OlD (October 24, 1980) indicates that:
The Staff position on [ cold shutdown] is that the licensee must identify and environmentally qualify the equipment needed to complete one method (path) of achieving and maintaining a cold shutdown condition. (emphacis added)
This position indicates that the Staff believec there is no other equivalent means of satisfying the applicable design criteria, and concurs with NECNp that the Applicant is required to identify and qualify at least one path to cold shutdown. This rcquirement is also reflected in correspondence from the Staff to the Applicant. See leters from Frank J.
Miraglia, Chief of Licensing Branch No. 3, to William C.
Tallman, Chairman of Public Service Co. of New Hampshire, dated April 21, 26, and 28, 1982.
Contrary to the Applicant's and Staff's position with respect to specificity it is the Applicant's duty, not NECNP's, to identify a path to cold shutdown and shou
l 15 that the required equipment meetc the applicable design criteria. NECNP does not contend that the Applicant must use a particular path to cold shutdown, only that the Applicant must identify and environmentally qualify at least 1
one such path. The specifics may be chosen by the Applicant.
NECNP has met its threshold burden of showing that no such path exists today.
I.J. Sabotage NECNP will shortly present the qualifications of several expert witncsses to testify on the Applicant's security plan. As acknowledged by both the Applicant and the Staff, NECNP cannot frame a proper contention until its cxport: - ----
have had an opportunity to review the security plan.
i i I.K. Instrumentation Both the Applicant and the Staff object to this contention as premature, because the Applicant has not yet selected Post-Accident Monitoring instrumentation. The Applicant proposes that NECNP should submit a " late filed contention"
, when description of the PAM instrumentation is cubmitted.
NECNP would object to any requirement that a PAM contention must satisfy the NRC standard for late filed contentions.
9 To do 3o would amount to penalizing NECNP for the Applicant's failure to complete the FSAR on time. NECNP therefore proposes to handle this contention in either of the following two b
, _ _ _ _ ,y , _ _ _
16 ways,-at the Board's pleasure:
- l. NECNP will maintain this contention and either amend it or drop it when the PAM instrumentation is' selected; or
- 2. NECNP will drop the contention now, with the option of presenting an instrumentation contention without prejudice when the PAM instrumentation is selected.
1 I.L. PORV Flow Detection
- The basis for this contention, NUREG-0578'c requirement
'that the plant-design must provide for direct indication of power-operated relief valve position, has been incorporated
- as Clarification Item II.D.1 into NUREG-0737 which identifies current TMI-related licensing requirements. The acoustic accelerometer system does not provido the required direct indication. We note that the deadline for compliance with this requirement was January 1, 1980.
I.M. Fire Protection In response to the objections of the Applicant and the 4 Staff, NECNP has reworded this contention as follows:
The Applicant's fire protection system does not meet the requirements of GDC 3 as implemented by the Commicsion in CLI-80-21.
I The basis for this contention lies in the Applicant's admission that it has not satisfied all terms of Branch Technical Position APCSB 9.5-1. FSAR at 1. 8-4 3. According to the Commission in CLI-80-21, BTP 9.5-1 Appendix A and the I
3---.,,-y ,-+- -e - , --, , w-,
i 17 Commission's proposed rule on fire protection " define the essential elements for an acceptable fire protection program !
or nuclear power plants docketed for construction permit prior to July 1, 1976, for demonstration of compliance with Cencral Design Criteria 3..." CLI-80-21, 11 NRC at 718.
Accordingly, noncompliance with Appendix A of BTP 9.5-1 constitutes sufficient basis for a contention that the applicable General Design Criterion has not been satisfied.
In particular, NECNP asserts that the following components or systems do not meet GDC 3 because they do not satisfy BTP APCSB 9.5-1, App. A:
A. General Guidelines for Plant Protection
- 1. Building design
- a. cable spreading rooms
- b. floor drains
- c. floors, walls and ceilings
- 2. Control of Combustibles
- a. reactor coolant pump lube oil system
- 3. Electric Cable Construction, Cable Trays and Cable Penetrations i'
- a. cable spreading rooms
- b. cable trays outside cable spreading rooms
- c. control room cabling
- 4. Ventilation
- a. discharge of products of combustion
- b. power supply and controls
- c. protection of charcoal filters
- d. stairwells
- e. smoke and heat vents l
l i
_. = - .. - . - . . . - . . . . .
18 i I
- 5. Lighting and Communications i a. fixed emergency lighting j B. Fire Detection and Suppression
- 1. Detection-alarm and annunciation 4
- 2. Water Sprinkler and Hose Standpipo Systems
- a. sprinkler and standpipe layout
- b. supervision of valves C. Guidelines for Specific Plant Areas
- 1. Primary and secondary containment--normal operation
- 2. Control room
- 3. Cable =preading room
- 4. Suitchgear rooms
- 5. Remote safety related panelc
, 6. Dicsel generator areas
- 7. . Diesel fuel oil storage areas 8.- Safety related pumps
- 9. New fuel area
- 10. Spent fuel pool area
- 11. .Radwaste building
- 12. Decontamination areas D. Special Protection Guidelincs
- 1. Wolding and cutting, acetylene-oxygen fuel gas systems
- 2. Storage-areas.for dry ion exchange resins
. . _ . . - . . ~ _ _ . , ..- _ ,_. _ . _ . , . _ _ - _ . . _ - _. _.- . __ _ ..
19 According to the Applicant, the above-listed cystems and components do not meet BTP APCSB 9.5-1, App. A. Letter from PSCNH to NRC dated August 29, 1977, Table 3.
I.N. Solid Waste Dispocal In response to the suggestions of the Applicant and the Staff, NECNP has reworded this contention ac follows:
The Applicant hac not provided a means to handle radioactive solid waste during normal reactor operation, including anticipated operational occurrences, as required by GDC 60.
I.O. Emergency Feodwater I.O.l. The Applicant and the Staff object to this contention on the ground that there is no regulatory requirement for the design changes requested by NECNP.
NECNP responds that the Preamble to 10 CFR Part 50, Appendix A, ac quoted in thic contention, and the fact that the discharge header is pressurized when in use, provide cufficient basic for the assertion that the emergency feedwater system should be single failure-proof with respect to a rupture in the diccharge header and that the design of the system should protect against passive failurcs. .
I.O.2 NECNP takes this opportunity to correct error in this contention. The first sentence should refer to a single failure instead of a locs of offcite power, and is
I -
20 corrected as follows:
The emergency feedwater cystem is inadequate in that a break in the common discharge header between Valve 65 and Valve 125 (sec FSAR, Figure 6.8-1), coupled with a single failure, would result in a loss of'feedwater to all steam generators. The Seabrook design must be revised to provide an emergency feedwater cystem that complies with GDC 17 with respect to the high-energy piping in the discharge header.
The basis for thic contention remains the same, i.e. that the common discharge header must be qualified as pressurized because under the conditions it will be called upon to operate in, it will be pressurized. The Applicant argues that the EFS is not high or moderate energy piping because it will not be precsurized during " normal operation" of the plant. NECNP submits that normal operating conditions, under which the discharge header will be empty, are irrelevant to the quection of whether the EFS can be relied upon during an accident, when it will be filled with pressurized water.
The Applicant cannot provide a reasonable ascurance that the EFG will operate safely, given this faulty reasoning.
I.P. Human Engineering The Applicant and the Staff object that NECNP has asserted no regulatory requirement in support of thic contention. The regulatory requirement is found in NUREG-0737, which sets forth current TMI-related licensing standards.
Clarification Item I.D.1 requires licensees to perform control room design reviews "to identify significant
21 human factors and instrumentation problems..." Among the aspects to be emphasized in the review are the " grouping of displays and the layout of panels", and " improvements in the safety monitoring and human factors enhancement of controls and control displays." Id. The placement of a monitoring instrument in a position where it is imposcible.
to read it without leaving one's position at the control panel is a classic example of poor human engineering which Task I.D.1 is intended to correct.
I.Q. Oystems Interaction The Applicant and the Staff object that there is no regulatory basis for this contention. Systems interaction is an "unrecolved safety insuc" identified in NUREG-0510.
In Virginia Electric and Power Co. ,( North Anna Nuclear Power Station, Units 1 and 2) ALAB-491, 8 NRC 245 (1978), the Appeal Board ruled that an applicant for an operating license must demonstrate that each applicabic generic safety issue has been resolved for that particular reactor, or demon =trate the existence of measures employed at the plant to compensate for the lack of a solution to the problem.
The persistence of this issue, which has not been resolved on the generic level, requires that it be dealt with on the individual plant licensing level.
In addition, 10 CRR 50.55a(h) incorporates-IEEE 279-1971, which requires, inter alia, the analysic of systems interactions.
22 The study of systems interactions ic also required by GDCs 20, 22, and 24, which address design bases for procection systems.
NECNP is not required to state in this contention all the ways in which the Applicant's analysis of systems interaction is deficient. Our assertion that the Applicant's particular systems interaction review is insufficient is based on the Staff's lack of satisfaction with it (letter from Robcrt L. Tedesco, Asst. Director for Licensing, NRC, to William C. Tallman, Chairman, Public Servico Company of New Hampshire, dated August 19, 1981); and on the NRC-commissioned " state of the art" review which concluded that adequate methods of analyzing systems interaction simply do no exist. See NUREG/CR 1901, NUREG-1896, and NUREG/CR 1859.
NECNP believes this contention'is acceptable as drafted since it is clear the entire systems interaction analycis is faulty. NECNP should not be required to identify-particular deficiences when overall failure is acknowledged.,
However, if this contention ic rcjected, NECNP would assert an alternative comparable to that accepted by the Licensing Board in Long Island Lighting Co. (Shoreham Nuclear Power Station Unit 1), Docket No. 50-322, filed March 15, 1982, s1. op. at 12.
. 23 I.R. Hydrogen Control The Applicant and the Staff object to this contention on the ground that NECNP has not postulated a " credible accident scenario" resulting in hydrogen releaces above the limits contemplated by 10 CFR 50.44, ac required by the Commission in Metropolitan Edison-Co. (Three Mile Island Station, Unit 1), CLI-80-16, 11 NRC 674 (1980). NECNP cubmits that the Three Mile Island case is not consistent with the Commission's subsequent decision in Duke Power Co.
(William B. McGuiro, Nuclear Station Units 1 and 2), CLI 15, 14 NRC 1, 2 (1981), where the " likelihood of an accident that would lead to generation of hydrogen in excess of the design limits in 10 CFR 50.44 and the effectiveness of mitigative measurec were " critical" to the Commicsion's deliberations on the effectiveness of an operating license.
The Commission's recognition of the inadequacy of 10 CFR 50.44 therefore given NECNP a sufficient basis for this contention. Further more, NECNP believes that the Three Mile Island accident adequately _ demonstrated the " credibility" of a hydrogen release in excess of currently ascumed proportions, and that further exposition of other specific accident scenarioc is an unnecescary exercise. Accordingly,,NECNP urges that this contention be adopted as proposed.
However, should the Board deny the contention as originally i
propoced, NECNP asserts the following contention, cupported i
i f
=
Y
24 by a credible accident scenario:
NECMP contends that the hydrogen control system at Seabrook is inadequate to protect the public health and safety in that it would protect againct the hydrogen produced by a metal water reaction involving only 1.5% of the fuel cladding and in that the manual operation of the hydrogen recombiners and other hydrogen control equipment has not been demonstrated to be adequate to assure that large amountc of hydrogen can be safely accommodated without a rupture of the containment and a release of substantial amounts of radioactivity into the environment.
As a credible accident scenario, NECNP asscrts the following:
- 1. a pipe break in the reactor coolant pressure boundary causes a LOCA, as defined by 10 CFR
- 50. 46 (c) (1) .
- 2. failure of the ECCS to maintain coolant inventory.
The cause of this failure may be: electrical or mechanical component failure; common mode failures resulting from the LOCA: design deficiencies which underminc ECCS effectiveness; and/or operator. error.
- 3. The Zircaloy fuel cladding melts; the zirconium reacts with water, liberating hydrogen gas.
- 4. The hydrogen concentration within the containment increases to the flammability limit before the combustible gas control system becomes effectivo, or said system never operates effectively.
- 5. Uncontrolled hydrogen-oxygen reaction (cxplosion) occurs.
- 6. Containment is breached; a substantial fraction of the core inventory ic released to the atomosphere, I resulting in offsite doses at the LPZ (low population zone) boundary which exceed the 10 CFR 100.11 guidelines of 25 rems whole body.
and 300 rcms thyroid.
l t
25
- 7. . This accident scenario should be construed as a TMI-2 type LOCA, with similar or equivalent hydrogen generation and explosion potential.
The acceptance of a hydrogen control contention as thus restated and accompanied by a credible accident scenario is supported by the recent Licensing Board and Appeal Board decisionc concerning the Perry rcactors. Cleveland Electric Illuminating Co. (Perry Nuclear Power Plant, Units.1 and 2), ALAB-675, Docket Nos. 50-440, 50-441, filed May 17, 1982. The Perry decision relied in part on the Commission'c Proposed Policy Statement related to Safety Coals for Nuclear Power Plants, which we have previously referenced, and on the Commission's proposed hydrogen control rule. In the case of Seabrook, the proposed rule would assume a hydrogen release resulting from a 75% metal-water reaction, cubstantially more than the Applicant now even discusses.
46 F.R. 62281 (December 23, 1981).
The Perry decision was also based in part on a concern for possible operator error that could delay the initiation of the hydrogen control system. The same situation exists at Seabrook, were both the hydrogen recombiners and the backup purge system depend upon operator action triggered by instrument readings or an alarm. FSAR at 6.2-76, 77, 80, 81.
It appears that operator action may also be required to start the hydrogen monitoring subsystem, which would provide the operator with the necessary information, although
26 this is not clear from the FSAR. FSAR at 6.2-74.
- In any event, there is more than ample room for operator error to render the hydrogen control system ineffecitve.
Finally, we urge the Board to focus on hydrogen control, rather than hydrogen generation. The scenario that we have presented, with the TMI-2 characteristics and relcase amounts, provides the essential information with respect to generation. The crucial safety issue is whether the hydrogen can be adequately controlled, and that should be the center of the Board's attention.
I.S. Loose Parts Detection System The Applicant and the Staff object that this contention improperly calls for compliance with a Regulatory Guide.
NECNP belicves this contention is properly worded, calling for conformance to the Regulatory Guide or an adequate alternative in order to comply with the regulations i themselves. The Staff also objects that a loose parts i
l detection system is not required by 10 CFR 50.36 or 10 CFR
- 20. l(c) . We refer the Staff to Reg. Guide 1.133, which discusses the manner in which it implements those regulations. Unless the Applicant submits an alternative means of satisfying the regulations which is equivalent to the protection afforded under the Reg. Guide, failurc to comply with the Reg. Guide constitutes a factual basic l
27 for an allegation that the regulations have not been met.
More important, it is a fundamental principlc that a reactor
?
must be maintained and repaired to the came level of safety as when it was built and licensed. No regulation exempts the Applicant from its continuing responsibility to assure compliance with the regulations and safety requirements.
I.T. Steam Generators The Applicant and the Staff challenge this contention on the ground that steam generator defects documented by NECNP cannot bc attributed to the Model F steam generator which has been installed at Scabrook. NECNP is not required at this point to state with particularity the defective characteristics of the Model F steam generator.
Although the Model F is of recent vintage and has not been subjected to the test of operation at full power, Westinghouse steam generators have a history of degradation problems that are not restricted to a single defective model. The three reactors where problems have been found -- Indian Point Unit 2 (Model 44), San Onofre Unit 1 (Model 27),
and Surry Units 1 and 2 (Model 51) -- all contain different Westinghouse steam generators. Furthermore, a recent " improvement" on Westinghouse steam generators - "Model D-4" -- is also experiencing problems. Nucleonics Week, Vol. 23 No. 11, March 18, 1982 at 7. Steam generator degradation can thereby i
28 be seen to cover a broad spectrum of Westinghouse models, which may indeed include Model F. This basis for NECNP's Contention is not vitiated by the fact that the Seabrook generator bearc a different model number than its predecessors.
NECNP requires the opportunity of discovery to discern those similarities which may cubject it to the same problems as previous models.
I.U. Turbine Missiles In responding to the Applicant's.and the Staff's objections, NECNP would first like to correct an error on page 50. The second paragraph, beginning "In-service inspection of steam generators..." should be deleted.
Contrary to the Applicant's ascertion that NECNP here seeks to elevate a Regulatory Guide to the status of a regulation, NECNP has properly contended that the Applicant must either comply with the Reg. Guide or precent an equivalent alternative means of satisfying GDC' 4.
The Staff objects that NECNP has not shown specifically how- the Applicant does not comply with GDC 4. NECNP referc the Staff again to FSAR Figure 3.5-1, which shows possible curbine missile trajectory paths intersecting the containment 3 and to the FSAR discussion referenced in our --
original filing of this contention. NECNP contends this violates GDC 4, which provides that structures, systems and
29 components important to safety must be protected against dynamic effects.
NECNP would also like to direct the Board's attention to a recent Appeal Board decision regarding risks posed by turbine missiles to safety-related components, which found that the chief cause of turbine disk cracking has been misjudged, and that corrective measures consequently may have exacerbated the situation. Virginia Electric and Power Co. (North Anna Nuclear Power Station, Units 1 and 2)
ALAD-676, May 26, 1982, slip op. at 30. The Board found that:
When we commenced our turbine missile inquiry in 1979, brittle fracture.was apparently thought to be the principal cause of turbine disc cracking. But such is not the case. Rather, it ic now recognized that the more serious concern is stress corrosion. Unlike brittle fracture, which generally is the product of turbine overspeed, stress corrosion cracking can occasion a disc failure at normal turbine operating speeds, as well as under startup stress. In this connection, the evidence suggests that the very empahsis on improved material to enhance brittle fracture toughnecs may have also increased the susceptibility of the discs to stress corrosion. (emphasis added)
Id. The Board noted that the causes of turbine stress are unknown, and allowed the plant to continue operation only with a strict inspection program in place. The Board also noted the need for more extensive rescarch on the subject.
The Applicant's analysis of turbine missile strike and damage probability is based on assumptions concerning
30 turbine overspeed and probabilities of failure mechanisms which have been put into quection by ALAB-676. FEAR at 3.5-10-11. Given the new, conflicting information on the risk' of generating turbinc missilen inside reactors, NECNP's contention is all the more important to the resolution of cafety issues at Scabrook.
I.V. In-Service Inspection of Steam Generator Tubes In responce to the objections of the Applicant and the Staff, NECNP rewords this contention as follows:
The Applicant has not demonstrated that it has met GDC 14, 15, 31, and 32 because it implements an insufficient program for in-scrvice inspection of steam generator tubes, i.e. Reg. Guide 1.83, Rev. 1.
The basic for thic contention remains that the long history of problems with steam generators and the recent accident at Ginna show a need for improved in-service inspection of steam generator tubes beyond the requirements of Reg. Guide 1.83. The Ginna inspection program, which was in compliance with Reg. Guide 1.83, failed to reveal the damage which caused the accident a short time after an inspection.
I.W. Scismic Qualification of Electrical Equipment In responce to the objections by the Applicant and the Staff, NECNP rewords this contention as follows:
The Applicant has not complied with Criterion III, " Design Control" of Appendix D to 10
31 CFR Part 50. In particular, the Applicant has not satisfied Reg. Guide 1.100 Rev. 1 or provided an equivalent alternative meanc of meeting Criterion III.
This contention has two major bases. First, the Applicant has stated that its seismic qualification program for NSSS safety-related equipment is delinated in a 1977 Westinghouse topical report which is " currently under review by the NRC." FSAR at 1.8-36. NECNP believes the fact that this topical report has been revised three times since then shows that the NRC is not satisfied with the 1977 report upon which the Applicant relies. Second, seismic qualification is an unresolved safety issue.
Task A-46 of NUREG-0705 stresses the importance of compliance with Reg. Guide 1.100 to provide the necessary
" margin of safety" in new plants. Since the Applicant admittedly has not qualified all safety-related equipment to Reg. Guide 1.100, NECNP has sufficient basis for asserting that the Applicant's seismic qualification program is inadequate. NECNP contends that the entire program, not just individual components, is inadequate insofar as it fails to meet the proper standard. Any necessary specificity is provided by reference to those items identified by the Applicant as not complying with Reg. Guide 1.100.
II. Quality Assurance II.A.1 The Applicant and Staff object that the Applicant's Quality Ascurance program was litigated at the construction
i l
32 permit stage and may not be raised again in the absence i
of either "significant cupervening developments having a possible material bearing upon previously adjudicated issues" or "the presence of some unusual factors having 4
special public interect implications." Alabama Power Co.
(Farley Nuclear Plant, Units 1 and 2), ALAB-182, 7 AEC 210, 216; remanded on other grounds CLI-74-12, 7 AEC 203 (1974).
NECNP has saticfied that standard. As discussed in NECNP's contention, the interpretation of Appendix B to 10 CFR Part 50 on which Seabrook's QA program was based was recently j reviewed by the Commission and found wanting with respect
- to major safety-rclated syctems and components. In particular, s
the Commission found that Appendix B criteria have not
- been fully implemented for in-core instrumentation, reactor coolant pump motors, reactor coolant pump powcr cables, and radioactive waste cystem pumps, valves, and storage tanks. None of these systems or components is covered by the Seabrook QA program. In addition, NECNP contends that I all equipment that removes heat from the steam generatora during shutdown should be subject to the QA program.
l The importance of qualifying heat removal systems was recognized after the accident at Three Mile Island in 1979, and has been identified as a new Unresolved Safety Issue. NUREC-0705, Tack A-45.
i i
I
33 l Both the change in the Commission's QA program and the recognition of decay heat removal a: an important safety function are significant developments which have occurred since the Seabrcok construction permit was issued in 1976. The make fundamental changes to the requirement: for Quality Assurance which must be satisfied in order to justify the issuance of a construction permit or an operating license. Furthermore, because these changes concern the qualifications of safety-related equipment, they have profound public interest implications.
NECNP has satisfied the Commission's requirements for raising these new issues, which are critical to a reasonable assurance of cafe operation of-the plant.
II.A.2 NECNP here contends that the Applicant has not complied with Appendix B to Part 50', which requires the implementation of an effective Quality Assurance program to assure that the design and construction of the plant conform with applicable regulations and provide a reasonable assurance that the public health and safety will be protected.
Contrary to the Applicant's assertion that this issue is precluded by res judicata, the implementation of-the Applicant's CA program, which includes design and construction, wac never litigated at the construction permit stage. The Applicant's further assertion that this issue can only be litigated in a show cause proceeding is simply erroneous.
The implementation of a quality assurance program is a
34 proper issue for an operating license proceeding. See Virginia Electric and Power Company, (North Anna Nuclear Power Station, Units 1 and 2), (LBP-77-68,,6 NRC 1127 (1977).
The Staff has suggested that NECNP's request for relief be limited to an independent audit of all QA documentation, a physical reinspection of the plant, and an engineering analysis of all aspects of the plant that cannot'be inspected now. We reassert that we consider thece remedies to be the minium that may be required, and that even those steps may not be sufficient given the previouc QA failures. In fact, as the I&E report indicate, the Applicant's recordkeeping on its QA program has been so poor that it is impossible to tell where all the QA problems lie; and an audit may be re' quired to make a threshold determination of which areas must be investigated.
As a basis for this contention, NECNP has provided thirteen examples of the I&E reports which cite the Applicant for QA deficiencies. The citations provide an initial indication of the pervasiveness and severity of the QA problem at Seabrook. Our point is not that the QA program falls in thirteen areas, but that thece failures in so many areas provide a basis for a contention that the entire QA program is faulty and must be examined by i the Board, The full scope of the plant's QA deficiencies
35 can only be discerned through discovery of such documents as nonconformance reports, audit reports, trending reports, and other records concerning the quality of construction at Seabrook which are otherwise unavailable. Where NECNP has established sufficient basis for the admission of a contention, it would be unfair to limit the scope of the contention to the restricted number of issues discernable without the benefit of discovery.
II.B. Quality Assurance for Operations The Applicant has no objections to contention II. B.
The Staff has objected to contentions II.B. 1, 2 and 4.
4 II.B.l. The Staff would limt this contention to Section 17.2 of the FSAR. NECNP accepts the limitation and rewords the contention as follows:
The FSAR addresses quality assurance for plant operation at Section 17.2. Section 17.2 fails to address each of the criteria in Appendix B in sufficient detail to enable an independent reviewer to determine whether and how all of the requirements of Appendix B and the guidance in all applicable regulatory guides will be satisfied.
III. Emergency Planning In response to the objections of the Applicant and the Staff, NECNP reasserts this contention in the following manner: we accept the wording proposed by the Applicant as a general statement of the contention. HowcVer, cach
36 of the sixteen items listed in our contentions as " specificity and basis" constitute separate subparts of the contention which NECNP asscrts must be individually addressed. This approach will reconcile the position of the Applicant and Staff, which are otherwise diametrically opposed to each other.
II.B.2 The Staff objects that NECNP has not specifically alleged QA deficiencies resulting from the overly restrictive classification of systems and equipment which are subject to quality assurance. NECNP contends that the Applicant has undertaken a Quality Assurance Program for a class of equipment it calls " safety-related", which is narrower than the class of equipment considered "important to safety" under CDC 1. Several examples of this discrepancy, including Applicant's' failure to perform QA for in-core instrumentation, reactor coolant pump motors, reactor coolant pump power cables, and radioactive waste system pumps, valves, and storage tanks, are given in Contention II.A.l. The cricir21 " result of the Applicant's failure to cover all systems and equipment important to safety is the Applicant's consequent inability to provide a reasonable assurance that those systems and
(
equipment will be able to perform their safety function.
l NECNP considers the examples given in contention II.A.1 l
to be indicators of a pervasive classification deficiency, and not to be limits upon this contention. We provide
37 sufficient specificity by establishing that the classification system is deficient.
II.B.4. The Staff objects that NECNP has cited no regulatory requirement for its assertion that the Applicant must provide a means of assuring that maintenance, repairs, and post-repair inspections at the Seabrook facility will be carried out in the manner necessary to protect the public health and safety.
The regulatory requirement supporting this contenion is found in Appendix B. In its definition of quality assurance, Appendix B states the fundamental principle that:
" quality assurance" comprises all those planned and systematic actions necessary to provide adequate confidence that a structure, system or component will perform satisf actorily in service. (emphasis added)
Appendix B, paragraph XVI, also requires that:
Measures shall be established to assure that conditions adverse to quality, such as failures, .
malfunctions deficiencies, deviations, defective material and equipment, and nonconformances are promply identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective
, action taken shall be documented and reported to appropriate levels of management. (emphasis added)
Further support for this contention is found in Appendix A, GDC, which requires that:
e 4
- - .e ,- -- . - - - . . - -
38 A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems and components will satisfactorily perform their safety functions. Appropriate records of the design, fabrication, erection and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit. (emphasis added)
The Nuclear Regulatory Commission may issue licenses valid for as long as forty years. The licensing concept would be rendered entirely meaningless if the responsibility of the license applicant to assure the quality of systems, equipment and components important to safety ended when the operating license was issued. The Commission cannot provide the public with a reasonable assurance that the reactor will not endanger the public health and safety throughout its license term unless there is some assurance that equipment will be maintained and repaired to at least the same standard that it met originally. The minimum requirements for this assurance would be to establish within the quality control program a system for maintaining and repairing defective equipment, inspecting the results of such actions, and keeping accurate records.
39 Respectfully submitted, William S. Jordan, III Diano J. Ct:rran liarmon & Weiss 1725 I Street, N.F.
Suite 506 Washington, D.C. 20006 (202) 833-9070 Counsel for New England Coalitica Nuclear Pollution DATED: June 17, 1982 m
\
w x %
e X_
N
~
s I P s
- - - -