ML20210C419
| ML20210C419 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Three Mile Island, Crystal River, Rancho Seco, Bellefonte, 05000000 |
| Issue date: | 04/06/1987 |
| From: | Mcgarry J BABCOCK & WILCOX OPERATING PLANTS OWNERS GROUP, BISHOP, COOK, PURCELL & REYNOLDS, BWR OWNERS GROUP |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML20206H055 | List: |
| References | |
| 2.206, NUDOCS 8705060181 | |
| Download: ML20210C419 (85) | |
Text
-
DOCKETF MPC UNITED STATES OF AMERICA '87 MH 14 PS :15 NUCLEAR REGULATORY COMMISSION
.:TF ' .
BEFORE THE . .:
DIRECTOR OF THE OFFICE OF NUCLEAR REACTOR REGULATION 4
)
In the Matter of )
)
Petition for Immediate Action )
to Relieve Undue Risk ) Docket No.
Posed by Nuclear Power ) (10 C.F.R. 52.206)
Plants Designed by the )
Babcock & Wilcox company )
)
)
PRINCIPAL RESPONSE OF B&W OWNERS GROUP TO PETITION FILED UNDER 10 C.F.R..
52.206 BY THE UNION OF CONCERNED SCIENTISTS J. Michaft McGarry
' Daniel F. Stenger BISHOP, COOK, PURCELL & REYNOLDS '
1200 Seventeenth St., N.W.
Washington, D.C. 20036 (202) 857-9800 Attorneys for B&W Owners Group April 6, 1987
[
G D '
l i
TABLE OF CONTENTS i
Page \
I.
INTRODUCTION........................................ 1 II.
COMMENTS............................................ 5 A. Legal Standard.................................. 5 B. The NRC Has Previously Ruled On The Matters Presented In The Petition............... 7 C. B&W Plants Are Safe............................. 11
- 1. Once Through Steam Generator................ 20
- 2. Pressurizer................................. 23
- 3. Auxiliary Feedwater System (ATWS)........... 26
- 4. Integrated Control System (ICS)............. 30
- 5. Non-Nuclear Instrumentation (NNI)........... 36 D. The Operating History of B&W Reactors Does Not Demonstrate An Undue Risk To
, Public Health and Safety........................ 40
- 1. Pre-TMI Events.............................. 40
- 2. The TMI Accident............................ 42
- 3. Post-TMI Events............................. 43 E. NRC's Reexamination of B&W Plant Safety Has Not Been Compromised................. 49 F. The C'urrent Reassessment Pro Not An Example ~'Of Delay..... gram.................... Is 63 G. NRC Possesses The Technical Capability To Assess Safety Implications Of B&W Plants...................................... 65 H. The NRC Has Provided an Adequate Basis To Justify Continued Operation............ 75 III.
CONCLUSION.......................................... 77
4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ,
l l
BEFORE THE l DIRECTOR OF THE OFFICE OF NUCLEAR REACTOR REGULATION
)
In the Matter of )
)
Petition for Immediate Action )
Docket No.
to Relieve Undue Risk )
Posed by Nuclear Power ) (10 C.F.R. 5 2.206)
Plants Designed by the )
Babcock & Wilcox Company )
)
PRINCIPAL RESPONSE OF B&W OWNERS GROUP TO PETITION FILED UNDER 10 C.F.R.
5 2.206 BY THE UNION OF CONCERNED SCIENTISTS I. INTRODUCTION The Union of concerned Scientists (UCS) has filed a petition ,
under 10 C.F.R. I 2.206 requesting the suspension of the operating licenses or construction permits for all plants utilizing the Babcock & Wilcox (B&W) nuclear steam supply system.1 In particular, the petition seeks the suspension of the operating licenses of Arkansas Nuclear One Unit 1, Crystal River Unit 3,
~
Davis-Besse, Oconee Units 1, 2 and 3, Rancho Seco and Three Mile Island Unit 1, and the suspension of the construction permits for Bellefonte Units 1 and 2.2 The petition further requests that
-1/ Petition for Immediate Action to Relieve Undue Risk Posed by Nuclear Power Plants Designed by the Babcock & Wilcox Company, dated February 10, 1987 (Petition).
2/ Bellefonte Units 1 and 2 are currently under construction.
l Trootnote continued to next page.)
l l
l
. ~ , - - _ - . _ _ _ _ _ _ _ . - _ -
the NRC conduct public hearings to determine the corrective actions necessary for B&W reactors to achieve adequate protection of the public health and safety.
By letter dated February 26, 1987, the Staff requested the B&W Owners Group 3 to provide a response to the petition. An (Footnote continued from previous page.)
The Bellefonte construction permits were issued on the basis of decisions rendered in a contested proceeding. A full airing of environmental, site suitability and safety issues resulted in a finding of reasonable assurance that the plant could be constructed and operated without undue risk to the
! health and safety of the public and that any remaining safety issues would be resolved prior to issuance of an operating license. In the Matter of Tennessee Valley Authority (Bellefonte Nuclear Plant Units 1 and 2), LBP-74-91, 8 AEC 1124, 1137 (1974), aff'd, ALAB-253, 8 AEC 1182 (1975). TVA has requested an extension to complete construction of Units 1 and 2 in 1993 and 1995 respectively. Under these 2
circumstances as to these units.
there is no need to address the issue of relief See Porter County Chapter of the Izaak Walton League v. NRC7 IO6 F.2d 1363, 1369 (D.C. Cir. 1979)
(construction.or nuclear plant "does not of itself pose any danger to the public health and safety"). Any questions that the Staff may have arising from the petition can be addressed in the operating license phase of the proceeding.
Throughout this response, the B&W Owners Group will refer to operating conditions and operating plants. It should be '
! noted that TVA and B&W have performed for Bellefonte a rather i-detailed design review with respect to many of the issues raised in the petition. All new requirements have been incorporated into the B&W 205 FA NSSS for Bellefonte through i
a change assessment program conducted by TVA and B&W. These improvements, combined with the safety of the design of the operating plants, support a conclusion that the Bellefonte units when completed can be operated without endangering the health and safety of the public.
3/ For purposes of these comments, the B&W Owners Group consists of Arkansas Power & Light Company (Arkansas Nuclear One Unit 1), Duke Power Company (Oconee Units 1, 2 and 3), Florida Power Corporation (Crystal River Unit 3), GPU Nuclear Corporation (Three Mile Island Unit 1), Sacramento Municipal Utility District (Rancho Seco), The Toledo Edison Company
! (Davis-Besse) and Tennessee Valley Authority (Bellefonte Units 1 and 2).
l l
m . __ . . _.
initial response was filed on March 6, 1987 limited to UCS' t
request for immediate action (i.e., suspension). By letter' dated March 13, 1987 the NRC denied UCS* request for immediate suspension.
In so doing, the NRC noted (at p. 2), among other things, that improvements implemented at B&W plants since 1980 have resulted in a reduction in the scram rate, and that transients in 1985 (at Davis-Besse and Rancho Seco) were "not :
indicative of safety problems in general at B&W-designed plants that would require immediate staff action." Moreover, the NRC pointed out that substantial improvements have previously been implemented, or are in the process of being implemented, as a result of a current reassessment program (at p. 2):
Since the B&W plant reassessment began, almost one hundred (100) recommendations have been referred to the B&W plant owners, who have implemented, or are im
. these recommendations.plementing, many These changes of have i already improved plant safety by improving the ICS and the performance of the main feedwater system to reduce the number of challenges to safety systems due to feedwater transients.
Because of these changes, the staff concluded that the plants are safer now than they were a year ago. The NRC found no undue risk in the utilities' interia operation a year ago, and, after taking into consideration the utilities' i
safety improvements, again finds no undue risk in allowing the utilities to continue to operate the plants.
With regard to the petition in chief, the case turns on whether the petition raises a " substantial health and safety issue." In the Matter of Consolidated Edison Co. (Indian Point, Unit Nos. 1, 2 and 3), CLI-75-8, 2 NRC 173 (1975). The petition alleges that certain " unique" design features of B&W reactors "make them inherently more dangerous than other pressurized water l
l
. .__ - - _ __, _ , , _ - _ _ , _ _ . _ . _ . . . , _ - - . _ ~ _ - . . . . . _ _ _ _ _ _ _ _ , . . . . , . - . - - -
reactors." Petition at 4-5. UCS alleges design deficiencies in five specific systems: (1) the once through steam generator, (2) the pressurizer, (3) the auxiliary feedwater system, (4) the integrated control system, and (5) non-nuclear instrumentation.
Petition at 9-13. UCS then points to the operating history of B&W reactors as confirming the existence of alleged design flaws.
Petition at 14-19. UCS also alleges that NRC has failed to take effective action to address this situation. For example, UCS takes issue with NRC's assignment of a reassessment of B&W plant transient responses to the B&W Owners Group; takes issue with NRC's alleged practice of studying problems at B&W plants but failing to resolve them; and takes issue with the NRC's technical l
capability to properly understand B&W plants. Petition at 19-38.
In the comments below, the B&W Owners Group responds to each of the allegations raised in the petition. It will be shown in 3
each and every instance that UCS' technical allegations are unfounded. Indeed, the information recited in the petition is well known to the NRC and has previously been determined not to justify individual plant licensing action, including shutdown of B&W plants. Letter from V. Stello to H. Tucker, dated January 24, 1986. In these circumstances, it is evident that no
" substantial health and safety issue" is present. The petition must therefore be denied.
l 1
II. COMMENTS A. Legal Standard )
j A Section 2.206 petition is a request that the NRC institute i
an enforcement (or show cause) proceeding pursuant 10 C.F.R.
5 2.202 to modify, suspend or revoke a license or for other appropriate relief. The Commission has interpreted Section 2.206 to permit issuance of a show cause order only when the Director finds that " substantial health or safety issues ha[ve] been raised. . . . (A] mere dispute over factual issues does not suffice." Indian Point, supra, 2 NRC at 176; see also In the Matter of Northern Indiana Public Service Co. (Bailly Generating' station, Nuclear-1), CLI-78-7, 7 NRC 429, 433 (1978). This standard has been recognized by the courts. Florida Power &
Light Co. v. Lorion, 470 U.S. 729, 732 (1985); Porter County
, Chapter of the Izaak'Walton League v. NRC, 606 F.2d 1363, 1367 (D.C. Cir. 1979).
As the D.C. Circuit recently observed, the phrase
" substantial health and safety issue" is "a term of art within the Commission, because it is the language reserved as a trigger for action rather than a description of the severity of the concern." Lorion v. NRC, 785 F.2d 1038, 1041 (D.C. Cir. 1986)
(on remand). The D.C. Circuit's decision in Lorion points out that the NRC's recognition of potential safety concerns does not equate with a finding of a " substantial health and safety issue."
The Commission's precedents make it clear that i it is not obligated to take enforcement action "whenever we receive information adverse to the integrity of existing nuclear power ww, , -,,,,,,----+-----,----w-. -,,,l-,< -
safety . . . ." In re Nuclear Regulatory Commission, 5 N.R.C. 16, 21 (1977), citing .
Nader v. Nuclear Regulatory Commission, 513 F.2d F.2d 1045, 1054-55 (D.C. Cir. 1975).
785 F.2d at 1041.
In resolving Section 2.206 petitions the Director has broad discretion. See Indian Point, 2 NRC at 175. The Commission will review the Director's decision only on its own motion and only to
" determine if the Director has abused his discretion." 10 C.F.R.
5 2.206(c)(1). As the Commission has stated:
. . . . we believe the question whether the Director has abused his discretion in denying a request for a show cause order to embody the following elements: (1) whether the statement of reasons given permits rational under-standing of the basis of this decision; (2) whether the Director has correctly understood governing law, regulations, and policy; (3) whether all necessary factors have been considered, and extraneous factors excluded, from the decision; (4) whether inquiry appropriate to the facts asserted h,as been made; and (5) whether the Director's decision is demonstrably untenable on ghe basis of all
, information available to him.
Indeed, when dealing with the question whether to institute enforcement proceedings (which includes action on a section 2.206 petition), an agency's discretion is accorded the greatest deference. The Supreme Court has recently held that an agency's decision to decline to take enforcement action is presumptively not subject to review by the courts. Heckler v. Chaney, 105 S.Ct. 1649 (1985).
l With these considerations in mind, we now turn to the question whether the petition raises a " substantial health and safety issue."
4/ Indian Point, 2 NRC at 175.
4
l
\
i B. i Matters Presented in the PetitionThe NRC Has '
The UCS petition presents nothing new .
falls into three categories. First, The information cited transients before the TMI-2 accident UCS claims that operationa overly sensitive to feedwater flow upsetsshow that B&W re Second, UCS maintains that the occurrenc Petition at 14-15.
1980 shows that the post-TMI modificatione of transients since !
alleged design problems. s have not cured the Petition at 16-19. 1 upon an ongoing reassessment of B&W reactors asFinally, UC !
proof that deficiencies still remain.
however, Petition at 19-23. In each instance, i
the NRC has investigated the concerns and h as found generically that continued operation of B&W rea t c ors is safe.
0667, The pre-TMI transients were reviewed by th e NRC in NUREG-
" Transient Response of Babcock & Wilcox-Desig (May 1980) at 4-1. ned Reactors" NUREG-0667 noted (at 2-2)
[ "not believe that complete plant shutdown of ththat the Staff did j either necessary or desired with regard to publie B&W plants [wa safety."5 i
c health and i
As for post-TMI modifications, the Commission required short-term design and procedural changes immedi tl a e y after the accident.
These changes included modifications to upgrade the reliability of the auxiliary feedwater system See, e.g.,
In the S
~/ of An assessment of feedwater transients at B&
made in NUREG-0560, reedwater Transients in Pressurized Wa" Staff Report on the Ge Designed by the Babcock was a short-term assessme& Wilcox company"ter (May Reactors 1979). This noted B&W that activities reactors. were in progress NUREG-0560 at 8-1.
to improve s fnt after the T a ety at
- - - - ~ ~ ~~
g- -"-- '~ _ _ _ __ ---
Matter of Sacramento Municipal Utility District (Rancho Sec,o Nuclear Generating Station), 44 Fed. Reg. 27779 (1979).
Additional long-term modifications to enhance safety were also required, including changes to the Integrated control System. 44 Fed. Reg. at 27779.
These long-term actions were further refined in NUREG-0737,
" Clarification of TMI Action Plan Requirements" (1980). NUREG-0737 applied to all power reactors 0 and was specifically approved by the Commission. In so doing, the Commission found that the actions set forth therein were -
generally sufficient to assure adequate response to the TMI-2 accident.I A review of the status of NUREG-0737 actions for B&W plants clearly shows that major improvements have been made. See NRC Letter to UCS, dated March 13, 1987.
In lifting the shutdown orders imposed on B&W plants
~
following TMI-2, the NRC noted that licensees had satisfactorily completed NRC prescribed actions, that such actions were acceptable and that resumption of operation was appropriate.
See, e.g.,
Sacramento Municipal Utility District, Docket No. 50-312, Notice of Authorization to Resume Operation, dated 1979, motion to stay denied, Friends of the Earth, Inc. v. United States, 600 F.2d 753 (9th Cir. 1979); Duke Power Company, Docket Nos. 50-269 et al., Notice of Authorization to Resume Operation ,
l dated May 18, 1979.
6/
~
NUREG-0737 was not directed exclusively to B&W plants. In all plants, all PWRs or all BWRs.nearly all cases NUREG-0737 require .
7/
- Statement of Policy:
Further Commission Guidance for Power i
Reactor Operating Licenses, 45 Fed. Reg. 85236, 85238 (1981).
In addition, the short-term and long-term TMI Action Plan requirements were endorsed in NRC adjudicatory proceedings which involved extensive hearings on two B&W plants. See Sacramento 1 i
Municipal Utility District (Rancho Seco Nuclear Generating Station), LBP-81-12, 13 N.R.C. 557 (19/81), aff'd, ALAB-746, 18 N.R.C. 749 (1983) (hereinafter cited as Rancho Seco);
Metropolitan Edison Co. (Three Mile Island Nuclear Station, Unit 1), LBP-81-59, 14 N.R.C. 1211, 1223-1423 (1981), aff'd as modified, ALAB-729, 17 N.R.C. 814 (1983), aff'd as modified, CLI-84-11, 20 N.R.C.
1 (1984) (hereinafter cited as TMI-1 Restart).
With regard to the reassessment, this program was initiated as a result of transients at Davis-Besse and Rancho Seco in 1985.
However, contrary to UCS' allegation, the conducting of a reassessment does not call into question the safety of B&W plants or the validity of the NRC's past findings. In response to questions by Representative Markey in hearings during April 1986, the NRC explained that "B&W reactors can safely continue to operate while the NRC reassess [es] the B&W plant design
(
requirements." The NRC described the Davis-Besse transient of June 9, 1985 as " plant-specific" and, with respect to the Rancho Seco transient of December 26, 1985, noted that "a loss of ICS power at other B&W plants would not have resulted in as severe an over-cooling event as that experienced at Rancho Seco due to plant design differences and previous actions taken to compensate for a loss of ICS power event." NRC Response, dated April 15, 1986, Question 18; see also Letter from V. Stello to H. Tucker, l dated January 24, 1986; and NRC's Letter to UCS, dated March 13, 1987.
_._.._J____.-_. _ ,- _ _ ._ _ . _ _ _
UCS has provided no new information which would disturb these findings. Indeed the substantive matters presented in the petition were also set forth in a rebruary 1986 report UCS submitted to the NRC. The NRC concluded at that time that suspension was not appropriate, noting that UCS had "not provided any information that would cause us to alter our conclusion that continued operation of the B&W plants, during the time it takes i to complete our reexamination, does not pose an undue risk to the public health and safety." Letter from V. Stello to R. Pollard and E. Weiss, dated March 25, 1986.0 .
In essence, UCS believes that whenever a fresh look is taken at safety (e.g., the B&W reassessment), industry and the NRC are implicitly conceding a present deficiency. This type of 8/
~ As discussed fully herein, the B&W Owners Group has provided new information confirming the correctness of the NRC's past findings. A Safety and Performance Improvement Program 4
(SPIP) was initiated by the B&W Owners Group following the 1985 transients at Davis-Besse and Rancho Seco. As information has been developed it has been submitted to the NRC. Currently, the NRC is reviewing SPIP, Rev. 03. The final conclusions and recommendations of the SPIP are expected to i be available by mid-1987. However, to date no substantial l safety issues have been identified by the NRC or the B&W j
Owners Group as a result of SPIP. See NRC letter to UCS, dated March 13, 1987.
Two important parts of SPIP are a sensitivity study commissioned by the B&W Owners Group from MPR Associates, an independent consulting firm, and a risk assessment review. '
The technical work performed in the sensitivity study, entitled "A Comparative Study of the Sensitivity of B&W Reactor Plants" (" Sensitivity Study"), was completed in March 1987 and is in the process of being documented for submittal to the NRC. This study examines the sensitivity of B&W reactors vis-a-vis other PWRs. The risk assessment review is set forth in SPIP, App. L ,(" Risk Assessment"). It was intended to assess the risk importance of.the more l significant transients that had occurred on B&W plants. ;
l l
~ ~ ' ~ ' ~ ~ ' ' ' ' " '
misconception has been rejected in the past. In the TMI restart proceeding, the Appeal Board rejected UCS' " notion that nuclear plant operation cannot be considered reasonably safe as long as scientific efforts are underway to develop new and better safety features." In the Matter of Metropolitan Edison Co.
(Three Mile Island Nuclear Station, Unit 1), ALAB-729, 17 NRC 814, 828 (1983).
In the comments below, the B&W Owners Group addresses each of the allegations raised and shows that the NRC's previous conclusions remain valid and that no substantial health and safety issue is present.
C. B&W Plants Are Safe.
UCS alleges that "[u]nique elements of the B&W plants make them inherently more dangerous than other pressurized water reactors" (Petition at 4); that "B&W reactors are extremely sensitive to events that would be innocuous in other PWRs."
Petition at 13. In UCS's view, certain inherent characteristics and hardware systems of B&W reactors cause them to pose an undue risk to public health and safety.
It is important at the outset to dispel this popular, albeit erroneous, theme. The B&W plants are safe. The safety of the plants is supported by a large body of technical information.
Some of the supporting information directly relevant to the UCS allegations include:
Final Safety Analysis Reports and Safety Evaluation Reports
PRA Results Severe Accident Investigations On-going B&W Owners Group Reactor Vessel Materials Program B&W Owners Group Safety and Performance Improvement Program (SPIP)
Information in safety analysis reports describes how the plants were designed and built before they went into operation.
The B&W plants met the applicable NRC criteria for nuclear plants built at the time they were licensed. They were built to the established conservative codes and standards. As with all U.S.
J nuclear plants they incorporated multiple levels of defense with conservative approaches being taken at each level. See NRC Letter to UCS, dated March 13, 1987 at 3; see also Oconee Unit 1, Safety Evaluation Report (SER) (1970) at 67-70. All the operating plants were licensed upon the definitive finding of safety required by Section 182 of the Atomic Energy Act.
Complementing the general conservative approach in the areas of regulatory and other technical requirements, a multitude of j deterministic accident analyses were performed and reviewed which
- utilized many conservative assumptions. See, e.g., Oconee Unit 1 FSAR at Section 14. These analyses dealt with the consequence of each of the design basis accidents. In the years since the original analyses were performed there have been numerous supplementary safety analyses conducted on the B&W plants. These include (1) extensive small break LOCA calculations, (2) extensive ATWS calculations, (3) extensive pressurized thermal shock analyses, (4) and a multitude of best-estimate l
l l
s .
13 -
calculations to support the important transition to symptom-oriented emergency operating procedures.9 The B&W plants were designed to perform differently from other pressurized water reactors (PWRs). They were designed to respond rapidly to changes in load demand. They were designed to produce superheated steam at the outlet of the steam generators.
4 There are numerous other differences as well. The characteristics which allow them to be more efficient and to respond to load changes quickly have been perceived by UCS as differences in safety margins. There are no studies which provide support for this perception. In fact, as discussed below, there are independent studies which indicate equivalent margins when comparing the B&W plants with other pressurized water reactors. One of the earliest was performed in support of the President's Commission after TMI-2.10 Furthermore, the B&W plants contain a combination of other features not found in other pressurized water reactors which enhance their safety. These include the following (as well as others discussed below):
High pressure emergency core cooling pumps which permit direct core cooling under the full range of operating conditions.
-9/ See (1) Evaluation of Transient Behavior and Small Reactor U331 ant System Breaks in the 177 Fuel Assembly Plants, Vol.
1, 2 and 3, May 1979; (2) Analyses of B&W NSS Response to ATWS Events, BAW-1610, January, 1980; (3) documentation cited below at note 13 for the pressurized thermal shock analyses; and (4) B&W Owners Group Emergency Operating Procedures Technical Bases Documents 74-1152414 (September 1985).
--10/ S. Levy and J.E. Hench, SLI-7904, "A Study of Simulation and Safety Margins in Light Water Reactors" (August 26, 1979).
l
Special features in the reactor internals to enhance core cooling (such as reactor vessel vent valves).'
No potential for noncondensible gases to collect in steam generator U-tubes and interrupt natural circulation.
high-point vents.(The10 U-bends C.F.R. in the two B&W hot legs have S 50.44(c)(3)(iii).)
Superior steam generator tube integrity.11 Some of the alleged sensitivity is a result of features that provide advantages for the B&W design. For example, for the particular B&W plants that are subject to greater cooldown rates from overfeeding of emergency feedwater, this results from the flow capacity of their auxiliary feedwater systems which were designed to ensure a reliable and sufficient quantity of water to prevent undercooling, an event of more consequence than overcooling since overcooling has little or no potential for core damage.
The potentially greater cooldown rates of these plants were found by the sensitivity Study to be such that no equipment would be endangered. Sensitivity Study at A-14 to A-15.
Nevertheless, SPIP is investigating actions that will reduce the frequency of over-cooling transients.
l A comparison of severe core-damage frequency is instructive in dispelling UCS' claim that B&W plants are "more dangerous" than other PWRs. While the B&W owners Group does not have access to the detail of Probabilistic Risk Assessments (PRAs) for other plants, a comparison of the published results of PRAs is instructive. The overall core-melt risk of B&W plants as 11/ See NUREG-0667 at 2-2 and 5-18.
__ - . - _. - - - . ,. .- =_ -.
determined by PRAs is similar to other PWRs and is consistent with Commission statements concerning risk.12 Actions in connection with the Commission's Severe Accident Policy Statement also lend support in showing B&W plants are safe.
In promulgating this Policy Statement, the Commission was aware of "11 (NRC/AEC-sponsored] plant-specific PRA's and numerous industry-sponsored ones." See 50 Fed. Reg. 32138, 32144 (1985). These PRA's included B&W plants.
See Memorandum from W.J.
Dircks to the Commissioners, " Safety Goals" dated January 5, 1983.
The Commission found, on the basis of such information, that
" existing plants pose no undue risk to the public health and safety." 50 Fed. Reg. at 32144.
This finding of no undue risk to the public health and safety from the operation of nuclear power plants applies to all commercial light water reactors l including the B&W plants. .
The B&W Owners Group Reactor Vessel Materials Program began more than ten years ago. The materials program is the most extensive program of its type in the United States. One of the 1 key objectives of this program is to monitor, understand and predict the changes occurring in the rector vessel materials as 12/ Compare Oconee PRA, A Probabilistic Risk Assessment of Oconee
-~
Unit 3, NSAC-60 (June 1984); Crystal River 3 Probabilistic Risk Assessment, Draft Report (March 1987); with Millstone Unit 3, Probabilistic Safety Study; Indian Point 2, Probabilistic Risk Assessment, Docket No. 50-247 (1982);
Indian 286 (1982); Point 3, Probabilistic Risk Assessment, Docket No. 50-Yankee Rowe, Reactor Safety Study, WASH-1400 (October 1975);
Probabilistic Risk Assessment, Docket No. 50-29 (1982);
50-352 (1983).
Limerick Probabilistic Risk Assessment, Docket No.
the plants progress through their lifetimes. This extensive Materials Program provides added assurance that conservative pressurized thermal shock (PTS) operating limits will be maintained throughout plant lifetime. Extensive documentation has been provided supporting the acceptability of PTS risk on B&W plants.13 The B&W Owners Group Safety and Performance Improvement Program (SPIP) is a program, based on operating experience data, to reduce the frequency and severity of complex transients and reactor trips and to ensure acceptable plant response during those trips and transients that do occur. SPIP, Vol. 1, at II-4.
SPIP has reviewed six years of operating experience data and reviewed specific systems -- ICS, NNI, Main Feedwater, EFW/AFW, Instrument Air, and Secondary Plant Pressure Control System --
which are important to trip reduction and post-trip response.
SPIP has reviewed six years of operating experience data and reviewed specific systems -- ICS, NNI, Main reedwater, EFW/AFW, Instrument Air, and Secondary Plant Pressure Control System --
which are important to trip reduction and post-trip response. As noted, SPIP has been provided to the NRC and the NRC has stated that it provides further confirmation of the safety of B&W i
plants. See NRC letter to UCS, dated March 13, 1987. To date, some 150 recommendations have been referred to the B&W plant owners for action. SPIP, App.J.
13/ See " Reactor Vessel Pressurized Thermal Shock Evaluation,"
l Duke Power Company, January 1982; B&W-1791, "B&W Owners Group l Probabilistic Evaluation of Press %rized Thermal Shock", June 1983; B&W-1859, " Evaluation of Steam Generator Overfill (Footnote continued to next page.)
I l
1 l
l Of particular relevance to the UCS petition are two programs of SPIP: the Sensitivity Study and the Risk Assessment of .
previous complex transients. The' Sensitivity Study, which has been completed and will be submitted to the NRC in the next few weeks, includes a
" review of the B&W [ nuclear steam supply system (NSS)], as well as NSS's with recirculating steam generators (RSG's). . . .
The study includes review of the thermodynamic characteristics, coolant inventories, transient response, the protection safety, and control system characteristics, and operator actions relative to PWR design differences."
(SPIP, Rev. 03, V-Il The Sensitivity Study has concluded that, relative to other PWRs, B&W units:
are not more sensitive to reactivity upsets; are not more sensitive to reactor coolant system flow upsets; are less likely, on average, to experience a leak leading to a not loss of reactor coolant; are somewhat less sensitive to steam demand i upsets such as load rejections and turbine trips l so that a reactor trip on a turbine trip is not required to ensure plant safety; are not more likely to overcool following a reactor trip; are more sensitive to main feedwater upsets (though the frequency of such upsets is not greater than in other PWRs); but are more capable of riding out limited reductions in feedwater, e.g. loss of a single feedwater pump; are in some, but not all, plants subject to greater cooldown rates from overfeeding of emergency feedwater; l
(Footnote continued from previous page.)
Leading to Steamline Break and the Relationship to Pressurized Thermal Shock," December 1984.
__ ~_ _. . _ _ _ _ . _ _ . . . _ _ _ _ - _ . . . , - . . - - - _ _ _ . - - - - - - - - - - - - - - - -
4, are equivalent to many other PWRs in terms of time available to use alternative means of decay heat removal, on a complete loss of all feedwater; '
for most reactor trips, do not impose greater control burdens on plant operators; and impose greater burdens on plant operators in diagnosing and responding gg failures of~
automatic control systems.
As would be expected the sensitivity Study found that B&W plants are more sensitive in some areas, less sensitive in other areas and about the same in others. The sensitivity Study makes recommendations to improve upon those areas of greater relative sensitivity. See sensitivity Study at i to 11.
4 The Risk Assessment project of SPIP analyzes a number of abnormal transients (including the 1985 transients at Davis-Besse and Rancho Seco) since February 1980. Using the PRAs for Crystal River-3 and oconee the Risk Assessment finds that these abnormal transients are not significant to core-melt risk at all the B&W units. SPIP, Rev. 03, App. I, 5-2, and App. L.
In short, though many have expressed their opinion that B&W 3
plants are significantly more sensitive to system upsets than other PWRs, the facts, as set forth in part above, demonstrate t
otherwise.
i I
Undeterred, UCS focuses on five specific B&W plant systems or i
components and alleges that they increase the risk of an accident. Petition at 8-13. Contrary to the UCS conclusion, different operating characteristics and different hardware 14/ Sensitivity Study at i.
l t
systems do not imply different safety margins. The types of nuclear steam supply systems used in commercial nuclear power plants vary widely. Valid safety' evaluations require that each type of NSSS be separately evaluated. The responses to equipment malfunctions or accidents vary considerably depending on the type of system being considered. Nevertheless, each system criticized by UCS is addressed below to lay to rest the perceived
" sensitivity of B&W plants" issue.
It should be noted that in the TMI-l restart hearings, UCS had the opportunity to challenge the adequacy of these systems.
Yet the UCS contentions did not include a challenge to the once Through Steam Generator design, pressurizer size, the ICS (although another intervenor did litigate this system), or NNI.
While UCS participated in hearings on the reliability of the TMI-1 Emergency reedwater System (ErWS),15 the issue was raised in the proceeding only because of a Licensing Board inquiry. UCS filed no contention challenging the ErWS. Thus the UCS contentions in the TMI-1 Restart proceeding were simply attacks on NRC licensing standards and its TMI Action Plan, applicable to other PWRs as well as to B&W plants. See generally TMI-1 Restart, supra, 14 N.R.C. 1211, 1225-33, 1258-94, 1312-18, 1328-1409 (1981). (Issues concerning the design of several B&W systems were also litigated in the Rancho Seco proceeding (in which UCS did not participate). 13 NRC at 568-604.) Further, UCS' petition to the U.S. Court of Appeals for the Third Circuit, 15/ The EFWS is another name for the auxiliary feedwater system (AFWS) which is discussed infra.
~~
1 seeking review of the NRC's restart authorization for TMI-1, did l not challenge any of the agency's findings on plant design'and procedures. See Three Mile Island Alert, Inc., et al. v. NRC, l 771 F.2d 720 (3rd Cir. 1985), cert. denied, 106 S. Ct. 32 (1985).
UCS has made no showing that it is entitled to yet another opportunity, in an NRC adjudicatory setting, to challenge the B&W design. Thus UCS should bear a heavy burden to show that new 4
information, unavailable during the TMI-l Restart proceeding, demonstrates that these systems are inadequate.
- 1. Once Through Steam Generator UCS alleges that the once through steam generator (OTSG) design has exhibited " extreme sensitivity to feedwater flow upsets (and] has turned events that would be innocuous at other PWRs into crises for equipment and operators at B&W plants."
Petition at 9. This allegation,.however, ignores basic safety advantages of the OTSG design and prior NRC findings that the
! OTSG is technically sound.
1 The B&W steam generator design is known as a once-through design in reference to the fact that the primary coolant transfers heat to the secondary coolant as the secondary coolant makes a single pass through the steam generator. The OTSG design and the B&W NSS enhance the response to load changes on the electrical grid. See, e.g., NUREG-0667 at 5-1. As a result of this enhanced ability to follow load, primary system parameters in B&W plants -- e.g., coolant temperature, pressurizer level and pressure -- will respond more rapidly to changes in feedwater flow than in other plants.
_ _ , _ . - - - , - - - -- - , - - - - - - - - - - - - - - - - - - - - - - - - ~ ' - - - - - ~ ~ - ~ ' " ' ' ' ~ - ' ~ ~ ' ' * * ~ ~ " ~ ~ ~
, The NRC is well aware of these facts, and has determined the OTSG, as originally designed, to be adequate. See, e.g., Oconee Unit 1, SER (1970) at 27.
i More recent reviews of the OTSG found it to be " technically sound." NUREG-0667 at 2-3.
The NRC's conclusion was based in part on the safety advantages of the OTSG, such as the fact that the steam generator tubes have superior integrity compared with those in other reactors. See NUREG-0667 at 2-2 and 5-18. Because the OTSG does not have U-tubes, there is also no potential for noncondensible gases to collect in the tubes and interrupt natural circulation.16 4
Issues concerning the alleged sensitivity of the OTSG at Rancho Seco and the adequacy of the short-term and long-term actions resulting from TMI-2 were litigated in In the Matter of Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station), 13 NRC 557 (1981). There the Licensing .
Board found that these actions were an appropriate response to improve the OTSG's secondary-to-primary sensitivity. 13 NRC at 580.
4 The significance of the shorter steam generator dryout times for B&W reactors was also addressed in the TMI-l restart proceeding.
The Appeal Board found that steam generator dryout does not represent a loss of decay heat removal capability and that sufficient time is available for operators to restore feedwater. 17 NRC at 832-33. In affirming the Appeal Board, the 16/ B&W
~~
gases could collect.
reactors do have a U-shaped hot leg where noncondensible venting of such gases. However, See 10 C.F.R.
high point vents allow the 550.44(c)(3)(iii).
- ,uyr-vw-99 -- p - g -www..,9,.-y-.--gm . -,,.---,...g.- --.-7 -g.yy---yz,ww .y-"w wwr= eg ' '
-wwe--- 'r-ww v-mm's+ ' -- -'"4-4-'v-r -'
WW - w-
Commission found that there is a "significant amount of i
additional time for operators to take corrective action." ~20 NRC at 10.
Because of the OTSG method for introducing feedwater, it is known that B&W reactors are less susceptible to water hammer events than other PWRs. During the time prior to 1982, the NRC found that 27 water hammer events occurred on the secondary side of U.S. nuclear plants, though none occurred on B&W plants.
NUREG/CR-2059, " Compilation of Data Concerning Known and Suspected Water Hammer Events in Nuclear Power Plants" (May 1982).
Landing support to these conclusions is the Sensitivity Study commissioned by the B&W owners Group. The Sensitivity Study found that the low inventory in the OTSG reduces the potential for overcooling or post-trip criticality for steam line break events inside containment. See Sensitivity Study, Vol. II, at A-10, citing previous study. The Sensitivity Study found that the OTSG is more sensitive to feedwater flow upsets than other PWRs; however, the peak reactor coolant pressure for B&W plants was found comparable to that for other plants. Sensitivity Study, Vol. II, at A-12.17 17/ A B&W owners Group report on high pressure trips indicates
~~
thst since 1980 the highest pressure reached by a B&W reactor at power was 2336 psig, with the exception of the Davis-Besse transient of June 9, 1985 where a pressure of 2400 psig was reached. BAW-1890 A, Justification for Raising Setpoint for Reactor Trip on High Pressure (August 1986) at Table 4-2.
The setpoint for reactor trip for B&W reactors was lowered from 2355 psig to 2300 psig shortly after TMI. The normal high pressure trip setpoint for other PWRs is 2400 psia. An increase in the trip setpoint to 2355 psig has been requested (rootnote continued to next page.)
4
- _ _ _ _ . _ , , - - - - . - - -_ , - , _ -- -.--y- _ . , , - - - . - -----,-,--w-,v,- _ . - . - - - , , , - , - , , , . - - - . - ----
The sensitivity Study also supports the view that despite the shorter steam generator dryout times for the OTSG, the overall 4
time available for corrective actions by operators is comparable to that for other PWRs because of the responsiveness of the primary system in B&W reactors. See Sensitivity Study at A-17.
This point is supported by the procedures and training for operators at B&W reactors which have focused on response to feedwater upsets. See NUREG-0737, Enclosure; 44 Fed. Reg. at 27779.
For these reasons, the NRC's prior conclusions as to the adequacy of the OTSG design remain valid.
- 2. Pressurizer
, UCS' allegation regarding the B&W pressurizer focuses on the allegedly small size of the pressurizer when compared with those of other PWRs. UCS goes on to claim that the " combination of the OTSG design and the small pressurizer makes reactor temperature i
and pressure in a B&W plant extremely sensitive to a change in l
feedwater." Petition at 9. This allegation, as shown below, is based on the misconception that B&W pressurizer volume is less than that for other PWRs on a per-megawatt basis.
In a PWR the pressurizer is designed to accommodate density changes in the coolant resulting from changes in temperature.
The pressurizer provides an indication to the operator of the coolant volume inventory. The pressurizer is typically operated (Footnote continued from previous page.)
by the B&W owners Group and approved by the NRC. See SPIP, Rev. 3, Vol. 1, at IV-8. Even with this change the setpoint for B&W reactors will be lower than that for other PWRs.
. _ . ~ . . - - _ _ _ _ _ . . . _ _ _ _ _ _ . _ . _ _ _ . . _ _ _ _ . . - - - - .
with about half of its volume filled with liquid and the remaining volume filled with steam. The liquid volume should neither empty n'or overfill during expected transients. See Rancho Seco, 13 NRC at 583-84.
All B&W plants were licensed after a review of the pressurizer. In addition, issues involving the pressurizer were addressed in the Rancho Seco proceedings mentioned above, and the pressurizer design was found adequate. Rancho Seco, 13 NRC at 582-83; aff'd as modified, ALAB-703, 16 NRC at 1540.
These findings of adequacy are confirmed by recent
( information. A B&W owners Group review of transients since 1981 (the TAP data base) reveals only one instance (Rancho Seco, December 1985) in which pressurizer level indication has been off-scale low. See SPIP, App. I, 0
$ 4.
In additibn, a c~omparison of pressurizer sizes shows that several non-B&W plants have smaller pressurizers than B&W plants l
and that B&W pressurizers are consistently sized with those of l
l l
i ,
l l --18/ It should be noted, however, that the entire lower hemisphere of the pressurizer is below the lower level tap on all B&W units, so that an off-scale low reading does not necessarily mean that the pressurizer is empty of water. See sensitivity Study, vol. II, C-24.
l 1
. o ether PWRs when considered in terms of the ratio of RCS volume to pressurizer. volume.19 All PWRs undergo a contraction in co,olant liquid volume following a reactor trip because of the reduction in reactor coolant temperatures. The contraction in the coolant volume results in a reduced volume of pressurizer liquid. Recent studies show, however, that the post-trip outsurge of B&W plants is not significantly different than that for other PWRs.
Sensitivity Study, Vol. II, C-23 to C-24. The B&W pressurizer design thus does not present a transient response appreciably different from other PWRs.
r i;
i 19/
~~
- Pressurizer RC System Volume Volume Ratio NSSS
. V p
V RCS Y RCSd3 p Vendor Plant
. 1500 ft 3 10,000 ft 3 6.67 B&W All operating plants 1800 ft 3 11,232 ft 3 6.24 W Larger operat-ing plant 900 ft 3 6,166 ft 3 6.85
- CE Smaller operat-ing plant 1200 ft 3 8,776 ft 3 7.31 CE Larger operat-ing plant 1500 ft 3 9,520 ft 3 4.3S CE Three larger operat-ing plant 2250 ft 3 11,800 ft 3 5.24 B&W All construction permit plants See alyo Sensitivity Study, vol. II, C-12, and Tab 7.
l
\
l t- -
t
e o 26 -
In these circumstances, it is clear that the NRC's findings as to the adequacy of the pressurizer design remain valid. '
Pressurizer design thus does not present a substantial health and safety issue.20
- 3. Auxiliary Feedwater System (AFWS)
UCS alleges that the AFWS design is unreliable. UCS also asserts that the AFWS has not been upgraded to " safety grade." Petition at 11.
A description of the AFWS is set forth in NUREG-0667 at 5-34 to 5-39.
Simply put, the AFWS provides make-up water to the secondary a
coolant in the steam generator when the main feedwater system becomes inoperable.21 It is important to note that the B&W plant design does not make the AFWS any more important than in other PWRs (i.e., it
- supplies cooling water when the main feedwater system fails).
20/ UCS suggests that the pressurizer design in combination with the OTSG may produce an overcooling event which could cause vessel failure.
scenario in whichIn particular, operation of the UCS postulates ECCS leads to anaovercooling pressurized thermal shock (PTS) event. Petition at 10. PTS is the phenomenon that may result from severe overcooling of the reactor vessel concurrently with high pressure. The Commission has expressly found that "the PTS risk for B&W plants is not significantly different from that of other PWRs." Final Rule: Analysis of Potential Pressurized Thermal Shock Events, 50 Fed. Reg. 29937, 29938 and 29941 (1985). Since the B&W reactors operate in compliance with the commission's PTS regulations, 10 C.F.R. S 50.61, there is no reason to second-guess the Commission's prior findings on the PTS risk of B&W reactors. For a further discussion of PTS see n. 33 infra.
21/ There is no single B&W AFWS design.
The AFWS for each plant is designed by the utility's architect / engineer. See TMI-1 Restart, 17 NRC at 832; 20 NRC at 10; Rancho Seco, 13 NRC at 598. As a result, the systems differ significantly. Comparisons of AFW systems collectively at B&W plants with those of other PWRs do not necessarily lead to definitive conclusions.
4
- ~. w ,,,-w--,,, --wn 4-----,-------=------ ,, ,,-,,,,-a,w---,--,-,-,.,--,,,--,,r,,-,----. wen,,,-,,-+,,-,,+-,,+--.a --n, - ,,--
o .
Post-TMI upgrades to AFW systems at TMI-l and Rancho Seco were extensively explored in hearings. '
In Rancho Seco, the Licensing Board found that the AFWS was reliable and would be further improved by completion of the long-term modifications. 13 NRC at 598-605, aff'd ALAa-655, 14 N.R.C. at 805-08; ALAB-746, 18 NRC at 753-56. In the TMI-l Restart case, the Appeal Board found that "the reliability of the (AFWS] for a small break LOCA or loss of main feedwater is not significantly different from [AFW] systems at other nuclear power plants and that it is sufficiently reliable to adequately protect the health and safety of the public." 17 NRC at 835, aff'd 20 NRC at 10.
Moreover, the time available to take corrective action following a failure of both the main feedwater system and the AFWS in a B&W plant is generally comparable to or better than that for other PWRs; the alternative means for decay heat removal capability are generally as good as or better than other PWRs. See NUREG/CR-4471, "Los Alamos PWR Decay-Heat-Removal Studies: Summary Results and Conclusions (1986);
see also sensitivity Study, vol. II, Tab 15. For those B&W plants that are subject to greater cooldown rates from overfeeding of emerg,ncy feedwater, the cooldown rates do not endanger any equipment.
- Sensitivity Study, Vol. II, at A-14 to A-15.
The operational history of the B&W plants has placed focus on the AFWS due to undercooling and overcooling events. (Some of these events are discussed in Section D.) As a result, in 1980 the Staff established an NRC Task Force to conduct a review of B&W operating plant AFWS reliability. The Task Force recommended that the ATWS be upgraded to meet safety-grade requirements or, as an alternative, that consideration be given to the addition of a dedicated ATWS (i.e., a
e .
separate train).
See NUREG-0667 at 5-41 to 5-44. Consistent wi,th NRC direction and contrary to UCS' allegations, the ArW systems at B&W plants now represent safety-grade systems. See NRC April 15, 1986 Response to Congressman Markey Question 27, Enclosure, and discussion below; see also Staff SER on Emergency Feedwater System Review for
- TMI-1, dated February 18, 1987. Davis-Besse and Bellefonte were i
originally licensed with safety-grade ArWS. The other B&W plants have provided redundant components, back-up power supplies, quality assurance and survivability in harsh accident environments for the ArWS.
The Task Force also recommended that the ArWS be automatically initiated and controlled by safety-grade features independent of non-i' safety systems to preclude steam generator dryout and overcooling of the RCS. See NUREG-0667 at 5-41 through 5-44., This has been accomplished.
See NRC April 15, 1986 Response to Congressman Markey Question 27, Enclosure.
NUREG-0737 Item II.E.1.2 required that the t
specific function for initiation of the ArWS be safety grade.
l As a parallel action, after the TMI-2 accident, additional short-term actions were required which resulted in upgrading the ArWS.
i These actions included plant-specific hardware changes (e.g.,
I automatic start features and ability to start pumps from a vital bus) as well as procedural changes (e.g., to control ATW independent of the ICS).22 See NUREG-0667, at 5-39 to 5-40 and Appendix A. Long-term I
22/ Also, NUREG-0578 recommended actions to remove automatic actuation
~~
of the ArWS from the ICS, since the ICS was not safety grade.
NUREG-0578 at A-30 to A-31. This change has been made at each See plant or procedures have been adopted to accomplish the same objective. (For Bellefonte, the ArWS was designed with safety grade instrumentation and controls independent of ICS and powered (rootnote continued to next page.)
. _ , _ , - - - - ' = " " ' ' ' * * - - * ' " ' " ~ " '
I actions were also taken such as (to take just a few examples) installation of two motor-driven ArW pumps per unit at Oconee and the addition of a new motor-driven ATW pump at Davis-Besse, installation of a safety grade ArW control system independent of the ICS and connecting the motor-driven ArW pump to a vital bus at ANO-1, Crystal River 3, TMI-l and Davis-Besse. Id.
As a result of an incident in 1985 (Davis-Besse), the NRC re-visited the ATWS reliability issue. See NUREG-1154, at 4-3. The B&W Owners Group embarked upon a Task Force Action Plan directed to NUREG-1154 and, among other things, the reliability of the ArWS. This Task Force effort lends support to the previous NRC findings of ATWS reliability.
The Task Force efforts reviewed the ATWS for the purpose i
of identifying any changes that may be necessary. The Task Force found, based upon a review of a significant amount of information, that aside from the Davis-Besse experience, there had been only one 1
ATW turbine overspeed occurrence (which was due to a valve position .
error) at all other B&W plants; that there had been no steam piping design problems or turbine governor reliability problems; that the reliability analysis confirmed a high degree of reliability; that the present ArW/ETW design is adequate; that ArW/ETW testing is being l
performed to required standards; and that common mode failures were properly addressed.
As a result the Task Force concluded that no (rootnote continued from previous page.)
by safety grade, diesel generator backed pov9e supplies (see Belefonte FSAR, Chapter 7.4.2.3.)). UCS attempts to imply that
{ the ICS controls the ArWS; it does not. See NRC April 15, 1986 t
Response to Congressman Markey, Question 777 -See Toledo Edison Company's dated September " Course of Action to Support Restart ol Davis-Besse,"
10, 1985.
.~ --- -.-.--- --. - - , - - - . . , . . . . - , , _ , . . , - _ . , - - ..,-.,,,--cw,,,_a---- ,,,-,-,.,,_-.,.,,,.e--,=--. , , - - - - - - , , , ,
30 -
additional immediate action need be taken with respect to the AFWS.
See SPIP Appendix H at 3-22 through 3-26.23 In sum, the AFWS' design was initially found acceptable by the NRC; operating history uncovered problems with the delivery of AFW cooling; the matter was thoroughly reviewed by the NRC, and where improvements were required they have been taken. At this time the safety-grade requirements have been incorporated into the AFWS for the B&W plants.
Thus, while events may transpire which involve the AFWS, it can be seen that the matter has been fully considered, and that actions have been taken to assure the system's reliability to a level where it is reasonable for the Staff to conclude that the system is adequate and that no substantial safety issue exists. Nevertheless, further review of the AFWS is underway as part of SPIP, and further recommendations for improvements may be developed from that review.
4 '. Integrated control System (ICS1 UCS alleges that the ICS design has " heightened" concern at B&W plants "because failures in the ICS cause the equipment it controls to malfunction." Petition at 12.
A description of the ICS design and function is set forth in NUREG-0667 at 5-49 through 5-50. Simply, the ICS coordinates the flow of steam to the turbine and the rate of steam production. It is designed to prevent reactor trips in many instances such as load changes or feedwater heating upsets, loss of a reactor coolant pump, loss of a main feedwater pump or (as originally designed) a turbine i
' 23/ In the SER issued February 18, 1987 for TMI-1, the Staff found the ArWS at TMI-l to be acceptable.
t 6n--,. ,- .,,_,,,,,-,_.g,.rg..,
p--- ..-,g_w,,. - - - - . , - ,-,,p_- --
31 -
trip from 100% power.24 The power for the ICS is not considered Class lE because the system is not part of the plant protection system; however, most B&W plants provide power to the ICS from the Class lE vital buses. This system was reviewed by the NRC in the initial licensing of B&W plants and found to be acceptable. See, e.g., Oconee Unit 1 SER, supra, at 50. Because the ICS is not a safety-grade system, no credit is taken for the ICS in accident analyses such as those in FSAR Chapter 14 or 15.25 That is to say, the plant is designed to shut down safely after all anticipated accidents without reliance on the ICS. The ultimate safety of the plant is assured not by the ICS but by other systems that are safety grade.
Operational events have brought attention to the ICS. In response to I&E Bulletin 79-27, the B&W plants reviewed buses supplying power to non-safety control systems such as the ICS and made appropriate modifications. See NRC April 15, 1986 Response to Congressman Markey i
Question 19.
This review was expanded as a result of a 1980 incident (Crystal River) with the NRC requesting information in six specific areas. NUREG-0667 at 5-60. As a result, several improvements were made. These included (1) improving the ICS/NNI power bus separation and signal path channelization, (2) improving ICS/NNI multiple l instrument failure indication, and (3) providing ICS/NNI power bus redundancy. See NRC April 15, 1986 Hesponse to Congressman Markey Question 27, Enclosure.
--24/ It is important to note that the value of integrating the control
- ' of key plant variables is not greater on a B&W reactor than on other PWRs, i.e., the ICS was not designed to mitigate the alleged sensitivity of the B&W plant's design.
25/ See, e.g., TMI-1 Restart, 17 NRC at 847.
---._-__-.-_,,-.y
. . . . . , - , ____.,.,__..e. .+ - , . . . , - , _ , - _ , _ , , . . . - . . , - . . . , _ _ . _ - , , _ , . _ , , . . , , - . - . _ , , . _ _ . , , - . y -, - - - - _ . - --.--,
In addition, one of the long-term actions ordered by the Commission following TMI-2 was that B&W licensees perform a reliability analysis including a failure mode and effects analysis of the ICS. This analysis is set forth in BAW-1564. BAW-1564 was reviewed by Oak Ridge National Laboratory. As a result, the Staff requested B&W plants to address two BAW-1564 recommendations (i.e.,
, ICS power supplies and ICS input signals). The Staff reviewed BAW-1564 and the B&W licensees' responses to the recommendations in BAW-1564.
Based on these reviews the Staff stated that it " believes the (B&W]
design meets all current regulatory requirements." See April 15, 1986 NRC Response to Congressman Markey, Question 22, Enclosure 4 (Staff SER relating to TMI Action Plan, Item II.K.2.9). The Staff also stated that:
"In addition, since the Staff has not identified any specific control system failures or actions that would lead to j
unacceptable consequences, the Staff does not believe that any additional immediate licensing action is warranted at this time." .
(Id.)
The adequacy of BAW-1564, the Oak Ridge review of it, the functions of the ICS, and relevant improvements following TMI-2 and the February 1980 Crystal River 3 event, were confirmed in -
l the Rancho Seco and TMI-l hearings. For Rancho Seco, see LBP ! 12, 13 N.R.C. at 568-73. For TMI-1, see LBP-81-59, 14 N.R.C. at 1282-93. UCS did not participate in the TMI-l hearings on the ICS.
UCS makes much of the fact that after the TMI-2 accident the l
Staff recommended that the ICS be safety grade. Petition at 11.
As noted in response to a Congressional inquiry, the NRC's l
- -.__ -.,.~-. _ - .
- - . . ~ . . . . . - .- .. _ __
e ..
1 recommendation was not supported by analysis, but rather "was offered for consideration by the NRC Staff subsequently assigned to review the issue". NRC April 15, 1986 Response to Congressman Markey Question 23. An in-depth review of B&W information (BAW-
- 1564) by oak Ridge led to the conclusion that it was not l necessary to make the ICS a safety-grade system. This status of j
the ICS as a non-safety system is the same as the corresponding non-integrated control systems at other PWRs. See NRC April 15, 1986 Response to Congressman Markey, Question 22, 4
Enclosure 4, at 4-5.
Recent programs of the B&W owners Group lend support to the conclusion that the ICS is adequate. Prior to the initiation of SPIP, the B&W owners Group Transient Assessment committee evaluated certain ICS design features. As a result,
- recommendations were made to-reduce reactor trips. See SPIP, I
Appendix C at 2-01 to 2-7. These rec,ommendations have now been incorporated into the SPIP recommendation tracking system.
The B&W Owners Group Transient Assessment Committee also initiated a program regarding the improvement of ICS response to j input failures. See SPIP, Appendix D.
! This effort _resulted in
{ recommendations to improve the reliability of ICS inputs. Sei' SPIP, Appendix D at 2-1 to 2-6. These recommendations are also factored into SPIP.
In addition, the B&W Owners Group assessment of NUREG-ll54
- resulted in an action item (13) to assure that plant maintenance procedures are adequate to prevent ICS malfunctions due to corrosion. While the investigation of this issue found that all l
l
- . \
I l
l B&W plants had some form of maintenance on the system, the Report concluded that the B&W owners Group Isc Committee should dev'elop generic guidelines for utility preventive maintenance programs.
These guidelines have been written and corresponding maintenance programs are being implemented. See SPIP at App. J.
As a result of the 1985 incident at Rancho Seco, the B&W owners Group initiated several reassessment actions, despite the fact that the incident was plant-specific (see section D). These include:
Conduct additional training on loss of ICS power; Assure that the transient that results from loss of ICS power has been properly assessed and any deficiencies corrected; i
Assure plant procedures have adequately addressed symptoms and consequences associated with the transient that results from loss of ICS power; 4
Reassess testing of loss and restoration of ICS power; Familiarize operators with 1985 incident; Perform root cause analyses; Review assumptions made regarding availability and use of ICS controls and instrumentation; Reevaluate selected component positions on loss of ICS power; Review specific ICS power actionsr-Review maintenance of ICS.
SPIP at IV-27 to IV-30. These action items have been incorporated into SPIP. SPIP at IX-2 to IX-13.
The SPIP includes a system review of the ICS. This review i
. includes substantial input from the B&W Owners Group I&C
- Committee (see SPIP at IX-3 to IX-4). The objective is to
I perform a comprehensive review of the ICS, to improve its reliability and to limit the consequences of failures and the ICS contribution to trip frequency and the severity of transients. Work is currently ongoing, and improvements have already been made, such as adding annunciation for power supply problems in the ICS and NNI.
The Staff was aware of this fact when it denied UCS' request for emergancy relief. See NRC letter to UCS, dated March 13, 1987. In addition to these improvements, the Sensitivity Study makes a further recommendation to improve ICS reliability. Sensitivity Study, Vol. II, at B-6 to B-7.
In sum, the ICS was originally found to be adequate; operating history has resulted in improvements which make the system even better. The more significant changes involve separating AFW from ICS, improving power supply and input signal reliability and causing ICS-controlled equipment to go to a known safe state on power failure. The B&W owners Group l
is continuing to examine the system with the thought of making additional improvements. The fact that the ICS is being l improved upon does not indicate that there are current deficiencies. This point-is supported by the Advisory Committee on Reactor Safeguards (ACRS) which concluded that "the ICS is an asset to plant safety and reliability."26 oak Ridge has likewise concluded that the ICS "is not a shortceming i
i 26/ Review and Evaluation of the Babcock & Wilcox Nuclear Steam
~-
i Supply System, prepared by a Task Force of the ACRS Staff and Fellows, dated April 7, 1980, at 20.
4
+ , - - - , , --
,--e.,.,.--,,-m,,---.n.,--n--.--- . . . . . - - - , , - , - . . _ ,.. ,.--,.- - --. .-,-n.- - .nn-----.----. - - - - - -
as might be inferred from current suspicion. . . , instead it is a significant asset to plant safety. . . .
"27 See also'a similar finding in the NRC's April 15, 1986 Response to t
Congressman Markey's Questions 21-23. Thus the system is adequate and being made even better. Certainly given this history, it cannot be said that a substantial safety issue exists.
5.
Non-Nuclear Instrumentation (NNI)
UCS alleges that the B&W plants are unsafe in part because the NNI is non-safety grade and has proven incapabl'e of compensating for the alleged " inherent instability of reactor 4
temperature and pressure after a disturbance in the main feedwater system." Petition at 13.
NUREG-0667 provides a description of the NNI. NUREG-0667 at 5-50 to 5-52. The NNI provides input to various plant control systems and supplies control room information to assist the operator in determining plant conditions. The NNI also provides signals for reactor coolant system make-up and pressure control.
The NNI was not designed or even intended to compensate for reactor temperature and pressure variations.
The power for the NNI system is not Class 1E because the system i
is not part of the plant protection systems. However, for l
27/ Report Review:
i Integrated Control System Reliability Analysis, Oak Ridge National Laboratory, Instrumentation and control Division (January 1980) (FIN-B-0731), quoted in the report cited in note 26 supra at 20.
reliability purposes, most B&W plants do provide NNI power from
{
Class 1E vital buses. NUREG-0667 at 5-50 to 5-52.28 The NNI system was described in each plant'sRFSAR and the plants were licensed after a review of the system. See, e.g.,
oconee Unit 1 SER at 49-50. operating history has led to the 4
conclusion that the NNI should be. improved upon and it has been. Specifically, I&E Bulletin 79-27 requested B&W licensees 4
to review buses supplying power to non-safety related 3
instrumentation that could affect the ability of the plant to shut.down. In response to I&E Bulletin 79-27, the B&W plants implemented specific design modifications, administrative controls and procedures.
See NRC April 15, 1986 Response to Congressman Markey Question 19(D).
NUREG-0667 required a set of safety-related indicators to be located in the control room. NUREG-0667 at 5-56. Such
] improvements have been implemented at those B&W plants that did not previously meet these requirements. See NRC April 15, 1986 Response to Congressman Markey, Question 27, Enclosure, at 7.
The staff's review of BAW-1564 resulted in a recommendation to improve the reliability of the NNI power supplies. This i
28/ It should be noted that information provided by NNI is not
-~
' essential to plant safety. Procedures and training are in place which enable the straight-forward identification of failures of NNI and enable continued plant control and operator awareness of major parameters. The B&W Owners Group provided answers to extensive questions related to the NNI in a January 11, 1985 letter to G. Lainas, NRR Division of Licensing. Monitoring of key plant variables is provided by Regulatory Guide 1.97 equipment and the Safety Parameter Display System.
l i
,c.- --- -., .n. - . - .. . - _ . - - - - . - - -
recommendation has been implemented. See, e.g., Staff SER for Crystal River Unit 3, NUREG-0737, Item II.K.2.9, at 2 (incidded in NRC April 15, 1986 Response to Congressman Markey, Question 22, Enclosure 4) (noting modification made at Crystal River 3).
Subsequent to the TMI-2 accident, the NRC and B&W licensees took a number of actions in the area of control systems.
NUREG-0667 at 5-57 to 5-58. The NRC also has sought information concerning the following (NUREG-0667 at 5-60):
- power event upsets on NNI t
susceptibility of B&W plants to the 1980 event at Crystal River information available to operators concerning a power event upset on NNI testing to determine reliability of information after a power upset of NNI corrective actions NUREG-0667 provided a list of conclusions and recommendations aimed at improvement of the reliability of the plant control ,
system, the establishment of a minimum set of parameters that will enable the operator to assess plant status, the provision of adequate indications of subcooling margin and installation of i
safety-grade containment high radiation signals. NUREG-0667 at 5-61 to 5-65. The status of actions is set forth in the NRC's April 15, 1986 Response to Congressman Markey, Questions 23 and 27.29 29/ Regulatory Guide 1.97 requires some NNI instruments to be made safety grade. These actions have been and are being accomplished in accordance with the Regulatory Guide 1.97 process.
<m-- --,--,- swv----,c ,w ,,,,-n,--,--- -,,-g.-,,,----,r- ..~-- ,w.w,
t s .
Recent programs support the conclusion that the NNI system is adequate. The B&W Owners Group determined as a result of the Rancho Seco transient in 1985 to proceed with the following reassessment actions regarding NNI (many of which had already been taken):
assure that B&W documents have adequately addressed the transient that results from loss of NNI power; assure that plant emergency procedures have adequately addressed the symptoms and consequences associated with the transient that results from the loss of NNI power; reassess loss and restoration of NNI power tests; familiarize operators with the 1985 Rancho Seco event; review root cause of NNI power failure of this event; review B&W assumptions regarding availability and use of NNI controls and instrumentation; reevaluate selected component positions on loss of NNI power; review maintenance of NNI.
SPIP, at IV-28 to IV-30. These actions have been taken. SPIP, App. J.
The SPIP includes a system review of NNI. This review, as in the case of ICS, includes substantial input from the B&W Owners Group IEC Committee (see SPIP, at IX-3 to IX-4). The objective is to perform a comprehensive review of NNI, to improve its reliability and to limit the consequences of failures and NNI contribution to trip frequency and the severity of transients.
Work is ongoing; however, improvements have already been made, such as adding annunciation for power supply problems in the ICS
and NNI. The NRC has been advised of these improvements and i referenced this fact in its denial of UCS' emergency petition. '
See NRC letter to UCS, dated March 13, 1987.
In sum, NNI was found to be adequate at the initial licensing stage; upgrades have been made in response to operating history which further improve the system. Therefore it cannot be said that NNI gives rise to a substantial safety issue.
D. The Operating History of B&W Reactors Does Not Demonstrate An Undue Risk To Public Health And Safety UCS claims that the operating history of B&W reactors shows that the plants are unsafe. Petition at 14-19. The events cited by UCS fall into three categories: (1) pre-TMI events, (2) the TMI-2 accident itself, and (3) post-TMI events. Each is addressed below. '
i
- 1. Pre-TMI Events UCS first cites a list of electrical power failures on the ICS or NNI dating from 1974 to 1979. Petition at 14. As noted above, these occurrences were reviewed by the NRC in NUREG-0667 at 4-1 to 4-15 and App. B. A special NRC Task Fcree known as the B&W Reactor Transient Response Task Force was established at that time by the Director of NRR to assess the generic aspects of operating experience at B&W reactors. The NRC did not find that
) these occurrences warranted a finding of a substantial safety issue and recognized that significant post-TMI requirements were being imposed on B&W plants as part of other NRC efforts, including those of the Lessons Learned Task Force and the TMI
.,._,,,,,..m___yr-,w_.._megy, _r .,y.,,y_,y-,.n.,. , , , , . , , _ , _ . , ,,,,-.,_,_.-_.,,,.....m .,,.,,,_.,___._m-...e.,
Task Action Plan which was later incorporated into NUREG-0737.
see NUREG-0667 at 3-1 to 3-10.
It is important that these occurrences be understood in their proper perspective. First of all, these occurrences were not the type of serious accidents for which the plants are designed. It is recognized that transients of this nature will occur in a
- facility as complex as a nuclear power plant, and thus the plant is designed to safely cope with these events. As a result, the occurrences cited by UCS cannot be said to compromise the public health and safety, second, while the data cited by UCs (Petition at 14) appear numerically significant, the facts show just the opposite. The list of ICs or NNI power supply failures actually indicates less than one per year per plant -- well within the frequencies indicated in ANSI N.18.2-1973. Similarly,' the list of automatic reactor trips (Petition at 15) indicates less than three per year per plant which is well below the current industry average,30 and the list of " loss of feedwater transients" (Petition at 15) shows only about one per year per plant for a transient that by itself does not pose a substantial risk to public health and safety.
Cf., Rancho Seco, 13 NRC at 574. Further, the frequent mention of PORV openings before 1979 fails to acknowledge that during that time the plants were designed for the PORVs to open to prevent reactor trips. See Rancho seco, 13 NRC at 580.
l I 30/ See Trends and Patterns Report of Unplanned Reactor Trips at UT3. Light water Reactors in 1985, prepared by the Office for Analysis and Evaluation of Operational Data (August 1986).
Third, the primary contributors to the transients in question were generally power failures rather than the systems themselves.
Such instrumentation power failures are not isolated to B&W plants.
Furthermore, as discussed above, improvements directed to the ICS and NNI power issues have been implemented as a result '
of I&E Bulletin 79-27 and NUREG-0667. See Section C.4 and C.5 supra. Additional efforts to improve ICS response to input failures and to address other effects associated with loss of NNI power are being evaluated as part of SPIP. Id.
In short, the petition places undue and unfounded emphasis upon these pre-TMI events and simply ignores the major efforts of the past six years to improve plant performance.
- 2. The TMI-2 Accident UCS claims that the TMI-2 accident " evidenced many of the same characteristics as'ptevious accidents at B&W plants. . . . "
Petition at 15. The implication is that the TMI-2 accident casts doubt on the safety of the B&W reactor design. Contrary to the i
UCS assertion, it is well known that the TMI-2 accident became serious because of factors unrelated to the nuclear steam supply system.
The events at TMI-2 were a result of equipment malfunctions, operator actions and, more fundamentally, a j
failure, not of analyses, but of understanding the symptoms of i
l breaks at certain locations (e.g. small-break LOCA transients).
TMI-l Restart, 17 NRC at 920-21. The TMI-l restart hearings, in t
which UCS participated, did not result in a conclusion that the TMI-2 accident called into question the basic B&W design. See
( ALAB-729, 17 NRC at 894-95.
l l
It is true that the Confirmatory Shutdown orders imposed on B&W plants required implementation of certain specific actions '
before the plants were allowed to restart.31 However, the implications of the TMI-2 accident, including operator actions as well as many design issues, were not restricted to asw reactors or even to PWRs. NUREG-0737 requirements in such areas as operating procedures, control room modifications, training, post-accident monitoring, upgrading safety and relief valves, and emergency planning, applied across the board to all or almost all plants (though some actions were vendor-specific). See NUREG-0737, Enclosure 1. In short, the TMI-2 accident did not
{
primarily involve alleged B&W sensitivity issues.
- 3. Post-TMI Events UC5 claims that the operating performance of B&W reactors since the TMI-2 accident "provides no basis to conclude that the hazards inherent in their design have been corrected." Petition at 16. Specifically, UCs cites the 10 " abnormal transients" that have occurred at B&W reactors since rebruary 1980 as evidence that problems continue to recur. Petition at 17.
The term " abnormal transient" is equivalent to the Category C transients described in SPIP. These are transients "where system conditions reach limits which require safety system response or operator response to mitigate." SPIP, Vol. 1, VI-3. There have been 10 such transients since February 1980. The three most 31/ see, e.g.,
-- NUREG-0578, "TMI-2 Lessons Learned Task Force 5tatus 4.
Report and Short-Term Recommendations" (July 1979) at l
l
~
. . l significant transients (Crystal River 3 (February 26, 1980),
Davis-Besse 1 (June 9, 1985) and Rancho Seco (December 26, 1985))
have been studied in depth by the'NRC and industry and necessary corrective actions have been taken or are under review. See NUREG-0667 at 1-1 to 2-15; NUREG-1154, " Loss of Main Auxiliary i
Feedwater Event at the Davis-Besse Plant on June 9, 1985" (July 1985); SPIP, Vol. 1, IV-25 to IV-30. As previously noted, the NRC has found that the 1985 transients at Davis-Besse and Rancho Seco were plant specific. NRC April 15, 1986 Response to Congressman Markey Question 18. Further, the NRC has noted that no offsite consequences resulted from these transients, or any of the other transients refstred to by UCS. See Letter from Chairman Palladino to Congressman Matsui, dated April 23, 1986, Enclosure 1, at 1.
All the abnormal transients have been reviewed by SPIP, and the findings of SPIP lend support to the hec's eravious findings.
i See SPIP Vol. 1, VI-17 to VI-20, VIII-1 to VIII-3 dad App. I, all 4 of which have been submitted to the NRC. In general it was found that such transients result from a failure.to balance heat removal and heat production under post-trip conditions.
Significantly, eight of the ten Category C transients resulted from excessive primary to secondary heat transfer (overcooling),
while only one resulted from inadequate primary to secondary heat transfer (undercooling). SPIP, Vol. 1, VI-19.
As noted, a risk assessment has been performed as part of i
SPIP by reviewing the PRAs for Crystal River-3 and Oconee against the Category C transients that have occurred. See"SPIP, Vol. 1, I
.-. . - - . -- . . . - _ _ _ = -
l Section VIII. This evaluation concluded that these Category C transients are not significant to core-melt risk at B&W plants.
SPIP, App. I,~5-2.
. Eachoof the category C events has been considered 7
individually in Appendix I of SPIP (see esp. 55 4 and 7). The
. sequence of each transient was reviewed, as well as the root l
causes of the trip and the response that caused Category C
{ criteria to be experienced. SPIP, App. I,
- IV-1 to IV-2. In each i instance the subject plant was brought to a safe shutdown j
condition so as not to compromise the public health and safety.
t SPIP, App. I, 57.
Two of the transients (ANO-1 (April 7, 1980) and Crystal
]
River 3 (June 16, 1981)) were triggered by loss of off-site
! power.
In addition, at least three of the transients were the h
result of plant-specific features (Davis-sesse 1 (March.2, 1984)
! (wiring error) and (June 9, 1985) (several equipment
! malfunctions) and Rancho Seco (December 26, 1985) (plant design
- differences). Four of the ten Category C events occurred with i
j the plant at low power during initial startup or at high power I
- (returning to full power) but with less than normal decay heat l load, so that the risk of offsite consequences was inherently l
i lower and design margins were greater. In each of the ten
! events, the plant was brought to a safe condition by following I normal procedures.
And in none of the events were there offsite f
I consequences adverse to public health and safety. See NRC Letter i
to UCS, dated March 13, 1987; NRC Letter to Congressaan Matsui, Enclosure 1, dated April 23, 1986.
l
-t m- - - - - - - - -. -+=+c,www-_ er,w,---.+--s-wei-ww,..e----e
j A different perspective on recent operating history is'also provided by reviewing the distribution of Category "C" tran'sients
- . in the past five years. Two have' occurred at Davis Besse, three H 4
at Rancho Seco and one at Crystal River. A major improvement i
program.has been completed at Davis-Besse (see Toledo Edison a Company's Course of Action, supra, dated September 10, 1985; and NUREG-il77 (restart SER)); a major improvement program is underway and will be completed at Rancho Seco before restart (see 4 Restart Report submitted December 15, 1986). The 1985 Category C transient at Crystal River was actually not significant in that only one parameter (OTSG 1evel) exceeded the C range by a slight amount. SPIP, Vol. I, at VI-7.
The SPIP review has identified detailed recommendations to i
improve plant response and reduce the frequency of Category C transients. SPIP, App. I, f 7. Thus steps are being taken to minimize the number and severity of these types of transients.
f To put the post-TMI operating history of B&W plants in yet another perspective, over the last three years the scram l
frequency of Bsw plants has been better than the industry average. Further, of six key indicators of safety performance ;
used by INPO, B&W plants are above average performers. Appendix
, A presents this information graphically. Furthermore, in 1984 l
Bsw reactors not only had the lowest reactor scram rate but the highest availability of any group of plants in the industry.32 These basic indications serve to underscore the safety of B&W reactors.
32/ In 1985 oconee 2 established a new world record for
~~
continuous days at power -- 439 days.
UCS also claims that the December 26, 1985 transient at Rancho Seco calls into question the adequacy of the short-te'ra post-TMI requirements. Petition at 18. UCS cites, in particular, the June 1979 SER which found that these short-term requirements had been satisfied by Rancho Seco. Petition at 19.
UCS alleges that the event on December 26, 1985 demonstrated that these requirements were ineffective. Id. The focus of UCS' allegation is on the procedures and operator training that had been adopted for controlling feedwater flow in the event of an ICS power failure. As the NRC has noted, the previous Staff findings as to procedures were based on the then current event-oriented emergency operating procedures at Rancho Seco. See NRC April 15, 1986 Response to Congressman Markey Question 24. In 1985, when the licensee adopted symptom-oriented procedures, existing procedure.s for loss of ICS power were not included. The NRC's Incident Investigation Team found that during the December 26, 1985 event operators were unable to terminate flow to the OTSG (before restoration of ICS power) because the ArW (ICS) flow control valve had failed open, and the manual isolation valve was stuck open. NUREG-1195 at 6-8. The NRC fully addressed the implications of this event in response to questions from Congressman Markey during hearings in April 1986. The NRC noted that "a loss of ICS power at other B&W plants would not have resulted in as severe an over-cooling event as that experienced at Rancho Seco due to plant design differences and previous actions taken to compensate for a loss of ICS power event." NRC April 15, 1986 Response to Congressman Mackey, Question 18. It
.=. . ._ . . . - _-
is evident, therefore, that the 1985 Rancho Seco transient did not have generic implications casting doubt on the adequacy of
- post-TMI changes.33 In sun, the operating history at B&W reactors should not lea'd the Staff to conclude that a substantial safety issue exists.
The most significant transients have been studied in depth by the NRC and industry and necessary corrective actions have been taken or are under review. Within the past five years the significant Category C transients have occurred at only two plants. Within that same period a comparison of key safety parameters indicates above average performance records for Bsw plants. Furthermore, even though the risk significance of category C transients has 4
been reevaluated and found to be bounded by previous PRA results, the SPIP has identified many detailed recommendations which should continue to improve plant response and reduce the I
frequency of Category C transients.
33/ UCS makes much of the fact that during the 1985 Rancho Seco transient the plant's limits for pressurized thermal shock l (PTS) were exceeded. Petition at 10 and 18. All asW
(
i reactors operate in compliance with the PTS regulations in 10 C.F.R. I 50.61 in order to ensure the design life of the l
reactor vessel. As the NRC has noted, current operational i PTS limits 1, XI-9. for B&W reactors are conservative. See SPIP, Vol.
These PTS limits represent supplemental guidance for the operator in maintaining plant conditions so as not to
, reduce the life of the vessel. The B&W owners Group is i reviewing this guidance to ensure that the conservatism in I the limits does not result in confusion to the operator.
This review is being conducted in conjunction with the B&W Owners Group Materials Program in order to assess the potential advantages of relaxing the PTS limits.
vol. 1, XI-10. See SPIP,
- With regard to Rancho Seco, while the PTS operating guidance was exceeded, the values reached were within the conservatism discussed above.
i l
4
- . . . . - - - - - . - - _ _ _ _ . . -- ~._ - .-,. - - -
l E. NRC's Reexamination of B&W Plant
, safety Ras Not Been Compromised UC5 alleges that the ongoing reassessment of B&W plants "has been compromised in scope and schedule by NRC's delegation of its regulatory responsibility.to the owners of the B&W plants."
Petition at 19. UC5 also advances a series of arguments directed
) to (1) bias,34 (2) scope,35 (3) substance,36 (4) schedule,37 and (5) backfit.38 Each of these points is discussed below.
At the outset it is important to note that the NRC has great discretion in deciding how to carry out its responsibility to l assure that the public health and safety is being adequately protected.
Baltimore Gas & Electric Co. v. Natural Resources Defense Council, 462 U.S. 87, 97 (1983).
4 In the exercise of this
- discretion the NRC determined that it would be appropriate to
)
34/ UCS alleges that the B&W owners Group will be biased since-they are on record as saying plants are safe now. Petition
, at 20-21.
j.
35/ UC5 relies upon NRC and ACRS statements that the previous B&W Owners Group program was not addressing safety, but rather 4
economic performance. Petition at 21.
k l
-/36UC5 relies upon NRC questions on the B&W Owners Group program and concludes that two fundamental objectives of the program are not being addressed: comparison of B&W plants with other i PWR's and a determination of whether the B&W design is i adequate. UCS also complains that B&W Owners Group has not committed to implement any of the program recommendations.
Petition at 21-23.
( 37/ UCs voices concern that because of the assignment of the i
reassessment to the B&W Owners Group the schedule has l slipped; that there is no schedule for implementation; and that there is no commitment to make modifications. Petition at 23-24.
38/ UCS asserts that the backfit rule (10 C.F.R. 5 50.109) will l be relied upon and that this will add time to any -
i implementation. Petition at 24.
l f
I j
assign initial reassessment responsibility to the Bsw owners Group. See Memorandum from V. Stello to Chairman Palladind, dated March 21. 1986. That it did so is certainly proper. Much of NRC regulation is based upon the concept that utilities are responsible for assessing the safety of their plants consistent with NRC guidance; that such information is provided to the NRC; and that the NRC thereafter reviews the material. See, e.g., 10 C.F.R. 5 50.34 which adheres to this process with respect to initial licensing documents (i.e., the safety Analysis Reports).
See also 10 C.F.R. 5 50.91 regarding the no significant hazards consideration analysis to be performed by licensees when ' seeking license amendments. The propriety of this procedural course has been recognized by the courts. See, e.g., Union of Concerned scientists v. AEC, 499 F.2d 1069, 1076 (D.C. Cir. 1974).
This procedure has been applied to the current reassessment.
The NRC provided initial guidance as to the content of the plan.
See NRC April 15, 1986 Response to Congressman Markey's Question
- 18. This was modified in joint discussion between the NRC and the B&W owners Group (note that the B&W owners Group did not arbitrarily determine the scope or content of the program). See SPIP, Rev. 03 at I-2 to I-3; see also the NRC's April 15, 1986 Response to Congressaan Markey Questions 16 and 18. The B&W Owners Group has provided and will continue to provide program results to the NRC for its review. Thereafter the NRC will review the program results to assess their adequacy. This t
in i i, i -
approach is not only consistent with the NRC regulatory scheme, it is the regulatory scheme.39 The NRC also offered other reasons for the exercise of its discretion in encouraging the B&W owners Group to "take a leadership role in this matter." See Stello Memorandum to Commission, March 21, 1986 wherein it is stated:
We did this (i.e., encouraged the asW owners Group to assume a strong leadership role) in consideration of expertise and resource availability, and in recognition of the fundamental responsibility of the utility members of the B&W owners Group for the proper design of the operating plants.
It is also important at the outset to' understand SPIP. The program is unique in technical scope and breadth. In general, its objectives are to enhance the safety and performance of the B&W plants by reducing the frequency of reactor trips and transients. SPIP, I-5. -
While there are tasks within SPIP which go beyond the use of actual operating experience, SPIP is based on the in-depth review of over six years of operating history on all B&W plants. The l access and ability to evaluate this operating experience is a unique aspect of the B&W owners Group. SPIP,Section III.
SPIP includes many analytical studies and embraces both hardware and non-hardware matters. It actively searches for 39/ Itrole.
-- should be emphasized that the NRC has not played a passive
! It has had, and continues to have, an extensive i
dialogue with the B&W owners Group. See for example the numerous questions the NRC has raised. SPIP Rev. 03 Section XI. The NRC and the asw owners Group have met in several announced meetings, and NRC representatives have attended selected SPIP meetings.
. _ . - ~
-o .
information by gathering fist-hand input from people most familiar with the plants. The diversity and breadth of its scope are illustrated by the included projects:
Sensitivity Study Operating Experience Review
' EFW/ATW System Secondary Plant Relief Rancho Seco Event Review Emergency Operating Procedures Operator Burden Project ICS/NNI Systems Davis-Besse Event Review MrW System Review Instrument Air System Risk Assessment Trip Initiator Project Operations /M41ntenance Personnel Interviews SPIP is supported by the B&W Owners Group Executive Committee which has the authority to make substantive decisions and commit the necessary resources to implement those decisions. The member utilities recognize their interdependence and have established l i
means of monitoring each other's actions. This is done in such a way as to receive executive attention. SPIP, II-l to II-6.
To provide further assurance of its technical adequacy, the
- B&W Owners Group Executive Committee has built independence into its program. This has been done by assembling an independent advisory board (IAB) comprised of four senior persons, each from i
different organizations and backgrounds and each with over 25 years experience in nuclear technology. This board consists of Professor N.E. Todreas of MIT, Mr. W.H. Layman of the Electric Power Research Institute, Mr. R.S. Brodsky of Beta Corp., and Saloman Levy of Levy Associates.
We now address each of UCS' specific allegations.
._. . . _ _ _ _ _ ___ __.__ _. _ _ _ _ _ _ ~ _ . _ _ _ _ _ _
Bias UCS' allegation of bias is so frivolous that little attention i
need be,given it. The B&W statement relied upon 40 has been supported by the NRC. See Stello letters of January 24, 1986 and March 25, 1986; see also NRC April 15, 1986 Response to Congressman Markey Question 18. Accordingly there is a sound basis for the B&W statement and no indication of bias.
Furthermore, the UCS allegation of bias completely disregards the fact that the B&W Owners Group program will be reviewed by the Staff, as well as the IAB.
Scope UCS' reliance upon statements of the NRC and ACRS are misplaced. The NRC statement 4l focused upon the ability to complete the project in 1986 rather than upon a concern over scope. However, given the level of depth that the reassessment was undertaking, the NRC subsequently recognized that a delay in the completion of the project was warranted given the nature of the information that would be provided. See Commission Meeting:
Briefing on Assessment of B&W Plants, November 6, 1986, Tr.
, 4, 78,;89-90, 91-92.
\
--40/ UCS alleges that a statement by a senior B&W official that he is "very confident" that the B&W reactors are " perfectly safe as designed" demonstrates bias. Petition at 20, 41/ UCS references the following Staff statement:
, [d]ue to the lack of specificity and the programmatic goals you have set for the BWOG
, program, the Staff is unable to conclude that t
x your program will fully respond to our concerns on the B&W design by (year-end]."
Petition at 21.
s 1
Ej .
(
l -
i l
Furthermore, if there was any Staff concern as to the ability to complete the project on a timely basis, it has been cured inasmuch as the Commission was made aware of the schedule in '
November 1986 and did not express any disagreement. Commission Meeting: Briefing on Assessment of B&W Plants, November 6, 1986.
Tr. 88-92. That is not to say that the Staff does not have questions, it does. See SPIP, Rev. 03,Section XI.
However, these are specific questions directed to details of the program and not questions as to the overall scope. The NRC's vigilance in overseeing the SPIP project demonstrates that the NRC has not abrogated its responsibility.
With regard to the ACRS statement, a concern was raised as to whether the reassessment's focus was upon plant safety, as opposed to on-line performance.42 While B&W had previously L
focused on a reactor trip reduction program in an effort to improve on-line performance, i.e., the B&W Owners Group Transient 42/ UCS references the following ACRS statement:
At the time of our Subcommittee meeting the B&
l W Owners Group program's main emphasis seemed i
to be directed at improving plant on-line performance, rather than addressing the safety objectives of the NRC-B&W reassessment initiative. Our review of this program indicates that it may lead to improved plant i
on-line performance, however, we are concerned that plant safety does not appear to be its central focus. We believe it should be.
While it is true that improved plant
' performance could represent safer operation, that is not an inescapable outcome.
Letter from D. Ward, ACRS, to V. Stello, dated July 16, 1986 at 1.
I i
Assessment Program,4 results of this program were integrated !
into SPIP, because it had focused on the frequency of reactor trips and transients. The prior existence of an on-line improvement program may have caused confusion within the ACRS and given rise to its statement. In any event, the ACRS, in a full committee meeting, essentially withdrew its prior concern, complimented the B&W Owners Group for correcting the deficiency noted by the subcommittee and acknowledged that SPIP was properly focusing on safety. See ACRS Meeting, September 11, 1986 at 243-249. The current ACRS view of the adequacy of the program is supported by the NRC. As noted by the NRC, the main focus of the SPIP reassessment is to " identify potential improvements related to reducing the frequency and severity of anticipated operational transients and thereby improving the overall safety of B&W plants."
NRC April 15, 1986 Response to Congressmen Markey j Question 16(A) (emphasis added). This point is underscored in the NRC's letter to the ACRS on this precise point. See Letter from H. Denton to D. Ward, dated August 14, 1986. Therein the Staff stated that it intends "to ensure that the scope of the specific projects in the program are broad-based . . . [and that]
this approach will ensure that the broader plant safety issues are addressed as well." Lastly, the latest revision of SPIP references interactions with the ACRS and states that as a 43/ In 1980 B&W initiated a Transient Assessment Program which was viewed as means of improving plant performance. See SPIP, Rev. 03 at IV-1 through IV-2 and Appendix B; Commission Meeting, Briefing on Assessment of B&W Plants, November 6, 1986, Tr. 15.
- f. _ _ _ _ _ _ _ . _ - ._ - -- - - - - - - - - - - - - - - -- ' '- ~~~ ~ ~~
result, "the focus of the program was broadened to emphasize improvements in safety associated with trip reduction and '
performance improvements." SPIP, Rev.03 at II-1.
Substance UCS raises five arguments to support its allegation that the B&W owners Group reassessment is deficient. Each is addressed 4
below:
- 1. UCS takes issue with the inability of SPIP to compare the overall safety of B&W reactors to other PWRs. Petition at
- 22. As noted in a March 21, 1986 v. Stello Memorandum to the Commission, the B&W owners Group was encouraged "to assume a strong leadership role in accomplishing key aspects of the overall effort required." (Emphasis added.) Implicit in the i Staff position is that it would be taking the lead in some aspects of the. effort. The B&W Owners Group has explained that it cannot make a comparison of B&W plants to other PWRs "because of a lack of criteria upon which the comparison can be made."
Letter from G.R. Skillman to D.M. Crutchfield, dated August 29, 1986. Under such circumstances it is reasonable to leave this matter with the NRC which has access to in-depth systems information for all PWRs.
However, it should be noted that SPIP has compared the key response characteristics between the B&W reactor and other PWR designs. SPIP, Rev. 03,Section V, and Appendix L. Indeed, SPIP has as its objective:
- a. to characterize quantitatively the response of B&W plants under normal conditions relative to other PWRs.
l
- b. to identify upsets and accidents where '
response of B&W plants may differ l substantially from the response of other PWRs. ~ '
- c. to evaluate the margins to various established safety limits in such upsets, identifying those upsets, if any, where B&W plant safety margins are significantly different than the margins of other plants.
- d. to recommend appropriate corrective action if the B&W safety margins are significantly smaller than other PWRs. [SPIP Rev. 03 at V-2].
To accomplish the above objectives, the dynamic performance of three B&W plants, two Combustion Engineering plants and two
- Westinghouse plants has been evaluated. SPIP, Rev. 03 at V-3.
- 2. UCS takes issue with the fact that the SPIP does not compare the PRAs of B&W plants with other PWRs. A review of the NRC's original reassessment plan indicates that it did not assume that the B&W Owners Group would take the lead on this issue. See Memorandum from V. Stello to the Commission, dated March 21, 1986, Enclosure. This NRC positon is consistent with the B&W Owners Group position that it does not have access or knowledge of generalized PRAs on other PWR designs from which to make a comparison. Furthermore, the B&WOG questions the validity of a direct comparison of core melt frequencies from PRAs that use different base assumptions and methodologies. [SPIP at XI-16].
While overall comparisons of final PRA results were mentioned earlier (in Section C), the B&W Owners Group does not feel that a direct detailed comparison with other PRAs is necessarily the
=ost meaningful approach; however, the NRC, who has access to the PRAs for other PWRs, is in the best position to make this comparison if it so chooses.
A
,_-.-m. . . . . .,
, - - . _ . ~ , _ _ . , _ , , , . -, . - , . . , , . . . _.
e,a . --- -
This is not to say that SPIP does not address the adequacy of PRAs for B&W plants. This matter is thoroughly explored in the SPIP Risk Assessment Review. See'SPIP Rev. 03 Section VIII and Appendix L.
- 3. UCS takes issue with the B&W owners Group statement in response to the NRC's original reassessment plan 44 that "it is unclear as to what the 'present set of requirements' refers."
Petition at 22, citing Letter from G.R. Skillman dated August 29, 1986, supra.
It is interesting to note that the NRC revised reassessment plan does not contain the questioned language. See Enclosure to Stello March 21, 1986 Memorandum to the Commission.
In any event, SPIP is addressing the current licensing bases requirements. See SPIP, Rev.03 at V-1.
- 4. UCS takes issue with B&W owners Grcup position that Staff recommendations will be "c'onsidered" as opposed to
" consolidated" in the SPIP. Petition at 22. This argument, like the bias allegation, deserves little attention. The simple answer is that the NRC has reviewed and will continue to review SPIP.
If the NRC determines that additional matters should be considered the NRC will take appropriate action. However, i t should be pointed out that the B&W Owners Group has reviewed the 44/ The
-~ Staff's original reassessment plan included a statement that the NRC "will determine whether the present set of requirements for B&W plants are appropriate for the long-term and lead to a level of safety at B&W plants that is comparable to other PWRs." See Petition at 22.
l l
l
.,-.7-_,-y-- _.-_.,._.,_,.-,-,--,___,,g, _
,_,_,_,-y -
,.,,_,_,-~-y - .-- . . , . , - , - , , , , , - , - , - . , , , - - - - , , , , , , . - - ,
59 -
Staff's reassessment plan, and with minor exceptions, included the areas defined by the Staff in SPIP. See SPIP, Rev. 03 Table XI-l at XI-16. Further,Section XI of SPIP, Rev. 03 provides the Bsw Owners Group response to Staff questions. A reading of this section clearly shows that SPIP has included (or " consolidated")
Staff concerns.
- 5. Lastly, UCS complains that the B&W licensees have not committed to implementing all the recommendations of SPIP.
Petition at 23. As can be seen from the description of the SPIP, implementation of SPIP recommendations is an integral part of the process. See SPIP, III-2. Indeed, as set forth in SPIP, "[t]he most important aspect of the SPIP Program is the implementation of the recommendations. . . ." SPIP, Rev. 03 at III-7. As noted, the recommendations will be screened for applicability and priority by each asW* utility. Thereafter, the individual utility '
will prepare an implementation plan and carry it out. Id. It should also be noted that SPIP recognizes that many of the recommendations will lead to plant changes and that "it is the intent of the BWOG to implement these recommendations on 'a priority basis" consistent with plant outages and other plant changes. SPIP, Rev. 03 at III-8. There is room, of course, for a utility to reject a recommendation that is not warranted at its facility, e.g. where past similiar actions make it unnecessary, equipment differences make it inappropriate or operating experience indicates it to be unjustified.45 45/ Ample evidence exists to invalidate UCS' allegation that the
~~
B&W Owners Group will be reluctant to make modifications (rootnote continued to next page.)
l schedule UCS feels that slippage in the original schedule is a '
significant matter. NRC's original schedule called for completion of its program by the end of 1986. At the time this date was set the problem was only qualitatively defined and the program scope was not defined at all. Since the development of the original plan by NRC, significant discussions have been held
, with the B&W Owners Group. While these discussions have been fruitful, it required approximately six months to reach a mutually acceptable definition of the problem and the program's
- scope. In addition, the SPIP is an extensive program which requires a reasonable period of time to enable it to properly fulfill its objectives. The schedule of work is set forth in
, Appendix A to SPIP, Rev. 03. Each of the items explored is i
, necessary and, as can be seen, the time is not excessive. As noted a'bove, the Commission has been kept fully apprised of this
- schedule. See Tr. 88-97 (November 6, 1986). Simply put, the additional six to seven months necessary to complete SPIP cannot j be viewed as excessive, especially when given the benefits that i
1 have been and will be gained.
In addition, the Staff's initial estimate of time necessary to complete the reassessment was exactly that, an estimate. As more details were provided and more focus was placed on the 1
(Footnote continued from previous page.)
i recommended by SPIP. As the Staff noted: "Since the B&W plant reassessment began, almost one hundred (100) recommendations have been referred to the B&W plant owners, who have implemented, or are implementing, many of these
! recommendations." NRC letter to UCS, dated March 13, 1987.
initiative the amount of work and the amount of time were extended. Certainly, extension of the project for an additional six to seven months can be viewed'as coming within the NRC discretionary powers of determining the length of time necessary to adequately complete a project.
The fact that three supplementary revisions of the SPIP report have already been provided to the NRC and that a fourth will be submitted in April shows that the B&W owners Group has proceeded with dispatch.
UCS also alleges that there is no schedule for implementation.
The easy answer to this allegation is that until the recommendations have been made, evaluated for applicability and prioritized, it is not meaningful to set a schedule.
<, However, the B&W Owners Group, as already noted, has stated its intent to implement recommendations on a priority basis
- ~ consistent with outage schedules and other work. See SPIP, Rev.
03 at III-8. Furthermore, many items have already been implemented despite the fact that the final recommendations have yet to be made. See,NRC letter to UCS, dated March 13, 1987.
Backfit UCS complains that once recommendations are made, the backfitting rule (10 C.F.R S 50.109) will come into play. The backfitting rule requires the NRC to demonstrate, on the basis of a systematic and documented analysis, that a proposed backfit will produce a substantial increase in the overall protection and be cost justified. 10 C.F.R. S 50.109(a)(3). UCS is fearful that this process will lengthen the time of implementation.
Petition at 24.
- o. .
UCS' backfit concern is not well-taken. The B&W Owners j 1
Group's aggressiveness in addressing the complex transient j concern is adequately demonstrated by having put forth already over 150 recommendations that can improve safety and performance.
Aside from those which are not appropriate for a specific plant, it would be unlikely that the B&W owners Group would use the ;
backfit rule protection to avoid acting on their own recommendations. Accordingly, it is not anticipated that the backfit rule will come into play to any appreciable extent.
However, if a utility feels that it is in a backfit situation it is perfectly within its rights to invoke section 50.109. UCS' position is simply a challenge to the regulations, which has no place here. Cf. 10 C.F.R. I 2.758. It should be noted that the NRC was aware that perhaps not all SPIP recommendations would be
~
justified at each particular plant when it noted, in response to l a Congressional inquiry, that "[clomplete implementation of the
, changes will depend en such things as the significance of the i
safety improvement, and the extent of modifications." NRC April 15, 1986 Response to Congressman Markey, Question 25.
l l
l
F. The Current Reassessment Program Is Not An Example of Delay ,
UCS claims that the current reassessment of B&W reactor design is nothing more than an example of the NRC using " study after study" to delay taking effective action. Petition at 25.
This claim both distorts the history of the NRC's actions and is inconsistent with the case law.
Similar claims have been rejected in the past. Where a
" substantial health and safety issue" is not present, the Commission has ruled that a systematic review of technical issues is appropriate. See, e.g., In the Matter of Petition for Emergency and Remedial Action, CLI-78-6, 7 NRC 400, 428 (1978).
The courts without hesitation have upheld the NRC's judgment not to take drastic action such as ordering shutdown of plants while potential safety issues are analyzed. Lorion v. NRC, 785 F.2d at 1041; Nader v. NRC, 513 F.2d at 1055. As the court observed in '
Nader, a Section 2.206 petitioner should not be allowed to disturb the orderly process that has been established for the resolution of technical issues. 513 F.2d at 1055.
What is more, UCS' claims of NRC inaction are incorrect.
Without belaboring the point, following the TMI-2 accident, the NRC promptly ordered the shutdown of B&W reactors. See, e.g., 44 red. Reg. 27779 (1979). The operating license for TMI-l was suspended pending resolution of various issues, including B&W design issues, in the course of the restart hearing.40 As noted
--46/ B&W design issues were resolved by the Commission in In the Matter of Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit 1), 20 NRC 1 (1984); see also the Appeal Board's resolution in ALAB-729, 17 NRC 814 (1983).
4 9
.- . _ _ . . _ . . , , . , . , _ . . _ _ _ . , _ . , _ . , , _ , _ , . _ , . , . - , , . _ + , _ . , . , ,.,,,,,,,,,.,__._,,,.,,;._...,.
G o above, the Commission, on an industry-wide basis, instituted a i
series of short-term actions to improve response to off-norm'a l conditions in the secondary system. These included decreasing the setpoint for high pressure reactor trip, raising the pressurizer PORV setting, and upgrading the reliability of the
, auxiliary feedwater system. See 44 Fed. Reg. at 27779. A variety of long-term actions to improve safety were also required. These were eventually incorporated into NUREG-0737.47
~
More recently, following the transients at Davis-Besse on i
June 9, 1985 and Rancho Seco on December 26, 1985, the NRC issued confirmatory action letters requiring those plants to remain l shutdown pending an investigation, implementation of necessary corrective actions and NRC authorization to restart. See NRC letters of June 10, 1985 and December 26, 1985 to the respective 4 utilities. See also NUREG-1154, at 8-1 to 8-3; NUREG-1195, App.4 In early January 1986, as noted above, the NRC set in motion the -
. comprehensive reassessment of B&W reactor design issues that is i
currently in progress and has been vigilant in overseeing implementation of the program. See section E, supra. The various program actions being considered as part of the reassessment have been prioritized and are scheduled for i
l l
l 4jb/ UCS alleges that TMI-2 modifications have yet to be made at
! B&W plants. Petition at 6. Contrary to the UCS allegation, the B&W operating plants have essentially completed all TMI-related modifications (over 100 modifications).
48/ In addition, the B&W Owners Group undertook its own review of the Davis-Besse and Rancho Seco transients of 1985 and developed recommendations for improvements. See SPIP, Vol.
1, IV-25 to IV-29.
l l
re-.-- -- w.~. - - e ----n,-,---,~,,,------.--a w ,-,, - - - - v- -.-.,-,n.$-..--,-,-,---,---. -,w--~w--n-,e- .e y ,- ---- m- - - ,v----es
completion in a timely manner. See SPIP, Vol. 1, X-1 to X-5, and App. A.4' The NRC has shown no hesitancy to see that prompt and e effective-action is taken when circumstances warrant. The current reassessment will result in timely actions to reduce the frequency of complex transients and to improve response to those transients that do occur. Where, as here, an orderly process has been established to address technical issues, a request for precipitous action under Section 2.206 should be denied.
G. NRC Possesses the Technical capability to Assess Safety Implications of B&W Plants UCS alleges that "NRC lacks the technical capability to accurately predict the complex behavior characteristics of B&W 49/ UCS spends fully five pages dwelling on the fact that i
- Unresolved Safety Issues A-17, " Systems Interaction," and A-47, " Safety Implications of Control Systems," have yet to be resolved, Petition at 25-30. UCS asserts that the B&W ICS t
is "known to be one of the prominent examples of unresolved i systems interaction questions." Petition at 25. This is not l the place, however, to raise concerns about the Staff's
' handling of Unresolved Safety Issues. Cf. In the Matter of Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-728, 17 NRC 777, 806-07 (1983)(final Staff resolution of unresolved generic safety issue license).not necessary precondition to grant of operating USI A-17 and USI A-47 will be resolved for all affected plants in the normal course of the NRC's USI program. In any event, Oak Ridge National Laboratories, which is reviewing the USI A-17 issue, has had the benefit of extensive participation by Duke Power Company on systems interaction studies for Oconee. This information will help determine apply to B&W theplants. extent to which the USI A-17 resolution will With respect to the ICS, we have addressed this system above and shown that no substantial health and safety issue is present.
9 e -, --- -
.,,---~,,--e -
2 "
plants in accident conditions. . . . Petition at 30. In support of this allegation, UCS states that "the NRC Staff 2 failed to demonstrate a factual basis for its claim that B&W plants can achieve adequate core cooling using two methods of cooling called the ' boiler-condenser mode' and '
feed and bleed' cooling." Petition at 31.
UCS also alleges that NRC "has not devoted the resources necessary to acquire (the requisite technical] capability."
Petition at 30. In this regard, UCS points to statements concerning the adequacy of NRC computer codes to predict B&W plant behavior during accidents or operational transients. UCS complains that the NRC has failed to conduct programs to a confirm the validity of the computer codes. Absent the confirmatory work, UCS implies that NRC cannot have the requisite confidence in its ability to analyze complex transients and accidents in B&W plants. Petition at 35-38.
UCS' allegations are addressed below.
UCS' allegations are based on mischaracterizations of the record in the TMI-l restart proceeding. They are also rebutted l by an examination of the credence given to the Staff by the l
l Commission, the Atomic Safety and Licensing Boards and the Atomic Safety and Licensing Appeal Board, as well as the courts
- regarding safety issues in general and B&W issues regarding accident and transient response analysis specifically. See I Union of Concerned Scientists v. Atomic Energy Commission, 499 F,2d 1069, 1072 (D.C. Cir. 1974); see also Matter of Metropolitan Edison Company (Three Mile Island Unit 1), ALAB-729, 17 NRC 814, 843-48 (1983).
l - _
Boiler-Condenser Cooling UCS alleges that during the TMI-l restart hearings the N'RC stated that natural circulation could be achieved by boiler-condenser cooling.50 However, UCS avers that the Staff, while acknowledging that "there are no experimental data . . .
confirming the boiler condenser mode of natural circulation,"51 relied upon NRC computer codes which were shown to be in error and that this demonstrates a lack of Staff technical competence. Petition at 32.
, 50 The Appeal Board described the concepts of natural
-~/
circulation and boiler-condenser cooling as follows in the j TMI-1 restart proceeding (17 NRC at 830)(footnote omitted):
If the emergency feedwater (ErW) system is available, core cooling may be accomplished by natural as liquid or two-phase) to the steam circulation of reactor coolant (either generators, where heat is transferred to secondary water which in turn converts to steam. Natural circulation is dependent upon the' difference in reactor coolant density in '
the reactor core and the steam generators. If the reactor coolant system is relatively free of steam bubbles, liquid (also called single-1 t
phase) natural circulation can be maintained.
I If there is substantial steam formation at the high points of the reactor coolant system, however, cooling would depend on the establishment of a type of natural circulation .
referred to as the " boiler-condenser" mode.
In this process, core decay heat generates steam, which rises through the hot legs to the steam generators, where it condenses. Water then flows through the cold legs to the core, where the process begins anew.
1 i 51/ As discussed later, the Appeal Board found that "the heat removal calculations include sufficient conservatisms to make a full scale test of the boiler-condenser process at TMI-1 unnecessary before restart." ALAB-729, 17 NRC at 848; see also n.124 at 846.
l l
l l
l
, _ _ . _ . . _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ , . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . . , _ _ _ _ . _ _ _ . . . . . . _ . - . . . . - . . . _ . . - . . - _ . , - . . _____._m. . . . .. _ _ -
An examination of the Appeal Board's decision shows UCS' allegation to be misleading. As reflected therein, the record had been reopened by a prior Appeal Board decision (ALAB-708, 16 NRC 1770 (1982)) to provide further evidence on "the ability of the boiler-condenser mode of natural circulation to remove enough decay heat to prevent core damage. . . .
ALAB-729, 17 NRC at 842. While the Appeal Board made a variety of statements as to the evidence presented, with respect to the specific point of the viability of the boiler-condenser mode, the Appeal Board relied upon analyses performed by both the licensee and the Staff. Indeed, with regard to the Staff's analysis, the Appeal Board stated that "we consider the Staff analysis credible." ALAB-729, 17 NRC at 845, n.119.
UCS' focus is not on the Staff analysis that was accepted
, by the Appeal Board, but upon an analysis performed by EG&G, a l ,
research organization that conducted core cooling studies for the NRC. While the Appeal Board found that the NRC witnesses were unable to adequately explain the EG&G analysis because I
they had not performed a detailed review, it found that "this does not disturb our conclusion that the boiler-condenser process can remove adequate core heat to allow HPI flow to prevent core uncovery." ALAB-729, 17 NRC at 848, n.118. The
~ Appeal Board decision was affirmed by the Commission. CLI 11, 20 NRC 1 (1984).
What UCS has done is to focus on a report that was not prepared by the NRC and show that the NRC could not adequately respond to questions on the report. UCS then leaps to the
e .
i conclusion that the NRC does not possess the requisite technical capabilities. What it ignores is that the Staff d'id perform an analysis, that it was able to answer questions on that analysis and that the Appeal Board found it to be credible. Under such circumstances it cannot be.said that the staff is technically unaware of these issues.52 UCS then turns to Staff statements that allegedly call the boiler-condenser mode of cooling into question, absent testing.
UCS contends that since testing is not being conducted, the NRC is not in a position to pass upon the adequacy of boiler-condenser cooling and thus should be viewed as technically incapable in this area. Petition at 31-33.
In response, as it has already been noted, the Appeal Board found the boiler-condenser mode of cooling to be capable of removing core decay heat and that full scale testing was not necessary given the conservatisms of the heat removal calculations. However, the Appeal Board was aware that " future experimental work is planned to investigate the boiler-condenser mode of cooling at an integrated systems test facility . . ." (ALAB-729, 17 NRC at 848, n.137), and in dicta, ,
encouraged that this work proceed. That the-NRC has subsequently stated that continued research is needed to assess 52/ It should be noted that not only did the Appeal Board find the Staff's analysis to be credible, it relied on the Staff with regard to other issues before it. ALAB-729, supra, 17 NRC at 843-848. This only goes to underscore the technical capabilities of the Staff, which capability was subjected to the scrutiny of the hearing process, and the cross-examination of UCS itself.
t i
_ _ _ . _ _ ~ . , _ _
the effectiveness of the boiler-condenser process does not contradict the Appeal Board findings, but rather should be viewed as a staff effort to gain even more knowledge about a response mechanism with which it is already familiar. In this regard, the NRC has encouraged the B&W Owners Group to participate in the Multi-loop Integral Systems Test (MIST) program. The B&W Owners Group is participating. This program acknowledges that B&W plant behavior in accidents and transients is confirmed by detailed calculations. However, MIST intends to confirm these calculations through a code confirmation process. See transcript of NRC meeting with B&W Owners Group, March 19, 1987. At the March 19, 1987 meeting, the Staff stated its agreement with the basic safety of the B&W reactors.
Under the above circumstances it cannot be said that the staff is unfamiliar with boiler-condenser cooling because it is calling for additional research, nor can it be said that a substantial safety issue is present. See Power Reactor Development Co. v. International Union of Electrical, Radio and Machine Workers, 367 U.S. 396, 408 (1961) (" nuclear reactors are fast-developing and-fast-changing. What is up to date now may not, probably will not, be as acceptable tomorrow"). See also Lorion v. NRC, 785 F.2d at 1041-1042; Nader v. NRC, 513 F.2d at 1055 (ongoing research does not mean there is a substantial safety issue with the system).
l
- e Feed and Bl'eed UCS alleges that the NRC has demonstrated a technical inability to properly analyze the feed and bleed mode of core ,
cooling.53 UCS refers to a staff analytical assessment that was questioned by the Appeal Board because of a subsequent NRC contractor report. Petition at 33-34. UCS also implies that the Appeal Board found that the NRC's knowledge of feed and bleed was deficient and that further analysis was necessary.
Petition at 34-35. Each of these points is addressed below.
With regard to the Staff's testimony concerning the adequacy of feed and bleed cooling, the Appeal Board did not i
find that the NRC's testimony "to be without merit" as stated by UCS. Petition at 34. Rather the Appeal Board recognized j
that the testimony provided "some support for the position that feed and bleed can provide adequate core cooling. . . . " ALAB-
--53/ The Appeal Board in ALAB-729 described the feed and bleed process as follows (17 NRC at 848-49):
The method of core cooling referred to as
" feed and bleed" relies on cool makeup water being added to the reactor coolant system at a sufficient rate to replace the hot coolant that escapes. Decay heat is removed by
- allowing the incoming water to absorb some of that heat and then be replaced by more cooling water. The makeup water is supplied by the HPI pumps initially from the Borated Water Storage Tank (BWST). When the BWST is emptied, the Decay Heat Removal (DHR) system can be used to supply water from the containment sump to the HPI pumps. The reactor coolant is expelled from the reactor coolant system through the break, the PORV, or the safety relief valves.
l l
1
O t 729, 17 NRC at 852. However, the Appeal Board focused on the uncertainties associated with the relevant analyses. Rather than totally rejecting the Staff testimony, it said that "we are unprepared to state conclusively that feed and bleed will successfully provide core cooling. . . . " Id.54 In reviewing the Appeal Board's decision it is clear that the NRC Staff fully understood feed and bleed, so much so that it was able to characterize the uncertainties involved in the analyses.
Thus it cannot be said that the Staff was technically incapable in this area. What remains at issue is UCS' second allegation, that absent additional analysis, the NRC remains deficient in its knowledge of B&W plants. The source of the Appeal Board's concern were two 1982 Board Notifications (BN-82-93 and BN-82-107) which reflected that work conducted by EG&G for the *.Z1C called into question the ability of feed and bleed to keep the core covered. See ALAB-729, 17 NRC at 849-850.
It was on this basis that the Appeal Board concluded that additional research was necessary if the NRC sought to rely upon the feed and bleed mode of cooling.
54/ It is important to note that neither GPU Nuclear nor the NRC Staff relied upon feed and bleed cooling. Feed and bleed became an issue in the TMI-l Restart proceeding only when the Licensing Board questioned the reliability of the EFW System, and thus concluded that feed and bleed
!, capability was necessary to remove decay heat in some circumstances. On appeal, the Appeal Board and the Commission upheld the reliability of the EFW System, making it unnecessary to consider feed and bleed as a vital ,
contributor 11, 20 N.R.C.
to decay heet removal capability. See CLI at 8-10. .
-_-.--_..-,,.--_,._..---_-_,-_,-,,.~,,--.-.,%- ,,e.,y-..r.- , . ,-..-- - -- ---,,,.y. wy .--.-,%,y-,-,,,.v,----,,-r--r-mm
0 9 Consistent with the Appeal Board's observation, additional feed and bleed analyses have been conducted by NRC contract 5rs in 1985 which confirm the adequacy of the feed and bleed mode of cooling. See NUREG/CR-4471, prepared by the Los Alamos National Laboratory, at 123 ("[t]he feed-and-bleed mode can be used successfully in all instances in which the feed mode has been determined to be successful"). Since such investigations have been conducted, it cannot be said that the Staff is without knowledge on this matter.
Underscoring the adequacy of the feed and bleed mode of cooling are additional tests which are being conducted as part of the MIST Program. This testing, conducted in late 1986, i
confirms testimony in the TMI-1-restart record,55 as well as NUREG/CR-4471, that feed and bleed is a viable mode of keeping the core cooled. This information will be provided to the Staff in the near future.56 In sum, while the above clearly demonstrates the Staff's technical capability the Staff has
- indeed pursued further analysis of feed and bleed. It has 4
factored in the Appeal Board's comments as to uncertainty and further examined the viability of the feed and bleed process.
Codes UCS also relies upon Staff statements that the codes which model complex transients at B&W plants are in need of i
55
-/ ALAB-729 adequacy of the feed and bleed mode.
referenced a series of analyses that supported the 17 NRC at 850-852.
1 56
-/ It should be noted that the feed and bleed issue is relevant to other PWRs and is not just a B&W reactor issue.
l
~.- - - _ - - _ - - _ _ - _ _ _ - - . - - _ _ _ _ _ _ . _ _ - . - . . _ - , _ - - - - _ . - - - . . .
. o 4
reassessment. This, UCS' argues, is an indication that the NRC is without knowledge of aspects of B&W plant behavior during accidents. Specifically, UCS relies upon a staff statement that "NRC safety analysis codes now have a limited ability to predict the outcome of B&W plant transients and accidents."
Petition at 37. This statement is based upon the claim that codes to evaluate B&W plants are based on data from other PWRs, not B&W plants, and that B&W data should be collected through actual testing.
This allegation fails to recognize that, irrespective of the codes used, the Staff is satisfied that existing B&W analysis provides sufficient detail of B&W plant behavior during accidents and transients. See March 19, 1987 NRC meeting with B&W Owners Group, Tr. 20. UCS also fails to recognize the existence of the B&W MIST Program. MIST and supporting facilities were specifically designed and constructed to address the unique features of the B&W design (i.e., '
the hot leg U-bend and steam generators). Data from MIST and other facilities are being used to benchmark the l
adequacy of system codes, such as RELAP-5 and TRAC, for predicting abnormal plant transients. This information will be provided to the NRC and any question as to the adequacy of the codes and NRC's understanding thereof will be addressed.
In sum, based on the above discussion of boiler-condenser i
cooling, feed and bleed and computer codes, it cannot be said~
the Staff is without knowledge of these matters. It follows that no substantial safety issue exists.
- o l
H. The NRC Has Provided_An Adequate Basis to Justify Continued Operation and Construction .
UCS asserts that the NRC has "no technical basis" for its decision to allow continued operation and construction of Bsw reactors pending completion of the ongoing reassessment program.57 Petition at 38. On the contrary, the NRC has a more than adequate basis to permit continued operation and
, construction.
First of all, the NRC has licensed the. operating B&W reactors upon making the definitive finding of safety required by section 182 of the Atomic Energy Act. As the Commission has i
recognized, a " presumption of safety" attaches to the issuance of an operating license.58 This presumption is not only based on the NRC's reviews and findings in the licensing process, but
- is also supported by the NRC's vigilant inspection and '
- enforcement activities. See Union of Concerned Scientists v.
l l NRC, 711 F.2d 370, 383 (D.C. Cir. 1983) ("the NRC maintains ,
constant vigilance over the safety of nuclear power plants and monitors compliance with safety requirements at each nuclear 4
j reactor on a day-to-day basis"). UCS' claim that the NRC lacks a technical basis to allow continued operation is merely an 57/ As noted above, the fact that there may be open safety issues does not warrant halting construction of a nuclear power plant. Porter County Chapter of the Izaak Walton
' League, 606 F.2d at 1369. There is therefore no reason to question the NRC's technical basis for allowing continued construction of Bellefonte Units 1 and 2. The opportunity for addressing open issues as part of the operating license proceeding provides an adequate forum.
l
--58/ See Final Rule: Revision of Backfitting Process for Power Reactors, 50 Fed. Reg. 38097, 38103 (1985).
i
,--.w,, ~ - , n.n n,, -_,-_,_.,_,,,,,...,.,_,,---..----.,_g-n.-,,.m..n_
, . - - . , , __ _, .-n--,-w_,, , w a p. , g--,, - - - -e..--
- , l I
I attempt to shift the burden to the NRC and licensees to justify continued operation, when the burden under Section 2.206 properly rests with UCS to show otherwise. See 10 C.F.R. I 2.732' ("the proponent of an order has the burden of proof") .
1 As noted in the NRC's March 13, 1987 letter denying UCS' request for immediate relief, the NRC is also aware that substantial efforts have been undertaken as a result of (1) the short-term and long-term actions arising out of the TMI-2 accident (see NUREG-0737), (2) the actions resulting from the February 26, 1980 transient at Crystal River 3 (NUREG-0667),
(3) the previous actions of the B&W Owners Group to reduce the frequency of reactor trips (see SPIP, Vol. 1, IV-1 to IV-25),
(4) the activities initiated by the B&W Owners Group as a result of the June 1985 Davis-Besse transient and NUREG-1154 (see SPIP, Vol. 1, IV-25 to IV-27), (5) the previous actions
- initiated by the B&W owners Group in response to the December 1985 trdnsient at Rancho Seco (see SPIP, Vol. 1, IV-27 to IV-30), and (6) the information developed as part of the Sensitivity Study, MIST and SPIP. It is essential to note that to date neither the NRC nor the B&W owners Group has identified a substantial safety issue during the course of the ,
reassessment. Moreover, as the NRC's March 13, 1987 letter noted (at p. 2), "[s]ince the B&W plant reassessment began, almost one hundred (100) recommendations have been referred to
, the B&W plant owners, who have implemented, or are implementing, many of these recommendations." All this demonstrates that safety improvements have been made and will
.I i
. . 1 continue to be made at plants that were licensed upon NRC findings of safety. l l
Further, the NRC is aware that transients in 1985 at Davis-Besse and Rancho Seco did not present generic safety issues.
As the NRC noted in the March 13, 1987 letter (at p. 2)
"[w]hile both events were significant, plant specific problems caused them, and the events are not indicative of safety problems in general at B&W-designed plants that would require immediate staff action."
It is also a telling point to note that UCS has presented no technical studies of its own to substantiate its allegations that B&W reactors are unsafe or to show the existence of a substantial safety issue. Instead, UCS has chosen to rely upon information well known to the NRC.
In these circumstanc'est there is a sound basis to permit continued operation and construction. The NRC has confirmed this point. See NRC April 15, 1986 Response to Congressman Markey, Question 18; Letter from V. Stello to R. Tucker, dated January 24, 1986; Letter from V. Stello to R. Pollard and E.
Weiss, dated March 25, 1986. The adequacy of these NRC findings was discussed in our Initial Response and is incorporated herein.
III. CONCLUSION A Section 2.206 petition may be granted only upon a showing
-- and Staff finding -- of a " substantial health and safety i
i issue." UCS has made no such showing here. Instead, UCS has a
, , y-n...-,. .- .---,---,--.=,.,.---,-.--s..nn-,----..,--a-- --.--,,----,n_. ---.,---.,.,,---,---.n_.-.,.------n __-n,-,-
presented old information with which the NRC is already intimately familiar and which previously has been found by the NRC not to warrant shutdown of B&W reactors or a suspension of construction permits.
As the foregoing discussion shows, there is a wealth of information demonstrating that the operation and construction of B&W reactors do not present a substantial health and safety issue.
This information includes initial licensing review, staff reviews of the operating experience of asW reactors, the implementation of upgrades in response to operational transients, the post-TMI improvements that were applied to all plants, and the information generated to date as part of the ongoing reassessment and other studies of B&W design issues.
The NRC is also aware that the causes of significant transients in 1985 d,id not indicate generic problems. While uncertainties may have existed sufficient to warrant instituting the current reassessment program, such uncertainties do not equate with a finding of a substantial health and safety issue. See Lorion
- v. NRC, 785 F.2d at 1041. Further, the reassessment program
shows that significant progress is being made which has the potential for allaying the NRC's concerns. ~
In these circumstances, the petition should be denied.
I Respectfully submitted, J. Michael Mc arry Daniel F. Stenger BISHOP, COOK, PURCELL &
REYNOLDS 1200 17th Street, N.W.
Washington, D.C. 20036 (202) 857-9800 April 6, 1987 Attorneys for the B&W Owners Group
\
9 l
l
4 b CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing " Principal Response of B&W Owners Group to Petition Filed Under 10 C . .
FR.S 2.206 by the Union of Concerned Scientists" were served by hand delivery (as indicated by an asterisk) or by deposit in the U.S.
mail, postage prepaid, on the following this 6th day of April 1987:
- Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ellyn R. Weiss Robert D. Pollard i
Union 1616 PofStreet, Concerned Scientists N.W., Suite 310 Washington, D.C. 20036
{ -.w
. Michael Mg ar I
Y 5 ,
d .w
- APPENDIX A
_ SAFETY SIGNIFICANT PERFORMANCE INFORMATION ~
Unpund himsk saa ah othd Wa* thpand ss% spisi Adarum 4 A=,
i-S- '
_i.. a u
' ' S 2-4.1 g,, 4.3 4- 1.4-1 7
. 3. 3
/ k goet (pe. ) ,,,, ,,..
(,,,, ) rear etyAng I&vAssy
.i i;. ,,
u i' '-
- ~~
~
.... . ..i ..
(,,,, f ...
G
7 9 s.
FuiIddar Volume silor4 mil Said Iniente M Mw MM 3..
EttII- 4.M et 5H. 'H 3.g.gg 459 5 '
ett
. it.61 3* 4N-
- . 3!7
= 1[.3ja 1300-t[.g8- 32H-
.' 1 Eltl5 3f00-i ett.n -
8
. 1985 till 1981 !!83 1988 Illi stas re.. i r. , tus (,.. 3 fusiIsioWih Yolume of low-tml Said Rodoa5n m I & V Awaje 8&WAway
=,4. ii
- sa-Sol in a-
- 5. .
= -
-iq.es.
a g
- '"I 190$ itM 8 1984 itg3 etal (f f 81.), f ees fece o
( ,,l g l
l
_, 's A n.
ced= wen > *N Se imi-r r , aniw m N he her m in- I'"' e.n in-in ns j'"' e.u
'82
,,e. sn 1 "'
g 1982 IH3 1914 1983 1g46 1982 1981 1134 a ggg- gyg
" I"n i ren fe .. . I h Raidell % 'edusfrial Safely lost-Tee kodent Role B & W Awaje ggyyg in. ...
l l
l Filah 'in ins im nu ine (sne i,
!ilh.. ,n . .u ,,..
(y,, j.
l l
l l
- - - - - - - - - - - - - - - - - , --