ML20195D645
| ML20195D645 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/03/1988 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19292H962 | List: |
| References | |
| CAW-88-056, CAW-88-56, NUDOCS 8806230161 | |
| Download: ML20195D645 (13) | |
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v Westinghouse Power Systems Neear tecneica 3'5'" b5*
Electric Corporation b 355 Pmseggh Pennsylvama 15230 0355 June 3, 1988 CAW-88-056 Dr. Thomas Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Co::nission Washington, D.C. 20555 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMAIIOri Fh0M PUBLIC LISCLOSU.%
Subject:
Sequoyah Nuclear Plant Reply to NRC Inspection Report Nos.
50-327/85-24 and 50-325/83 Appendix R
Dear Er. Murley:
Be proprietary inforr.ation for which withholding is being requested in the enclosed letter by Tennessee Valley Authority is further identified in an affidavit signed by the owner of the proprietary infor=ation, Westinghouse Electric Corporation. Tne affidavit, wnich accompanies this letter, sets forth the basis on which the information may be withheld frce public disclosure by the Cocnission and addresses with specificity the considerations listed in paragraph (b) (4) of 10CFR Section 2.790 of the Co==ission's regulations.
he proprietary material for which withholding is being required is of the same technical type as that proprietary material previously submitted as Affidavit AW-80-68.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Tennessee Valley Authority.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-88-056, and should be addressed to the undersigned.
Very truly yours, W -
h Rober A. Wiesemann, Manager
.. aatory & Legislative Affairs Enclosures cc: E. C. Snomaker, Esc.
Office of the Ger.eri. I:7.sel, NRC
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8806230161 880617 PDR ADOCK 05000327 i O DCD l
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PROPRIETARY INFORMATION NOTICE TRANSMITTED HEREWITH ARE PROPRIETAFY AND/0R NON-PROPRIETARY VERSIONS OF DOCUMENTS FURNISHED TO THE NRC IN CONNECTION WITH REQUESTS FOR GENERIC AND/0R PLANT SPECIFIC REVIEW AND APPROVAL.
IN ORDER TO CONFORM TO THE REQUIREMENTS OF 10CFR 2.790 0F THE COMMISSION'S REGULATIONS CONCERNING THE PROTECTION OF PROPRIETARY INFORMATION SO SUBMITTED TO THE NRC, THE INFORMATION WHICH IS PROPRIETARY IN THE PROPRIETARY VERSIONS IS CONTAINED WITHIN BRACKETS AND WHERE THE PROPRIETARY INFORMATION HAS BEEN DELETED IN THE NON-PROPRIETARY VERSIONS ONLY THE BRACKETS REMAIN, THE INFORMATION THAT WAS CONTAINED WITHIN THE BRACKETS IN THE PROPRIETARY VERSIONS HAVING BEEN DELETED. THE JUSTIFICATION FOR CLAIMING THE INFORMATION 50 DESIGNATED AS PROPRIETARY IS INDICATED IN BOTW VERSIONS BY MEANS OF LOWER CASE LETTERS (a) THROUGH (g) CONTAINED WITHIN PARENTHESES LOCATED AS A SUPERSCRIPT IMMEDIATELY FOLLOWING THE BRACKETS ENCLOSING EACH ITEM OF INFORMATION BEING IDENTIFIED AS PROPRIETARY OR IN THE MARGIN OPPOSITE SUCH INFORMATION. THESE LOWER CASE LETTERS REFER TO THE TYPES OF INFORMATION WESTINGHOUSE CUSTOMARILY HOLDS IN CONFIDENCE IDENTIFIED IN SECTIONS (4)(ii)(a) through (4)(ii)(g) 0F THE AFFIDAVIT ACCOMPANYING THIS TRANSHITTAL PURSUANT TO 10CFR2.790(b)(1).
AW-80-68 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared Robert A. Wiesemann, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (' Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
< l/ <lM:a Robert A. Wiesemann, Manager Regulatory and Legislative Affairs Sworn to and subscribed before me this 6 day of 9 980.
NotaryP5bl[c/
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AW-80-68
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t (1) I am Manager, Regulatory and Legislative Affairs, in the Nuclear Technology Division, of Westinghouse Electric Corporation and as 7
such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public dis-closure in connection with nuclear power plant licensing or rule-making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Water, Reactor Divisions.
(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790 of the Concission's regulations and in con-
. junction with the Westinghouse application for withholding ac-companyir.g this Affidavit.
(3) I have personal knowledge of the crit'eria and procedures utilized by Westinghouse Nuclear Energy Systems in designating information as a trade secret, privileged or as confidential commerical or financial infomation.
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the in-forgation sought to be withheld from public disclosure should be withheld.
(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
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C -68 (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Wertinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.
The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence
- it falls in one or more of seve.a1 types, the release of which l might result in the loss of an existing or potential com-petitive advantage, as follows:
(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.)
where prevention of its use by any of Westinghouse's competitors without license from Westinghouse consti-tutes a competitive economic advantage over other companies.
l (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, I method, etc.), the application of which data secures a' compet* tive economic advantage, e.g., by optimization ,
or improved marketability. l l
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(c) Its use by a competitor would reduce his expenditure of resources or impr ve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
i j (d) It reveals cost or price information, production cap- '
I acities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
I (e) It reveals aspects of past, present, or future West-inghouse or customer funded development plans and pro-grams of potential comercial value to Westinghouse.
(f) It contains patentable ideas, for which patent pro-tection may be desirable.
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(g) It is not the property of Westinghouse, but must be l treated as proprietary by Westinghouse according to agreements with the owner.
There arc sound policy reasons behind the Westinghouse system
~ which include the following:
(a) The use of such infomation by Westinghouse gives Westinghouse a competitive advantage over its com-petitors. It is, therefore, w*thheld froni disclosh. e to protect the Westinghouse competitive position.
, - AW-80-68 (b) It is information which is marketable in many ways.
The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c) Use by our competitor would put Westinghouse at a competitive disadvantagt by reducing his expenditure of resources at our expense.
(d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary infer-mation, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
l (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition in those countries.
(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success
! in obtaining and maintaining a competitive advantage.
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AW-80-68 (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
(iv) The information is not available in public sources to the best of our knowledge and belief.
(v) The proprietary information sought to be withheld in this sub-mittal is that which is appropriately marked in the attach-ment to Westinghouse letter number NS-TMA-2337, Anderson to Tedesco, dated November 7, 1980, concerning information re-lating to NRC t eview' of the Westinghouse treatment of the Baron Dilution transient at Comanche Peak. The letter and attachment. are being submitted in response to the NRC request at the October 15, 1980 NRC/ Westinghouse Texas Utility Services Inc. meeting.
This information enables Westinghouse to:
(a) Justify the Westinghcuse design.
-(b) Assist its cristomers to obtain licenses.
(c) Provide operational flexibility to customers when not at power assuring them of. safe and reliable operation.
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Further, the information gained from the Westinghouse treat-ment is of significant commercial value as follows:
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$-80-68 l (a) Westinghouse uses the information to perform and justify analyses which are sold to customers. l (b) Westinghouse sells analyses and the resulting electronic l devices.
Public disclosure of this information concerning analysis methods and functional requirements is likely to cause sub-stantial ham to the competitive position of Westinghouse because competitors could utilize this information to assess
, and just.ify their own designs without comensurate expense.
The parametric analyses performed and their evaluation re-l present a considerable amount of highly qualified develop-ment effort. This work was contingent upon a program which .
has been underway for the past four months. Altogether, a substantial amount of money and effort has been expended by Westinghouse which could only be duplicated by a competitor if he were to invest similar sums of money and provided he had the appropriate talent available.
Further the deponent sayeth not.
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LNCLOSURE 2 TVA Submittal Dated May 23, 1988 (L44 880523 809)
(NONPROPRIETARY]
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144 880523 809 SN 157B Lookout Place WAY 231988 1 U.S. Nuclear Regulatory Comnission ATTN: Document Control Desk Washington, D.C. 20555 l
Gentlemen:
In the Matter of ) Docket Nos. 50-327 l Tennessee Valley Authority ) 50-328 SEQUOYAP NUCLEAR PLANT (SQN) - NRC INSPECTION REPORT NOS. 50-327/88-24 AND 50-328/88 APPENDIX R By letter dated April 19, 1988 TVA responded to the referenced inspection report. On May 10, 1988, an enforcement conference was held in Rockville.
Maryland, at which TVA presented further information focused, at NRC's request, on the single issue of a postulated fire in the annulus. At the conclusion of the conference, NRC requested that TVA submit information for the record. Indication was also given that enforcement action at Severity Level III was under consideration. TVA maintains that enforcement action is not supported by the facts of this case, least of all enforcement action at an escalated leu l.
l In this letter, TVA provides the technical information requested by the NRC regarding the fire in the annulus issue and demonstrates thrt safe shutdown was assured at all times for SQN rnit 2 (see enclosure). A..achment 1, "Analysis Following Loss of Normal RCS Letdown and Boric Acid Makeup" IPROPRIETARY) and Attachment 2, "Appendix R Reactivity Calculations" ,
I IPROPRIETARY), of the enclosure contain information which is proprietary to Westinghouse Electric Corporation. Accordinf,l", we request that this )
l information be withheld from public disclosure.
In order ta not delay this submittal of information to the connission in connection with the review of SQN Appendix R related calculations, we will )
comply with the requirements of 10 CFR 2.790 to provide proprietary and nonproprietary versions, together with an affidsvit, as soon as the proprietary information contained in the submittal has been specifically identified and the proprietary and mproprietary versions have been prepared.
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i We will submit the recuired number of copies of the proprietary and nonproprietary time. versions of the information and required affidavit at that ,
In the meantime, we have provided sufficitnt copies for you to initiate your review.
It is our understanding that E. Shomaker, Esq., of the NRC Office of the General Counsel, has advised Westinghouse that he concurs with this procedure.
We expect to be able to fully comply with the requirements for the proprietary and nonproprietary versions of the information and an ,
accompanying affidavit within four weeks,
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U.S. Nuclear Regulatory Comission MAY 231988 TVA'S POSITION TVA's position regarding the cable-to-cable interactions issue was stated in the April 19, 1988 letter and reemphasized in the enforcement conference. TVA maintains that a method was proposed by SQN that was technically sound and considered by past NRC review to be satisfactory. On the basis of statements made by NRC in the May 10, 1988 enforcement conference, it is TVA's l understanding that, while NRC does not consider a change in NRC staff position to have occurred, the interactions specifically reviewed and accepted ;
previously by NRC will not be cited as examples of the all ged violation. The remaining item under consideration for enforcement action concerns the adequacy of TVA's shutdown analysis fc; a postulated fire in the annulus of SQN unit 2. In particular, NRC stated its position that enforcement would be based on the asserted inadequacy of the TVA analysis at the time of the March
, 1988 inspection. With regard to this remaining item, it is TVA's position that:
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Questions concerning TVA's analysis of the annulus fire should have been treated as an unresolved item to be resolved on a technical basis following the inspection.
NRC interpretation of section III.G.2 of Appendix R with respect to cable interactions (including spurious actuations and high-low pressure interfaces) has evolved since 1981 and has not been consistent in all cases.
Even if enforcement action is determined to be appropriate in this case, l application of the NRC Enforcement policy results in no greater than a Severity Level IV violation.
As indi:ated during the recent SQN Appendix R enforcement conference, NRC provisunally agreed that public health and safety were never in question for SQN unit 2, pending NRC review of the TVA analyses. The remainder of this letter setr fcrth TVA's position regarding enforcement action.
TREATMENT AS AN UNRESOLVED ITEM A. Generic Letter (GL) 86-10 Guidance l
GL 86-10 (page 3, section C) discusses fire protection documentation requirements.
The NRC intends to initiate enforcement action where, for a given fire area, compliance with AppendiM R is not readily demonstrable and the licensee does not have available a written fire hazard analysis for the area. (emphasis added)
4 U.S. Nuclear Regulatory Commission MAY 231988 l
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Indeed, the Commission's statement in the letter suggests that no enforcement action at all should be taken for the annulus fire issue at SQN. TVA performed a comprehensive Appendix R evaluation of SQN in 1984, including development of a shutdown model, tracing of safe shutdown and associated circuit electrical cabling, and development of documentation demonstrating how Appendix R was satisfied for each area. Additionally, other supporting calculations and evaluations were performed for specific Appendix R-rclated issues, including the reactivity issue caused by a fire in the at.nulus. Thus, TVA had the documentation called for in GL 86-10, paragraph C. Before the March 1988 inspection, TVA had identified a potential deficiency in the 1987 analysis and was in the process of resolving this deficiency at the time of the inspection by performing confirmatory analyses, the details of which are provided in the enclosure.
, Not all potential deficiencies or nonconformances are reportable or constitute violations of regulations or the license. It is important to )
evaluate identified deficiencies /nonconformances to determine reportability as well as whether the condition is outside the licensing basis of the plant. Key to this determination is the operability of components or systems as a result of the condition, i.e., safety significance. ]
The bi*ntif hetion of a deficiency does not automatically indicate inoperability or noncompliance with regulatory requirements; the l deficiency must be evaluated to reach this determination. TVA has performed reviews for the annulus fire issue and did not reach a point where a violation involving equipment operability or regulatory )
requirements existed. TVA's initial determination in this respect has 1 been confirmed by the recently completed analyses contained in the enclosure.
Clearly, SQN is not in the situation described in GL 86-10; TVA had evaluated and performed area analyscs that were available at the time of l the inspection. This analysis showed that letdown was not required to '
achieve and maintain safe shutdown, and the questions raised concerning l the analysis did not revise the analysis results. The issue is not whether TVA had performed and documented the Appendix R evaluat3.on, but whether a potential deficiency in a certain calculation assumption automatically equates to a violation of 10 CFR 50, Appendix R.
B. Unresolved Item Examples Numerous examples exist wherc documentation questions were raised in NRC inspections of Appendix R programs and were designated as unresolved items. This treatment is also consistent with the NRC Inspection and Enforcement Manual, Chapter 0400, "Enforcement program " which defines an unresolved item as, "If sufficient information does not exist to determine if a violation of a requirement occurred, the item may be identified as an unresolved item in an inspection report." A few examples of Appendix R, documentation-related unresolved items are:
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U.S. Nuclear Regulatory Cornission lAAY 231988 Fort Calhoun, July 1, 1983: The licensee had not considered the potential affect of spurious opening of an auxiliary feedwater (AFW) recirculation valve on the ability of the AFW pump to perform its postfire shutdown functions, pending the licensee's review, this was considered an open item.
Trojan, August 31, 1983:
The inspection report noted that a documented high-low pressure interface analysis had not been performed and that a documented analysis of the effects of spurious signals on safe shutdown had also not been performed. This item was indicated as not being an item of noncompliance or a deviation and would be verified once a complete analysis has been performed.
The licensee's calculation for shutdown with loss of service water pumps was questioned by the inspection team. The calculation assumed that both fire pumps would be available; however, a fire in the intake structure could cause loss of one fire pump in addition to loss of all three service water pumps. This was left as an open item pending completion of a licensee reanalysis. i I
i Fort St. Vrain, March 16, 1984: No spurious signal or common enclosure associated circuit analysis had been performed. This was designated as an unresolved item. {
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Davis Besse, August 30, 1984: The inspection report noted that the l I
licensee had not addressed the issue of pressure control while unintaining hot standby for a fire at the alternate shutdown panel.
This was carried as an unresolved item in the inspection and not a violation or noncompliance. ,
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palisades, November 14, 1986: -
The licensee had not performed a breaker coordination study f or non-safe-shutdown circuit on a common power source with safe shutdown equipment. The licensee committed to complete the analysis within approximately one year. This was an unresolved item.
During the inspection, the NRC requested technica! ..:stification for certain systems (heating, ventilating, and air cont _.oning [HVAC),
seal cooling, heat tracing) not protected as safe shutdown systems.
The licensee committed to provide documented analyses to the NRC after the inspection. This was an unresolved item.
Pilgrim, June 18, 1987: The NRC inspection team had difficulty in following the licensco's documented Appendix R analysis. The licensee committed to prepare a document to more clearly describe the detailed analysis. This was designated as an unresolved item.
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Arkansas Nuclear One (ANO), units 1 and 2, September 30, 1987:
The licensee had not performed an analysis to demonstrate that safe shutdown could be achieved without protecting the HVAC systems to meet Appendix R. This was designated as an open item pending completion of ventilation analyses. This was an unresolved item.
During the inspection, the licensee failed to provide marked-up drawings to demonstrate routing of electrical cables. The documentation apparently existed, but the licensee failed to present the requested information during the inspection. Cable separation routing was designated as an unresolved item.
Several high-low pressure interfaces were found to not have adequate interface protection, including certcin ones that were found to be unprotected for shorts and hot shorts in several areas. This was an unresolved item.
During the inspection, the licensee indicated that documented qualification test reports were not available for three different fire barrier penetration seal designs. This was an unresolved item.
Salem unit 2, January 26, 1988:
The licensee adequacy didrating)
(fire not have fire test of fire barrier documentation penetrationtoseals.
demonstrate the Thi= was cited as an unresolved item.
The licensee relied on use of pneumatic jumpers to prevent spurious operation of valves in achieving hot shutdown. The licensee pointed outthe by that this had previously been cpecifically reviewed and accepted NRC.
The inspectors felt this was a repair and thus not i
allowed for hot shutdovn. This was cited as an unresolved item.
These examples identify numerous cases where documentation deficiencies were not designated as violations, but rather as unresolved items. Several of these are items more significant than the subject TVA item in that many performed. are cases where an analysis of a particular issue or area was not examples As noted before, TVA had performed an analysis. Designation of the SQN issue as an unresolved item would be consistent with GL 86-10, other regulatory precedents, and regulatory requirements on the identification and disposition of nonconforming items.
ENFORCEMENT ACTION TVA maintains its position, stated in the April 19, 1988 letter, that previous NRC review and approval of TVA methodology for dispositioning certain cable interactions renders enforcement action inappropriate in this case. A meaningful enforcement policy is not well served by the imposition of
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U.S. Nuclear Regulatory Commission hk h h jhhh sanctions based upon the disagreement of one inspection team with the findings of three prior inspection teams reviewing the same issues. This is especially true when the technical basis for the alleged violation lies in a regulatory
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area not specifically addressed by the applicable regulation, 10 CFR Part 50, ;
Appendix R, and where NRC guidance has evolved throughout the period of j Appendix R implementation and NRC inspections.
Should NRC determine, however, that enforcement action will be taken for the single issue of fire in the annulus, it is clear that the violation should be classified, at the very worst, no higher than Severity Level IV under the Commission's current enforcement policy,10 CFR Part 2. Appendix C. revised ;
September 28, 1987. TVA has shown conclusively that an incorrect engineering judgement used in support of a 1987 safe shutdown analysis would have had no impact on safe operation of SQN unit 2 even using extreme assumptions (see enclosure). Moreover, the 1987 analysis deficiency (and thus the alleged violation based on thic analysis deficiency) represents a single, non-safety-signifierat deficiency in an extensive fire protection program.
TVA addrs.sses the basis for this position in more detail below.
A. Inconsistent Staff Interpretations of Circuit Failure Modes Required to be Addressed in the Fire Hazards Analysis
- 1. Appendix R and Initial Safe Shutdown Guidance Appendix R became effective on February 17, 1981*.Section III.C.2 requires protection of circuits needed to operate safe shutdown equipment in the event of a fire and associated nonsafety circuits, wherever fire damage to those circuits ". . . could prevent operation of cause maloperation due to hot shorte, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and )
maintain hot shutdown conditions . . . ." ,
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Three days after the effective date of Appendix R, CL 81-12 was issued to clarify the meaning of this provision. With respect to this letter, the following should be noted:
Section 4 of the "Staff position" states that reactivity control for pressurized water reactor (pWR) hot standby includes a "letdown system if required." (emphasis added) **
Although Appendix R does not apply by its terms to SQN unit 2, the unit 2 license required compliance with specific sections of Appendix R. !
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' ** TVA's basis for achieving and maintaining safe shutdown without letdown is set forth in enclosure 1. Examples are also provided of NRC acceptance of this approach on other f acilities. ,
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U.S. Nuclear Regulatory Commission n /( 231988
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Cable interactions and high-low pressure interf aces are discussed in Enclosure 2 to CL 81-12--but only in connection with the application of Section III.G.3 on alternative shutdown.
On April 7, 1982, a "clarification" of GL 81-12 was issued, which stated the following in regard to associated circuits:
It is important to note that our interest is only with those circuit (sic) (cables) whose fire-induced failure could affect shutdown. The guidelines for protecting the safe shutdown capability from the fire-induced failures of associated circuits are not requirements.
These guidelines should be used as guidance only when needed. These <
guidelines do not limit the alternatives available to the licensee for protecting the shutdown capability. All proposed methods . . . will be evaluated by the staff for acceptability.
- 2. Fitzpatrick Exemption In November 1982, the Power Authority of the State of New York (PASNY) requested an exemption "for 3 Phase AC and DC Power Circuits" (Fitzpatrick docket, November 22, 1982). This exemption was requested on a generic basis for application "to the Authority's entire (Appendix R) reevaluation." PASNY had made clear in an earlier Fitzpatrick submittal dated February 26, 1982, that, while it had identified multiple spurious actuations at high-low pressure interf aces. It considered 3-phase, alternating-current (ac) and 2-wire, direct-current (de) power circuit "reconnection" f ailures to be incredible. However, PASKY believed that 2-wire de control circuit reconnection failures were credible and thus would be evaluated for spurious actuation potential.
The NRC granted t',se requested methodological exemption--with no exception f or high-low pressure interf aces--in April 1983 (48 FR 19963). The staff accepted the licensee's argument that circuit f ailure modes of these two types (3-phase ac and 2-wire de power circuits) were not "cousidered credible." This exemption was described to the Commission in SECY-83-269 (July 5, 1983; pages A-2, -3).
- 3. Generic Letters 85-01 and 86-10 Nearly two years after the Fitzpatrick exemption was issued, the staff released GL 85-01 (issued January 9, 1985). In section 5.3 of the "Questions and Answers," safe shutdown issues are covered.
Question 5.3.1 reads, "What circuit failure modes must be considered in identifying circuits associated by spurious actuation?" The two-sentence response reiterates sections III.G.2 and III.L.7 of Appendix R and notes that circuit failure modes "could equipment. be bypassed by assuming all possible failure states for the
s U.S. Nuclear Regulatory Commission hlf(fhhlhhh The staff position regarding 3-phase ac and 2-wire de power circuit shorts established in the Fitzpatrick exebption is not included. In the response to Question 5.3.10 regarding assumed plant transients, the staff restates its position that simultaneous spurious actuations must be addressed for high-low pressure interface actuations. As in GL 81-12, however, this position is established for alternative shutdown under section III.L.7.
The latest staff guidance document, GL 86-10, reexamines these issues in the response to Question 5.3.1 (the same question as answered in GL 85-01). This response attempts to fit together the various pieces described above. First, section III.G.2 is cited in connection with the high-low pressure interface issue. Second, the "Fitzpatrick doctrine" is acknowledged for 3-phase ac and de ungrounded shorts but, as to the latter, is apparently extended beyond power circuits to control circuits as well. Third, the "Fitzpatrick doctrine" is disallowed entirely for high-low pressure interfaces without any explanation of the technical or regulatory basis for this position.
- 4. Other Examples of Inconsistent Interpretation It is clear from the above that, throughout the period 1981-1986 and probably beyond, staff interpretttion of the circuit failure requirements of sections III.G.2 and III.L.7 has undergone a steady, and sometimes inconsistent, evolution. Examples of inconsistency later than the Fitzpatrick exemption can eacily be located.
Comanche Peak Safety Evaluation Report (SER)
In a Supplemental Safety Evaluation Report (SSER) issued to Comanche Peak in 1985 (SSER No. 12, October 1985), the staff approved the licensee's disposition of a residual heat removal (RHR) shutdown isolation valve interaction (high-low pressure interface) as follows:
Power for the RHR shutdown isolation valves has been removed at the circuit breaker; therefore, spurious operation as a result of fire damage of control cables is not possible. Because the valves are powered by 3-phase A80-V ac, spurious operation as a result of fire-damate_d_
power cables between the circuit breaker and the valve motor is not considered credible. (page 9-9; emphasis added)
This SSER was written while GL 86-10 was in preparation.-yet it is clearly inconsistent with the staff response to Question 5.3.1 in that it allows grounds t disposition of a high-low pressure interface on the however, hat a 3-phaso short "is not considered credible." It is, consistent with the Fitzpatrick exemption.
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-9 U.S. Nuclear Regulatory Commission Haddam Neck Exemption In an exemption issued to the Haddam Neck plant in 1984 (49 FR 46516; November 26, 1984), the staff approved deviation from section III.G.2 separation requirements for the control room on the theory that a fire would damage at most two adjacent control panels before being extinguished. Thus, the licensee had to address circuits, including spurious operation, only by analyzing two adjacent panels at a time., No mention is made of high-low pressure interfaces, though it is obvious that a fire in the control room involves circuits for all such plant interfaces. If the concern regarding such interfaces is based on risk, some additional requirements might be imposed if two adjacent panels contained circuits that, if both were involved in a fire, could spuriously open a high-low pressure interface. The exemption imposed no such additional review requirement.
- 5. Sumary This detailed review of staff positions on circuit failure modes has been undertaken to demonstrate that these positions have evolved gradually, ano not always consistently, during the period 1981-1986.
It was during this same period that SQN unit 2 sought to implement the backfit provisions of Appendix R and was reviewed by three staff inspections.
It should not be surprising that these previous inspection teams found TVA's disposition of cable interactions acceptable, considering that (1) it is a technically sound approach, and (2) "staff positions" were neither solidified nor supported by a clearly expressed technical or regulatory basis. TVA was therefore fully justified in concluding that its approach to cable interactions constituted another method proposed by the liceasec and was considered by the staff on its own merit and was found to be satisfactory (reference CL 86-10).
TVA does not dismiss the views of the March 1988 inspection team on technical grounds and in fact agreed with the inspection team that provision of assured letdown for a fire in the annulus was a superior technical resolution. Other inspection team findings are under TVA review for long-term technical resolution; but while the inspection team was clearly entitled to reach its own technical judgments in this area of Appendix R, the unsuitability of this case for enforcement action--particularly objective at an review of the escalated level--should be clear upon f acts.
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-10 U.S. Nuclear Regulatory Commission j[gg]Qgg B. Application of Enforcement policy
- 1. 10 CFR part 2 Appendix C Statements made by NRC during the enforcement conference suggested that a Level III violation was being considered for the fire in the annulus issue. Given the facts in this case, particularly at this level, enforcement action would directly contravene the Commission's enforcement policy contained in 10 CFR part 2. Appendix C revised September 28, 1987. Given a reading of the facts least favorable to TVA, no more than a Severity Level IV violation can be found.
The Commission generally characterizes violations below Level II as follows:
Severity Level III vio'.ations are cause for significant concern. Severity Level IV violations are less serious but are of more than minor concern; i.e., if Icft uncorrected, they could lead to a more serious concern. Severity Level V violations are of minor safety or environmental concern.
As TVA demonstrated at the May 10, 1988 enforcement conference and as verified by the enclosure to this letter, the incorrect depressurization rate assumption in the 1987 peak Xenon analysis did not create a significant safety concern. In fact, safe shutdown could have been achieved using the least favorable combination of operational events and assuming that operacors did not follow prescribed maximum-boration procedures. Thus, had the problem gone uncorrected, even if this conclusion is disputed and NRC concludes that the analysis error might have led to a more serious safety concern, a Sevetity Level IV violation at most would be called for.
These conclusions are supported by application of the Commission's "Reactor Operations" guidance in Supplement I of the enforcement Policy. The examples most apropos are those involving 10 CFR 50.59 l because the alleged violation involves a written analysis as opposed to an operational event.
A Level III violation is suggested where the licensee has "(Failed) to meet the requirements of 10 CFR 50.59 such that a required license amendment was not sought." This implies that the licensee modified tne facility in such a way that an unreviewed safety question was raised or a technical specification violated without performing the appropriate analysis and requesting a license amendment. With respect to the SQN annulus fire issue, this would have i been equivalent to failure to perform a fire hazards analysis for the ares at all or to have performed only the most cursory analysis.
To the contrary, TVA performed a detailed documented analysis I addressing the annulus fire issue in 1984 and 1987. While the latter calculation was found (following SQN shutdown) to contain an incorrect assumption, which ultimately did not negate the conclusion of the 1987
4
-11 U.S. Nuclear Regulatory Commission hffkfhhlhhh calculation (i.e., safe shutdown could be achieved and maintained),
both analyses were adequate to demonstrate shutdown capability. In terms of 10 CFR 50.59, this would be similar to having performed two e
reviews of a single plant modification, the second of which was incorrect in some respect not significantly affecting safety or outcome of the 10 CFR 50.59 evaluation. This set of facts would fall under the Severity Level IV example of 50.59 deficiencies. Such deficiencies, to be less than Severity Level III, would involve an error in a 50.59 analysis that would not have affected the overall assessment of the potential for an unreviewed safety question or technical specification violation. That is, correction of the analysis would lead to the same conclusion. In this case TVA has shown that correction of the 1987 analysis leads to the same conclusion reached based on that analysis--that achieving and maintaining safe shutdown in the event of an annulus fire were, at all times, assured.
- 2. Fire protection Escalated Enforcement Precedent A review of previous Appendix R violations issued by the Commission demonstrates that Severity Level III has been assigneo for progra=matic and hardware deficiencies far more serious than the SQN ;
annulus fire analysis issue. A few examples are:
Trojan, September 29, 1983: failure to provide adequate separation of RHR pumps; f ailure to protect redundant cabling at i containment penetrations; failure to provide adequate monitoring instrumentation for alternative shutdown; and inadequate lube oil collection system.
Susquehanna, June 15, 1987: no automatic suppression system to protect RHR pumps; inadequate separation of cabic raceways in the containment access area; redundant shutdown trains not adequately separate in the Equipment Removal Room; and inadequate emergency lighting system.
peach Bottom, July 29, 1987: "numerous instances" of failure to separate redundant safe shutdown equipment and cables; and "apparent lack of management attention."
Surry, July 15, 1987: power and control cables for diesel generators not adequately separated, causing loss of reactor coolant makeup capability for both units if offsite power were lost in a fire.
Salem unit 2, March 24, 1988: "r.umerous instances" of failure to separate redundant safe shutdown equipment and cables; and "multiple examples of these significant violations."
I U.S. Nuclear Regulatory Commission g $ $hh i
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Summary ;
All of these cases, and others that could be cited, involve fire protection deficiencies that exceed in both number and potential safety impact, the single issue of the annulus fire at SQN and, within that issue, the single deficiency in the 1987 calculation.
CONCLUSION Application of the Commission's general Severity Level guidance, the specific enforcement policy guidance applicable to reactor operations, the discussion of documentation requirements in GL 86-10, and Appendix R enforcement precedents result in the same consistent conclusion: enforcement action is not warranted in this case. No regulatory purpose is served by finding a violation based upon a licensee-identified calculational deficiency not affecting plant safety, promptly addressed upon identification, and representing a very small aspect of a fire protection program found to be in co=pliance with regulatory standards in all umjor respects.
Very truly yours, TENNESSEE VALLEY AUTHORITY original signed By R. L. Gridley R. Gridley, Director Nuclear Licensing and Regulatory Affairs Enclosure cc (Enclosure): See page 13
O U.S. Nuclear Regulatory Commission h4/VY 231988 )
cc (Enclosure):
Mr. K. P. Barr, Acting Assistant Director for Inspection Programs TVA Projects Division U.S. Nuclear Regulatory Commission '
Region II 101 Marietta Street, NW, nuite 2900 !
Atlanta, Georgia 30323 Mr. G. G. Zech Assistant Director for Projects TVA Projects Division U.S. Nuclear Regulatory Commission !
One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 '
E. Shomaker, Esq., Office of the General Counsel, NRC U.S. Nuclear Regulatory Commission Mail Station 15-B-18 l Washington, DC 20555 i
R. A. Wissemann, Mgr. Regulatory &
Legislative Affairs (w/ Attachments) i Westinghouse Electric Corporation P.O. Box 355 ,
Pittsburgh, Pa. 15230 Sequoyah Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Read Soddy Daisy, Tennessee 37379 l
l l
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_. - ~ -
Enclosuro l May 10 1988 Enforcement Conference STATEMENT OF THE ISSUE The focus of TVA's presentation made during the May 10, 1988 enforcement conference centered on the annulus fire issue. Simply stated, the scenario is as follows: A fire occurs in the annulus requiring safe shutdown of the plant. The fire damages or otherwise renders inoperable reactor coclant system (RCS) letdown path equipment. The plant must then be borated to a safe shutdown condition to prevent return to criticality. Because SQN unit 2 did not have a protected letdown path before the inspection, NRC questioned SQN's past capability to have been able to borate the RCS to achieve and maintain a safe shutdown condition. Accordingly, NRC considered that TVA had f ailed to identify the letdown interaction. TVA had not identified this as an interaction because SQN's safe shutdown logic did not require a letdown path to achieve and maintain safe shutdown. It is TVA's position that, at all times before the March 1988 inspection, safe shutdown could have been achieved and maintained given a fire in the annulus.
ORIGINAL ANALYSIS Before discussing the basis for this position, it is important to review the documented analyses that were in place before the March inspection. The initial Appendix R analysis was conducted as part of the SQN 1984/1925 Appendix R reevaluation effort and was the analysis of record through the time of the SQN shutdown (August 1985). The analysis assumptions included 100-percent power, equilibrium Zenon, and consideration of beginning, mAddle, and end of core life. The results of this analysis showed that sufficient boron could be added to the RCS to maintain hot standby conditions indefinitely. Establishment of a letdown path was not required to add this boron to the system. Hot standby was achieved through reactor trip, boration to offset Xenon decay and moderator temperature coefficient, makeup provided by the charging pumps, heat removal from the steam generators, and in the longer term passive heat loss from the pressurizer.
Because the analysis was conducted for a specific core cycle (unit 1 cycle 3),
the recommendation was made that a letdown path be established so as to avoid reanalysis for each fuel cycle. System Operating Instruction (SOI) 26.3 was developed to establish this letdown path using the RHR system if the plant was to be taken to cold shutdown (CSD). A copy of this analysis is provided in attachment 1.
The use of equilibrium Xenon in the analysis is a reasonable assumption, consistent with other fundamental assumptions used in Appendix R evaluations such as normal operation, 100-percent power, and all equipment initially available for service. Additionally, it should be noted that the basis for SQN Technical Specification 3/4.1.2, "Boration Systems " states, "the maximum expected boration capability requirement occurs at 'end of lif e' from full power equilibrium Xenon conditions . .
." This is the basis for the
~
a reactivity control system, establishing the initial conditions for all design basis accidents. It would not be reasonable to require Appendix R design / analysis to exceed that capability required to mitigate design basis accidents. In summary, the equilibrium Xenon calculation provided a representative analysis, it was the documented analysis in place through the SQN shutdown, and it showed that a letdown path would not have been needed to achieve safe shutdown.
1987 ANALYSIS Well after the SQN shutdown, a second analysis was undertaken assuming peak Xenon for the initial condition. The condition was postulated to result from a fast restart. While certainly more limiting than the previous 1984 analysis, this would amount to utilization of a low probability and physically implausible transient condition as the initiation point for the Appendix R analysis.
Initial conditions were chosen to minimize boron concentration in the core. The condition starts with a reactor trip from full power with equilibrium Xenon, completion of posttrip review, a management decision to restart within 2-3 hours of the trip, maximum dilution rate initiated to counter Zenon buildin, and a step power increase from 0- to 100-percent power coincident with Xenon peaking and control rods at maximum insertion rate. An immediate reactor trip is then postulated to occur under these conditions, thereby resulting in 0-percent reactor power coincident with peak Xenon and minimum conditions.boron concentration. A fire is then assumed to occur at these Initial evaluation of this peak Xenon case indicated that safe shutdown could still be achieved; and, so as to conservatively resolve all questions, the 1987 Appendix R analysis was performed using peak Xenon. In addition to the initial conditions described above, consideration of beginning, middle, and {
end of core life and a depressurization to permit RHR cut-in in approximately 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> following reactor trip was also assumed. A copy of this 1987 analysis is provided in attachment 2.
Following the March 9, 1988 public meeting (held to address Appendix R related issues) and before the March 14, 1988 inspection, questions were raised by TVA regarding the validity of the depressurization rate used in the 1987 !
analysis. These questions were in the process of being addressed at the time of the March inspection, specifically to determine if any of the questions j might affect the validity of associated shutdown capability.
1 1988 POSTINSPECTION ANALYSIS ~
I At this point. TVA had not determined that shutdown capability was affected (by the questions) but rather initiated studies to verify that it was not. As previously discussed, TVA considered the equilibrium Xenon analysis to be ,
representative of realistic shutdown scenarios and verified that the questions raised concerning the peak Xenon analysis did not impact the validity of the equilibrium Xenon analysis results. This provided a high level of confidence l
that safety during operation was not at issue regardless of the outcome 1 regarding the depressurization rate questions. However, to fully evaluate the
more severe postulated conditions, TVA analyzed two peak Xenon cases to verify that, even assuming an unrealistic peak Xenon initial condition, safe shutdown capability input to thewas maintained in spite of the incorrect depressurization rate analysis.
20,000-parts-per-million (ppm) Boron Case The first case involved evaluation of the response of both equipment and operators that would have occurred if a fire in the annulus rendered letdown paths inoperable.
the following: TVA verified that hot standby was still achievable based on SOI-26.2 called for use of maximum boration of 20,000-ppm boron using the boric acid tank; although the 1987 analysis had assumed 2,000 ppm, the instruction clearly would h1ve directed the operators to initiate maximum boration. TVA verified that this boric acid system, which is in the Auxiliary Building, would have been unaffected by the fire in the annulus. It was determined that less than 2,000 gallons of 20,000-ppm boron would be adequate to achieve hot standby; and 12,000 gallons of volume in the RCS was available because of "shrink" resulting from the cooldown (547 to 350 degrees Fahrenheit). A conservative boration flow rate of one gallon per minute (gal / min) was assumed.
It should be noted that required boration volume and flow rate were determined by final calculation to be even smaller than cited in the enforcement conference, i.e.,
versus 20 gal / min. 2,000 versus 6,000 gallons, one gal / min As a result, it was again shown that no letdown was required to achieve and maintain safe shutdown.
It should be noted that CSD could be achieved in the longer term through passive heat loss from the pressurizer, dropping pressure down to RHR "cut-in." The supporting analysis for this case is provided in attachment 2 with input from attachments 3 and 4
Verification of this case ensured that, even above and beyond the more been provided. equilibrium Xenon case, safe shutdown capability had always representative 2,000-ppm Boron case The second peak Xenon case evaluated went one step further than the previous case.
TVA verified that, using the same assumptions from the 1987 analysis (designated shutdown equipment and 2,000-ppm boron),
have been achieved and maintained. TVA verified that hot safe shutdown could still standby was achievable through boration to replace RCS shrink volume using a boration path provided through reactor coolant pump seal flow and safety valve relief.
Boration to CSD also would have been achievable in the longer term. The su,t,)orting analysis attachments 3 and 4 for this case is provided in attachment 2 with input from OTHER PLAllT EXAMPLES Although this information was not presented at the enforcement conference, other PWR plants had similarly determined that a letdown path is not required to allow adequate boration for safe shutdown. For example:
ANO, units 1 and 2:
ANO-1 isolates all letdown paths for postfire safe shutdown. The licensee determined that adequate boration for cooldown and Xenon decay can be provided using high-pressure injection pumps from the borated water storage tank (2,270 ppm) to offset RCS inventory shrinkage (reference NRC Inspection Report 50-313/87-14, 50-368/87-14 September 30, 1987).
ANO-2 isolates all letdown paths for postfire shutdown. The licensee determined that adequate boration can be provided using the charging pumps taking suction from the boric' acid makeup tanks (8.750 ppm) or from the refueling water storage tank (1,731 ppm) to offset RCS inventory shrinkage (reference NRC Inspection Report 50-313/87-14, 50-368/87-14, September 30, 1987).
point Beach, units 1 and 2: The licensee analysis isolates normal and excess letdown paths. The required shutdown margin is maintained by the addition of water from the refueling water storage tank (2,000 ppm) using the chemical and volume control system thereby offsetting RCS volume shrinkage (reference NRC Inspection Report 50-266/87007, 50-301/87007, July 7, 1987).
It should be noted that these inspections did not raise any issues with the licensee's method of acco=plishing this safety function.
SUMMARY
For the three previously described cases--(l) equilibrium Xenon, (2) peak Ienon using real-life response (20,000 ppm), and (3) peak Xenon using the 1987 analysis assumptions (2,000 ppm)--TVA has verified that the depressurization rate calculation did not alter the validity of previous analysis results in demonstrating the ability to achieve and maintain safe shutdown.
Two options were examined by TVA at the time of the audit to resolve the annulus fire issue: (1)
(2) possible hardware upgrade to ensure letdown.further analysis and possible procedure chang Upon evaluation of these two options, it was determined that, to expedite restart of SQN unit 2 and to simplify postfire operator actions, the hardware option was preferable.
actions were prompt--the inspection took place March 14-18, 1988; are the The modification was implemented and inspected by MRC on March 23, 1988. It should be noted that the modification involved addition of approximately 10 sprinklers to approximately 7,000.
TVA later without determined plant that technical resolution could have been achieved modification.
However TVA maintains the decision made at the time was appropriate considering the need to expedite resolution for restart; the timeframe involved in reaching technical and regulatory resolution with NRC staff; and the relative simplicity in effecting a quick, conservative resolution through the modification.
While TVA understands this has been a complicated technical and regulatory issue, enforcement action is not warranted.
9 pai o
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k LICENSING TRANSMITTAL TO NRC
SUMMARY
AND CONCURRENCE SHEET DATETRANSMITTNE TO RLc MAY 198 DATE DUE NRC 5/20/88 ,
~'
SUBMITTAL PREPARED BY Clenn W. Morris FEES REQUIRED YES NO. X PROJECT / DOCUMENT I.D. Sequoyah Nuclear Plant (SON) - NRC Inspection Report ,
Nos. 50-327/88-24 and 50-328/88 Appendix R l PURPOSE /
SUMMARY
To provide Appendix R-related calculations requested during the May 10, 1988 enforcement conference and to provide further information supporting WA's position that the SON fire in the annulus issue does not j warrant escalated enforcement action. l
)
RESPONDS TO (RIMS NO.) COMPLETE RESPONSE YES 1 NO PROBLEM OR DEFICIENCY DESCRIPTION There is no problem or deficiency associated l l
with this submittal.
l CORRECTIVE ACTION / COMMITMENT There are no commitments made in this submittal. ,
CONCURRENCE SIGNATURE OR NAME ORGANIZATION LETTER PEFERENCE DATE ,
H. L. Abercrombie SQN Site Director < /B/W lo' S. J. Smith SON Plant Manager - I'/
- 8 A J. B. Hosmer SON DNE Project Engineer i .43. A h M/f/F M. R. Harding SON Site Licensing Mgr 7?I M d~/h &
Independent Verification Required? Yes No X Independent Verification Performed by Date
- Name/ Organization APPROVED MMM a
'DSLRA MANA" , '
_ DATE f!A O ff
/ /
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Sequoyah Site Licensing Concurrence Sheet i
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DUE TO RLG 5/19/88 DATE DUE h7C 5/20/88 PROJECT / DOCUMENT I.D. Sequoyah Nuclear Plant (SON) - NRC Inspection Report Nos. 50-327/88-24 and 50-328/88 Appendix R i
CONCURRENCE SIGNATURE OR NAME ORCANIZATION LETTER REFERENCE DATE G. k'. Morris SQN Licensing Engineer b' 7[M /0!bb J. S. Smith SON Site Licensing - . I k I M. A. Cooper SON Complianee Manager % <*-,M. N::,W E J. E. Sullivan SON PORS
[ //Nsa J. J. Pierce SQN DNE-MEB F. A. Koontz SQN DNE-NTB / f-418 l J. F. Lemons SON DNE-NTB E. S'//9//6
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DATE DVE NRC 5/20/88 !
PROJECT Nos. / DOCUMENT I.D. _Sequoyah Nuclear Plant (SCN) - NRC\
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50-327/88-2t. and 50-328/86-2t. - Appendix R i CONCURRENCE NAME SIGNATVRE OR OR0/WH47_ICN_
LETTER REFERENCE L E Morris SON Licensing,,Epgineer DATE My g .q _^p 7 k g S[/8!b8 l._ S. Seith SON Site tieencing
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,' '.!: ANALYSIS FOLLOWING LOSS OF NORMAL RCS LETDOWN AMD BORIC AC '
.,g As requested per Mr. Sullivan, the attached report was prepared by the Reactor Engineering Unit to assist in ascertaining whether or not the Scquoyah Nuclear Plant Ply with RCS letdownRand Appendix normal boric acid makeup systems need to be made to com-requirements.
plant's ability to obtain and maintain safeThe problem presented was to evaluate the with a station blackout, shutdown conditions if, concurrent
.'y a fire was to render these systems inoperable, and
',- boric acid makeup could only be from the Refueling Water Storage Tank (RWST). U'
,?. Safe shutdown conditions were defined as Mode 5 (K-EFF < 0 99, TAVC < 2000 , F). - ~'
- 7 The results of the analysis show that given the scenario above, l
- l h and, normal boric acid makeup systems are not the RCS 1ctdown J.j Unit necessary to place and maintain the -
? ,,' 1 Cycle 3' core in Made 5, provided pressurizer icvel is raised to 757.
within 24 to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> following reactor trip. (This analysis includes a 25 ppm .,"'-
conservation in the X-EFF calculations.) -
Because Unit 2. Also, it hasWestinghouse been in a refueling has not outage, the analysis was not performed for . "
, an evaluation. Thus, yet provided the dati necessary for such' '
Unit 2 core, and to any future coresthe applicability of the report's results to the present
(. for either unit, may be in question. For this reason, and the fact that pressurizer level must be maintaiaed in a manner that requires specific operator actions, the following suggestions are offered: -
First, that Westinghouse could be contracted to perform an FSAR type "limiting analysis"
would substantiate these or similar resulte for the lile of the plant.
Secondly, since there is a significant '
time period (more than a day) before K-EFF approaches 0.99000, procedures could be developed to specify actions to ,
be impler.ented path within that time by the plant which would ensure the restoration of a letdown frame.
- Once letdown ,is established, the RCS boron con- ~-6 centration could be increased to at least that of the RWST. It ,. ;. _ ,
probabic that a makeup path from the boric acid tanks could be restored.is also very a either or both, suf ficient With <
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final possibility is to perform this analysis for each cycle. This is the 1 cast .
attractive of the three alternatives, . "; <~ -
could result since the time to perform such an analysis '
new proceduresin unwanted delays, and the results may indicate the need to develop ' '
for each core, as indicated above.
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