ML17108A808

OEDO-16-00783 - Closure Letter for Samuel Miranda, Citizen, Email 2.206 - Enforcement Petition Regarding Exelon'S Byron and Braidwood Stations

ML17108A808
Person / Time
Site:	Byron, Braidwood
Issue date:	06/23/2017
From:	Michael Case Plant Licensing Branch III
To:	Miranda S Public Commenter
Wiebe J, NRR/DORL, 415-6606
Shared Package
ML16327A598	List: ML17108A808
References
2.206, OEDO-16-00783
Download: ML17108A808 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 June 23, 2017 Mr. Samuel Miranda 2212 Forest Glen Road Silver Spring, MD 20910

Dear Mr. Miranda:

This letter responds to your November 15, 2016, Petition addressed to U.S. Nuclear Regulatory Commission Executive Director for Operations (EDO), Victor M. McCree, regarding Exelon Generation Company's Byron and Braidwood (Byron/Braidwood) Stations (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17010A051). Your Petition was supplemented by the following documents: *

• transcript of your meeting with the Petition Review Board (PRB) on February 1, 2017 (ADAMS Accession No. ML17059C395)

• e-mail (ADAMS Accession No. ML17034A183)

• written statement (ADAMS Accession No. ML17034A179)

• excerpts from the updated final safety analysis report for Byron and Braidwood Stations (ADAMS Accession No. ML17034A181)

• transcript of your meeting with the Petition Review Board on March 15, 2017 (ADAMS Accession No. ML17089A581)

• e-mail and abstract for International Conference on Nuclear Engineering paper ICONE24-60472 (ADAMS Accession No. ML17089A582)

The EDO referred your Petition to the Office of Nuclear Reactor Regulation under Section 2.206, "Requests for Action Under This Subpart," of Title 10 of the Code of Federal Regulations (10 CFR 2.206). In your Petition, you requested that the NRG take the following actions:

1. Revoke the Licensee's authorizations to operate Byron and Braidwood Stations at any uprated power level.

2 Impose a license condition on current operations requiring the Licensee to provide an acceptable demonstration of compliance with the aforementioned design requirement

[preventing anticipated operational occurrences from developing into more serious events].

3. Require the Licensee to file a 10 CFR §21, "Reporting of Defects and Noncompliance,"

report regarding its statement of no significant hazards.

As the basis for your request, you state that the licensee's compliance rationale has omitted or mistaken important points, which you describe as 8 omissions and 11 errors. You also provided examples of what you consider to be errors in the licensee's no significant-hazards consideration. Exelon submitted the no significant hazards consideration in its July 5, 2000, power uprate license amendment request (ADAMS Accession No. ML003730544).

S. Miranda You met with the NRC's petition review board (PRB) to discuss your Petition on February 1, 2017, and again on March 15, 2017. The PRB has considered those discussions in determining whether or not the Petition meets the criteria for consideration under 10 CFR 2.206.

After careful consideration, the PRB has concluded that your Petition does not meet the criteria for consideration under 10 CFR 2.206 because the issues you raised have either already been the subject of staff review, evaluation, and resolution or do not present significant new information. For example, the staff has already reviewed some of the issues in the safety evaluation (SE) dated May 4, 2001 (ADAMS Accession No. ML033040016), the SE dated August 26, 2004 (ADAMS Accession No. ML042250531), and the report of the Backfit Appeal Review Panel (BARP) dated August 23, 2016 (ADAMS Accession No. ML16236A208). The SARP report stated that the contribution of the current configuration at Braidwood and Byron to overall plant safety is very small and the report resolved issues of retrospective assessment.

Based on the very small contribution of the issue to plant safety (i.e., significance), the new information brought up in the petition would not change the NRC determinations that conclude that the Braidwood and Byron licensing basis and current configuration complies with the applicable regulations and provides adequate protection of the public health and safety. The enclosure to this letter provides further detail regarding the NRC's decision on your petition.

While not meeting the criteria for consideration under 10 CFR 2.206, consistent with the SARP report, the underlying technical issues raised in the petition appear to represent a conservative approach that could provide additional safety margin but do not constitute significant safety issues. By memorandum dated September 15, 2016 (ADAMS Accession No. ML16246A247),

the NRC Executive Director for Operations tasked the Director of the Office of Nuclear Reactor Regulation (NRR) to inform him of his plan to respond to assess pressurizer safety valve performance after water discharge and to assess RIS 2005-29, as well as its proposed Revision 1, through the appropriate generic process. By memorandum dated January 3, 2017 (ADAMS Accession No. ML16334A188), the Director of NRR provided the plan details and target dates for implementation of the plan. To the extent the issues in the petition are applicable, they will be considered during the plan implementation.

Thank you for bringing these issues to the attention of the NRC.

Sincerely, Micha J. Case, Director Divisi n of Systems Analysis Office of Nuclear Regulatory Research Docket Nos. 50-454, 50-455, 50-456, and 50-457

Enclosure:

As stated cc: Distribution via Listserv

, PKG ML16327A598 Incoming ML16327A599 Closure Letter ML17108A808 *via e-mail OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA Tech Editor(QTE) -* -- -

NRR/DE/EPNB NAME JWiebe SRohrer Via e-maillcomments included JBillerbeck*

DATE 05/11/17 05/01 /17 04/25/17 04/20/17 OFFICE NRR/DSS/SRXB/BC OE NRO/DSRA/SRSB NRR/DORL/

NAME EOesterlie* G Figueroa-Toledo* TDrzewiecki* LBanic*

DATE 5/24/17 5/16/17 5/17/17 5/16/17 OFFICE OGC/NLO NRR/DORL/LPL3/BC RES/DSA/D NAME SKirkwood DWrona MCase DATE 06/01/17 6/08/17 6/23/17 BASIS FOR NOT ACCEPTING THE PETITION REGARDING EXELON'$ BYRON AND BRAIDWOOD STATIONS Petition Issue Basis Sunnortina Discussion The licensee's unnecessary Not There is an acceptance criteria for overpressure analysis significant overpressure contained in NUREG-0800, reveals a lack of new Standard Review Plan (SRP), 15.5.1, understanding of the information. "Inadvertent Operation of ECCS [emergency inadvertent operation of the core cooling system]," Revision 1,Section II, emergency core cooling "Acceptance Criteria" which cites general system (IOECCS). {Error 1} design criteria (GOG) 15, as it relates to the reactor coolant system (RCS) being designed to assure that the pressure boundary will not be breeched during anticipated operational occurrences (AOOs). The NRC staff in its safety evaluation (SE) dated May 4, 2001, related to the Braidwood/Byron uprate, acknowledged that the acceptance criteria included ensuring that the peak RCS pressure remain less than the safety limit of 110 percent of design and the licensee demonstrated that it met the criteria through use of an analysis.

The staff considers performance of an analysis an acceptable means of demonstrating compliance with acceptance criteria and does not consider it to be indicative of licensee misunderstanding of the IOECCS event.

The licensee's unnecessary Not There is an acceptance criteria contained in departure from nuclear significant NUREG-0800, SRP, 15.5.1, "Inadvertent boiling rate (DNBR) analysis new Operation of ECCS," Revision 1,Section II, reveals a lack of information "Acceptance Criteria" which cites GDC 26, as understanding of the it relates to the reliable control of reactivity IOECCS. {Error 2) changes to assure that specified acceptable fuel design limits are not exceeded, including AOOs. The NRG staff in its SE dated May 4, 2001, related to the Braidwood/Byron uprate, acknowledged that the licensee included a DNB analysis in its application for uprate dated July 5, 2000 (ADAMS Accession No. ML003730536), and demonstrated that the calculated DNBR remains greater than the safety limit. The staff considers performance of an analysis an acceptable means of demonstrating compliance with acceptance criteria and does not consider it to be indicative of licensee misunderstanding of the

-- IOECCS event. . ------

Enclosure

Petition Issue Basis Suooorting Discussion The licensing basis (Exelon Previously The licensee did address the non-escalation letter dated July 5, 2000, nor addressed criteria for the IOECCS event. Exelon's the updated final safety and July 5, 2000, letter states in Section 6.2.20.2 analysis report (UFSAR), resolved. that the criteria for Condition II events include Revision 15 (ADAMS not generating a more serious plant condition.

Accession No. In response to a request for additional ML14363A393)) does not information, Exelon, in a letter dated provide an analysis or January 31, 2001 (ADAMS Accession No.

evaluation to demonstrate ML010330145), states that the Electric Power that the non-escalation Research Institute (EPRI) testing showing that requirement is satisfied. ability of pressurizer safety valves to reseat

{Omission 1} following liquid discharge supports the conclusion that the inadvertent SI event would not transition to a higher condition event and provided supporting information. The NRC staff concluded in its May 4, 2001, SE, that the licensee's crediting of the pressurized safety valves (PSVs) to discharge liquid water during the spurious SI event to be acceptable. In addition, the Updated Final Safety Analysis Report (UFSAR), Section 15.5.1.2, states, "The Inadvertent Operation of the ECCS During Power Operation event does not progress into a stuck open Pressurizer Safety Valve LOCA event. All three valves may lift in response to the event, but they will reclose.

The resulting leakage from up to three pressurizer safety valves that are seated is bounded by flow through one fully open valve.

The consequences of the event are bounded by the analysis described in UFSAR Section 15.6.1, "Inadvertent opening of Pressurizer Safety or Relief Valve." This event is also classified as an event of moderate frequency."

The Backlit Appeal Review Panel (BARP)

I report concluded that "the standard in place in 2001 and 2004 and at present is simply that I the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment" and "in the absence of a PSV failure to reseat, the Panel concluded that the concerns articulated by the NRC staff in the Backfit SE related to event classification, event escalation, and I compliance with 10 CFR 50 34(b) and GDCs 15, 21, and 29 are no lonqer at issue." I

Petition Issue Basis Sunnortina Discussion The missing non-escalation Previously As discussed above in Omission 1, the SARP case analysis reveals a lack addressed report concluded that in the absence of a PSV of understanding of the and failure to reseat, event escalation is no longer IOECCS. !Error 3\ resolved. an issue.

The IOECCS evaluation is New issue, If the PORVs are assumed to function either non-conservative, or not normally (and credited in the safety analysis),

based upon a requirement to significant. the pressurizer would fill faster than with the prevent the PORVs from use of the PSVs, given the lower setpoint opening. Either of these pressure of the PORVs relative to the PSVs.

interpretations indicates the This is due to the flow characteristics of the licensee lacks an ECCS pumps delivering more flow at lower understanding of the RCS pressure. However, the difference in IOECCS. {Error 3) time to fill the pressurizer (i.e., time for the operators to take action to prevent liquid discharge) will be small and is not considered significant since operator action is not credited in the IOECCS analysis to stop the ECCS flow until liquid has already passed through the valves.

There is no description of New issue, As noted above, if the analysis were done how the PORVs would not crediting the PORVs (instead of the PSVs),

respond to an IOECCS. significant. the pressurizer would fill slightly faster,

{Omission 2} however, the end result would still be some liquid passing through a valve into the pressurizer relief tank as there is no operator '

action credited in the IOECCS analysis to stop the ECCS flow until after liquid has passed through the valves.

Petition Issue Basis Sunnorting Discussion The licensee does not justify Not There is no requirement to justify the use of the use of PSVs, in lieu of significant PSVs as opposed to PORVs. The PORVs, to respond to AOOs. new Byron/Braidwood UFSAR, Section 5.4.13.1,

{Omission 3} information states, "The pressurizer power-operated relief valves are not required to open in order to prevent the overpressurization of the reactor coolant system. The pressurizer safety valves by themselves are sized to relieve enough steam to prevent an overpressurization of the primary system." There is no statement that the PSVs cannot open during an AOO.

Petitioner refers to the American Nuclear Society (ANS), "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," N18.2, 1973 (ANS N18.2-1973), statement that AOOs "shall be accommodated with, at most, a shutdown of the reactor with the plant capable of returning to operation after corrective action." This does not imply that relief or safety valves for other systems cannot function during an AOO.

There are many AOOs where relief or safety valves (in both the primary and secondary sides) are credited, including events such as excessive increase in secondary steam flow, loss of external electrical load/turbine trip and loss of normal feedwater flow. The NRC staff interprets the ANS standard to implicitly mean that no damage to reactor systems occurs while the worst thing occurring is a reactor shutdown.

The licensee makes an Not The licensee does not state that these two invalid comparison between significant events are similar or directly comparable, two dissimilar events new rather that one event leads to the other.

(inadvertent PSV opening information and the IOECCS, with a While they aren't the same, it can be shown stuck open PSV) {Error 4) that the IOECCS can lead to an event similar to an inadvertent opening of a PSV (with different initial conditions) resulting in similar consequences (i.e., releases to the containment).

The NRC staff understood the licensee's comparison by stating in its May 4, 2001, SE that, "The licensee states that the resulting leakage from up to three PSVs is bounded by

[ flow through one fully open PSV, which is an analvzed ..event."

Petition Issue Basis Sunnortina Discussion The licensee claims that Not Based on the text of the petition, the basis for ECCS flow will match PSV significant identifying this as an error is predicated on water relief rate. {Error 4} new assuming the PSVs fail open following liquid information discharge. In its May 4, 2001, SE, the NRC staff states, "A review of the above stated EPRl test data indicates that the PSVs may chatter for the expected fluid inlet temperature but that the resulting PSV seat leakage following the liquid discharge would be less than the discharge from one stuck-open PSV, which is an analyzed event. Therefore, the NRC staff finds the licensee's crediting of the PSVs to discharge liquid water during the spurious SI event to be acceptable." The SARP report states, "The Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment." The flow out of the PSV will approximately match the flow from the ECCS provided the PSVs cycle open/closed to maintain pressure and do not leak. While the licensee assumes the valves reclose (consistent with the SARP report), they also conservatively assume the three PSVs may leak with an equivalent flow area of a single stuck open PSV (analyzed in FSAR Section 15.6 1) In this case, the flow out of the PSV will initially exceed the ECCS flow as the petitioner states. However, after some time the RCS pressure will decline to reach an equilibrium where flow out of the PSV is annroximatelv eaual to flow in from the ECCS.

The licensee fails to use due Not This statement is based on the issue in {Error diligence when passing on significant 4}, above, which was based on a vendor-supplied information new Westinghouse document (NASL-93-013).

to the NRG. {Error 5} information Although this is a new issue, it is not safety

s~.~~ci.n.~.a~ -~*().tt?_g__ above in {Error 4}. I

Petition Issue Basis Sunnortina Discussion The licensee claims that the Not The licensee does not explicitly state what the ECCS is a normal RCS significant petitioner claims. In Section 15.5.1.2 of the makeup system. {Error 6} new UFSAR the licensee indirectly makes the information claim that ECCS is a normal makeup system as they use this logic to demonstrate compliance with the acceptance criteria. The licensee provides an example (from ANS 51.1/N18.2-1973) of a Condition II event as a "minor reactor coolant system leak which would not prevent orderly reactor shutdown and cooldown assuming makeup is provided by normal makeup systems only." The licensee then states "operation of the ECCS maintains RCS inventory during the postulated event." However, given that the inadvertent actuation of the ECCS is the initiating event, by definition, the ECCS will be operating but not providing a normal reactor coolant makeup function as described above. Therefore, this issue is not considered safety significant as inventon1 will be maintained in the RCS.

The licensee failed to identify Not Under certain conditions, the NRG staff and correct the Idaho significant considers it possible to challenge both PORVs National Engineering and new and PSVs during the same transient if the Environmental Laboratory information event lasts long enough. For example, if (INEEL) error in stating that credited, the PORVs would open first, then, if the IOECCS will challenge they don't have power/instrument air and the both the PSVs and PORVs. N2 tanks deplete, they fail closed. At this The licensee transmitted point, the PSVs would be relied upon for INEEL's report to the NRG pressure relief. Both the licensee in its July 5, staff without verifying its 2000, letter, as supplemented by its January accuracy. {Error 7} 31, 2001, letter, and the NRG staff in its May 4, 2001, SE, mentioned the PORVs and their role in the IOECCS analysis. Both the PSVs and the PORVs may be challenged. Based on the above, the information is not sionificant.

Petition Issue Basis Sunnortino Discussion The licensee did not provide Previously The BARP report concluded, "Given the NRC the valve test results needed addressed staff's resolution of TMI [Three Mile Island]

to qualify the PSVs for water and Action Plan, Item 11.0.1, and the NRC staff's relief. {Omission 4} Resolved. prior approvals reviewed by the panel, the panel concludes that the Office of Nuclear Reactor Regulation (NRR) staff's current application of the American Society of Mechanical Engineers (ASME) Code is not supported by the historical record." In addition, "The panel did not find any evidence that the licensee had claimed or the NRC staff had believed that the valves were "qualified" in an ASME BPV [Boiler and Pressure Vessel]

Code certification sense; rather, the record shows thorough consideration of the testing conducted on valves of the type installed at the plant and a well-informed technical judgment that this testing provided appropriate aualification."

The licensee analysis Previously The backfit appeal review panel " ... concluded requires the PSVs relieve addressed that in 2001 and 2004 and at present, the water, and then reseat. and known and established standard of the

{Error 8) Resolved. Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Commission has not established a more detailed or prescriptive standard."

Petition Issue Basis SuooortinQ Discussion The licensee does not Not The Byron/Braidwood UFSAR, Section describe the design change significant 5.4.13.1, states, "The pressurizer power-process it used, including new operated relief valves are not required to open quality controls, to determine, information in order to prevent the overpressurization of and specify the functional, the reactor coolant system. The pressurizer and component requirements safety valves, by themselves, are sized to for PSVs, when operated relieve enough steam to prevent an during AOOs (e.g., the overpressurization of the primary system."

IOECCS). {Omission 5) There is no statement that the PSVs cannot open during an AOO. There are many AOOs where credit is taken for relief and safety valves, including events such as excessive increase in secondary steam flow, loss of external electrical load/turbine trip and loss of normal feedwater flow.

BARP report, Section 4.2, states, "The Panel concluded that in 2001 and 2004 and at present, the known and established standard of the Commission is that the failures of PSVs need not be assumed to occur following water discharge if the likelihood is sufficiently small, based on well-informed staff engineering judgment. The Commission has not established a more detailed or prescriptive standard." Based on the above, the issue is not significant and does not imply a new desiqn function for the PSV.

The licensee fails to meet the Previously The SARP report found, "The determination GDC 21 single-failure addressed that application of the single failure criterion is requirement. {Error 9} and necessary first appears in the draft Revision 1 Resolved. to RIS [Regulatory Issue Summary] 2005-29, which is still under development, and is not included in any final NRC requirement or guidance document reviewed by the panel."

In addition, the SARP report stated, "Finally, in the absence of a PSV failure to reseat, the Panel concluded that the concerns articulated by the NRC staff in the backfit SE related to event classification, event escalation, and compliance with 10 CFR 50.34(b) and GDCs 15, 21, and 29, are no lonaer at issue."

Petition Issue Basis SunnortinQ Discussion The licensee does not Previously By letter dated August 18, 1988 (ADAMS evaluate potential damage to addressed Accession No. ML003772409), the NRC staff the PSVs. {Omission 6} and provided its Technical Evaluation Report Resolved. (TER) regarding the performance testing of relief and safety valves for the Byron (pages 161-188 of the file), and Braidwood Stations (pages 189-217 of the file) The TER discussed the chattering of the valves and evaluated damage found during subsequent inspection of the valves. The TER concluded that the valves oerformed satisfactorilv.

Application of the PSVs Not There are no requirements to limit PSVs to comes too late to meet the significant only operate during Condition Ill or IV events.

non-escalation requirement. new The Byron/Braidwood UFSAR, Section

{Error 1O} information 5.4.13.1, states, "The pressurizer power-operated relief valves are not required to open in order to prevent the overpressurization of the reactor coolant system. The pressurizer safety valves, by themselves, are sized to relieve enough steam to prevent an overpressurization of the primary system." This does not reference ANS categories as a requirement and does not say that the PSVs cannot open during an AOO.

There are many AOOs where credit is taken for relief and safety valves, including events such as excessive increase in secondary steam flow, loss of external electrical load/turbine trip and loss of normal feedwater flow There is no evaluation of the New issue. The petition does not provide facts to support number of pressurization Petition did the statement. As stated in MD 8.11, the NRC cycles against the plant's not provide staff will not review a petition if the petition limit. {Omission 7} sufficient "fails to provide sufficient facts to support the facts to petition but simply alleges wrongdoing, support violations of NRC regulations, or existence of conclusion_ safetv concerns" The licensee creates a new Not The opening of a PSV during an AOO is not accident and does not significant by itself, a new accident. Additionally, BARP address the new accident in new report, Section 5 - states, ". . in the absence its no significant hazards information of a PSV failure to reseat, the Panel statement. {Error 11, concluded that the concerns ... related to Omission 8} event classification, event escalation, and compliance with 10 CFR 50.34(b) and GOCs 15, 21, and 29 are no longer at issue."

Therefore, in the absence of assuming the PSV fails open, there is no possibility of a new accident.

Petition Issue Basis Suooorting Discussion The licensee employed a Not BARP report, Section 5, states, " ... in the circular logic that failed to significant absence of a PSV failure to reseat, the Panel demonstrate that the Byron/ new concluded that the concerns ... related to Braidwood plant design information event classification, event escalation, and meets all of its design compliance with 10 CFR 50.34(b) and fPnUirements GDCs 15, 21, and 29 are no lonaer at issue."

The technical review staff of Previously The BARP report " ... concluded that in 2001 the NRC's NRR had addressed and 2004, and at present, the known and approved the licensee's and established standard of the Commission is applications for power Resolved. that the failures of PSVs need not be assumed upratings for the Byron and to occur following water discharge if the Braidwood plants that likelihood is sufficiently small, based on claimed it had complied with well-informed staff engineering judgment. The a key design requirement, Commission has not established a more which requires nuclear plants detailed or prescriptive standard."

to be designed in a way that prevents AOOs from In addition, the BARP report, Section 5 states, developing into more serious " ... in the absence of a PSV failure to reseat, events. The licensee's claim the Panel concluded that the concerns ...

relied upon its plants' PSVs related to event classification, event to perform safety functions escalation, and compliance with 10 CFR that are outside their design 50.34(b) and GDCs 15, 21, and 29 are no basis lonoer at issue."

The licensee submitted, Not The opening of a PSV during an AOO is not, under Oath and Affirmation, a significant by itself, a new accident. Additionally, BARP statement of no significant new report, Section 5 - states, "... in the absence hazards, as per 10 CFR information of a PSV failure to reseat, the Panel Section 50.92 concluded that the concerns ... related to event classification, event escalation, and compliance with 10 CFR 50.34(b) and GDCs 15, 21, and 29, are no longer at issue."

Therefore, in the absence of assuming the

, PSV fails open, there is no possibility of a new accident. Without the possibility of a new accident, the statement of no significant hazards isn't in error. .

Petition Issue Basis Sunnortina Discussion The CVCS [chemical and Not The mass addition as a result of a malfunction volume control system] significant of the eves is bounded by the IOEees malfunction does not lead to new because only the charging pumps would be an immediate reactor trip. An information putting mass in, since the SI pumps would not analysis is necessary to start and inject. The referenced malfunction demonstrate that the reactor (UFSAR Section 15.4.6) shows that mass is automatically tripped injection is limited to 205 gallons per minute.

before any fuel clad damage If no operator action is taken, the over can be incurred. Exelon temperature delta temperature or the high does not provide one. neutron flux trips the reactor before DNBR Instead, points to another, limits are exceeded. Therefore, the issue is dissimilar event analysis, the not significant.

eves malfunction that increases reactor coolant inventory. This event is a reactivity anomaly, not a mass addition event. It cannot be used to address a mass addition event