Text

NRDCs Response in Support of Foe Petition to Intervene, San Onofre Units 2 and 3
	ML121850554
Person / Time
Site:	San Onofre
Issue date:	06/27/2012
From:	Fettus G; NRDC
To:	Annette Vietti-Cook; NRC/SECY
	SECY RAS
References
	50-361-CAL, 50-362-CAL, Confirmatory Action Letter, FFF-2
	Download: ML121850554 (48)
	v • d • e

DOCKETED USNRC N R D C June 27, 2012 (3:42 p.m.)

THE EARTH'S BEST DEFENSE OFFICE OF SECRETARY RULEMAKINGS AND

.June 27, 2012 ADJUDICATIONS STAFF Docket Nos. 50-361 and 50-362 Via Electronic & Overnight Mail U.S. Nuclear Regulatory Commission Office of Secretary of the Commission Ms. Annette L. Vietti-Cook, Secretary of the Commission Sixteenth Floor One White Flint North 11555 Rockville Pike Rockville, MD 20852 RE: NRDC's Response in Support of FOE Petition to Intervene, San Onofre Units 2 and 3

Dear Ms. Vietti-Cook:

Please find attached NRDC's Response in Support of FOE Petition to Intervene and NRDC's Notice ofIntent to Participate, filed this day via electronic and overnight mail to the service list created for this matter by Friends of the Earth (FOE) on June 18, 2012. There are two attachments to the document, a certificate of service and an entry of appearance included as well.

As we understand it, FOE filed original pleadings under the license numbers for the San Onofre Nuclear Generating Stations 50-361 and 50-362. FOE was denied an approval code to obtain the digital identification certificate necessary to file with the Commission's electronic filing system.

As there is no docket and in order for our response to be timely, we file this day via electronic and overnight mail. We respectfully request notice via electronic mail at the address listed below if and when a docket is created for this matter.

If you have questions, please do not hesitate to contact me at (202) 289-2371. Thank you for considering our views on these important matters.

Sincerely, (signed electronically) Geoffrey H. Fettus Geoffrey H. Fettus Senior Attorney Natural Resources Defense Council 1152 15th Street, NW, Suite 300 Washington, DC 20005 (202) 289-2371, gfettus@nrdc.org

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the Matter of:

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SOUTHERN CALIFORNIA EDISON

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License No. 50-361

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License No. 50-362 (San Onofre Nuclear Generating Station, Units 2 and 3))

June 27, 2011 NATURAL RESOURCES DEFENSE COUNCIL'S (NRDC)

RESPONSE IN SUPPORT OF FRIENDS OF THE EARTH PETITION TO INTERVENE AND NRDC'S NOTICE OF INTENT TO PARTICIPATE Introduction The Natural Resources Defense Council ("NRDC") respectfully submits this timely response in support of the Petition to Intervene and Request for Hearing filed by Friends of the Earth ("FOE") on June 18, 2012 (hereinafter "FOE Petition for Hearing") and the associated Application To Stay Any Decision To Restart Units 2 Or 3 At The San Onofre Nuclear Generating Station Pending Conclusion Of The Proceedings Regarding Consideration Of The Safety Of The Replacement Steam Generators (hereinafter "FOE Stay Application"). See 10 C.F.R. § 2.323.

We believe that under current circumstances in which the NRC has found that San Onofre Nuclear Generating Station in San Clemente, California ("San Onofre") is not safe to operate at its licensed maximum thermal power limit given the rapid deterioration of its steam generator tubes, which resulted in a leak in one tube, law and precedent require NRC Staff to amend San Onofre's license to establish a new reduced maximum safe operating thermal power I

limit. The new limit should correspond not only to the known impaired heat rejection capacity of the replacement steam generators (RSGs), but also to a new safe operating point at which a recurrence of the recent rapid tube erosion phenomena can be precluded prior to restart of one or both units at San Onofre. In short, specification of a new maximum safe thermal power limit requires a license amendment and notice of opportunity for public hearing with all procedural rights that accrue in such an instance.

NRDC therefore supports FOE's Petition for Hearing and requests NRC institute a license amendment proceeding with a notice of hearing allowing NRDC and other affected parties opportunity to participate.1

Background

The original facts that led to FOE's Petition for Hearing are well documented. On January 31, 2012, San Onofre suffered a steam generator tube leak in Unit 3 that resulted in the release of radioactive material into the environment. SCE thus performed a rapid shut down of the unit. Prior to the leak in Unit 3, SCE discovered excessive wear in Unit 2, which was offline for a refueling outage. Subsequently, advanced deterioration of many tubes was discovered in the replacement steam generators, which had been in operation for eleven months in Unit 3 and less than two years in Unit 2. Both units are currently shut down while NRC and SCE continue investigations. Both NRC and SCE have made periodic assurances that the matter is being addressed, but there has been no meaningful opportunity for public involvement that included any procedural or substantive rights.

One of the public actions taken with respect to the situation was the issuance of a We incorporate by reference the background, assertion of timeliness, supporting documentation and the single contention filed by FOE in its Petition for Hearing and Stay Application.

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Confirmatory Action Letter (CAL) from the NRC staff to SCE on March 27, 2012. The CAL directed SCE to keep San Onofre Units 2 and 3 shut down until SCE has taken, and NRC has reviewed, certain actions related to the investigation of the rapid tube degradation that was detected in both units and which caused a radioactive release in Unit 3. As FOE described in its petition, the CAL does not require SCE to propose a license amendment, nor does it specify that it will allow for a public adjudicatory hearing process as provided for by 10 C.F.R. § 2.309, or for one prior to restarting the units. Instead, the CAL restates SCE's description of the steam generator problems and the commitments SCE made as of March 23, 2012 to address the issues at Units 2 and 3; it does not show any independent analysis by the NRC, nor require more of the licensee beyond the actions for which SCE has volunteered.

On June 18, 2012, FOE filed its Petition for Hearing with the Commission. FOE's Petition asserts that under 10 C.F.R. § 50.59 the San Onofre replacement steam generators may not be operated without one or more amendments to the San Onofre operating license. The FOE Petition then asks that the Commission either recognize that the current CAL process instituted by NRC to address the situation at San Onofre is in fact a license amendment proceeding under 10 C.F.R. § 2.309 and 42 U.S.C. § 2239, or convene such a license amendment proceeding under these authorities or under the Commission's inherent supervisory authority over the nuclear industry. FOE further requested status as a party in any such proceeding, and that, pursuant to 10 C.F.R. § 2.309, the Commission provide an adjudicatory public hearing with respect to the causes and potential remedies for the failure of the replacement steam generators at San Onofre.

Contemporaneous with FOE's Petition and Stay Motion, NRC Staff disclosed its preliminary findings to the press, five months after the break that precipitated the shutdown of 3

Unit 3. See, Feds Say Design Flaw Led To Calif Nuke Plant Woes. Michael R. Blood, Associated Press, Jun. 18, 2012, http://www.sacbee.com/2012/06/18/4570064/ap-exclusive-feds-design-led-to.html. In the article, it is apparent that substantial changes were made regarding structural and material elements that may have direct implications in the accelerated degradation of steam tubes at San Onofre.

Discussion Precedent Requires a License Amendment Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include Technical Specifications ("TSs") as part of the license. 42 U.S.C. § 2232. The licensee provides TSs in order to maintain operational capability of structures, systems and components that are required to protect the health and safety of the public. The Commission's regulatory requirements related to the content of the TSs are found in 10 CFR § 50.36, "Technical specifications," which include the following categories: (1) safety limits, limiting safety systems settings and control settings (§ 50.36 (c)(1)); (2) limiting conditions for operation (LCOs) (§ 50.36 (c)(2)); (3) surveillance requirements (SRs) (§ 50.36 (c)(3)); (4) design features (§ 50.36 (c)(4)); and (5) administrative controls (§ 50.36 (c)(5)). In general, there are two classes of changes to TSs: (a) changes needed to reflect modifications to the design basis (TSs are derived from the design basis), and (b) voluntary changes to take advantage of the evolution in policy and guidance as to the required content and preferred format evolve TSs over time. The situation at San Onofre is within the regulatory language contemplated by changes needed to reflect modifications to the design basis.

As an example of bases for a hearing, by letter dated June 25, 2009 (ML091670298), the 4

NRC issued Amendment No. 220 to the operating license for San Onofre Unit 2 and Amendment No. 213 for San Onofre Unit 3. These amendments revised the inspection requirements and tube plugging criteria for the replacement steam generators. It is being widely reported that the licensee may plug more tubes than required by the steam generator plan described in TS 5.5.2.1 1.2 In that case, the steam generator program as found in the TSs appears deficient, requiring that the licensee adopt more stringent measures for safe reactor operation than prescribed within the program.

Regarding the tube plugging criteria or some other technical matter of which we are currently not aware, the San Onofre circumstance is analogous to a situation at the Nine Mile Point Nuclear Generating Station in 1997. There, the licensee had a consultant evaluate cracking identified in the reactor core shroud. The consultant's report concluded that crack propagation rates would not undermine necessary safety margins, but the consultant's evaluation assumed better reactor water chemistry than defined by the TSs.3 By letter dated July 2, 1997, the licensee submitted a license amendment request to replace the non-conservative water chemistry measures that had existed in the TSs with the far more conservative parameters established in the consultant's report.4 By letter dated September 18, 1998, (MLOI 1030259), NRC issued 2

See, e.g., "[a]ny tube that satisfies the SG Program repair criteria will be removed from service by plugging. Preventative plugging and stabilization of specific tubes potentially susceptible to degradation is also planned." May 10, 2012, Licensee Event Report (LER) 2012-002-00, at 6. Found at ML 12136AO65, Letter from Douglas R. Bauder, SiteVice President & Station Manager, San Onofre Nuclear Generating Station, Southern California Edison to NRC.

3 See Attachment 1, April 17, 1997 Letter from David A. Lochbaum, Nuclear Safety Engineer, Union of Concerned Scientists to Mr. S. Singh Bajwa, Acting Director, Project Directorate 1-1, Division of Reactor Projects - IAI, United States Nuclear Regulatory Commission.

4 See Attachment 2, Niagra Mohawk Power Corporation's Application to Amend License, July 2, 1997.

5

amendment No. 163 to the Nine Mile Point Unit I operating license.

Thus, it is apparent that at minimum there should be a modification to the design basis at San Onofre as the Steam Generator Program and its repair criteria purport to establish the line between safe and unsafe operation with degraded tubes. If SCE plugs more tubes than required by the plan in order to ensure safety, SCE could argue it has a legal basis for safe operation with degraded tubes (the current SG Program) but is actually employing different criteria for safety.

Thus, the SG program as it is currently written under its TSs is likely not sufficient to ensure safety. Thus, a license amendment with opportunity for full public involvement is required.

If, despite this history and the publicly reported NRC staff finding that design deviations in the RSGs appear to be the source of the current technical problems, the Commission remains uncertain on the question of whether these changes required and still require a license amendment, the Commission could exercise its inherent authority to convene a public evidentiary hearing to consider this question before ruling on the contention raised in FOE's petition. But there can be no doubt that a Staff enforcement proceeding, which by definition seeks enforcement of the terms of an existing license, is legally insufficient for weighing the merits of a restart of one or both units at San Onofre that inherently involves amendment of the current operating license in order to ensure adequate protection of the public health and safety.

Thus, NRDC believes it in the interest of all the prospective parties for the Commission to determine that consideration and final determination of a restart of one or both units at San Onofre must take place in the context of a public license amendment proceeding.

Opportunity for Public Hearing Prior to Restart of Reactors FOE's precise suggestions with respect to opportunities for a public hearing were that the 6

Commission either recognize that the current CAL process is in fact a license amendment proceeding under 10 C.F.R. § 2.309 and 42 U.S.C. § 2239, or convene such a license amendment proceeding under these authorities, or under the Commission's inherent supervisory authority over the nuclear industry. See FOE Petition at 2. We adopt and incorporate FOE's assertions and further note that given the compromised and uncertain heat rejection capacity of the impaired RSGs, any restart of one or both units at San Onofre will require a determination for each unit of whether recurrence of the recent rapid tube erosion phenomena can be precluded by a new reduced maximum safe operating thermal power limit that corresponds not only to the known impaired heat rejection capacity of the RSGs, but also a new safe operating point. It has not yet been determined that the units can be safely operated even at a reduced power level, an issue that will be among those vetted in the public hearings.

Further, if the NRC does determine there exists a new thermal power limit at which the units can be safely operated, it will almost certainly be less than that specified in the existing San Onofre license, and therefore a staff level enforcement proceeding based on the terms of the current license is no longer relevant, even in the unlikely event that the Commission should determine that the design changes to the RSGs do not rise to the level of requiring amendment of the existing license. The specification of a new maximum safe thermal power limit for an impaired commercial reactor must be open to the detailed technical justification and scrutiny of a public adjudicatory proceeding. The public cannot be assured that restart of one or both units at San Onofre can be achieved with adequate protection of the public health and safety without adequate participation in the review process as required under the Atomic Energy Act (AEA) and the Commission's rules. Reaching such a fully informed and reasoned judgment on whether a 7

new safe operating regime at reduced power can be identified and adequately supported by SCE and NRC Staff is precisely the objective of the public adjudicatory licensing proceeding in which FOE and now NRDC seek the opportunity for "party" status.

Under the AEA, the Commission must grant a hearing on a license application upon "the request of any person whose interest may be affected by the proceeding, and shall admit any such person as a party to such proceeding." 42 U.S.C. § 2239(a)(1)(A). In this particular matter, while there has been no application for an amendment to a license, it seems apparent that under 10 C.F.R. § 50.59, a licensee is required to obtain a license amendment if the proposed modification meets any one of eight criteria affecting the existing safety analysis as enumerated in subpart (c)(2) of §50.59. The criteria, in part, require an amendment when the proposed changes would: (1) create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report [(FSAR)] (as updated); (2) create a possibility for a malfunction of an SSC [system, structure, or component] important to safety with a different result than any previously evaluated in the final safety analysis report (as updated); or (3) result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses. As FOE demonstrated in its submission, the design of the replacement steam generators at San Onofre met the criteria that trigger a license amendment thirty-nine separate times. See Gundersen Expert Decl. at ¶ 32. Thus, replacement of the steam generators at San Onofre should have triggered an obligation that the NRC determine through a license amendment proceeding whether the new design was safe.

Now, however, with the current shut down of the units and the ongoing NRC investigation, there is an opportunity for both the NRC and the public to evaluate the effect of 8

such changes on the safety of the plant in a public proceeding so that the public may evaluate the safety risks and technical basis for proposals for restarting the reactors and offer the opinion of its own independent expert(s).

Notice of Intent to Participate and Standing NRDC is a national non-profit environmental organization with offices in Washington, D.C., New York City, San Francisco, Chicago, Santa Monica, and Beijing. NRDC has a nationwide membership of over 357,000 (plus hundreds of thousands of online activists),

including 63,996 members in California, at least 3,386 members living within 30 miles of San Onofre and approximately 440 members living within 10 miles of the facility. Among its missions, NRDC seeks to maintain and enhance environmental quality, to safeguard the natural world for present and future generations, and to foster the fundamental right of all people to have a voice in the decisions that affect their environment. Since its inception in 1970, NRDC has sought to improve the environmental, health, and safety conditions at the nuclear facilities operated by the Department of Energy and the civil nuclear facilities licensed by the NRC and their predecessor agencies. To that end, NRDC utilizes its institutional resources, including legislative advocacy, litigation, and public outreach and education, to minimize the risks that nuclear facilities pose to its members and to the general public.

By granting FOE the relief it requests and designating an adjudicatory hearing on the technical and safety basis for restarting the reactors, NRDC would have an opportunity to enter an appearance as a party, enter individual standing declarations, and obtain redress via a public, transparent and legally sufficient proceeding to protect NRDC members whose concrete interests may be harmed by the actions at San Onofre. See Lujan v. Defenders of Wildlife, 504 U.S. 555, 9

572, n.7 (1992) ("[P]rocedural rights are special: The person who has been accorded a procedural right to protect his concrete interests can assert that right without meeting all the normal standards for redressability and immediacy.") (internal quotations omitted). Thus, we write this day to serve notice that (1) we support FOE's Petition and Stay Application; (2) enter an appearance of counsel; and (3) officially notice the Commission that we intend to participate when a notice of hearing has been issued.

Conclusion For the reasons stated above, prior to restart of one or both units at San Onofre, NRC Staff must determine that a new safe operating point exists capable of precluding a recurrence of the recent rapid tube erosion phenomena. If a new maximum safe thermal power limit is found to exist, then specification of such requires a license amendment and notice of opportunity for public hearing. Thus, NRDC requests that a public hearing be noticed and an opportunity to intervene be provided.

Respectfully Submitted, (signed electronically)/ Geoffrey H. Fettus Geoffrey H. Fettus Senior Attorney Natural Resources Defense Council 1152 15t' St. NW Suite 300 Washington, D.C. 20005 (202) 289-2371 gfettus@nrdc.org Filed this date of June 27, 2012 10

UNION OF CONCERNED SCIENTISTS April 17, 1997 Mr. S. Singh Bajwa, Acting Director Project Directorate I-i Division of Reactor Projects - I/lI United States Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

POTENTIAL UNANALYZED OPERATION OF NINE MILE POINT UNIT I WITH CORE SHROUD VERTICAL CRACKS

Dear Mr. Bajwa:

UCS reviewed the letter dated April 8, 1997, from Niagara Mohawk Power Corporation (NMPC) to the Nuclear Regulatory Commission (NRC) regarding the core shroud at Nine Mile Point Unit 1 (NMP-1). We have identified an apparent violation of Section 50.59 to Title 10 of the Code of Federal Regulations. Specifically, it appears that NMPC is proposing to operate NMP-1 without obtaining a necessary change to the technical specifications on reactor coolant chemistry. to NMPC's submittal dated April 8, 1997, contained a non-proprietary version of GE Report No. GE-NE-523-B13-01869-043 Rev. 0, "Assessment of the Vertical Weld Cracking on the NMPI Shroud." Page 9 of this GE document states:

"The experience in BWRs has shown that IGSCC [intergranular stress corrosion cracking]

initiation and growth is related to operating time. The initiation process is a stochastic process and with time the probability of cracking increases. This process can be accelerated if the water conductivity is higher because impurities aid crack initiation and accelerate crack growth. The characteristics of the coolant environment are also known to promote IGSCC on both the outside and the inside of the shroud."

Clearly, water chemistry is an important factor in controlling shroud weld cracking caused by IGSCC.

On page iii of the GE document, it is stated that a "bounding crack growth rate of 5x10"5 inches per hour" was assumed in the analysis. On page 11 of the GE document, this bounding crack growth rate is stated to be "characteristic of higher reactor water conductivity environments (-0.3 gS/cm)." On page iii of the GE document, the bounding crack growth is said to be conservative because of excellent water chemistry at NMP-1 (<0.1iiS/cm).'

According to a published conversion table, a seimen (S) can be converted to a mho by multiplying.by. 1.00. Therefore, all future references to conductivity will be in terms of pimho/cm for conformance with the NMP-1 technical specifications.

Washington Office: 1616 P Street NW Suite 310

Washington DC 20036-1495 ' 202-332-0900 - FAX: 202-332-0905 Cambridge Office: Two Brattle Square

Cambridge MA 02238-9105 ' 617-547-5552

FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203

Berkeley CA 94704-1567

510-843-1872 - FAX: 510-843-3785

UNION OF CONCERNED SCIENTISTS NMPC's letter dated April 8, 1997, indicates that the enclosures, including the GE document, "establishes the acceptability of the as found vertical weld cracking for a minimum of 10,600 operating hours (above 200'F)."

The analysis supporting the proposed operating of NMP-I with the core shroud vertical weld cracking appears to rely on reactor coolant chemistry limits that are significantly more limiting than the NMP-l Technical Specifications. By letter dated July 14, 1993, the NRC issued Amendment No. 142 to the NMP-I Operating License. Technical Specification 3.2.3 limits the reactor coolant conductivity to <2 gmho/cm with steaming rates less than 100,000 pounds per hour and to <5 gimho/cm with steaming rates greater than 100,000 pounds per hour.

Amendment No. 142 implemented some "cosmetic changes" (e.g., correction of typographical errors, repagination, etc.) to the NMP-1 Technical Specifications. The limits on reactor coolant chemistry remained unaffected by this amendment from the values implemented by issuance of Amendment No.

9 to the NMP-1 Operating License by NRC letter dated April 28, 1976.

Thus, the bounding crack growth rate assumed in GE's analysis appears to be based on reactor coolant chemistry limits that are over 10 times more restrictive than the NMP-I Technical Specifications. It would seem that NMPC should obtain a license amendment before it operates NMP-1 with reliance on the 0.3 tmho/cm conductivity value.

According to the Bases for NMP-1 Technical Specification 3.2.3:

"Materials in the primary system are primarily 304 stainless steel and the Zircaloy fuel cladding. The reactor water chemistry limits are established to prevent damage to these materials. Limits are placed on chloride concentration and conductivity. The most important limit is placed on chloride concentration to prevent stress corrosion cracking of the stainless steel."

Technical Specifications are intended to provide reasonable assurance that the facility can be operated safely. The core shroud crack growth rate analysis supports NMPC's conclusion that NMP-1 can be safely operated for a minimum of 10,600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months with the identified cracking. Since the crack growth rate analysis relies on conductivity limits that are significantly more restrictive than the existing Technical. Specification limits, it appears that an amendment is warranted.

NMPC could, of course, submit an evaluation of the crack growth rate using the reactor coolant conductivity limit in its current Technical Specifications. Based upon the qualitative analysis quoted from the GE document, it is reasonably assumed that the 10,600 hour0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months minimum operating time could be significantly shortened. Without such an analysis, NMP-1 operation with reactor coolant conductivity above the 0.3 igmho/cm value assumed in the GE analysis represents an unanalyzed condition.

In any event, NMPC's restart of NMP-1 would seem to constitute a violation of Section 50.59 to Title 10 of the Code of Federal Regulations unless a license amendment is obtained to lower the Technical Specifications' limits on reactor coolant chemistry to the value assumed in the crack growth analysis or an analysis is performed of the crack growth rate at the current Technical Specifications' limit. The April 17, 1997 Page 2

UNION OF CONCERNED SCIENTISTS NRC should issue NMPC an order or a confirmatory action letter requiring that an amendment be obtained or an analysis be completed prior to restart of NMP-I.

UCS was unable to attend the public meeting in New York on April 14, 1997, due to the very short notice on the meeting's rescheduling from its original April 10, 1997, date. However, from discussions with individuals who were able to attend this meeting, it appears that the meeting produced three unanswered questions. UCS respectfully requests a formal response from the NRC to the following questions prior to the restart of NMP-l:

1) As indicated by Figure 5-6, "V-9 Crack Depth after 10,600 Hours," in GE's Report GE-NE-B13-01869-043 Rev. 0 (contained in Enclosure 8 to NMPC's April 8, 1997, submittal), it is expected that nearly 24 inches of continuous through-wall cracking will be encountered. Is the crack growth rate after progressing through-wall the same as prior to becoming through-wall? Does through-wall cracking create the potential for vibrations that can increase the propagation rate?

2) As indicated by Appendix C, "Shroud Inspection Summary," in GE's Report GE-NE-B13-01869-043 Rev. 0 (contained in Enclosure 8 to NMPC's April 8, 1997, submittal), the heat affected zones (HAZs) for the vertical welds were inspected during the current refueling outage. There is no indication that areas other than the HAZs were examined. During the public meeting, the question of crack propagation beyond the HAZs was posed. Were areas outside the HAZs inspected? If not, why not? Have cracks propagated beyond the HAZs?

3)

Page iii of GE's Report GE-NE-B13-01869-043 Rev. 0 (contained in Enclosure 8 to NMPC's April 8, 1997, submittal) states that "no credit was taken for any portion of horizontal welds; it is assumed that each section of the shroud is a free standing cylinder." For the purposes of evaluating the integrity of the vertical welds, this appears to be a non-conservative assumption.

If a horizontal weld were through-wall cracked its entire circumference except for two points that are 1800 apart, then it is conceivable that forces acting on the shroud might tend to bow the shroud outward at the 900 and 270' locations since the intact weld portions would act to "pin" movement. If a vertical weld location coincided with these "bow" locations, the stress might be concentrated or higher than if the horizontal welds were totally non-existent as assumed in GE's analysis. Is GE's analysis non-conservative?

UCS understands that NMPC is anxious to resume operation of NMP-1, but we feel that the reactor coolant chemistry issue should be resolved and the above questions should be formally answered before this plant can be restarted safely.

Sincerely, David A. Lochbaum Nuclear Safety Engineer April 17, 1997 Page 3

NIAGARA MOHAWK G E N E R AT I10 N IT ud 1*

BUSINESS GROUP P. RNUYH SYLVLA IfNt.WWWMr Idly 2, 1997 CMeM;Offar NMP1L 1232

13. S. Nuclear Replaory CoMMio A=m: Documet Control Desk Washington, DC 20555 R*:

Nine Mile Point Unit 1 DoCke No. 50-220 Gontlcm=n:

During the 1997 refixeing oule at Nine Mie Point Unit I (NMP1), izn c

of te core shrud vedical wvldx revealed cvac* in excess of the =ccuinj criteria. By letter dated April 8, 1997, Niapz Mohawk Power Corporato (NMPC) providce design documndonz aud cval)Uadons to dcmoastMt the =ccepabAity of t as-found vctical wd cmckns in the NMPI cor shroud for at least 10,6W00 hours of hot (above 200 degrees F) qpeaton. By letter dated May 8, 1997, the NRC Issed a Safety EValuation appoving th rePI t of NMPI eontingeit on: 1) maintaing reactor coolant chemistry widin the gui n set forth in dte Electric Power Rmearch bstitute (ýP techniml report TR-103515-1l1 (BWRVIW-29),

-DWR Watr= Ch=eitry Guidcli= -199 Reviion,' and 2) the requimment that NMPC submit an application for a license amendment to addr the difference betwee the crarent TS conducttyt limits far ector coolant chemistry and the analyss assumpdi for core shroud crack growth rates. The NRC approved the NMPC analysis prsdiaed on the condition that NMP1 is opwaed in AC0oidam with tie BWR watcr chemistry piidcinc.

This appicaieon for ammdnt is being mb*,wd to addren the NRC'. mcond coning*ency.

NMPC hereby transmits an ApplicatiOn for Amndmnt to NN71 Opmating Lie= DPR-

63. Also enclosed as Attmenmt A is the proposed dnge to the Tefhnical Spentons CTS) set forth in Appendix A to the above menioned Hlic.

Supporting informfion and analyscs which demonstraft thm the proposed change involves no sismificalt hazards conaidctia pursant to 10CFR50.92 are indled as Aacbh'ent B. A n ed-up copy of the affcted TS pap=s is pruviiLkl as A'rU, aiue C to assist your mview.

Page 2 Th7c proposd Chu"* rvis SeOtles 3.2.3 md 4.2.3 to Neflct t" RW water chuenitry Sudclines. In addtion, the Bases for 3.2.3 and 4.2.3, wCoolant Chemistry', has been revised. These changes addms ft differences between th current TS wnducLivity limibL for reactor coolant chemitry and the anaysis a.umpions for core shroud crack growth

rates, Pirsuant to lOCFP09 1(b)(1), NMPC has prvjdW a copy of this liceme amendmet reques and the associawd analyrib g-,ing no siicant hazards consderation to the approprla% Oft rcWrcintative.

Very

y yours, Chief Nuclear Officer BRS/TRE/Cmk Attachments xc:

MW. H. 7. Miller, NRC Regioni Adminhisiamr Mr. A. W, Dromerlck, Acting irector, Project Directorate, 1-1, N.R Mr. B. S. Norxi3, Senior Reident Inspector Mr. D). S. Hood, Swnim Ptoject Munaet, NRR Mr. 1. P. Spath NYSERDA 2 Empire Plaa, Suite 1901 Albany, N 12223-1253 Rcords Management

UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of taagseft Mohawk Power Conlrution I

DoIket No. 50-220 Nine MWe Poin*t Unit I APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Commission, Niagara Mohawk Power Corporation (NMPC), holder of Facility Operating LUc"se No. DPM-63, hereby request that Section 3.2.3 and the asuociated surveillance Section 4.2.3 of the Technical Specifications (TS) set forth in Appendix A to that license be amended. The Irooosed changies have been reviewed in accordance with Section 6.5, "Review and Audit,' of the Nine Mile Point Unit 1 (NMP1) TS.

The proposed change revises the NMP1 TS Section 3.2.3 to reflect the "BWR water chemistry guidirlnca, 1998 rcvision (EPRI TR-103S57-Rl, BWRVIP-2t). Seotions 3.2.3a and 3.2.3b define new conductivity limits when the reactor water is >. 200 degrees F and thermul powvr l* _< 10%1, avid whtur Lthumial power is > 10%. The new conduvctiviLy limit is now 1 #mholcm compared to the existing limits of 2 pmho/cm and 5 pnhoicm. The chlornle Ion limit trom Section 3.2.3a remains at the same level but it is listed as 100 ppb instead of 0.1 ppm. The chloride ion limit from Section 3.2.3b is changed from 0.2 ppm to 20 ppb. Sulfate ion limits are added to Sections 3.2.3a and 3.2.3b at 100 ppb and 20 p*b, rexpectiveiv. from Section 3.2.3c the maximum conductivIty limit is changed from 10 pm ho/cm to 5 pmho/cm, the maximum chlirida ion concmntration limit Is changed from 0.5 ppm to 100 ppb and 200 ppb. and the maximum sulfate ion oonoentration of 100 ppb and 200 ppb is added.

The proposed change tevises NMP1 TS Section 4.2.3 to tncludo sulfate ions as a component to be lncluded In the sample analysis.

Included in this TS change is a change to the Bases for 3.2.3 and 4.2.3. "Coolant ChemistryW. The Bases has been changed to reflect the purpose of the specification which is to limit ;ntargranular stress corrosion cracking (IGECC) crack growth rates through the control of reactor coolant chemistry. The Bases describes the NMP1 operating philosophy of maintaining average levels for conductivity and chloride and sulfate Conoentrations over an operating cycle. Operation af the plant with*n these average values will ensure that the crack growth rate is bounded by the core shroud analysis.

The proposed change wig not authorize any change in the types of effluents or in the authorized power level of the facility in conjunction with this Application tor Ucense Amendment. Supporting information and analyms which demonstrate no significant hazards considerations pursuant to 10CFRS0.92, are included as Attachment B.

WHEREFORE, Applicant respectfully requests that Appendix A to Facility Operating License No. DPR-63 be amended in the form attached hereto as Attachment A.

NIAGARA MOI IAWK POWER CORPORATION By B. R. Sy, Chief Ncdear Officer Subscr~td sod Sworn lpbefore me On 1h13

.Rd day of N6TARY PUBLId Or I&MY W. RlfCA WaLIND~kuw 5UmnV k" QftM1mEIgJI

ATTACHMENT A NLAGARA MOHAWK POWER CORPORATION UCENSE NO. OPR-43 DOCKET NO. 50-220 PrqogAd Changes to Tecvaica SeC~ificotion.s Reglace t* eXIs.1fln pages 86, 97, and 98 with the atrached ruvised pages 9., 97, and

98. The pages have been retyped in their entirety with marginal marking3 to indicate changes.

UNSTW46 CONJITI ON FOR OPERATION US~HIIWGC~OZ)IT0IW OR PERAIONSURVEILLANCE RELIREMENT 3.2.3 COOLANT CHEMISIRY Applies to the reactor coolant syste-n demicel requirements.

0bbative:

To assue the chemical purity of the reactor coolant water.

a.

The reactor coolant water shall not exceed the following mitks with. the coolant temperature 200 degrees F and reactor thermal power

_% 10%. except as specified In 3.2.3c:

4.2.3.

OLANT CHEMISTRY Applies to the pedodic tessing requirements of thie reactor coolant chemistry.

To determine the chemical purity of the reactor coolant water.

Samples ahelf be taken end analyzed for conductivity.

chloride and stfate Ion content et least 3 times per I

week with a maximum tie of 96 hotrs between arnples. In adflition, If 1he conductivity becomes abnormal 1other then short term spikes) as indicated by the continuous oonductivity monitm, samples shall be tlken and analyzed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and daily thereafter until conduLtivity returns to normal levels.

When the contlnuoLs conductivity monitor Is inoperable, a reactor coolant sample shall be taken and analyzed for coiductivfty. ch:orde and sullate ion content at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Conductivity ChIcride Ion Sulfate Ion 1 jmiho/cm 100 ppb 100 ppb

b. The reactor coolant water shad not exceed the following limits with reactor thermal power

> 10%, except as specified In 3.2.3c:

Conductivity Chloride ion Sulfate ion 1 jnhofom 20 ppb 20 ppb AMENDMENT NO. U1 9G

LIFAMNO OONDITION FOR OPERA71ON SURVBLLALNCE HIMUIREMENT LIMIliNO GONOITION FOR OPERATION SURVEILLANCE PEQUIREPAENT

c.

The limits specified in 3.2,3a and 3-2.31; may be exceeded for a period of time not to excsed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In nn cane shal the reactor coolmt exceed the foatowfng ibmits at the specified corditions:

I.

With reactor cooalnt tern peratume _ 200 degrees ze the conductivltV has a mmxirnum lirvit of ryanh*/cm, or

2.

With reactor coolent temperature Ž> 200 degrees and rem-tor therell power

_e 10%. the mxiumum limit of chloride or sulfate ion ooncenrtraton is 20C ppb, or

3.

With rmector thermal powoo > 10%. the mwimum limit of chloride or acfate ion concentration is 100 ppb.

d.

If Specifications 3.2.3m, b. and c are not met.

normal ordedV shutdown shah be initiated within one hour and the reactor shall be shutdown and reactor coolrit tenperatur. be reduced -o

ý< 200 digitas F with n ten hours.

9. if the cotxtnuouý conductiuity monitor is Inoplrable for more than seven days. the reactor shal' be shutdown and reactor coolant temperature be reduced to < 200 degrees F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

AMENDMENT NO. Uii 97

BASES FOR 3.2.3 AND 42.3 COOLANT CHEMISTRY Tlhi specification Is being submitted to address art NI;C safety evaluation requirernant. In its Nioy 8. 1907 letter, the NBC requed that NMPC submit an application for amendment to address the differences between the current TS conductivity limita for reactor coolant chemistry and the analysis essumptons for rte core ahrou1 crack growth evaluations. The purpose of this specification is to lifit hnte-grarular stress corrosion csacking IIGSCC) crack growth rates throu3h the control of rmactor coolant chemistry. The LCO values ensure that transient canditieos are acted on to restore reactor coolant chemlt'~y values to normal in a reasonable tkne frame. Under transient conditions, potential crack growth rates could exceed analytical asmumptiona, however, the duration w3l be limited so that any effect or.

potential crack growth is minimized and the design basis assumptions -ase maintained. The plant is normnally operated such that the average chemistry for the operating cyde is mairtainad at the conservative values of < 0.2 poi'holcm for conductivity and < 5 ppb for chloride ions

, 5 ppb for sulfate Wio.s. This wil! ensure that the crock growth raie is bounded by the core shroud analysis assumptions (the analysis shows the crack growth to be < 2.2E-5 Tn/lh for these levelsi. Since these ire averae values, there are no specific LCO actions to be taken if these values are exceeded at a specific point in time.

Spezification 3.2.39, I), and c Is consistent with the BWR water coolant chemnistry guidelines, 1998 revisoon (EPRI TR-103515-Ri, BWRVlP-29). The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months action time period for exceeding the codant chemistry limits descibed in 3.2.3a and 3 ensures that prompt action, Is taken to restore coolant chomietry to normal operating levels. The requi'ement to comtmence shutdown wit in I hour. an4 to be shutdown and reactor coolant temperature be reduced to < 200 degrees F within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months mhninlzes the potentla for IGSCC crack growth.

A short lerm spike it deffned as a rise In conductivity I> 0.2 pmho/fm) such as that which coad arse fPor injecion of additional feedwater flow for a duration of approximataly 30 minutes in tire.

When conductivity is in Its profer normal range. chlorkie, sulfate, and other impurities affecting conductivity must alse be within their nonnal range. Whtn and iV coduuctvity becomes abnormal, then chloride ard sulfate enisurenarns are made to detumIre whether or not they are also out of their nxmnd operating values. Significant changes provide the operator with a warning mechanismn so he can Investigate and remedy the condition causing the change and ensure that no'mal operating average conditions are maintained within the bounds of the core shroud caw* growth analytical assumtplions.

Meithods available to the operator for correcting the off-standaid condition include, operation of the reactor clean-Lp systerr. reducing the input of impurities. end placing the reactor in shutdown ond rmdLcing reactor coolant temperature to < 200 degrees F. The major benefit of reducing reotor cooliant tempereture to

< 200 degrees F is to reduce the temperalure deaendant corrosion rates and provide time for the clean-up system to r-establish the purity of thle mort codent.

The-conductivity of the reactor coolant in continuously monitored. The samples of the coolant which ars analyzed for corductivt evary 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months wil saewe as a comparison with the continuous *oncductivity monitor. The reactor coolant samples will also be used to determine :he chloride mnd sulfate concentrations. Therefore, the sampling frequency ;s considered adequate to detect long-term ahanges" in the chloride and sulfate )on content. Howaver, If tte conductivity becomes abnormal (> 0.2 lrnho/cmt, chloride and sulfate ieoeurements will be made to assure thr. the normal rrnrts (< 5 ppb of chloride or sulfate) are nair*taTned.

AMENDMENT NO.

93

ATTACHMENT 8 NIAGARA MOHAWK POWER CORPORATION LICENSE NO. DPR43 DOCKET NO. 50-220 SuMortina Information and No Siardfigant Hazards Considerwaton Anadlysi INTRODUCTION The praposed Nina Mila Point Unit I (NMPI) Taehnical Speifieation (TS) chang eaentined herein presents a revision to NMPI TS Sections 3.2.3 and 4.2.3, end the Bwes for 3.2.3 and 4.2.3, "Coolant Chemistry*,

By letter dated April 8,. 1997, Niagara Mohawk Power Corporation (NMPC) provided design duiuvWiLMlii and evvtllunLls tu duitmuooLfrdL Lim wacv.wpLtavTity ul Lim aw-fuuwd vertical weld cracking in the NMP1 core shroud. for at least 10,600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months of hot (above 200 degrees F) operation. In its May 8, 1997 letter, "Modifications to Core Shroud Stabilizer Lower Wedge Retaining Clip and Evaluation of Shroud Vertical Weld Cracking, Nine Mile Point Nuclear Station. Unit 1," approving the restart of NMP1. *the NRC required that NMPC submit an azndication for a license amendment addressing the difference between the current TS conductivity limits for reactor coolant chemistry and the enalysis assumptions for cure shroud crack growth rutes.

This proposed Change incororutas into the TB the reactor cooulat chemistry ma3umptionr that were used for the core shroud weld crack evaluations.

EVALUATION The proposed revisions to TS Sections 3.2.3a, b, c, d. and e incorporate the analytical assumptions that were used by NMPC to evaluate the vertical weld cracking found in the NMP1 onrA Ahmid darin0 the 1997 refueling outage. The TS clhnaes. establish limits for condutiviMty and chloride and sulfate Ion concentrations that are equal to or more restrictive than the oxisting TS values. As a rosult of the analysis, on overage value of 0.2 umho/cm has been chosen for conductivity which is less than the BWR guideline fs

mu livul I value f1 urxuuutivity of 0.3 pmho/cm.

I he purpose of this -1 change is to limit IGBCG cmrak growth rates through the control of reactor coolant chemistry. The proposed LCO values ensure that bunsient conditions are acted on to restore reactor coolant chemistry values to normal levels in a reasonable time frame. Under transient conditions, potential crack growth rates could exceed analytical assumptions, however, the duration will be limited so that any effect on potential crack growth Is minimized and the design basis assumptions are maintained. The plant Is operated such that the average coolant chemistry values for the operating cycle are mnaintained at the conervativo values of C 0.2 pmho/om for conductivity end < S ppb for Page 1 of 3

chloride or sulfate ions.. This will ensure that the crack growth rate is bounded by the 5E-5 in/hr core shroud analysis assumptions, since the analysis shows a crack growth rate of < 2-2E-5 in/hr for these chemistry levels. Since the conductivity and chloride and sulfate ion values are average values, there are no specific LCO actions to be taken If these values are exceeded at a specific point in time. However, plant prmdures will ensure that actiors are tkinn to rpedtice the chemistry levels to the.pproptiate levels within a reasonable time frame.

The NMP1-specif'c analysis has established that the BWRVIP-14, Section 6.1.1 stimss intensity independent crack growth reLe of 2.2E-5 in/hr is conaervative for NMP1, PaovMcd that the average reactor coolant conductivity Is maintained < 0.2 /rnho/cm. The reactor coolant conductivity applied in the analysis derived a "model' conductivity which considers that reactor coolant is at the 5 ppb limits associated with the dhloride end sulfate Ion concentrations& Typically conductivity Is maintained below 0.1 pmho/cm on a cycle average' basis. This ensures that the NMPI -specific shroud analysis calculated crack growth Is bounded by the 2.2E-5 in/hr growth rate as determined by the 9WRVIP-1 4 disposition.

CONCLUSIONS The desigija docuivitniutiut und evuluations provided by NMPC to demonstrate the acceptability of the as-found vertical weld cracking in the NMP1 core shroud for at least 10,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months ot hot (above 200 degrees F) operation were accepted by the NRC.

However, the NRC's safety evaluation was contingent on maintairing reactor coolant chemistry within the BWR water chemistry guidelines, 1996 revision, and on the submittal of an apallcation for amendment that addressed the difference between the current Ts conductivity limits for reactor coolant chemistry and the analysis assumptions for core shroud crack growth rates. These proposed changes, which are equal to or more restrictive than the present TS values, will asure that NMP1 Is operated within the requiremento of the onallysi u*ed for the NRC's safety evaluation.

ANALYSIS No S*gnifi*eat Hazards onsideration Analysis 1 0CFR50.9 1 requires that at the time a Becanse requests an amendment, it must provide to the Commismion its analys;es using the standards in 10CFR50.92 concerning the issue of no significant hazards consideration. Therefore, in accordance with 10CFRSO.91, the following anslyces havc been performed with respect to the requested change:

The ooerafion of Ning Milei P4Ent Unit 1. in accordance with tt)B mo2gased arendment will not involve a slinificant increaUa in th-rffobabllitv or conseauences of an a9cident oreviously Ave luated.

The changes to the conductivity and chloride Ion action levels and the addition of sulfate ion levels as an action level in reactor water chemistry are being made to make the TB and its Bases consistent with the va!ues used in the core shroud vertical weld cracking evaluations. These new values reflect the BWR water chemistry guidelines, 1996 revision (EPRI TR-10351 5-RI. BWRVIP-29) and are equal to or morm restictive than the present TS vcluaes. No physical modifiection of the plant is involved and no changes to the method*.

in Page 2 of 3

which plant systeams ae operated are required, None Of the precursors of prevIously evaluated accidents are affected and therefore, the probability of an accident Previously evaluated is not Increased. These changes to the coolant chemistry TS are more restrictve limits-and n6 new failure modes are introduced. Therefore, these changes will not involve a significant ;ncmreM in the consequences of an accident previously evaluated.

Yhe operation of Nine Mile Point Unit 1. in accordance with the proood amendMrnntwii not groat@ the poasgbility Of

new gr diftMUt kind Wf accdent from any acaidorT previously evaluated.

The changes to the conductivity and ch¢lode ion action levels and the additon of sulfate Ion levels as an action level In reactor water clhemistry are being made to make the T3 and its Bases consistent with the values used in the core shroud vertical weld cracking evaluations. These new values reflect the BWR wate chemistry guidelines, 1996 revision (EPRI TR-1 03515-R1, BWAVIP-29) and are equal to or more restrictive than the present TS values. No physcel modification of the plant is involved and no changes to the methods in which Plant systems ate operatod rort rarijirmd. Thet change does nnt intrndtreA *n new failure modes or conditions that may create a new or different accident. Therefore. this change do". not create the possibility of a new or differont kind of accident proviojly evaluated.

The operation of Nine Milo Point Unit 1. in accordance with the oroposed amendment. will not involve a slanlflont redUCti!n in a marain ot satetv.

The changes to the conductivity and chloride ion action levels and the addition of sulfate Ion levels as an action level in reactor weter chemistry are being made to make the TS and its Bases consistent with the values used in the core shroud vertical weld cracking evaluations. Thes.. new vAlhun reflect the RWR water chemistry guideline.,.1996 revision (EPRI TR.10351 5-R, 1BWlVIP-29) and are equal to or more restrictive than the present TS values. No physical modification of tho plant is involved and no changes to the methods in which plant systems are operated are required. ThW change does not adversely affect any physicail beirrier lW Lim itlwsi uf radiation to plam personnel or the public. Therefore, tile change does not involve a significant reduction in 0 morgin of safety.

Page 3 of 3

ATTACHMENT C NIAGARA MOHAWK POWER CORPORATION UCENSE NO. DPR-63 DOCKET NO. S&.220 Narked Cotw of Pmn.sed Chylym to wmwnt Technical Specification The current version of pages 96, 1 7, end 9 of ft NMP1 Teohnfcal Specitleations have been hand marked-up to reflect the proposed changes.

LMITM COMMON FOR 01111311ATION SURVEILLANCE REGURUWAIT UMITIUG CONDITION FOR OFillATION SURVEILLANCE REOUUIEWJT 3.2.3 59LN HMSR Appmabjm.

Applies to the reactor coolant avateim chemnical Isquirements.

To asstir the chemical pualty of the reacotr coolant water.

L. The -mortor eodent weter shall not wmeed the 3.2.2c*.

4.2.3

COOLANT CHEMUB Applies to the periodic testin requiirment.0of the reector ©odoeb chemistry.

To deterrnine tdo chlTlulcd purity of the reactor coinlart water.

Samplea be taken eu en1lysed for conduct",

-a chwidiuI; coIteM at boat 3 tlins per week with a madmtzm time of 96 hour0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months . betwen samples.

a" n

ition. If the conductivity becoms abnornsl o0the then short teem volleu as indioated by the

":rJ o-ntinuous aonuuctvity m ior.

m*plu sha" be tuk end snalied wIthin £ hours and delly therwafter until conductivity returns to normal levels.

the contfrwoi otiruchivity monitor Is

.)1o7 Inorperable, a reactor coolant semlple shall be taken and analyzed for conductiV*t6e0t`lk 11dd6Au0n content at least once par 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months . *1-.A

%y~od.h

$U&F*I& Ion jIMIFano

b. The reactor ceohent water dull not eared thes flolow~reg "ito with

~1610.810 vini ti V-01 we *00,900 psands w~ lowr except-as specifled In 3.2.2c=

Conductivity ChiDMG bIo SV&PAE *1d) 4ý.ocrn ANOM4MENT NO. "f2 a

Be

URGTMG CONDMON FD% CIMATION SURVOLLAMM REQUINMNT.

WAliNG CONDI~1ON FO~ OPATION SLMVE)LLARcE RECUIREMINT.

In Im I...

nm c.The limt.spsvlfld In 2-23' an3.2-3b may be wxeuded for a psidi of time niot to exceed 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In no cese shal I t ii..

w U WW~~VNEW UYI~

y Jll~UlU~~l ~I IPf 4N I------

t!,n oM~',~

M lvii ii~~I'm~f 11M~

M

"=-..*.....=.=% _.._..rlm *.*

WWU

lVpV gr

d. 11 Spec ifications 3.2.3, b, end c aer not met, normal

~Vrl shutdcown shofl be initiated wit~n wes hout and the neactar shat be In thscold-

s.

the continouscu onduclivty nmonwo Is In

~apvbi for more. Omu 7 dao" dow remotar WWI be plnai ithi v.J..utdawneewww~w" --wftn 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

<A.ZE

06 F~

/ %ap&

/Cdlu e

2 Donis'&t-j*~~~~

~~

,,4A4 zocL A4 d

.~PD

~~tta~~&

>~-'F 4~g 7-4 97

L~q'f 4T~4Cr.~:

BASES FOR 3.2.3 AND 4.2.3 COOLANT ChEMSTRY It -

Matmials In the primary systam are iwimidly 304 sai1nreas stet sand the Zrvaloy fuel cladding. The reactor wetVW chemletrV Ilmits omr

/ etibUshed to prevent damage to thse mintelidL Umils are placed on chlorde corcentration and conductivity. Thu mast Importnt limit Is tht Pieced an chloride concentration to prevent strues crrosion cracking of the stainles steel. When the etoemIrV rate Is hssMan 100.000 pounds per hour. a more restrictive Uimit of 0.1 ppm has ben etamblished. At stemming retes of at lest 100,000 pounds par mf boding occurs causing daersarton of the rouctor w~ater, thus maintaluulng oxygen cnown trion at low levels.

A v1,ort term spike Is defined as a rise In conduclIci"such as thut which iouhL arlse from kisction ol gtcltionin loedwatair flow for a duration of approximstaly 30 minutes in time.

W h.,l Gonductivity Is In its proper nounul range.4FA--Peorid

&MM-Surii effecftd condct~ivity must also be within ther n~ormal irange. When and It conduc~tivi becomnes; abnormal. then c'tl Orlemaff ainlts ars made to determfine whe&(het or not tWhaey ainh out of their normal operaitlo values. JThe -veul A-a' n"es*c 1

t.i oa;-. Coodw~e -C~ ---

Wfu PIR m..r...,,.-sa,,

ie.,.

1~hwMntt Wu soc~

7' pH 0

cc el, CUM 900110 fs, 1111 j a hGO&O~~i, elwu is net a capum fy7 eotdowi-4.

IligyWqF lypSug wammwauu r"809101 Oesi 3ndLoid's; Mi In but~ high dIL4 to epasetdue JtiYU.I;.ea fOV'e invsWWIl t~na~h OM prto

~

hViudtesndpaing the rea-A cr acs devaendento h cOrrs~rates ari 0wletm o h la-psse ov-aslhteprt ftenca olr~)ma-t*

fOfiltt-JiyUn~

A

.!rn~ar4, shwu 4

fth rat fI Wvvmtteit" placing the.

.*6W~QUJM~w

=r4

%F"4.

W.e LamW

-M

~ua~~yeCees-jawpaf depndet orrsin Tie or p~deti e

fruo.kwu ythernt s~amples the puiyoo f th wic reao ehsnAevry s burnwi m*

e I--ie wil auleo6T bfse0P1OO ea V-rtthi nemalrneym.umo 2reuctor coolant r..uwp-e Cwhil We Modeain o thee idoif wea akare.

ofm dnved-metsk.-OIs co~*Iperiodm~et of 1"etectA Ion.0en changVSxes In te cmiorkahe1ir'"

be-toentoasTs-I,.IMp ofC r,"1~ff-P soxor Kftmnitred Thesemhs ir coolnt hic ar U~ei eery26 ourswarsere m0

=

1"IN-0 S

U-z*

60=0 he a tm d

w~p---tf.

ontuoMV I

it

ýL-t-thir-nnna-FArWL he oto co~nt ampes ill hla6beuse AMENDMENT NO.,41' 98

This spedcUicatn is beiit MubMiMtc o addes an NRC Sfy vualitiWm if n its May.9. 1997 letter, the NC required that N itbnit a a1cation for am"M iment to addtm the diffeaences betwee the curzent TS oondumvity linits for reactor coolant chemistry and the analytis assumptions for the com z--rmd crzck-growth evaluatiom. The purpose of this qxzficafimo is to limit ingranular M

M ki c

ng (IOSCC) craek gruwth raws (tbuh the comrol of Va= coolant chmir-ry. The LC0 valUCS anSure that transent co otIs ao acWd m to rcstoe ractor coolant chemistry values to normal in a reasonable time fmame. Unh*r tUamimt conditions, potential crm growth rates could exceed analyfica mznpfons, howe., the duration will be limited th3 t any effect on Potential crack growth is minimzed and the design basgs MonVtms are minainaaed. The plant is normally opasatd such that t= avcragc chemistry for t qpcruing oyce is maib nM at the conurvaive values of < 0.2 gmholcm for daintevity ad < ppb for chloride ions < 5 ppb for sulfat ions. This will enmsu ha the crack prowth raw i bounded by fhe ca shroud analys assumptions (the aalysis shows the arak nowth to be < 2.2*-5 inAhr for these levls). Since these ame avrage values, them am no ic& LCO actios to be tak=

if these values am exceeded at a apdfr point in ime.

specification 3.2.3a, b, and c is onsi*eat wv4M t BWR waw coolant chezmisuy guidelinms, 1996 rWv (EPRI TR*03515-1., BWRVIP-29). 7W 24 hout actio tim period for exceeding the coolant chemisry limits described in 3.2.3a and b ea=res that prompt action is tlk= to MWzC Coolatt ChMiSf to nraMal opegrAdU levels. The requirement t commence shutdown widi 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , aw to be shutdown and reactor oolant *em*pture be reudwcd tW < 200 dipas F within 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months ,jzn izti Lbk pyu tLi L. IGSCC rack growth.

NIAGARA MOHAWK G E N E R AT I O N BUSINESS GROUP P. PALP SYLVnA off m

July 2, 1997 CH Nudcar efNr M

L 1232 U. S. Nuclear Reguhaory Commisim A=: Docurnent Control Desk Wa~shington, DC 20S55 RE:

Nin Mile P*oit Unit 1 Dockt No. 50-220 aotnthien:

During the 1997 refieling outli al Nine Wile Point Unit I (NMP1), inspecdon of the core shroud vertiesi welds revealed cnich ia excess of the 3crening criteda. By lette dated April 8, 1997, Niapr Mohawk Power Corporo (NMPC) provided design documnation and cvaluations to dcmoamatat the acceptablity of thm as-found v=*A1 wZd acWns in tht N

l core shrod for at least 10,6W0 hours of hot (above 200 deees F) operation. By letter dated May 8, 1997, the NRC iued a Safety Evahailm aproving ft re Ptort of NMPI'

,ontingedt on: 1) mainling reactor coolant chemistry within the giiý act forth in the Electric Power Reamch Institute (Et* 1) technical report TR-103515-Rt (BWRVIP-29),

DWR Wa= Chcnziitry GuWdclin= - 196 Rviu*on,' and 2) the requirement that NMPC submit an application for a license ammudmnt to add= t difereace between the crrent TS conductivity limits for reactor coolant a=mty and the analysis asunphons for core shroud rack growth ratrm. The NRC approved the NMP anlysis preuized on the condition tfat NMPI is opWAW in acCoSISn with ft BWR waWr chumistry piddmcs.

This splleation for nmendm.

is being submittd to address the NRC'& secnd a*oina-cy, NMWC hreby transmits an Application fdr Amendment to MAPL Ope' License DPR-

63. Also enclosed as Armbment A is the proposed change to t TechnicW Specifca (MS) set forth in Appendix A to the above metiond licens.

Supporting information and analyms which demoastraw that tht pro*osed chame involves no sisnflant hazards con*idctaio puuant to 10CFR50.92 ar included as AUacmie~n B. A markW-up copy of the afflcucd TS lp-gs is provided is Altbtach=L C to assist yow review.

Page 2 7%C PrO 1

P Comae mvisU SoOdtw 3.2.3 WA 4.2.3 to lefifct the SRW water chemstry guidelines. In a4"dti, the Bases for 3.2.3 and 4.2.3, "Coolat Chemisy', haI bee=

rcvsa4d. These changes addresS the differencs between thu curent TS cunducLivity liniLb for reactor coolant chemistry and the analysis assump=

ns for core shroud crack growth rates.

Pursuant to IOCFSO,91(b)(1), NUPC has podiftl a copy of this license amendmemt reque.t and the associwxd analysis irg-,m no siniicant hazards consderation to the appropriate state reprentaive.

Very Mu#y yOu, Chef Nuclear Officer BRS/TRE/cmk Attachments xc:

Mr Miller, NRC Regional Administraor Mr, A, W. Otromerick, Acting Di or, Project Directorate, 1-1, NMR Mr. B. S. Norris, Senior Resident Inspector Mr. D. S. Hood, Senior Project Mauiet, NRR Mr. L. P. Spath NYSE1RDA 2 Empire Plaza, Suite 1901 Albany, NY 12223-125S kecor* Management

UNITED STATES NUCLEAR REGULATORY COMMISSION in the Maeter of

)

Magara Mohawk Power CorDoration Docket No. 50-220 Nine MUG Point Unit I APPUCATION FOR AMENOMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Commission, Niagara Mohawk Ilower Corporation (NMPC), holder of Facility Operating LUc"sM No, DPA-153, rarrby requse thaot Section 3.2.3 and the associated surveillance Section 4,2.3 of the Technical Specifications (TS) set forth in Appendix A to that license be amended. The Prooosmi changes have been reviewed in accordance with Section 6.5, "Review and Audit,' of the Nine Mile Point Unit 1 (NMP1) TS.

The proposed change revises the NMPI TS Section 3.2.3 to reflect the "BWR water chcmistry guldolince, 1996 rcvision* (EPRI TR-105T5-.R1, BWRVIP-21). Sootlons 3.2.3a and 3.2-.3b defrm new conductivity imite when the reactor water is > 200 degrees F nd thermaul powvr lI _£ 10%. uvid whlsn LhumiaI power is > 10%. The new conduwtiviLy limit is now 1 pmho/cm compared to the existing limits of 2 pmholcm and 5 pmho/cm-The chlorlee ion limit trom Uection 3.2.3a remains at the same level but it is listed as 100 ppb instead of 0.1 ppm. The chloride ion limit from Section 3.2.3b is changed from 0.2 ppm to 20 opb. Sulfate ion limits are added to Sections 3.2.3a and 3.2.3b at 100 ppb and 20.pb, respectiveiv. Riom Section 3.2.3o the maximum conductivity limit is changed from 10 pnho/cm to 5 pmho/cm, the maximum chlorida ion concentration limit is changed from 0.5 ppm to 100 ppb and 200 ppb. and the maximum sulfate ion oonoentrrattor of 100 ppb and 200 ppb Is added.

The proposed change revises NMP1 TS Section 4.2.3 to include sulfate ions as a component to be rInluded In Me sample analysis.

included in this TS change is a change to the Bases for 3.2.3 and 4.2.3. "Coolant Chemistry'. The Bases has been changed to reflect the purpose of the specification which is to limit intergranular stress corrosion erciting (IGSCC) crack growth rates through the control of reactor coolant chemistry. The Bases dascribes the NMPI operating philot&ophy of mainuiing average levels for condiuct*ty and chloride and sulfate oomentratons over an operating cycle. Operation of the plant within these average values will ensure that the crack growth rate is bounded by the core shroud analysis.

The proposed change will not authorize any change in the types of effluents or in the authorized power level of the tacility in conjunction with this Applicstion tor Lcense Amendment. Supporting information and analyw which demonstrate no significant hazards considerations pursuant to 10CFR50.92, ere included as Attachment B.

WHEREFORE, Applicant rwpeCtfully requests that Appendix A to Facility Operating Ucense No. DPR-63 be amended in the form attached hereto as Attachment A.

NIAGARIA MOI IAWK POWER CORPORATION By 6

B. R. SyIa Chief Nuclear Officer IMLmY X. BIWA w" ^*uhwb UA101 Xu Y#

WMA amp V

ATTACHMENT A NIAGARA MOHAWK POWER CORPORATION UCENSE NO. DPR43 DOCKET NO. 50-220 Procand Changaes to TholviCa Snecficotions Replace tr* existing pages 86, 87, and 98 with the atached revised pages 9., 97, and 98, The pages have bean retyped in their entirety with marginal maorkin*

to indicate changes.

MATING 001413M ON FOR OPERATION SURVEILLANCE IMMUREMENT UWUTEWO C~ONZ~ITI0N FOR OPERATION SURVEILLANCE IEOJJIREMBIT 3.2.3 COOLANT CHEMISTRY Applies to the reactor coolant systemn ct-emdice requirementa.

To eseum the chemical purity of the reactor coolant water.

a.

The reactor coolant water shall not exceed the fMllowing irnkts with. the coolant temperature

> 200 degrees F and reactor thermal power

__ 10%,. except as specified In 3.2.3c:

4.2.3 Apples to the periodic tMsUnM requirements of the reactor coolant chemistry.

9w)ective.'

To determne the chemical purity of the reactor coolant warm.

Samples sall be taken end analyzed for conductivity, j chloride and saulfate Ion content at least 3 times per I

week with a maximun time of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months between samples. in addition, If,he conductivity becomes abnormal (other than short term spikes) as indicated by the conrinuous conductivity monitm, samples shall be taken and nailyzed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months end doily thereafter until conductivity returns to normal levels.

When the contnuoLBs conductivky monitor Is inoperable, a reactor coolant sample shall be taken and analyzed for co-ductlvity, ch:oride and suilate ion I content at least once par 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Conductivity Chlcrde (on Sulfate Ion 1 pmho/crn 100 ppb 100 ppb

b.

The roaoaam coolant water shal not exceed the following limits w0th reactor thermal power

> 10%, exceptr as specified In 3.2-3c:

Conductivity Chloride ion Sulfate ion I pirnho/fttn 20 ppb 20 ppb AMENDMENT NO. U1 96

LIMMNQ CONDITION FOR OPERA'nON SURVISILLANCE REQUIREPANT LIMniMO ~OPLO~0N FOR OPERATION 6Ui~VEflA.AtICE KEQUIREMENT 4

c.

The limits specified In 3.2.3a and 3.2.3b may be exceeded for a period of time not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In nn case shad the reactor coolant exceed the falowfng limits at the specified conditions:

1.

With reactor coolant tern perature Ž_ 200 degrees :, the conductIvrtv has a m~ifmum limit of 5ph4/ncm, or

2.

With reactor coolent temperature > 200 deorees ý and rean'tor thermal rower

.<. 10%. the maximum lmkt of chloride or sulfote Ion concentration Is 20C ppb. or

3.

W*ith rmactor thermal powow > 10%. the manimum rmnit of chloride or sicfate ion concentrotion is 100 ppb.

d.

if Specifications 3.2.3m, b. and c are not ueet.

normal ordudiy shutdown shlal be initiated within one hour and the reactor shall be shutdown and reactor caolvit tenperature be reduced :v

< 200 degrees F within ton hours.

a. If the continuous conduovty monitor is Inoperable for more then seven days, the reactor shel' be shutdown and reactor coolmt temnperature be reduced to < 200 dogrees f within 24 howrs.

AMENDMENT NO.

97 97

BASES FOR 3.2.3 AND 42.3 COOLANT CHEMISTRY This specification Is being submitted to address art NF;C safety evaluation requirernent. In its May 8. 1997 Iettor, the NBC required that NMPC submit an application for amendment to address the differences between the current TS conductivity lindta for reactor coolant chemistry andi the analysis assmpt one for tie core shroud crack growth svaluaions. The purpose of this specification is to li.lt inte.-grarular stress corrosion cracking IIGSCC) crack growth rates through the control of reactor coolant chemlstry. The LCO values ensure that transient conditoios *is acted on to restore reactor coolant chemlstVy values to normal in a reaonable time frame. Under transient conditions, potential coick growth rates could exceed analytical assumptions, howeiver, the duration wil be limited so that any effect or potential crack growth is minirrized and the design basis assumptions sia maointaind. The plant is normally operated such that the average chemistry for the operating cycle Is malirined at the conservative values of < 0.2 pnholcm for conductivity and < 5 ppb for chloride ions

< 5 ppb for sulfate lo-s. This wil ensune that the crack growth rate Is bounded by the core shroud analysis assumptions (the analysis shows the crack growth to be < 2.2E-5 Tn/lt for these leelis). Since these are average values, there are no specific LCO actions to be taken if these values are exceeded at a specific point in time.

Spowificatlon 3.2.3a, b, and c Is consistent with the BWR water coolant chemistry guidelinee, 1996 revision (EPRI TR-103515-R1, BWRV1P-29). The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months action time period for exceedina the coolant chemistry limits described in 3.2.3. mnd 3 ensures that prompt action is taken to restore clant chemistry to normal operating levels. The requirement to coninence sautdown witlifn I hour, and to be shutdown and reactor coolant temperature be reduced to < 200 degrees F within 10 houers rmnis-nzes the potentse for IGSCC crack growth.

A short lerm spike Is defined as a rise In conductivity (> 0.2 pnhol:m) such as that which codd arlse f-vom injocian of additional feedwater flow for a duration of approximately 30 minutes in tirre.

When conductivity is in Its profer normal range. chloride, sulfate, and other impurities affecting conductivity must also be within their nornal rangeo.

When and i conductivity becomes abnormal, then chloride ard sulfate measurenierds are made to detumlr;e whether or not they are also out of their nnrmal operating values. Significant changes provide the operator with a waning mechanismn so he can investigate and remedy the condition causing the change and ensure that no-mel operating average conditions are maintained within, the bounds of the core shroud cra'c growth analytical assumptions. Methods wallable to the operator for correcting the off-standard condition include, operation of the reactor clean-Lp systemr, reducing the input of impurities. and placing the reactor in shutdown and reda.cing reactor coolant temperatue to < 200 degrees F. The major benefin of reducing reator coolant temperature to

< 200 degreas F is to reduce the temperature deaendent corrosion rates and provide time for the clean-up system to re-estabrish the puialt of the reactao coolant.

The-conductivity f the reactor coolant In continuously monitored. The samples of the coolant which aer analyzed for coaductivity every 96 houm will save as a comparisorn with the continuous -oonductivity monhor. The fesctor coolant samples will also be used to determine :he chloride end sulfate concentrations. Therefore, the sampling frequency ;s considered mloquate to detect long-term changes" in the chloride aid sulfate len content. Howaver, if the conductivity becomes abnormal (> 0.2prnho/cmt. chloride and sulltes measurements will be made to assure ths. the nornmal irnits (< 5 ppb of chloride or sulfate) are mnaintaTned.

AMENDMENT NO. iii 08

ATTACHMENT 8 NIAGARA MOHAWK POWER CORPORATION LICENSE NO. DPR43 DOCKET NO. 50-220 8unoortina Information and No Si)arfiynt Hazards Coomdeq*otin Anrjyula INTRODUCTION The proposed Nine Vila Point Unit I (NMPI) Technical Specification (TS) change contained herein presents e revision to NMPI TS Sections 3.2.3 and 4.2.3, and the Bases for 3.2.3 and 4.2.3, "Coolant Chemistry" By letter dated April 8, 1997, Niagara Mohawk Power Corporation (NMPC) provided design dv.U;rlwnstLaufI Wid uvulualuiis Lu delumiLlradLe t im a,;,ptabLi[ty uf Ltu lat-fuwid vertical weld craeking in the NMP1 core shroud, for at least 10,600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months of hot (above 200 degrees F) operation. In its May 8, 1997 letter, "Modifications to Core Shroud Stabilizer Lower Wedge Retaining Clip and Evaluation of Shroud Vertical Weld Cracking, Nine Mile Point Nuclear Station, Unit 1," approving the restart of NMP1, the NRC required that NMPC submit an a-imlcation for a license amendment addressing the difference between the current TS conductivity limits for reostor coolant chemistry and the analysis a;Sumptions for cQre shroud crack growth rates.

This *rOpOsed Change incorparatea into the TO the reactor coolait chemistry assumptions that were used for the core shroud weld crack evaluations.

EVALUATION The proposed revisions to TS Sections 3.2.3a, b, c, d, and e incorporate the analytical assumptions that were used by NMPC to evaluate the vertical weld cracking found in the NMP1 nrrA *hmijd rir'eing the 1997 refuerin0 outage. The TS chongeh establish limits for conductivity and chloride and sulfate Ion concentrations that are equal to or more roestritive then the axistang TO volues. As a rosult of tho anolysls, on bvorage value of 0.2 urnho/cm has been chosen ior conductivity which is less than the BWR guide4ine SuLiUu latvu!I v waluat fw undueut~ivity of 0.3 ormhobm.

I he purpose of this 1 5 change is to limit IGSCC crack growth rates through the control of reactor coolant chemistry. The proposed LC0 values ensure that transient conditions are acted on to restore reactor coolant chemistry values to normal levels in a reasonable time frame. Under transient conditions, potential crack growth rates could exceed analytical assumptions, however, the duration will be limited so that any effect on potential crack 0rowth Is minIrmized and the design basis assumptions are maintained. The plant Is operated such that the average coolant chemistry values for the operating cycle are maintained at the cgrmrvwtve vakJee of - 0.2 prho/om for conductivity and < S ppb for Page 1 of 3

chloride or sulfate ions.. This Will ensure that the crack growth rate is bounded by the 5E-5 in/hr core shroud analysis assumptions, since the analysis shows a crack growth rate of < 2.2E-6 in/hr for these chemistry levels. Since the conductivity and chloride and sulfate ion values are average values, there are no specifio I-CO actions to be taken If these values are exceeded at a specific point in time. However, plant procedures will ensure that actiors r

tAtiPkn to rwdtLiee the chemistry levels to the *ppropriste levels within a reasonable time frame.

The NMP1-specific analyes has established that the RWRVIP-14, Section G.111 sre intensity independent crack growth reLe of 2.29-5 in/hr is conservative for NMPI

'eovied that the average reactor coolant conductivity Is maintained < 0.2 pmho/cm. The reactor coolant conductivity applied in the anelysis derived a "model" conductivity which considers that reactor coolant is at the 5 ppb limits associated with the chloride and sulfate ion concentrations. Typically coriductv;'ty Is maintained below 0.1 #mho/cm on a cycle average' baes.

This ensures that the NMPI -specific shroud analysis calculated crack growth is bounded by the 2.2E-5 inthr growth rate as determined by the BWRVIP-14 dlsposltion.

CONCLUSIONS T*e de,*iyz docuimauiLiuni und evuluations provided by NMPC to demonstrate the acceptability of the as-found vertical weld cracking in the NMP1 core shroud for at least 10,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of hot (above 200 degrees F) operation were accepted by the NRC.

However, the NRC's safety evaluation was contingent on maintaining reactor coolant chemistry within the BWR water chemistry guidelines, 1996 revision, and on the submittal of an apolication for amendment that addressed the difference between the current T-conductivity limits for reactor coolant chemistry and the analysis assumptions for core shroud crack growth rates, These proposed changes, which ire equal to or more restrictive fthi the present TS values, will mure that NMP1 Is operated within the roquiremento of the enatlyi3 used for the NRC's safety evaluation.

ANALYSIS No Snifieart Hazards Consideration Analysis 1 OCFR50.9I requirems that at the time a licensee requests an amendment, it must provide to the Commission its an/rlyiys tieng the standards in IOCFRSO,02 concerning the issue of no signhflcant hazards consideration. Therefore, in accordance with 1 OCFRSO.9 1, the following analyfts have boon pcrformed with respect to the requested change:

The operation of NI*n MiW PC'nt UnIt 1, in accordance with the orooasuki amendment. will not Involve a siagnficant inerea" in the -robability or gonseauences of an agident greviouuly Aluated.

The changes to the conductivity and chloride Ion action levels and the addition of sulfate ion levels as an action level in reactor water chemistry are being made to make the TS and its Bases consistent with the va!ues used in the core shroud vertical weld crackigr evaluations. Thoe new values reflect the 1WR water chemistry guidelines. 1996 revision (EPRI TR-10351 5-RI. BWRVIP-29) and are equal to or morm restrictive than the present TB values. No physical modifterono of the plant is involved and no ¢hangr*

1 to tho methods in Page 2 of 3

which planT systwa are operated are required, None of the precursors of prOviously evaluated accidents are affected and therefore, the probability of an accident previoWly evaluated is not Increased. These changes to the eoolant chemistry TS are more restrictive limits-and nO new failure modes are introduced, Therefore, these changes will not involve a significant ;ncrese in the consequence& of an accident previously evaluated.

The operation of Nine Mile Point Unit 1. in accordance with the proooged armiendMrnt*, wi not eresto the peatibll*ty of _a Ow Qr diftMrnt kind 6f sccidant forom any acciden previously evaluated.

The changes to the conductivity and chloride ion action levels and the addition of sulfate Ion levels as an action level In reacor water chemistry are being made to make the TS and its Bases consistent with the values used in the core shroud vertical weld cracking evaluations. These new values reflect the BWR water chemistry guidelines, 1996 revision (EPRI TR-1 035 1 S-R1, BWAVIP-29) and are equal to or more restrictive than the present TS values. No physioal modification of the plant is involved and no changes to the methods in which plant systems am operated arn mvraiirad. The Mhange drin nt intrndtrA any new failure modes or conditions that may create a new or different accident. Therefore. this change do" not create the possibility of a new or different kind of accidont proviely evaluated.

The operation of Nine Mile Point Lnit 1. in accordance with thf oropsed amlendment. will n ot involve !I sll-gfloant reduSIOQn in a margin al sptcty, The changes to the conductivity and chloride ion action levels and the addition of sulfate ion levels as an action level in reactor water chemistry are being made to make the TO and its Bases consistent with the values used in the core alsoud vertical weld cracking evaluations. These npw vAlthe. reflect the RWR water chemistry guidelines, 1996 revision (EPRI TR.103515-RI. BWRVIP-29) and are equal to or more restrictive than the present TS values. No phyalcal modification of the plant is involved and no changes to the methods in which plant systems are operated are required. Ths change does not adversely affect any physical border W Lim itilause uf radiation to plant personnel or the public. Therefore, thle change does not involve a significant raduction in e margin of safety.

Page 3 of 3

ATTACHMENT C NIAGARA MOHAWK POWER CORPORATION UCENSE NO. DPR-63 DOCKET NO. SO.220 farked Caon of Pmaosed CthasionC-went Technital Specification The current version of pages 96*,V, ond 9* of ft NWV1 Technlcal Specitletions rAve been hand marked-up to reflect the proposed changes.

LWIUTID CONDITION FOR pigs1ATON BURVIEILANCE REGIIRUWMET 3.2.3 OLN HM8 Appiehs to the goe.tor coolant sstMn chemical isquirwiments.

To amatx Ohe chemical purity of the roeao coolant aL The - e ow codlet water shall wot snoeodth fallowig Armits with oftwzl eg jis lose V.

g IG9A6O Peude Per how wexetuam specified In 7

3.2.3c:

4.2.3-GLN HU O

Applies to 1he pedoiodi testing requirarnent:of thu reactor coallar chemistry.

To detarrrning the chumnicoll purity of the reecor coolart water.

Somke betakm mid ana lyzed for cornducth~t~

ýSw 01W ctetM at kheat 3 times per week with a movdmum tim* of 9S &our@ bstween samples.

-f'n ddtim I

couductivimy'

.w bnormd othr ten hontamspies)asbuWiaged by doe continuous conductivity nmonior, ameni.

"Ia be takien i d anolyzed witgi 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and delly themuuter unti Doncivi~tyvtt returns to nminal levels te onthuaota cottduivity monitor Is

>/0 Moerable, a reactor ocolant. sorple shell be taken and arealyzed for conuudctdtwn~cw~td0e~lsi content at meust once pr 13 Hours.

a Ch*q~e Ion SUuA.F4172

.bu

'a.',11(

b. The reactor ceds~t water "vit not exceed the following "iito vith sommooft nto w

wfrowr tioin

- -'a

$11 GuO~l

e. 4 ww hiex ecmpt-@au speclified In 3.2,ac:

Conductivity C1111011210 IM SLkf4,r tw AMeOMMENT NO. W4r9 Be

UNTIMG CONDITION FO OIPEATION UTINR CODITIN F~t O~flONSLIMVEILIANCE REafUIEEMT

c.

The liits. "mottled In &2.34 end 3.2-3b ay abe wecedud for an priod of time not to eioead 24 hews In no cab* s m f!)"ftco w a-m& -4

-a me W vn

%A A

=

sW~

U W44EI~qiWIWI u'vUBiB ~V uv Jp.;..rwu,..,H I4LI L.=...

"M*'*'

Ju, uuu

.-U5t*1 WIC

~Ad et~-s.

T ý -1 1 -L ý -- I. A WV--..

d.

It Saclfkcmtlons 3.2.38, b, and c ar not met, normal o~*rty shudown shol bI Initiated within mw hour and the nacta shag b In Met -cu3d

e.

the onatinwusu LIonhfuetwity noultar Is Inoperable for more 1m 7 do" doe memor shalt 4---Aý 4 e Zf; Af n m

ý amun i4

'hDuFS.

e-

<acs P

W./

C*

U

,.=

f 4.,J AMJMENT NO. *J2 97

L-09 1, 4c(,C BASES FOR 3.2.3 ANDO 4.2.2 COOLANT CHEMISTRY rMa-teriahs In the Vdimary systm are prlmadly 3D4 "Fels~r s atiel arnd the 33rcaloy fuel cladding. 7h. reactor w*Wtai herrItry 6ni~ts ore aetfllished to Prevent damage to fthse mitoelti.

Umila are pliced an chloide concentration and conductIvity. Thu rmot Important limit is dhat placed ant choride concmntration to prevent stress corrosion cracking of OwestaunlesssMeal. When tde utemmimi raeti Ieshs Men 100.000 pounds per hour. a more restrictive Uimt of 0.1 ppm has boon etamblisled. At steeinIg ratio of at Weest 1O00,O pounds per howu.

bollifi coccrs coukig deatration of the emoteor wavter, thto maintaoakrg o"ygmn oonoentrafion at low levels.

A short term splice Is defined as a rise In conduictiti'such ei tht

-Ih could arise from kiection ot e"dton al 1emdwateak flow 11orm duxatlo df approxinstely 30 aliwtes in ftif.

Whenr conductivity Is Ini ts proper nomal rongeAviv )c oridg.

1 urhites atlecting conductivity must also be within their normal irange. When and It couiduiclhty becomes abnormaul. then c'ilorlde'msi~aiNCInts aer made to determines whslher or not t!"

are adsm out or thir normal operellniveluas. Is M

nusN,1t-as.CdL IA Mt.i tu y pwg..4 ua.Jtft (O.N.Jo sia~o.

"et~e wow noth Ii~.un W10:114 011emo

-on P61a or l

i-weat chlrd.(i...r

.L.

~jA 7

d~

,. 1w.

wwtcu

.wu fs, ~jwJiAfl bowePoucc -

widy soe usda W wim votrs p! f4 is mainwi oenduWAYPmadetis awygo w ouato11

-ftvsk of-Owt caus~~e~n paing the d.,

morart deedatcorefo Tls adpoid time fo Vh la~u ytmt emsb h uiyo h eco otnJu~

ga.

I.anpudtlteprondum~

2bn theokn tieaarunh 0h

m. f dndvWidrotlarfC~

ing

,.1 uJ pola o~flW dis e&WuI

V_=cede-jimlwistftsr ow~

w.-

moo Term

_11. =IwNUU m

W/

ILL fte a

5 to r

nt nvoullbt mnonitored. The samples of Owe coolanit which are U6mn every 99fuswl carvs 54.F W -

r~i-u iU W

J¶jI 98-0 I UUU4.1UU4WW-.EUL-IFIUWO L.

widi Ws.,e rienvol rON94 r 33 r-Ijv~

will also be, wk~rhaft-ielr mal-rangs. TMe reator coolant samleswM b usd~

to ateurde te ooil TdW&?f~WI~Wi Ui~C Isco~iXLJejIt~ate to udetect long-term changesi In tin f. fvw contqnt. Hlow ver. If the Mo~c~yo-lf~eavlv, ctidorIds,4QiiiiAaxm,& will be made to mssu'e that tf4~&linis4t(6.~~

o6"eefcen &.2.8 Ore...1i.,.;de

('3M C

AMENDMENT NAO.,14T 8

o

This spedcl tln is beifg submim to addre an NR sa~cy ccug rziuim u I b

l*

May.8, 1997 letta, the NRC rqued that NUW abniC an apCplicati03 for amendment to addr mn the differences beweef the o

uenz TI d vy limits for a

actOr coolant cheAmistry and the a*alysis zassumin for e Ocom,rm,,ind CM'"

growth ValUat,,ons. The purpose of this spedficatiou is to limit intrmnulaw sess omnssm cicking (IOSeC) cGck growth ramt Lhrouh td coatrol of wawbor cixlmnt chcmisry. The LCO valucs ansure that transient conditions are &,dan to s.. I f coolant cbemisty values to normal in a rcasonable time frame. Under tmansicnt conditions, potetial cra*

rowt ntes could exceed Analytical aumptis, howev*, the duration wiMl be limited so tha any effct on potential crak growth is ini and te deign basis assumptOs = maintained. The plat is normally opcW such *at the.,-,-c chcamnisty fo ft qpc:%6q cycle is maintained at the consrvtive value, of < 0.2 gmho/.m for omutivity and <5 ppb for chlaid ous <5 ppb for sulfate ions. This will ansure, ta the crack prowh rate is bounde by the cora shroud analyds asaumptlom (the analysis show t c'ack wowth to be < 2.2E.5 ift for these lev-s). SiT*n these are average va1e tham arc no qm&

LCO actio to be tae if these values am excee, at a mpcif; point in

i*e.

specifLication 3.2.3a, b, and c is Consistcat witt ft BWR w&e CoOlant cheitdry guiddinos, 1996 rCvisMi (EPRI TR-10351S-R1, SWRVIP-29). The 24 hom acdoe time paiad for exceeding the coolant chenstry lmnits descibed in 3.2.3a aWd b

es

that prompt acdon is taLken to EMtM coolant ahemisz to normal operting Ieveds. The rquizrcmet to commence shutdown widhn 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and to be shutdown and f'ador coolant pt¢aaux be rcdu to < 200 dwZpms F wibn 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months inudmiaz le put=tlal ftx I(SCC Crack growth.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE COMMISSION In the Matter of:

))

SOUTHERN CALIFORNIA EDISON

)

License No. 50-361

)

License No. 50-362 (San Onofre Nuclear Generating Station, Units 2 and 3))

NOTICE OF APPEARANCE OF GEOFFREY H. FETTUS The undersigned, being an attorney at law in good standing admitted to practice before the court of the District of Columbia, hereby submits this notice of appearance in the above-captioned matter to indicate that he is counsel for Natural Resource Defense Council, Inc. (1152 15th Street NW, Suite 300, Washington, DC 20005).

Respectfully submitted, Executed in Accord with 10 C.F.R. §2.304(d)

/Signed (electronically) by Geoffrey H. Fettus Geoffrey H. Fettus Senior Attorney Natural Resources Defense Council 1152 1 5th Street, NW Suite 300 Washington, DC 20005 (202) 289-2371 June 27, 2012

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing NRDC's Response in Support of FOE's Petition to Intervene and NRDC's Notice of Intent to Participate in this matter were served via electronic mail to the addresses listed below on the 2 7th day of June 2012.

U.S. Nuclear Regulatory Commission Chairman Gregory B. Jaczko Mail Stop O-16G4 Washington, DC 20555-0001 CHAIRMAN@nrc.gov U.S. Nuclear Regulatory Commission Office of Commission Appellate Adjudication Mail Stop: O-16C1 Washington, DC 20555-0001 OCAAMail.Resource@nrc.gov U.S. Nuclear Regulatory Commission Commissioner George Apostolakis Mail Stop O-16G4 Washington, DC 20555-0001 CMRAPOSTOLAKIS@nrc.gov U.S. Nuclear Regulatory Commission Commissioner William D. Magwood Mail Stop O-16G4 Washington, DC 20555-0001 CMRMAGWOOD@nrc.gov U.S. Nuclear Regulatory Commission Ms. Marian Zobler Acting General Counsel Office of the General Counsel Mail Stop: 0-15 D21 Washington, DC 20555-0001 MZOBLER@nrc.gov U.S. Nuclear Regulatory Commission Office of the Secretary of the Commission Ms. Annette L. Vietti-Cook Secretary of the Commission Mail Stop O-16G4 Washington, DC 20555-0001 NRCExecSec@nrc.gov U.S. Nuclear Regulatory Commission Commissioner Kristine Svinicki Mail Stop O-16G4 Washington, DC 20555-0001 CMRSVINICKI( nrc.gov U.S. Nuclear Regulatory Commission Rulemakings and Adjudications Staff One White Flint North 11555 Rockville Pike Rockville, MD 20852 hearingdocket@nrc.gov U.S. Nuclear Regulatory Commission Commissioner William C. Ostendorff Mail Stop O-16G4 Washington, DC 20555-0001 CMROSTENDORFF@nrc.gov Southern California Edison Company Russell C. Swartz Senior Vice President and General Counsel 2244 Walnut Grove Ave.

Post Office Box 800 Rosemead, CA 91770 russell.swartz@sce.com

/Signed (electronically) by!

Geoffrey H. Fettus 11

ML121850554

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ML121850554

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Dear Ms. Vietti-Cook:

Background

SUBJECT:

Dear Mr. Bajwa:

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