A04824, Submits ATWS Rule Schedule Required by 10CFR50.62(d),per Generic Ltr 85-06.Facilities Participating in NRC Integrated Safety Assessment Program.Justification for Absence of Schedules for Haddam Neck & Millstone Unit 1 Provided

Submits ATWS Rule Schedule Required by 10CFR50.62(d),per Generic Ltr 85-06.Facilities Participating in NRC Integrated Safety Assessment Program.Justification for Absence of Schedules for Haddam Neck & Millstone Unit 1 Provided

ML20133K567
Person / Time
Site:	Millstone, Haddam Neck, 05000000
Issue date:	10/11/1985
From:	Fee W, Opeka J CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Butcher E, Zwolinski J Office of Nuclear Reactor Regulation
References
A04824, A4824, B11738, FL-85-6, GL-85-06, GL-85-6, TAC-59114, NUDOCS 8510220158
Download: ML20133K567 (7)

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,, (203) 665-5000 October 11,1985 Docket No. 50-336 A04824 Bl1738 Director of Nuclear Reactor Regulation Attn: Mr. Edward J. Butcher, Chief Operating Reactors Branch //3 Mr. John A. Zwolinski, Chief Operating Reactors Branch //5 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

Millstone Nuclear Power Station, Unit Nos. I and 2 Haddam Neck Plant ATWS Rule Schedule required by 10 CFR 50.62(dj On April 22, 1985, Northeast Nuclear Energy Company (NNECO) and Connecticut Yankee Atomic Power Company (CYAPCO) received Generic Letter 85-06(1) concerning quality assurance guidance for ATWS equipment that is not safety related. Generic Letter 85-06 indicated that issuance of the quality assurance guidance shall be considered the reference date initiating the schedule in 10 CFR 50.62(d). Section 30.62(d) of the ATWS rule requires each licensee to develop and submit a proposed schedule for meeting the requirements of the rule within 180 days, af ter issuance of the QA guidance. Additionally, the ATWS rule requires any schedule for completion, beyond the second refueling outage af ter July 26, 1984 to be justified. The purpose of this letter is to satisfy the requirements of 10 CFR 50.62(d) for Millstone Unit Nos. I and 2 and the Haddam Neck Plant.

Millstone Unit No.1 Millstone Unit No. I is participating in the NRC's integrated Safety Assessment Program. On May 17, 1985,(2) NNECO proposed that the scope of work required by 10 CFR 50.62 be explici considered in the context of the ISAP program. By letter dated July 31,1985, the NRC formally documented the results of the ISAP screening raview process for Millstone Unit No.1, and identified the issues or " topics" which the Staf f had concluded should be consider ed in ISAP.

Enclosure 2 to the July 31,19851ctter specifically identifies the requirements of 10 CFR 50.62 as 15AP Topic 1.18 for Millstone Unit No.1.

(1) H. L. Thompson Generic Letter 85-06 to All Licensees and Applicants, dated April 16, 1985.

(2) 3. F. Opeka let ter to C. I. Grimes, dated May 17,1985.

(3) H. L. Thompson letter to J. F. Opeka, dated July 31,1985.

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j 10 CFR 50.62(d) requires that all modifications necessary to address the ATWS

issue be completed by the second refueling outage af ter July 26,1984, or, if the i schedule calls for implementation later than this time, that justification be

provided and a final schedule mutually agreed upon by the Commission and the

, licensee. In the case of Millstone Unit No.1, the second refueling outage af ter 1 July 26,1984 is currently scheduled to begin in August,1987.

! On August 13, 1985,(4) NNECO provided a review of the current Millstone Unit i No. I design agqinst the requirements of 10 CFR 50.62. Additionally, on j August 23,1985,W NNECO provided an analysis of the public safety impact of i l the proposed ATWS modifications based on the Millstone Unit No.1 Probabilistic Safety Study which was performed in support of our ISAP ef forts. These  !

i evaluations concluded that there would be a slight reduction in risk resulting  !

j from implementation of these modifications. However, in keeping with the j philosophy of the ISAP program, this issue must be evaluated with respect to all other ISAP issues before a final determination and schedule for implementation can be proposed. As such, we are unable to provide a firm schedule for l

implementation at this time.

As interim justification, until a decision and schedule for implementation can be i determined through ISAP, we refer you to the NRC's ISAP Program Plan, [

l SECY-84-133, which states: )

I "To provide a stable environment to conduct ISAP, the Commission j has authorized the (NRC) Staff to suspend specific existing l implementation schedule requirements for the plants to be reviewed. '

} Each affected license will be expected to propose and justify deferras t

for specific implementation requirements that warrant further

evaluation. The associated implementation requirements and other safety issues will be evaluated collectively in an integra ted  !

assessmen t."

i Additionally, the NRC's raf t report on the risk-based evaluation of ISAP issues l

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for Millstone Unit No.1(6 concludes that "the maximum potential risk reduction l i from the installation of an 86 gpm St.CS is rather small." The report also states i

that "a plant-specific ATWS analysis should be performed to finalize the priority j of this issue." And finally, the report concluded that the analysis would demonstrate that the issue should have a low priority.

Itased on the fact that Millstone Unit No. I currently complies with two of the  !

three specific requirements for boiling water reactors, and that the one i remaining issue has been shown to not be a significant safety concern requiring near-term action, we conclude that the absence of a firm schedule at this time is justified. As stated earlier, a schedule for 1rnplementation of all outstanding

, plant backfits will be mutually agreed upon by NNECO and the NRC Staf f in the l context of the ISAP program.  ;

(4) 3. F. Opeka letter to C. I. Grimes, dated August 13,1985.

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($) 3. F. Opeka letter to C. I. Grimes, dated August 23,1985. )

(6) C.1. Grimes letter to R. M. Kacich, dated September 23,1985.

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Millstone Unit No. 2

' 10 CFR 50.62 requires that the following systems be installed at Millstone Unit No. 2 and be diverse from the Reactor Protection System (RPS):

1. Diverse Scram System (DSS) i

' 2. ATWS Mitigating System Actuating Circuitry (AMSAC) consisting 1 of:

a. Auxiliary Feedwater Actuation System (AFAS)

b. Turbine Trip (TT) l NNECO intends to comply with the ATWS requirements at Millstone Unit No. 2 with the systems that are described below: ,

1 A. Diverse Scram System

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NNECO proposes to install a DSS at Millstone Unit No. 2 which will utilize  !

the existing pressurizer pressure sensors. The change will involve the j

installation of four channels of pressurizer pressure instrumentation devices consisting of Bistables and logic circuitry arranged in a two out of '

four logic to trip the Motor Generator output contactors in the event of an ATWS (see figure 1). Provisions will be included to bypass any of the sensors and convert the logic to two out of three to allow for maintenance of the sensors. The DSS will have the following design attributes:

1. Environmental Qualification - The components of the system will generally be located in mild environments. Where a component is located otherwise, it will be designed and qualified to the j environment that it is located in. l

2. Seismic Qualification - When a component is located where its failure i

due to a seismic event will adversely affect other safety related i equipment, that component will be seismically quallfled.

3. Power Supply - The DSS will be supplied from reliable power sources.

l j 4. Testing and Calibration - The DSS will be desig:.ed and installed to allow on-line calibration and testing.

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5. Annunciation and Display - System status indication will be provided in the control room and DSS trip alarm will be annunciated at the i main control board.

6. Failure Mode on Loss of Power (LOP) - The DSS will not trip in the event of LOP since the DSS circuitry will be designed such that tripping relays are energized to trip.

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7. Diversity from Reactor Protection System (RPS)- The DSS will share 4

the existing Pressurizer Pressure Sensors with the RPS (sensors include transmitter and signal conditioning circuitry). All other

, components such as the Bistables (B/S), Matrix circuitry, actuating relays and final actuation devices will be diverse from the RPS.

8. Location of Equipment - The Bistables and logic matrix circuitry will be located in the control room. i The trip hardware is located at the Motor Generator control cabinet in the DC switchgear room.

9. Reliability - The DSS will have a reliability comparable to the RPS.

Reliability calculations will be completed at a later date.

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10. Setpoints and Response time - The DSS will have a trip setpoint of approximately 2400 psia increasing. Actual setpoint will be determined later to coordinate the DSS tripping with the RPS.

i The response time will be determined by the response times of l sensors, actuating relays and MG output contactor dropout time.

l 11. Quality Assurance - As stated in our December 10,1984 le t ter,(7) we l believe that our existing quality practices now applied to non-safety-i related equipment would be adequate for non-safety-related

equipment encompassed by the ATWS rule, however, to meet the l

guidance contained in Generic Letter 85-06 the system will be procured, designed and installed under the Northeast Utilities Quality Assurance Program Topical Report, Revision 7.

i j B. Auxiliary Feedwater Actuation System (AFAS)

The existing Millstone Unit No. 2 AFAS will be modified to satisfy the ATWS requirement. The present circuitry includes a timer (3 minutes) which delays the initiation of AFW af ter detection of low steam generator

] (S.G.) level. This circuitry will be modified to have the means to i automatically bypass the timer in the event of an ATWS. It is diverse from i the RPS from the sensor output to the final actuation device. The AFAS is initiated by four channels of 5.G. Instrumentation arranged in a two out of four logic. The trip setpoint is 12% decreasing level. The system is a l Category lE system which was installed in 1981 in response to NUREG-1 0737 item i1.E.1.2.1b.

(7) W. G. Counsil letter to S. M. Goldberg, dated December 10,1984.

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C. Turbine Trip j

With the implementation of the DSS at Millstone Unit No. 2, the diverse f turbine trip requirement will also be satisfied. This is because the existing RPS initiated turbine trip is accomplished by undervoltage relays  !

7 monitoring the 240 VAC reactor trip bus voltage. While the turbine trip in .

j an ATWS event is accomplished by the same undervoltage relays, the '

initiation of .the undervoltage condition is diverse from the RPS. (See r figure 1). [

j 10 CFR 50.62 requires that the scheduled completion for Millstone Unit No. 2, to i meet the requirements of 10 CFR 50.62, be accomplished prior to restart from <

I the planned August,1986 refueling outage. In light of the lengthy delay in the [

issuance of Generic Letter 85-06, the long lead time required to adequately plan '

and schedule outage activities, the time and effort necessary to perform detailed j J

engineering and design of systems and the long lead times required in the i procurement of equipment, full compliance with the ATWS requirements for l Millstone Unit No. 2 can be achieved by the end of the first refueling outage 18 "

months af ter the Staf f's concurrence with our proposed approach.  ;

! Accordingly, pursuant to 10 CFR 50.56(d), justification for scheduling final l implementation later than the second refueling outage af ter July 26,1985 is '

provided for the above stated reasons. NNECO is prepared to be in full  !

compliance with the requirements of 10 CFR 50.62 for Millstone Unit No. 2 by r i the end of the refueling outage scheduled to start in January,1988, provided i NRC Staff concurrence with NNECO's approach for meeting the requirements of 10 CFR 50.62 is received by March 1,1986.

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t Haddam Neck Plant The Haddam Neck Plant is also participating in the Integrated Safety Assessment t Program. The NRC's July 31, 1985 letter to CYAPCO documenting the results i of the ISAP screening review process specifically identifies the requirements of i j 10 CFR 50.62 as ISAP Topic 1.16 for Haddam Neck. For the same reasons given ,

above for Millstone Unit No.1, we are unable to commit at this time to a firm i schedule for implementation of these modifications at Haddam Neck. Rather, a y decision and schedule for the proposed modifications will be arrived at through l the ISAP process.  !

, Evaluations performed to date for the Haddam Neck Plant have shown the plant's transient response to an ATWS event to be significantly less severe than the generic Westinghouse plant evaluated in the development of the final rule, i

A plant specific analysis of the limiting ATWS event for the Haddam Neck Plant  !

has previously been provided to the Staff.(8) This event was the loss of  ;

feedwater ATWS.  !

j i (8) D. C. Switzer letter to A. Giambusso, dated February 28,1975.

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Since that time, several plant modifications have been implemented that would further reduce 'the consequences of an ATWS event. For. example, new Power Operated Relief Valves (PORVs) have been installed, and they have been upgraded to safety grade. The new PORVs each have a relieving capacity of more than three times that assumed in the earlier analysis. When normalized to core power, the PORY relief capacity is significantly greater than that assumed for the generic Westinghouse plant, which would result in a substantial reduction in calculated peak pressure from the earlier analysis.

There are other additional factors which would result in less severe consequences for the Haddam Neck Plant than for the generic Westinghouse plant. Haddam Neck operates at a lower core power level, only 53% of the Westinghouse generic value. Also, Haddam Neck has a more favorable moderator temperature coefficient and operates with a lower nominal primary system pressure and average temperature. All of these factors result in a more mild plant response to an ATWS event. In fact, our analyses predict a peak RCS pressure in the range of 2800 to 2850 psi for an ATWS, which is well within original system and vessel hydrostatic test limits. Also, our earlier ATWS analysis predicted no clad damage and thus minimal offsite radiological consequences.

We believe that the above information provides adequate technical justification to conclude that ATWS is not a significant safety issue requiring prompt action for Haddam Neck, and that the absence of a firm schedule at this time is acceptable. As with Milistone Unit No.1, a decisi;n on modifications and a schedule for implementation will be determined duong the ISAP review for the Haddam Neck Plant.

We trust you will find the above information satisfactory, and look forward to receiving your concurrence with ciur approach.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY NORTHEAST NUCLEAR ENERGY COMPANY 3QlOpeka f Senior Vice President By: W. F. Fee

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