ML20203H101

Forwards Request for Addl Info Re Individual Plant Examination of External Events at Millstone Nuclear Power Station,Unit 2.Requests Response within 60 Days of Receipt of Ltr

ML20203H101
Person / Time
Site:	Millstone
Issue date:	02/24/1998
From:	Mcdonald D NRC (Affiliation Not Assigned)
To:	Bowling M NORTHEAST NUCLEAR ENERGY CO.
References
TAC-M83642, NUDOCS 9803030155
Download: ML20203H101 (8)

Text

. _ - . _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . -

. Mr. Martin L Bowling, Jr. February 24, 1998 Recovery Officer Unit No. 2 Northeast Nuclear Energy Company clo Ms. Patricia A. Loftus Director Regulatory Affairs P. O. Box 128 Waterford, CT 06385

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATING TO THE INDIVIDUAL PLANT EXAM' NATION OF EXTERNAL EVENTS (IPEEE) MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 (TAC NO. M83642)

Dear Mr. Bowling:

The NRC staff is continuing its review of Northeast Nuclear Energy Company's IPEEE submittal and has determined that additional information is needed in several areas. The information needed is identified in the enclosure to this letter.

We request that NNECO provide its response within 60 days from receipt of this request. If you l

have any questions relating to this request, please contact me at (301) 4151408.

Sincerely, Original algned by:

Daniel G. Mcdonald Jr., Senior Project Manager Special Projects Office . Licensing Office of Nuclear Reactor Regulation Docket No. 50 336

Enclosure:

Request for Additional Information .

cc w/ encl: See next page DISTRIBUTION:

Dochet Filef ~

"PUBLIC' SPO L Reading i SPO Reading i WTravers - 'I PMcKee LBorry DMcDonald OGC ACRS JDurr, RI - kO/

ll l lll 1,11,11 Ill DOCUMENT NAME: G:\ MCDONALD \M83642.RAI i . . . . . , .e ,w . i, . 6. . 6. . c . cory .n w,, .,,.%o,,.rio. . e . co,y on,....e ,c . ,4 n..oeio.u .on, OFFICE SPO L:PH;_ SPO-L:LA h 'y SP04h )

NAME DMcDona br 02/9y98 T U 02/@!98 0

QM LBerry W

PM$40 ' --

DATE 021%/98 02/ /98 02/ /98 0FFICIAL RECORD COPY 98o303o133 98o22436 e e 7 -- " 7m4T-0 g p y DR ADOCK 05 ,,,,,m, <mm

y f,...% k o UNITED STATES

. O q NUCLEAR REGULATORY COMMISSION

$ WASHINGTON, D.c. 30ee6-CJ01 o,

• s e + + f February 24, 1998 Mr. Martin L Bowling, Jr.

Recovery Omcor Unit No. 2 Northeast Nuclear Energy Company c/o Ms. Patricia A. Loftus Director Regulatory Affairs P. O. Box 128 l Waterford, CT 06385

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION RELATING TO THE INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVENTS (IPEEE) MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 O AC NO. M83642)

Dear Mr. Bowling:

The NRC staff is continuing its review of Northeast Nuclear Energy Company's IPEEE submittal and has determined that additionalinformation is needed in several areas. The information needed is identified in the enclosure to this ictter.

We request that NNECO provide its response within 60 days from receipt of this request. If you have any questions relating to this request, please contact me at (301) 415-1408.

Sincerely,

, v e-Daniel G. Mcdonald Jr., Senior Project Manager Special Projects Office Licensing Office of Nuclear Reactor Regulation Docket No. 50 336

Enclosure:

Request for Additional Information oc w/ encl: See next page

REQUEST FOR ADDITIONAL INFORMATION INDIVIDUAL PLANT EXAMINATION OF EXTERNAL EVEMTS (IPEEE)

MILLSTONE NUCLEAR POWER STATION. UNIT NO _2 DOCKET NO. 50 336 Selsmic

1. The development of the safe shutdown equipment list (SSEL) for the Millstone Unit 2 (MP2)IPEEE is discussed briefly in Section 3.2.2.4 of the submittal (Systems ?nd Equipment Information). However, because of the lack of information provideo in the IPEEE submittal, it cannot be determined whether the selection of the SSEL and the treatment of the associated issues (e.g., nonseismic failures and human actions)in the IPEEE are consistent with the process described in NUREG 1407.

(a) Please provide, as doscribed in EPRI NP 6041, two plant specific success path logic diagrams (SPLDs) . one for transients wherein the reactor coolant system pressure boundary is intact and the other for a seismically induced smallloss of coolant accident. Please discuss in some detail the systems and the operating modes ci these systems used in the SPLDs. Please include in the discussions any requirements on the recirculation modes of operation of the emergency core cooling system and the ways they are met at MP2 as well as nny requirements for feed-and-bleed cooling in the success pathr.

(b) Please address the nonseismic failures and human action issues as described in Section 3.2.5.8 of NUREG 1407. Please include in the discussion the needed huinan actions and the consideration givan to their failure probabilities in the selection of the r,uccess paths.

(c) Please provide th6 dependency matrix for the systems used in the success paths and include a description of the support systems listed in the dependency matrix.

(d) It is stated in Section 3.2.2.4 of the IPEEE submittal that *In a few cases, PRA (probabilistic risk assessment) modeled equipment was removed from the SSEL if it had a low seltr .ic capacity and negligible contribution to risk as determined by wious PRA importance measures assessed using the MP2 intemal events PRA nodel." Please discuss the equipment reatoved from the SSEL based on the above consideration. Pleese include in the discussion the justification for their removal, thtir seismic capacities (e.g., their high confidence, low probability of failure (HCLPF) values), and the effects of their removal on the success paths (e.g., their effects on i the ava!! amity of the frontline systems in the success paths).

(e) Please provide the actual SSEL (i.e., the equipment list) including the buildings and floor locations of the components in the SSEL.

2. Please provide in structure response spectra (IRS) for locations that have SSEL equipment attached to them.- Furthermore, screening guidance in Tables 2 3 and 2-4 of EPRI NP-6041 are primarily intended for components mounted fairly low (less than 40 feet above grade)in stiff structures. Please identify the SSEL components that are b

Enclosure I

?; )

2-mounted significantly h!aher than 40 fest above grade, and discuss how these components were screened or evaluated. The SSEL components that cannot be screened out during walkdowns were selected for an HCLPF capacity evaluation. Please discuss in more detail (e.g., prcviding a few representative sample calculations) the component HCLPF evaluations. Also, if a calcul.ated HCLPF is below the review level earthquake (RLE), discuss the specific steps taksSo resolve this issue and the effect of -

= the resolution on the plant's HCLPF capacity.

3. Please provide a list of the low mggedness relays which are being treated under the USl A-46 program. Please also identify and provide a list of the IPEEE low ruggedness relays

- that hva not been subsumed in the Unresolved Safety lasue A-46 program.

! 4 In the section describing the plant walkdowns, if the Generic implementation Procedure-

- (GlP) was utilized, provide a specific account of how the procedure was implemented at MP2 Please also provide a sample Seistnic Evaluation Work Sheet (SEWS). Please also discuss wheiher seismic induced floods have been addressed by the seismic review team walkdowns, and if not, justify their exclusiori.

5. In Section 3.2.5.2, the submittal states that "[An) HCLPF of 0.13g was determined for battery rock DB1, 'B2 will have about the same or slightly higher capacity, and was assigned a 0.13g wpacity." It further states that even if some bolt failures occuved, redistribution of loads would take place and otner supports would take up the redistributed loads and enable the racks / batteries to remain operable. First, please identify the "other supports" that would take up redistribution of the loads._ Second, please discuss how the loads rediswibute to other supports and quantify the redistributed.

' loads on the other supports. Last, please demonstrate that the other supports identified above possess sufficient seismic capacity to support the redistributed loads.

'Eka

1. Section 4.8.2.1 of the submittalindicates that the peak heat release rate for electrical

. ca'uinet fires in the auxiliary building is 53 BTU /s, extrapolated from the EPRI Fire PRA implementation Guide. The staff believes that the control cabinet heat release rates bast.d on the Sandia test results have been misinterpreted and inappropriately

. extrapolated in the EPR; Fire PRA Implementation Guide. The range of heat release rates reported by Sandia range from 23 to 1171 BTUls.

Considering the range of heat release rates that could be applicable to different control cabinet fires, and to ensure that cabinet fire areas are not prematurely screened out of the ana'ysis, a baat release rate in the mid range of the currently available axperimental data (e.g.,550 Btu /s) should be used for the anal'; sis.

Please provide a technical basis for the selection of 53 BTU /s as the peak' heat release rate end assess the impact on the IPEEE fire analysis results if the peak heat release -

rate from a cabinet fire is increased to 550 BTU /s.

2. The submittal states that transient combustibles and ignition sources have been considered only in the auxiliary building. In general, the fire risk associated with a given compartment is composed of contributions from fixed and transient ignition sources.

Neglect of either contribution can lead to an underestimate of the compartment's risk and, in some cases, to improper screening of fire scenarios. The EPRI Fire PRA

3 Implementation Guide allows the screening of transient ignition sources in compartments -

where all fixed ignition sources have been screened out, for example. Based on this approach, a cable spreading room or a cable shaft that does not contain any items other than IEEE 383 qualified control and instrumentation cables, and access to the compartment is strictly controlled, can be screened out, if such compartments contain the cables for all redundant trains of important plant safety systems, a major vulnerability

- may be overlooked, without sufficient analysis of potential accident sequences and needed recovery actions, in compartments where all fixed lenitions sources have been screened out, has the possibility of transient combustible fires been considered? For each compartment where transi6nt fires have not been considered, please provide the justification for this conclusion and provide a discussion on compartment inventory in terras of system trains i

and associated components (i.e, cables and other equipment). Please explain whether or not the conditional core damage probabilities, given damage to all cables and equipment in these compartments, are significant (i.e., cables from redundant trains are present), if '

the conditional core damage probability for a compartment is considered significant, please provide justification for assigning a very low likolanood of occurrence of transient fuel fires for the compartment.

3. Both fire induced damage and automatic suppression system activation times for several of the fire scenarios are so short as to be physically unrealistic. While it can be reasonably assumed that predicted short times to fire damage tre bounding, the combined effect of short times for both critical damage and automatic suppression system activation may not be bounding.

Please provide an assessment of the impact on fire induced core damage frequency if automatic suppression system activation times are uniformly increased to 5 minutes.

4. The analysis of Fire Scenario M2TBL (Special Cata Study), pages 4-84 to 4 86, indicates that the loss of the turbine building will result in a loss of safe shutdown equipment (i.e.,

AFW, SW, AC power) and lead to core damage. The analysis takes credit for automatic and manual supptession, as well as the likelihood of " catastrophic" fires. This approach appsers to take excessive credit for the effectiveness of suppression.

A 10 reduction factor is already used to account for the observation that 0 out of 39 turbine building fires were " catastrophic." Presumably, the effoctiveness of suppression efforts (manual and automatic) is already reflected in this observation.

In using a generic automatic suppression reliability estimate (0.05) for the probability that the automatic suppression system extinguishes the fire, the analysis is implicitly assuming that the conditional probability of extinguishment,' given actuation, is 1.0.

Since the Millstone 2 turbine building is only partially sprinklered, it is not clear that this assumption is valid.

Important dependencies between automatic and manual suppression (e.g., missile damage from a catastrophic turbine or main feedwater pump failure, fire protection water supply failure, difficulties in extinguishing very large fires) have apparently not been addressed.

It therefore appears that the analysis of s s 1ario could be rather optimistic. ,

C Please provide detailed information on the separation of all safe shutdown equipment and circuits located in the turbine building, the proximity of all safe shutdown equipment to significant fire hazards, and an euuation of the approved exemptions in this fire area.

For all fire sources postulated in the turbine building, please describe the likely ,

progression of fire growth and damage if it is assumed that fire suppression activities l (automatic and manual) fall to prevent fire spread. In this assessment, p aase consider and discuss potential barrier failures and the potential for fire to spread to the protected side of the barrier causing subsequent damage. For each of the unsuppressed fire scenarios, please provide an assessment of the conditional core damage probability (CCDP) given the postulated fire damage in particular, do any of these fire scenarios lead directly to coro damage (CCDP=1)?

^

5. Several fire areas and targats sets were apparently dropped during the analysis and not discussed in the submittal. Please provide either a justification for eliminating, or an estimate of the fire cumulative damage function (CDF) contribution for the following:

Areas surviving screening, but not discussed further; A-1A, A 1C, A-1H, A 9.

Target sets identified, but not evaluated: M2A18-3, M2A1G 2, M2A1G-3, M211 A 2, M211A 3, M211 A-4, M2TB-4, M2TB 5, AND M2TB-6.

The following fire areas were screened based on a comparison of fire-induced unavailabilities with intemal events unavailabilities, A-15, A-16, A-19B, A-30, A 31, T-3, T-4, T-8, T-9, and T-10 6.9kV and 4.16kV switchgear rooms.

This screening does not apparently consider the likelihood of an initiating event induced by the fire. Please provide either additionaljustification for eliminating these areas, or an estimate of thelt fire CDF contribution.

6. The attemate shutdown methodology utilized by the licensee requires that subsequent to the evacuation of the main control room, all sources of electrical power and instrument air are manually tripped by deenergizing all DC sources except the battery feeds to the Fire Shutdown Panel (FSP). This places the plant in a self-induced station blackout (SBO) condition (SISBO). The licensee's procedures indicate that pressure and temperature are controlled using the steam driven AFW pump, secondary side code safety valves, and the pressurizer power-operated relief valves. Following a trip of the diesel genentors and loss of all DC power, except the feed to the FSP, AC power is supplied via the Unit 1 cross tie Bus 14H. The licensee has determined that 4 hot.rs is required to establish the feed from Unit 1. This scenario has not been evaluated in the licensee's IPEEE submittal. A plant that implements an SISBO may experience spurious operation of equipment prior to the deenergizing of circuits. The SBO may create conditions that can uncover the core in about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Please provide an assessment of the attemate shutdown methodology considering the SISBO and also assess all equipment or systems credited following a fire in Millstone Unit 2 that are shared or common with either Millstone Unit 1 and/or Millstone Unit 3.

Specify the equipment or systems that are utilized, evaluate the fire vulnerability of these systems and assess the impact on the fire induced core damage frequency for Unit 2.

1 5-

7. ' Page 4-45 of the submittal states that the probability of an inadvertent operation of equipment, or the probability of a hot short was estimated to be 7.0E-2 based on

. NUREG/CR-2258. The value reported in tne referenced NUREG (which was published in 1981) is based on the judgement of the authors, not on experimental data. Fire tests of po. wor, coaxial, and multiconductor cables sponsored by the Depattment_of Energy and published in UCRL-ID 110598 dated September 1992, concluded that in some cases

-(more than the 7% stated in NUREG/CR 2258) spurious actuation of components could occur in less than 5 minutes due to direct shorts between cables, intermittent shorts between cables, high impedance shorts between cables and electromagnetic fluxes. The

!ikelihood of hot shorts (and other failure modes) depends on a variety of factors, including the design and function of the affected cables. Please identify thos;. scenarios p

' whose fire risk contributions are sensitive to the hot short probability, and provide a detailed basis for the assertion that the hot short probability is small,

8. Tha licensee's analysis for the main control room in Sction 4.8.2.3 assumes that a fire involving one or two cabinets would not require an evacuation of the control room. No t

' basis has been provided for this assumption. Full scale electrical cabinet fire test data published by Sandia in NUREG/CR- 4527 indicates that the smoke from a fine involving a r single cabinet can totally obscure the visibility throughout an enclosure in 6-15 minutes, it l would be reasonably expected that operations personnel would transfer control to the FSP prior to this point. Transfer of control to the FSP would also be expected to occur in

- the event of a fire in a single critical control panel such as main control board (MB)

Panels 2, 3, 5, 6, or 8.

Please revise the an6 lysis to consider the potential for control room evacuation as a result of a fire involving a single cabinet includo an evaluation of the potential for a fire involving er exposing the FSP, which was not included in the submittal.

9. Page 4-66 states that for Target Set M2 tbs (Transformer 15G-2Y) the fire ignition -

frequency was based on the pump fire ignition frequency. Please provide a basis for

. using a pump ignition frequency for an oil-filled transformer or revise the analysis, accordingly.-

10.

The submittal uses the failure probability vales for automatic detection and suppression systems that are provided in the FIVE methodology. These data are acceptable for systems that have been designed, installed, and maintained in accordance with appropriate industry standards, such as those published by the National Fire Protedion Association. Please verify that the systems credited in the analysis are in compilance with the appropriate industry standards, or identify any deviations and revise the failure probability to reflect the ac'ual anticipated system performance.

Seismie. Fire. and Hiah Winds. Floods. and Other Entarnal Evside (HFOs) 1.

Please provide the current status of the items in Table 7.1 1 of the submittalin the fire, seismic, and HFO areas that are Identified in the table as items to be resolved. For each item please statt whether:

(a) it was implemented and the date of implementation, or (b) it will be implemented and the planned date ofimplementation, or (c) It has been dropped from consideration and the reason why it was dropped.

MJ

O

.- Northeast Nuclear Energy Company Millstone Nuclear Power Station UM 2 i I

cc: 1 Lillian M. Cuoco, Esquire Mr. F. C. Rothen Stinior Nuclear 9 ansel Vice President - Work Services Northeast U'.ilities Service Company Northeast Utilities Service Company P. O. Box 270 P. O. Box 128 Hartford, CT 06141-0270 Waterford, CT 06385 Mr. John Buckingham Ernest C. Hadley, Esquire Department of Public Utility Control 1040 B Main Street Electric Unit P.O. Box 549 10 Liberty Square West Wareham, MA 02576 New Britain, CT 06051 Mr. D. M. Goebel Mr. Kevin T. A. McCarthy, Director Vice President - Nuclear Oversight Monitoring and Radiation Division Northeast Utilities Service Company Department of Environmental Protection P. O. Box 128 79 Elm Street Waterford, CT 06385 Hartford, CT 06106-5127 l Mr. Davd Amerine Regional Administrator, Region l Vice 'l resident - Nuclear Engineering U.S. Nuclear Regulatory Commission and Support 475 Allendale Road Northeast Utilities Service Company

, King of Prussia, PA 19406 P. O. Box 128 l

Waterford, CT 06385 First Selectmen Town of Waterford Mr. Atlan Johanson, Assistant Director Hall of Records Office of Policy and Management 200 Boston Post Road Policy Development and Planning Waterford, CT 06385 Division 450 Capitol Avenue - MS# 52ERN Mr. Wayne D. Lannir'g P. O. Box 341441 Deputy Director of Inspections Hartford, CT 06134-1441 Special Projects Office 475 Allendale Road Mr. M. H. Brothers King of Prussia, PA 19406-1415 Vice President - Operations Northeast Nuclear Energy Company Charles Brinkman, Manager P.O. Box 128 Washington Nuclear Operations Waterford, CT 06385 ASB Combustion Engineering 12300 Twinbrook Pkwy, Suite 330 Mr. J. A. Pr%

Rockville, MD 20852 Unit Director . ..mione Unit 2 Northeast Nuclear Energy Ccmpany Senior Resident inspector P.O. Box 128 Millstone Nuclear Power Station Wateiford, CT 06385 clo U.S. Nuclear Regulatory Commission P.O. Box 513 Niantic, CT 06357 m __

I

• Northeast Nuclear Energy Company Millstone Nuclear Power Station UnM2 cc:

Mr. B. D. Kenyon Chief Nucler Officer- Millstone No1heast Nuclear Energy Company P.O. Box 128 Waterford, CT 06385 Citizens Regulatory Commission ATTN: Ms. Susan Perry Luxton 180 Great Neck Road Wa;arford, CT 06385 Dcborah Katz, Pret! dent Citizens Awareness Network P. O. Box S3.

Shelbume Falls, MA 03170 The Honorable Terry Concannon Co-Chair .

Nuclear Energy Advisory Council Room 4035 Legislative Office Building Capitol Avenue Hartford, CT 06106 Mr. Evan W. Woollacott Co-Chair Nuclear Energy Advisory Council 128 Terry's Plain Road

. Simsbury, CT 06070 Little Harbor Consultants, Inc.

Millstone -ITPOP Project Office P. O. Box 0630 Niantic, CT 06357-0630 Mr. Daniel L. Curry _

Project Director Parsons Power Group !nc.

2675 Morgantown Road Reading, PA 19607 Mr. Don Schopfer Verification Team Manager Sargent & Lundy 55 E. Monroe Street Chicago,IL 60603

.