ML20137U485

Responds to Requesting Addl Questions Re

ML20137U485
Person / Time
Site:	Seabrook
Issue date:	04/10/1997
From:	Deagazio A NRC (Affiliation Not Assigned)
To:	Doughty J SEACOAST ANTI-POLLUTION LEAGUE
References
NUDOCS 9704170014
Download: ML20137U485 (20)

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j UNITED STATES NUCLEAR REGULATORY COMMISSION PLY l 2 WASHINGTON, D.C. mas mang k...../ April 10, 1997

Ms. Jane Doughty j The Seacoast Anti-Pollution League l P.O. Box 1136

! Portsmouth, NH 03802 l l

Dear Ms. Doughty:

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i This is in response to your letter of October 11, 1996, in which you asked several additional questions relating to my letter to you dated September 9, 1996. To avoid lengthy repetition, a number of your questions or comments I contained in your . letter have been paraphrased. They are shown below in J italics with our responses and discussion following.

Question How is it that you maintain the licensee met the ACTION requirements of \

Technical Specifications 3.11.2.1 and 3.3.3.10 and assured that the dose 1

[ rates were decreased to the specified limits within 15 minutes? Isn't l deliberately performing a containment purge, with no equipment in i i

operation that permits one to know what the dose limits are within 15 1 l minutes, flagrantly in violation of these Technical Specifications?

Discussion i

! -Referring to my September 9, 1996, letter, my response therein to your I i Question 1 explained the compensatory actions the licensee must take to j continue releasing radioactive gaseous effluents while the number of i operable radioactive gaseous effluent monitoring instrumentation ,

l channels are less than the minimum number required to be operable. In )

{ -your October 11, 1996, letter, your discussion leading to the above i a

question attempts to link the Limiting Conditions for Operation (LCO) )

and associated ACTION of Technical Specifications (TSs) 3.3.3.10 and l 3.11.2.1 in a manner such that if effluent monitoring instrumentation is '

inoperable the release of gaseous effluents monitored by that

! instrumentation must be suspended. This interpretation of these specifications is not correct. TS 3.3.3.10 identifies the operability i

a requirements, including setpoints, for the monitoring instrumentation j- and the actions to be taken if the instrumentation is inoperable. TS i

i 3.11.2.1 specifies dose limits for determining the setpoints for the instrumentation and the actions to be taken when the limits are g exceeded.

4 In the event dose rate limits of TS 3.11.2.1 are exceeded, the ACTION requires decreasing effluent release rates within 15 minutes to reduce i the dose rates to within the limits. The requirements of this ACTION are entirely consistent with continued effluent release with inoperable

monitoring instrumentation as permitted by TS 3.3.3.10 provided the grab i-samples .(for noble gas activity) are taken and analyzed and alternate j continuous sampling is performed (for iodine and particulates) as yOl i 7

NRC FRE CENTER COPY 9704170014 970410 PDR ADOCK 05000443 H PDR ,

i Ms. Jane Doughty  !- ,

i. specified in Table 3.3-13 ACTION 33 and 35. When Table 3.3-13 ACTION 33

and/or 35 apply and if the analysis results of grab samples or auxiliary i

continuous samples indicate the limits of TS 3.11.2.! are exceeded, then TS 3.11.2.1 ACTION must be followed. However,-if the analysis results  ;

do not indicate the dose limits are exceeded, then TS 3.11.2.1 ACTION 1 1

does not apply.

, Question.

i Doesn't ACTION 35, which applies to subparts b. and c. of the system .

require continuous collection of samples with auxiliary sampling \

[ equipment as required in the Offsite Dose Calculation Manual? Wasn't 1 the licensee in violation of the requirements of ACTION 32 and ACTION 35?

l l Answer  !

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! The response provided in my September 9, 1996, letter focused mostly j i upon the requirements for release of radioactive noble gaseous effluents i since the C-10 Radiological Monitoring System findings of November 29,

! 1995, which raised the issue of containment purging with plant vent wide

! range gas monitor (WRGM) inoperable, did not relate to the release of

iodine or particulate material. My response should not have been so

, limited.

You are correct that ACTION 32 and 35 also apply when the plant vent 8

wide range gas monitor is inoperable. For a period of time on 4 November 28 - 29, 1995, a power supply failure affected certain WRGM

. instruments that are listed on Table 3.3-13 of the Seabrook TS. The  !

i licensee declared the entire WRGM inoperable until maintenance was completed. During the period of time the WRGM was inoperable, the licensee complied with all applicabl6 ACTIONS listed in Table 3.3-13.

Question t

In light of the C-10 findings, which indicated that things night not be as "normally" expected, wasn't the loss of the TLD at 243 degrees 1.2 miles from the plant cause for concern by the NRC, and shouldn't the NRC l

require the licensees or the state to provide a regular documented inventory that the TLDs are in place with no evidence of tampering?

) l i Discussion

{ As discussed in NRC Inspection Report 50-443/96-05, dated June 25, 1996,

! NRC concluded that Seabrook Station activities were not the cause of the  :

. instrument indications observed previously by C-10's radiation monitors.

As we previously stated in our September 9,1996 response, the loss of the NRC thermoluminescent dosimeter (TLD) at station 5 did not warrant

an investigation and the consequent loss of TLD measurement did not have

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, any significant impact on the quality of the overall data, or its use and application for radiological assessment. ,

1 In addition to the NRC's TLD network, the licensee and the State of hew Hampshire maintain their own independent TLD networks around the site.

The State of New Hampshire's TLDs are installed at different locations than the licensee's, but some may be colocated with NRC devices. As needed, these independent TLD networks would serve to augment the data base in the case of an emergent or off-normal condition at Seabrook Station. Relative to interpretation of the TLD measurements as reported in NUREG-0837 report for the period in question, we found nothing significant relative to any of the measurements that was out of the range of normally expected values.

A regular documented inventory is conducted for the NRC by the State of  !

New Hampshire on a quarterly basis while exchanging the TLDs. Upon  !

being sent to the NRC for processing, we consistently document )

discrepancies, such as missing dosimeters. Given the purpose and  ;

objective of the TLD network program, we are confident that our  ;

processes and controls are sufficient.

Question Given the licensee's laival error wherein the Ifcensed thermal power was exceeded and the 1cuy time span involved in the second incident, how is it that the NRC has confidence that the licensee correctly determined i that 3418 Mwt and 3413 Mwt were the highest thermal power levels reached

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at any point in time during these two incidents?

Discussion The first occurrence of overpower resulted from deliberate reactor operator actions to restore the reactor to full power following an indicated power decrease. The operator action was taken without a full understanding of the cause for the indicated power decrease. It was not until the next shift questioned the earlier power adjustment that it was determined that an instrument transmitter calibration had invalidated the normalizing constants used in the calorimetric, and that the previous shift's action to adjust the reactor to full power actually caused the power to exceed the licensed limit.

The second event occurred when the main plant computer was restarted following maintenance. The operators failed to recognize that upon restart, the main plant computer defaulted the calorimetric to the same mode that was invc1Wd in the earlier overpower event. At that time, l the evaluation of the first event was not yet complete, and the incorrect normalizing constants had not yet been updated.

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Ms. Jane Doughty The detailed evaluations which afforded a full understanding of both events, including the power levels reached during the events, were not fully completed until 4 days after the second event. The NRC was involved in considerable discussions with involved personnel starting the day after the first event, and an NRC inspector reviewed the completed evaluations.

Question Did Yankee Atomic Electric Company do t: calculations providing the basis for Seabrook Statfon's 100% rating being 3411 Mat?

Discussion The reactor design for the Seabrook Station is discussed 11 Chapter 4 of the Seabrook Station Updated Final Safety Analysis Report (UFSAR).

Specifically, the thermal and hydraulic design is discussed in Section 4.4 of the UFSAR. The basic thermal and hydraulic design providing the basis for Seabrook's 100% rating was performed by the Westinghouse Electric Corporation.

In addition to the basic thermal and hydraulic design of the reactor, other analyses are required to demonstrate the adequacy and safety of the design of the reactor and other systems during transient and accident conditions. The large-break and small-break loss-of-coolant accident (LOCA) an@es as described in Chapter 15 of the UFSAR have been performed by Westinghouse Electric. The non-LOCA transient and accident analyses described in Chapter 15 of the UFSAR have been performed by Yankee Atomic Electric Company. Analyses relating to containment performance as described in Chapter 6 of the UFSAR have been performed by Raytheon Engineers and Constructors.

UFSAR Section 4.4.7 identifies the various reference documents which, in part, identify the methodology used to establish the reactor thermal and hydraulic design parameters. Additionally, Section 6.8.1.6.b, Appendix A of the Seabrook Station Technical Specifications identifies the approved analytical methods to be used in the establishment of core operating limits.

On November 23, 1994, the NRC issued Amendment 33 to the Seabrook ,

operating license. This amendment authorized operation of Seabrook with I an expanded axial flux difference band and authorized certain fuel i design enhancements. The Yankee Atomic Electric Company provided the nalytical methodology and performed the calculations to support North Atlantic's license amendment application, except for the LOCA analyses provided by Westinghouse.

The Seabrook UFSAR, Technical Specificatiens, and Amendment 33 are available at the Local Public Document Room located at the Exeter Public Library, Founders Park, Exeter, NH. For your convenience, I have

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enclosed the reference list to UFSAR, Revision 4, Section 4.4 and i Section 6.8.1.6.b (as revised through Amendment 33) of the Seabrook Technical Specifications. References bearing the designation YAEC-i followed by a sequence of numbers and letters identify documents

[ prepared by the Yankee Atomic Electric Company.

] Questton Has the NRC verffled that the sampling and analysis required by

\' Technical Specifications 3/4.4.8 and Table 4.4-3 were carried out by the licensee, and has the NRC independently reviewed the findings?

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Discussion TS 3/4.4.8 and Table 4.4.3 require, in part, periodic sampling and analysis of the reactor coolant for gross radioactivity at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and for radiciodines at least once every 14 days. More

, frequent sampling and analyses are required if the specified limits are a

exceeded. The results from the analyses of routine samples of reactor j coolant, taken in conformance with TS 3/4.4.8 after the overpower events of October 19 and 26,1995, did not exceed.the specified limits and do 1-not indicate that any fuel clad damage occurred. The licensee includes a report of the most recent reactor coolant specific activity determinations at the daily Station Director's Meeting which normally is attended by NRC personnel at the site. Additionally, the NRC resident

-inspectors review the results on a routine basis.

Questions.

Who did the visual examinations of the eight fuel assemblies? How was the sample size of eight assenbiles selected, and were the assemblies randomly selected? Does the NRC consider eight fuel assemblies an adequate sample size in these circumstances? Would visual examination detect pinhcle leaks in fuel cladding?

Discussion Updated Final Safety Analysis Report Subsection 4.2.4.6.g. states, in i part, the following: l

" Visual irradiated fuel inspections will be conducted as necessary during each refueling. Selected fuel assemblies may be inspected for fuel rod failure, structural integrity, crud deposition, rod  !

bow and other irregularities. Fuel.  !

assemblies will be selected for inspection l

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based upon performance history and  !

recommendations made by the fuel supplier.

j "The fuel inspection program will be )

i expanded to include more fuel assemblies 1 j or greater detail of examination if high

] coolant activity is experienced during i

operation, irregularities are noted in fuel performance, irregularities are noted during routine inspections, or if a new

fuel design is incorporated."

l I indicated in my September 9, 1996, letter, the amounts of overpower in j- both events were not significant from the standpoint of inducing fuel a

cladding failure. I indicated, further, that extensive examination for fuel clad damage was unnecessary absent an increase in the gross I i radioactivity of the reactor coolant which is an excellent indicator of i

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fuel clad integrity. Considering the at,ove along with the overall excellent performance history of the nuclear steam supply system for the i cycle that had just ended, and the lack of any other contraindications, 4

! the size of the sample selected by the licensee was not unreasonable. i

! The licensee selected the assemblies to be examined based upon the  !

position in the core and the assembly irradiation history. Licensee l personnel performed the visual examinations.

The visual examinations are intended to reveal any physical manifestations of unexpected or unusual conditions. The detection of l 1 pinholes in fuel cladding would be most' difficult, if possible at all. l l Questions i  !

f SAPL asserts (with regard to the September 9,1996, response to SAPL  :

l Question 5 relating to the radiological conditions encountered during

the reactor head funnel guide inspection) "that the exposures tuLrg  ;

1 representative of actual radiological conditions" Why were exposures '

t higher than expected? Why does the NRC permit operation of the plant

.i when this inspection is incomplete?

l Discussion I The basis for SAPL's assertion is not provided. If SAPL has any l information that supports this assertion, it should be provided.

1 As stated in the September 9, 1996, response, the licensee's estimate of the radiological conditions to be encountered during the reactor head 4 funnel guide inspection was based on dose measurements under the reactor vessel. When the funnel guide inspection was being planned, the reactor head was in place on the reactor vessel and the reactor was operating, i

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so the planners relied upon dose rate measurements made during a previous outage in the space under the reactor. The planners  :

assumed that the exposure conditions under the reactor would be '

representative of the conditions workers would encounter during the inspection. Actual dose rates encountered were much higher.

As was previously explained, the licensee recognized that the entire inspection would involve considerably more personnel exposure than originally expected, so further inspection was suspended.

NRC Information Notice 94-40 (May 26, 1994) and Information Notice 94- l 40, Supplement 1 (December 15,1994) were issued to alert licensees of l several events involving detached guide funnels underneath the reactor head. NRC Information Notices, in part, are for the purpose of alerting licensees of conditions that have been discovered at other facilities.

Information Notices do not impose NRC requirements, and no specific action or written response from licensees are required. Thus, there is no restriction for continued operation of Seabrook because the licensee has not completed inspection of the reactor head guide funnels. There are no technical specification requirements applicable to this inspection.

Question Does the NRC believe the number of samples for entrained noble gas, as identified in the response to Question 6, represents an adequate program, and does this form a solid brsis for concluding that the measurements reported by C-10 can not be attributed to the release of noble gases from Seabrook?

I Discussion The conclusion reached in Inspection Report 50-443/96-05 that operating activities at Seabrook could not have caused the indications observed by the C-10 monitoring stations is based upcn much more information than merely the results of the liquid samples taken and analyzed for entrained or dissolved noble gases. The conclusion, instead, is supported by information obtained from a number of diverse sources including in-plant area radiation measurements, reactor coolant sampling for gross radioactivity, in-plant process (gaseous and liquid) radiation

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measurements, personnel dosimetry, operational logs, and off-site  ;

environmental monitoring. No data was uncovered from these l sources for the relevant period that is in conflict with the conclusion of the inspection report.

Sincerely, Original signed by Albert W. De Agazio, Sr. Project Manager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Docket No. 50-443 Distribution l Docket File ACRS  !

Enclosures:

As stated PUBLIC JRogge, RI '

PDI-3 RF cc w/encls: See next page SVarga )

PMilano l OGC l DOCUMENT NAME: G:\DEAGAZIO\SAPL10ll.96 To etcxive e copy of this document, indicate in the box: "C" = Copy without ettschment/ enclosure *E' = Copy with attachment / enclosure *N* = No copy 0FFICE PM:PDI-3/7hM/ LA:PDI-1,0l (A)D:PC1 4 l l l NAME ADeAgazio"' Slittle F PMil anc /V" ' ~

DATE 04/9/97 04/9/97 04 AD /97 OFFICIAL RECORD COPY i

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This Enclosure contains the references to the Seabrook Station, Unit No. 1 UFSAR, Revision 4, Section 4.4.7. I 1

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ENCL 0SURE 1

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_ 4.4.7 References  !

i 1. Christensen, J. A., Allio, R. J. and Biancheria, A., " Melting Point of i Irradiated U0 ," 2 WCAP-6065, February 1965.

2. Hellman, J. M. (Ed.), " Fuel Densification Experimental Results and Model for i j

Reactor Application," WCAP-8218-P-A (Proprietary), March 1975 and WCAP-8219-

' A, March 1975. 1 L 3. Tong, L. S., " Boiling Crisis and Critical Heat Flux," AEC Critical Review j Series, TID-25887, 1972.

j 4. " Evaluation of Westinghouse Request for Generic Approval", ENCLOSURE to i

• letter from Carl Berlinger (NRC) to E. P. Rahe, Jr.,-(Westinghouse), " Request i for Reduction in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty", dated June 18, 1986.

$ 5. Letter, L. A. Tremblay (Vermont Yankee Nuclear Power Corporation) to USNRC, "FROSSTEY-2 Fuel Performance Code - Vermont-Yankee Response to Remaining Concerns", BVY 92-54, May 15, 1992.

6. Letter, P. Sears (USNRC) to L. A. Tremblay (Vermont Yankee Nuclear Power Corporation), " Vermont Yankee Nuclear Power Station - Safety Evaluation of  :

FROSSTEY-2 Computer Code (TAC No. M68216)", September 24, 1992.

7. WCAP-8762-P-A, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles With Mixing Vane Grids", Westinghouse Electric Corporation, F. E. Motley, July 1984.

8. Tong, L. S., " Prediction of Departure from Nucleate Boiling for an Axially Non-Uniform Heat Flux Distribution," J. Nucl. Enerav, 21, 241-248 (1967).

9.- YAEC-1849P, " Thermal-Hydraulic Analysis Methodology Using VIPRE-01 for PWR Applications", Yankee Atomic Electric Company, F. L. Carpenito, October 1992.  ;

SEABROOK UPDATED FSAR REVISION 4

10. Letter from T. C. Feigenbaum (North Atlantic Energy Service Co.) to USNRC,

" Response to Request for Additional Information (TAC M86957 and TAC M86958)",

March 9, 1994. .

11. NP-2511-CCM Volumes 1-5, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores", Electric Power Research Institute.

12. Cadek, F. F., Motley, F. E. and Dominicis, D. P., "Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid," WCAP-7941-P-A (Proprietary), January 1975 and WCAP-7959-A, January 1975.

13. Rowe, D. S.,- Angle, C. W., " Crossflow Mixing Between Parallel Flow Channels During Boiling, Part II Measurements of Flow and Enthalpy in Two Parallel Channels," BNWL-371, part 2, December 1967.

14. Rowe, D. S., Angle, C. W., " Crossflow Mixing Between Parallel Flow Channels i During Boiling, Part III Effect of Spacers on Mixing Between Two Channels,"

l BNWL-371, part 3, January 1969.

! 15. Gonzalez-Santalo, J. M. and Griffith, P., "Two-Phase Flow Mixing in Rod

Bundle Subchannels," ASME Paper 72-WA/NE-19.

16. Motley, F. E., Wenzel A. H., Cadek, F. F., "The Effect of 17x17 Fuel Assembly

Geometry on Interchannel Thermal Mixing," WCAP-8298-P-A (Proprietary),

January 1975 and WCAP-8299-A, January 1975.

, 17. Cadek, F. F., "Interchannel Thermal Mixing Vane Grids," WCAP-7667-P-A i (Proprietary), January 1975 and WCAP-7755-A, January 1975.

l 18. Hochreiter, L. F., " Application of the THINC-IV Program to PWR Design," WCAP-8054 (Proprietary), October 1973, and WCAP-8195, October 1973.

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! 19. Hochreiter, L. E., Chelemer, H. and Chu, P. T., "THINC-IV An Improved i Program for Thermal-Hydraulic Analysis of Rod Bundle Cores," WCAP-7956, June l 1973.

20. Dittus, F. W., and Boelter, L. M. K., " Heat Transfer in Automobile Radiators of the Tubular Type," Calif. Univ. Publication In Ena., 2, No.13, 443461 (1930).

. 21. Weisman, J., " Heat Transfer to Water Flowing Parallel to Tube Bundles," Nucl.

Sci. Ena., 6, 78-79 (1959).

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SEABROOK UPDATED FSAR REVISION 4 I 22. Thom, J. R. S., Walker, W. M., Fallon, T. A. and Reising, G. F. S., " Boiling in Sub-Cooled Water During Flow up Heated Tubes or Annuli," Prc. Instn. Mech.

j Enars., 180, Pt. C, 226-46 (1955-66).

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23. Hetsroni, G., " Hydraulic Tests of the San Onofre Reactor Model," WCAP-3269-8, June 1964.

! 24. Hetsroni, G., " Studies of the Connecticut-Yankee Hydraulic Model,"

] NY0-3250-2, June 1965.

25. Idel'chik, I. E., " Handbook of Hydraulic Resistance," AEC-TR-6630, 1960.

) 26. Moody, L. F., " Friction Factors for Pipe Flow," Transaction of the American 1

4 Society of Mechanical Enaineers, 66, 671-684 (1944).

27. Maurer, G. W., "A Method of Predicting Steady State Boiling Vapor Fractions 1 in Reactor Coolan. Channels," WAPD-BT-19, pp. 59-70, June 1960.

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28. Griffith, P., Clark, J. A. and Rohsenow W. M., " Void Volumes in Subcooled j Boiling Systems," ASME Paper No. 58-HT-19.

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29. Bowring, R.=W., " Physical Model, Based on Bubble Detachment, and Calculation j'

of Steam Voidage in the Subcooled Region of a Heated Channel," HPR-10, December 1962.

30. Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee), " Acceptance for Referencing of Licensing ' Topical Report, EPRI NP-2511-CCM, 'VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1, 2, 3, and 4'",

Attachment:

" Safety Evaluation Report on EPRI NP-2511-CCM VIPRE-01", May, 1986.

31.. Letter from A. W. De Agazio (USNRC) to T. C. Feigenbaum (NAESCO), " Acceptance for Referencing of YAEC-1849P, ' Thermal-Hydraulic Analysis Methodology using VIPRE-01 for PWR Applications', for the Seabrook Station, Unit No.1 (TAC M86958)", August 15, 1994.

32. Duncan, R. N., " Rabbit Capsule Irradiation of U0 ,"g CVTR Project, CVNA-142, June 1962.

33. Gyllander, J. A., "In-Pile Determination of the Thermal Conductivity of UO z in the Range 500-2500 C," AE-411, January 1971.

1 SEABROOK UPDATED FSAR REVISION 4  !

34.- " Partial Response to Request Number 1 for Additional Information on WCAP-8691, Rev.1," letter from E. P. Rahe, Jr., (Westinghouse), to J. R. Miller l l- , (NRC), NS-EPR-2515, dated October 9, 1981; " Remaining Response to Request l

Number 1" letter, from E. P. Rahe, Jr., (Westinghouse), to J. R. Miller J (NRC), NS-EPR-2572, dated March 16, 1982.

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35. . Stora, J. P., "In-Reactor Measurements of the Integrated Thermal Conductivity of U02 - Effect of Porosity," Trans. ANS, 13, 137-138 (1970).

36. International Atomic Energy Agency, " Thermal Conductivity of Uranium Dioxide," Report of the Panel held in Vienna, April 1965, IAEA Technical i

Reports Series, No. 59, Vienna, The Agency, 1966.

! 37. Carter, F. D., " Inlet Orificing of Open PWR Cores," WCAP-9004, January 1969 1

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(Proprietary) and WCAP-7836, January 1972 (Nonproprietary).

38. Novendstern, E. H. and Sandberg, R. O., " Single Phase Local Boiling and Bulk l

' Boiling Pressure Drop Correlations," WCAP-2850 (Proprietary), April 1966 and WCAP-7916, June 1972.

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[ 39. Owens, Jr., W. L., "Two-Phase Pressure Gradient," International Developments in Heat Transfer, Part II, pp. 363-368, ASME, New York, 1961. j References 40 through 68 are not used. l

69. J. A. Boure, A. E. Bergies, and L. S. Tong, " Review of Two-Phase Flow Instability," Nucl. Engr. Desi f 25 (1973) p. 165-192. l

70. R. T. Lahey and F. J. Moody, "The Thermal Hydraulics of a Boiling Water '

Reactor," American Nuclear Society, 1977.

71. P. Saha, M. Ishii, and N. Zuber, "An Experimental Investigation of the Thermally Induced Flow Oscillations in Two-Phase Systems," J. of Heat Transfer, Nov. 1976, pp. 616-662.

72. Virgil C. Summer FSAR, Docket #50-395.

73. Byron /Braidwood FSAR, Docket #50-456.

74. South Texas FSAR, Docket #50-498.

d SEABROOK UPDATED FSAR REVISION 4

75. S. Kakac, T. N. Veziroglu, K. Akyuzlu, O. Berkol, " Sustained and Transient Boiling Flow Instabilities in a Cross-Connected Four- Parallel-Channel Upflow 1: System," Proc. of 5th International Heat Transfer Conference, Tokyo, Sept. 3-7, 1974.

76. H. S. Kao, C. D. Morgan, and W. B. Parker, " Prediction of Flow Oscillation in Reactor Core Channel," Trans. ANS, Vol. 16, 1973, pp. 212-213.

77. Ohtsubo A., and Uruwashi, S., " Stagnant Fluid due to Local Flow Blockage," L Nucl . Sci . Technol ., 9, No. 7, 433-434, (1972).

78. Basmer, P., Kirsh, D. and Senultheiss, G. F., " Investigation of the Flow Pattern in the Recirculation Zone Downstream of Local Coolant Blockages in Pin Bundles," Atomwirtschaft, 17, No. 8, 416-417, (1972). (In German).

79. Guimond, P. J., " Core Thermal Limit Protection Function Setpoint Methodology for Seabrook Station", YAEC-1854P, Yankee Atomic Electric Company, October 1992.

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This Enclosure contains Section 6.8.1.6.b of the Seabrook Station, Unit No.1  !

Technical Specifications as issued with Amendment 33 dated November 23, 1994.

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i ENCLOSURE 2 1

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I ADMINISTRATIVE CONTROLS 1 6.8.1.6.a. (Continued)

5. Shutdown Rod Insertion limit for Specification 3.1.3.5.

6. Control Rod Bank Insertion limits for Specification 3.1.3.6.

7. AXIAL FLUX DIFFERENCE limits for Specification 3.2.1. ,

8. Heat Flux Hot Channel Factor. F"'l and K(Z) for Specification 3.2.2.

9. Nuclear Enthalpy Rise Hot Channel Factor, and F"'l, for Specification 3.2.3. ,

The CORE OPERATING LIMITS REPORT shall be maintained available in the Control i Room.

6.8.1.6.b The analytical methods used to determine the core operating limits

, shall be those previously reviewed and approved by the NRC in:

, 1. WCAP-10266-P-A. Rev. 2 with Addenda (Proarietary) and WCAP-11524-A (Nonproprietary), "The 1981 Version of tie Westinghouse Er.CS Evaluation j Model Using the BASH Code". August. 1986 Methodology for Specification:

3.2.2 -

Heat Flux Hot Channel Factor

2. WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonproprietary), "NOTRUMP:

A Nodal Transient Small Break and General Network Code". August, 1985 Methodology for Specification:

3.2.2 -

Heat Flux Hot Channel Factor

3. YAEC-1363-A. "CASMC-3G Validation." April 1988.

YAEC-1659-A " SIMULATE-3 Validation and Verification." September 1988.

Methodology for S3ecifications:

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SHJTDOWN MARGIN for MODES 1. 2. 3. and 4 3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 3.1.1.3 -

Moderator Temperature Coefficient 3.1.3.5 -

Shutdown Rod Insertion Limit 3.1.3.6 - Control Rod Insertion Limits 3.2.1 -

AXIAL FLUX DIFFERENCE 3.2.2 -

Heat Flux Hot Channel Factor 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor

4. Seabrook Station Updated Final Safety Analysis Re) ort. Section 15.4.6.

" Chemical and Volume Control System Malfunction Tlat Results in a Decrease in the Boron Concentration in the Reactor Coolant System".

Methodology for S)ecifications:

3.1.1.1 -

SHJTDOWN MARGIN for MODES 1. 2. 3. and 4 3.1.1.2 -

SHUTDOWN MARGIN for MODE 5 SEABROOK - UNIT 1 6-18A Amendment No. 33

ADMINISTRATIVE CONTROLS 6.8.1.6.b. (Continued)

5. YAEC-1241. " Thermal-Hydraulic Analysis of PWR Fuel Elements Using the CHIC-KIN Code", R. E. Helfrich, March 1981 Methodology for Specification:

3.2.1 -

AXIAL FLUX DIFFERENCE 3.2.2 -

Heat Flux Hot Channel Factor 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor

6. YAEC-1849P, " Thermal-Hydraulic Analysis Methodology using VIPRE-01 For PWR Applications, " October 1992 Methodology for Specification:

2.2.1 -

Limiting Safety System Settings 3.2.1 -

AXIAL FLUX DIFFERENCE 3.2.2 -

Heat Flux Hot Channel Factor 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor

7. YAEC-1854P. " Core Thermal Limit Protection Function Setpoint Methodology l For Seabrook Station, " October 1992

{

Methodology for Specification: )

2.2.1 -

Limiting Safety System Settings l 3.1.3.5 -

Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits  !

3.2.1 -

AXIAL FLUX DIFFERENCE l 3.2.2 -

Heat Flux Hot Channel Factor l 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor 8 . YAEC-1856P, " System Transient Analysis Methodology Using RETRAN for PWR Applications." December 1992 Methodology for Specification: .

2.2.1 -

Limiting Safety System Settings 3.1.1.3 -

Moderator Temperature Coefficient 3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits 3.2.1 -

AXIAL FLUX DIFFERENCE 3.2.2 -

Heat Flux Hot Channel Factor 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor

9. YAEC-1752. " STAR Methodology Application for PWRs. Control Rod Ejection. l Main Steam Line Break," October 1990- - -

l Methodology for Specification:

3.1.1.3 - Moderator Temperature Coefficient 3.1.3.5 - Shutdown Rod Insertion Limit 3.1.3.6 -

Control Rod Insertion Limits 3.2.1 -

AXIAL FLUX DIFFERENCE

~3.2.2 -

Heat Flux Hot Channel Factor  !

3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor I SEABROOK - UNIT 1 6-18B Amendment No. 33

- . . -. .- - - - _ - __- __- - - . - . - _ _ -_...- _. _ =- -

1 .

!- ADMINISTRATIVE CONTROLS ,

6.8.1.6.b. (Continued)

10. YAEC-1855P, "Seabrook Station Unit 1 Fixed Incore Detector System Analysis," October 1992

Methodology for Specification:

3.2.1 -

AXIAL FLUX DIFFERENCE

3.2.2 -

Heat Flux Hot Channel Factor

3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor

11. (AEC-1624P, " Maine Yankee RPS Setpoint Methodology Using Statistical Combination of Uncertainties - Volume 1 - Prevention of Fuel Centerline Melt," March 1988 Methodology for Specification:

3.2.1 -

AXIAL FLUX DIFFERENCC 3.2.2 -

Heat Flux Hot Channel Factor 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor 6.8.1.6.c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle.. including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to .

the Regional Administrator and the Resident Inspector. l SEABROOK - UNIT 1 6-18C Amendment No. 33

_. ._. _ _ _ _ _ . _ . _ . _ _ _ . . . _ _ _ . . _ _ _ _._.___ _ .~ _ __ _ _

~

.~  !

.- i

! Northeast Utilities Service Company Seabrook Station, Unit No. 1 (

I-j CC 1

4 Lillian M. Cuoco, Esq. Mr. Dan McElhinney . l i Senior Nuclear Counsel Federal Emergency Management Agency 4 Northeast Utilities Service Company Region I l P.O. Box 270 J.W. McCormack P.O. &

Hartford, CT 06141-0270 Courthouse Building, Room 401 Boston, MA 02109 Mr. Peter Brann Assistant Attorney General Mr. Peter LaPorte, Director State House, Station #6 ATTN: James Muckerheide Augusta, ME 04333 Massachusetts Emergency Management Agency Resident Inspector 400 Worcester Road U.S. Nuclear Regulatory Commission P.O. Box.1496 Seabrook Nuclear Power Station Framingham, MA 01701-0317 P.O. Box 1149 Seabrook, NH 03874 Jeffrey lloward, Attorney General G. Dana Bisbee, Deputy. Attorney Jane Spector General Federal Energy Regulatory Commission 33 Capitol Street 825 North Capital Street, N.E. Concord, NH 03301 i Room 8105 Washington, DC 20426 Mr. D. M. Goebel I Vice President-Nuclear Oversight; ,

Town of Exeter Northeast Utilities Service Company '

10 Front Street P. O. Box 270 Exeter, NH 03823 Hartford, CT 06141-0270 ,

Mr. George L. Iverson, Director ~Mr. J. K. Thayer New Hampshire Office of Emergency Recovery Officer, Nuclear Engineering Management and Support State Office Park South Northeast Utilities Service Company  ;

107 Pleasant Street P.O. Box 128 Concord, NH 03301 Waterford, CT 06385 Regional Administrator, Region I Mr. F. C. Rothen U.S. Nuclear Regulatory Commission Vice President - Nuclear Work Services 475 Allendale Road Northeast Utilities Service Company King of Prussia, PA 19406 P.O. Box 128 Wr.terford, CT 05385 Office of the Attorney General One Ashburton Place 20th Floor Mr. A. M. Callendrello Boston, MA 02108 Licensing Manager - Seabrook Station North Atlantic Energy Service Corp.

Board of Selectmen P.O. Box 300

-Town of Amesbury Seabrook, NH -03874 Town Hall Amesbury, MA 01913 iS

_.- - - , , . - , y- ,

t Northeast Utilities Service Company Seabrook Station, Unit No. I cc:

Mr. Peter S. Paiton, Supervisor Mr. W. A. DiProfio Emergency Response Section Nuclear Unit Director Radiological Health Bureau Seabrook Station Division of Public Health Services

' North Atlantic Energy Service Corporation Health and Welfare Building P.O. Box 300 6 Hazen. Drive Seabrook, NH 03874 Concord, NH 03301-6527 Mr. Frank W. Getman, Jr. Mr. M. Nawoj Cocheco Falls Millworks Chief, Technical Hazards 100 Main Street, Suite 201 State Civil Defense Agency Dover, NH 03820 State Office Park South 107 Pleasant Street Mr. B. D. Kenyon Concord, NH 03001-6527 President - Nuclear Group Northeast Utilities Service Group 1 P.O. Box 128 Waterford, CT 06385 l Mr. B. L. Drawbridge Executive Director Services &

Senior Site Officer  !

North Atlantic Energy Service Corp. j Seabrook, NH 03874 Mr. R. Hallisey - Director Department of Public Health Commonwealth of Massachusetts Radiation Protection Program 305 South Street - 7th Floor Jamaica Plain, MA 02130 Commonwealth of Massachusetts SLO Designee A. David Rodham - Director

. Attn: Mr. James B. Muckerheide State Nuclear Engineer Massachusetts Emergency Mgmt. Agency 400 Worcester Road Box 1496 Framingham, MA 01701-0314 Diane Tefft - Administrator Bureau of Radiological Health Division of Public Health Services 6 Hazen Drive Concord, NH 03301-6527 Mr. Thomas O'Connell Radiation Scientist.

Radiation Control Program Department of Public Health 305 South Street, 7th Floor Concord, NH 03301-6527