ML20236A339

Affidavit of DG Bridenbaugh & Sc Sholly.* Discusses Spent Fuel Pool Expansion.Supporting Documentation Encl

ML20236A339
Person / Time
Site:	Vermont Yankee
Issue date:	03/01/1989
From:	Bridenbaugh D, Sholly S MHB TECHNICAL ASSOCIATES, NEW ENGLAND COALITION ON NUCLEAR POLLUTION
To:
Shared Package
ML20236A334	List: ML20236A330 ML20236A336 ML20236A339
References
OLA, NUDOCS 8903170095
Download: ML20236A339 (196)

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UNITED STATES 0F AME_RICA . .. NUCLEAR REGULATORY COMMISSION ~ @ ggg _7. P5 '26

                                           .Before the Atomic Safety and Licensing Board                       gg              .

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                                                                                      )-

In the Matter of )'

                                                                                      )

VERMONT YANKEE NUCLEAR ) Docket No. 50-271-OLA

       . POWER CORPORATION                                                            ) (Spent Fuel Pool Amendment)
                                                                                      )

(Vermont Yankee Nuclear ) Power Station). ) - March 1,1989

                                                                                      )

AFFIDAVIT OF DALE G. BRIDENBAUGH AND STEVEN C. SHOLLY A. BACKGROUND

1. . Dale G. Bridenbaugh is a principal of MHB Technical Consultants. He received a B.S. in Mechanical Engineering from the South Dakota School of Mines and Technology in 1953. From 1953 to 1976 he worked as an engineer an'd manager with the .

General Electric Company on many aspects of power generation equipment design, manufacture, and operation. During his last 10 years at General Electric, he held manage-ment positions in the Nuclear Energy Division where he managed the monitoring of nuclear power plant operations,ge implementation of solutions to nuclear plant opera-tional problems, and the development of a master performance improvement plan for power reactors. Prior to his management assignment in the Nuclear Energy Division, he spent several years as a field engineer at the first large scale commercial nuclear power plant built by General Electric for Commonwealth Edison Company at Dresden, near Chi-cago. There he supervised the construction, start-up, modifications, and repair of various ) portions of the plant. During this time, he was General Electric's site manager during the i 8903170095 PDR ADDCK890301 0500 1;l G m

r 1 2-first major refueling and maintenance outage at the Dresden Plant. In his capacity as con-sultant with MHB Technical Associates, Mr. Bridenbaugh has advised various governmen-tal bodies and individual groups on subjects related to the design, operation, and economic aspects of commercial nuclear power plants. He has served in various consulting capacities to the United States General Accounting Office; the state consumers' agencies and other state offices of the states of California, Illinois, Massachusetts, New Jersey, Ohio, Pennsyl-vania, Vermont, Maine; to Suffolk County, New York; and to the governments of Sweden and Norway, all in the evaluation of nuclear plants or programs. Mr. Bridenbaugh is a reg-istered professional nuclear engineer in the state of California and a member of the Amer-ican Nuclear Society. Further details of his experience and training are summarized in his resume, appended to this affidavit as Attachment 1.

2. Steven C. Sholly is an Associate Consultant with MHB Technical Associates, 1723 Hamilton Avenue, Suite K, San Jose, California 95125. He received a B.S. in Educa-tion, with a major in Earth and Space Science and a minor in Environmental Education, from Shippensburg State College, Shippensburg, Pennsylvania,in 1975. He has been an Associate Consultant with MHB Technical Associates since September 1985. In his capac-ity with MHB Technical Associates, Mr. Sholly has assisted in providing technical analysis and advice to local and state governments and independent organizations on a variety of issues related to the construction, operation, and safety of nuclear power plants. A princi-pal area of that analysis and advice is the assessment of (and interpretation of assessments of) the risks posed by the operation of nuclear power plants. He has served as a panelist and presenter in several NRC-sponsored forums related to regulatory uses of risk assess-ment, containment pulormance design objectives, and external events risk assessment. He has also presented expert testimony to the NRC and the Sizewell Inquiry (Un...d King-

_ . _ _ _ _ _ _ _ m

l L  ! I dom) concerning risk assessment and its applications. In connection with the Vermont Yankee Nuclear Power Station, he has provided technical advice to the New England ,

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Coalition on Nuclear Pollution and the Department of the Attorney General, Com- t I monwealth of Massachusetts, concerning the instant proceeding. In addition, he has per- i formed a brief technical review of the licensee's 1986 Vermont Yankee Containment Safety Study for NECNP. Further details of his qualifications are provided in Attachment 2.

3. The Vermont Yankee Nuclear Power Station (" Vermont Yankee") is located five miles south of Brattleboro, Vermont,in Windham County, Vermont. The plant received an Operating License on March 21,1972, and is operated by Vermont Yankee Nuclear Power Corporation. Vermont Yankee utilizes a General Electric, BWR/4 NSSS ,

with a Mark I pressure suppression containment. The authorized thermal power level for Vermont Yankee is 1,593 MWt.

4. By letter dated 25 April 1986, Vermont Yankee Nuclear Power Corporation

("the licensee") requested a change to Section 5.5.D of the Technical Specifications for the Vermont Yankee Nuclear Power Station which would increase the number of fuel assem-blies which could be stored in the spent fuel pool from 2,000 to 2,870.2 This change would be accomplished by installation of new spent fuel racks in the spent fuel pool which provide a closer packing of the spent fuel assemblies (i.e., installation of a high-density rack design).2 (Installation of the new racks was approved by the NRC staff by issuance of 2 Vermont Yankee Nuclear Power Corporation Letter No. FVY-86-34," Proposed Tech-nical Specification Change for Spent and New Fuel Storage", letter from Warren P. Murphy (VYNPC) to H.R. Denton (NRC), dated 25 April 1986. Attachment 3. 2 At the time oflicensing, the Vermont Yankee Nuclear Power Station technical specifi-cations provided for storage of 600 spent fuel assemblies. In September 1977, the licensee received approval to rerack the spent fuel pool via installation of high-density spent fuel storage racks. This license amendment provided for storage of up to 2,000 spent fuel assemblies in the spent fuel pool. See, Vennont Yankee Nuclear Power Cor-f poration (Vermont Yankee Nuclear Power Station), LBP-77-54,6 NRC 436 (1977), affirmed, ALAB-455,7 NRC 41 (1978).

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1. 1

(.. . License Amendment No.104 and the associated Safety Evaluation Report on May 20, 1988.).

5. .'On 26.May 1987, the Atomic Safety and Licensing Board issued a prehearing conference order'which granted NECNP's request for hearing and petition for intervention, and admitted a contention designated as Contention #1, with text as follows (as revised by the Appeal Board):

The . spent fuel pool expansion amendment'should be denied, because, through the necessity to use one train of the reactor's residual heat removal system (RHR) in addition'to the spent fuel cooling system in order to maintain the pool water within the design limit of 150 F, the single failure criterion as set forth in the General Design Criteria, and particularly Criterion 44, will be violated. The Applicant has not established that its proposed method of spent fuel pool cooling ensures that both the fuel pool cooling system and the reactor cooling system are single failure proof.

6. In a letter dated 2 March 1988 (FVY 88-17,' and attachments), Vermont Yankee Nuclear Power Corporation (VYNPC) " committed to design, install, test and make operational, a redundant seismically designed Spent Fuel Pool Cooling System prior to the time Vermont Yankee exceeds the existing 2,000 spent fuel assembly storage limit" for the spent fuel pool at Vermont Yankee. Attachment 4. The licensee committed to instali the enhanced system and make it operational prior to placement of more than 2,000 spent fuel assemblies in the spent fuel pool.

B. PERTINENT VERMONT YANKEE DESIGN DETAILS

7. The Vermont Yankee Nuclear Power Station (" Vermont Yankee") includes in its design a pool for the storage of spent fuel assemblies.3 The spent fuel pool provides 3 Vermont Yankee Nuclear Power Corporation, Final Safety Analysis Report, Vennont Yankee Nuclear Power Plant, updated 20 July 1982, Section 10.5.

for storage, shielding, and cooling of spent fuel. The pool is at an elevated location in the reactor building (secondary containment), adjacent to the primary containment and exposed to the refueling deck atmosphere. The spent fuel pool is a reinforced concrete structure supported by the reactor building walls. The pool is approximately 26 feet wide by 40 feet long by 39 feet high, and is lined with stainless steel to provide leakage protec-tion. Spent fuel is stored in the pool in free standing racks which are supported by the spent fuel pool floor.

8. The spent fuel assemblies contain irradiated uranium dioxide fuel which gen-erates heat as a result of decay of Ession products produced during operation of the reac-tor. This decay heat must be removed from the spent fuel assemblies stored in the spent fuel pool in order to prevent eventual damage to the fuel due to overheating.

9. Decay heat removal from the spent fuel pool is a safety-related function. If the decay heat removal function fails, the integrity of the spent fuel cladding cannot be maintained, and a radiological release from the damaged spent fuel assemblies can occur.4 For example, as set forth in NRC Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Revision 1, December 1975, at page 1.13-1, "Unless protective measures are taken, loss of water from a fuel storage pool could cause overheating of the spent fuel and resultant damage to fuel cladding integrity and could result in release of radioactive materials to the environment."

10. This is true both generally for BWR Mark I plants, and for Vermont Yankee specifically, as recognized in the updated FSAR. At Vermont Yankee, the spent fuel pool

• U.S. Nu: lear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Plants (LWR Edition), NUREG-0800, Rev.1, July 1981, Section 9.1.3," Spent Fuel Pool Cooling and Cleanup System", at 9.1.3-1. Attachment 5.

is cooled by the Spent Fuel Pool Cooling System (SFPCS).5 According to the Vermont Yankee Updated FSAR, the " safety objective"6 of the SFPCS is to maintain fuel pool water i at a temperature which will prevent damage to the fuel elements.' Accordingly, the SFPCS j at Vermont Yankee is safety-related.

11. The Vermont Yankee SFPCS consists of two identical trains of equipment.

Each train includes one 450 gpm8 centrifugal pump, and one 2.23 MBtu'/ hour tube-and-shell heat exchanger.10 The two trains of the SFPCS are headered together on the suction side of the pumps and at the discharge of the heat exchangers. The SFPCS is a non-seismic i 5 Vermont Yankee Nuclear Power Corporation, Final Safety Analysis Report, Vennont Yankee Nuclear Power Plant, updated 20 July 1982, Section 10.5. 6 According to the Vermont Yankee updated FSAR, when the word " safety"is used to , modify words such as objective, this indicates that the objective "is related to concerns considered to be of primary safety significance, as opposed to the station mission -- to generate electrical power. Thus, the word safety is used to identify aspects of the station - which are considered to be ofprim'ary importance with respect to safety." See, Vermont Yankee Nuclear Power Corporation, Final Safen' Analysis Report, Vermont Yankee Nuclear Power Station, updated 20 July 1982, at 1.2-3; emphasis added.

         '    Vermont Yankee Nuclear Power Corporation, FinalSafety Analysis Report, Vennont Yankee Nuclear Power Station, updated 20 July 1982, at 10.5-1.

8 Gallons per minute. Million British thermal units. 1 The heat removal capability of one train of the Spent Fuel Pool Cooling System is 2.23 MBtu/Hr according to the plant updated FSAR. For the proposed 2,870 fuel assembly limit on pool size, the NRC Staff estimates that the normal heat generation rate is 10.1 MBtu/Hr, and the abnormal heat generation rate (for a full core offload) is 21.46 MBtu/Hr. (See, U.S. Nuclear Regulatory Commission, Status Report on Review of l Expansion of the Spent Fuel Pool Storage Capacity, Vennont Yankee Nuclear Power Sta-tion, Docket Number 50-271, at 4-5, undated enclosure to letter dated 21 January 1988 , from Steven A. Varga (NRC) to Warren P. Murphy (VYNPC). Attachment 6. These heat generation rates are wellin excess of the capabilities of both trains of the SFPCS. Even assuming that the pool size remains at 2,000 assemblies, the conclusion that the SFPCS is inadequate to remove the pool decay heat rema:ns viable. ) l l L _ _ _ _ _ --_-__ _ _ _ _ _U

I Category I", non-Class IE22 system. The SFPCS cools the spent fuel pool by transferring decay heat from the spent fuel assemblies to the water in the pool and subsequently by pumping the pool water through heat exchangers to the reactor building closed cooling water system (RBCCWS). Heat from the RBCCWS is rejected to the ultimate heat sink via the residual heat removal service water system. The spent fuel pool is filled by, and makeup to the pool is supplied from, the condensate transfer system. The SFPCS pumps and heat exchangers are qualified for a temperature of 150 F.D

12. The residual heat removal (RH'R) system can be used to supplement SFPCS capacity when necessary. Connections are provided on the shutdown cooling piping por-tion of the RHR system for making a connection to the fuel pool system for this purpose.  !

1 In addition, the two loops of RHR are cross-connected by a single header, making it pos-sible to supply either loop of RHR by pumps in either loop."

           "        The pool structure and piping are designed to maintain their structuralintegrity for earthquakes up to the safe shutdown earthquake (SSE), but the system is not otherwise designed to seismic Category I standards. The SSE is one of the design basis accidents against which the design of safety-related nuclear power plant structure, system, and components is reviewed for adequacy.

12 U.S. Nuclear Regulatory Commission, Safety Evaluation by the Office of Nuclear Reac-tor Regulation Supporting Amendment No.104 to Facility Operating License No. DPR-25, Vennont Yankee Nuclear Power Corporation, Vennont Yankee Nuclear Power Station, i Docket No. 50-271,20 May 1988, at 2. Attachment 7. See, also, NRC SER, attachment to letter dated 14 October 1988 from Vernon L Rooney (NRC) to R.W. Capstick (VYNPC), at 2. Attachment 8. D U.S. Nuclear Regulatory Commission, Status Repon on Review of Expansion of the Spent Fuel Pool Storage Capacity, Vennont Yankee Nuclear Power Station, Docket Num-ber 50-271, undated, at 4 & 6, enclosure to letter dated 21 January 1988, from Steven A. Varga (NRC) to Warren P. Murphy (VYNPC), Attachment 6. See, also Final Safety 1 Analysis Report, Vennont Yankee Nuclear Power Station, updated 20 July 1982, at 10.5-2. M Vermont Yankee Nuclear Power Corporation, Final Safety Analysis Report, Vennont Yankee Nuclear Power Station, updated 20 July 1982, at 4.8-1 through 4.8-3. i _ _ _ - _ _ - _ _ _ _ _ .__ -_-_D

p , L C.EVA TION OF COMPLIANCE OF SFPCS WI SINGLE FAILURE CRITERION

13. The original intent of design provisions for the use of the RHR system in a spent fuel pool cooling mode represented a prudent step to provide an additional, emer-gency means to cool the spent fuel pool in the event that the primary cooling system (the spent fuel pool cooling system) fails. This is fully consistent with'the defense in-depth con-cept which underlies the Commission's safety regulations. Use of the RHR system to pro-vide routine backup to an inadequate spent fuel pool cooling system is not, however,in accord'with'the intent of the design provisions which allow cooling of the spent fuel pool by I

the RHR system, since routine reliance on the RHR system under these circumstances unacceptably degrades the reliability of the RHR system's primary function, which is decay heat removal from the reactor vessel and pressure suppression pool.

14. The RHR system is not meant to act as a second train for a spent fuel pool y

cooling system which cannot itself meet the single failure criterion (See paragraphs 20-22 . 1 and 24, below). Under these conditions, it is improper to rely on the residual heat removal system. Rather, the proper action is to upgrade the capabilities of the spent fuel pool cool-ing system (such as by providing a separate, redundant cooling train), or to reduce the heat load on the spent fuel pool to a level which can be supported by the existing pool cooling system (such as by, for example, utilizing an away from reactor (AFR) fuel storage facility, constructing an additional spent fuel pool and related auxiliary systems, or storing older spent fuel in a dry storage facility). l I

15. Pursuant to the Standard Review Plan (NUREG-0800), the NRC staff reviews spent fuel pool cooling systems to ensure that loss of function does not occur due to a single active failure. This " single failure criterion" applies to the design of the Vermont

9 Yankee spent fuel pool cooling system, as well as to the residual heat removal system and the service water system.15

16. According to an NRC Staff evaluation M, the Vermont Yankee SFPCS does not have sufficient capacity to cool the normal spent fuel pool heat load and maintain the pool water temperature below 140 F in the event of a single active failure. A similar con-clusion applies to conditions involving a full core offload. We agree with these conclusions, and believe that they necessarily apply to a higher temperature of 150 F.

17. Furthermore,in single failure analyses it is standard practice to evaluate the most limiting single failure (i.e., that single failure having the most severe impacts). The licensee claims to have evaluated "the most critical single active failure."" However, by assuming that the most critical single active failure for the SFPCS results at the worst in a situation where one pump and two heat exchangers remain available, the licensee has identified the most benign single failure possible.

18. In contrast, there are several postulated single failures for Vermont Yankee which would result in the availability of only one train of Engineered Safety Feature (ESF)

         "         NRC Staff Response to NECNP's First Set ofInterrogatories and Document Request to the NRC Staff,5 August 1987, Response to Interrogatory 26, attested to by John N.

Ridgely, at 3. See, also,10 CFR 50, Appendix A, Criterion 44," Cooling water". GDC-44 requires that the cooling water systems which transfer heat from structures, systems, and components important to safety (which applies to the Vermont Yankee spent fuel pool) be capable of removing the heat load without offsite power given a single active failure. Read together with the Vermont Yankee FSAR description of the safety objective of the SFPCS, GDC-44 quite clearly applies to spent fuel pools.

         "        NRC Staff Response to NECNP's First Set ofInterrogaton'es and Document Request ;o the NRC Staff,5 August 1987, Response to Interrogatory 32, attested to by John N.

Ridgely, at 6; and Response to Interrogatory 34, attested to by John N. Ridgely, at 7.

         "           " Memorandum of Vermont Yankee Nuclear Power Corporation in Response to Memorandum and Order of 10/24/88 and Motion for Leave to File the Same",10 November 1988, at 3.

equipment (i.e.,1 train of RHR,1 train of service water,1 SFPCS pump, and 1 SFPCS heat exchanger) being available, clearly a more severe condition than postulated by the licensee.

19. Two such failures are obvious, both in a loss of offsite power transient (one of the anticipated operational occurrences for nuclear power plants): (a) failure of one e train of safety-related AC power (due, for example, to failure of a diesel generator to start, failure of a diesel generator to continue to run once started, failure of safety-related switchgear to pick up the load, or unavailability of a diesel generator due to maintenance or surveillance testing), or (b) failure of one train of safety-related DC power (due, for example, to failure of a battery charger, unavailability of a DC battery due to maintenance or surveillance testing, or failure of a DC battery due to maintenance error). Under these failure conditions, only one train of service water will be available. Thus, the fact that the SFPCS heat exchangers can be cross-connected is largely irrelevant since one of the two SFPCS heat exchangers will not have water flowing past the secondary side of the heat l exchanger to serve as a mechanism for heat transfer across the heat exchanger tubes.

20. Moreover,it is apparent that the limiting failure for the SFPCS is the occur-rence of a Safe Shutdown Earthquake (SSE). Since the SFPCS is not designed to SSE I

l design standards, it must be assumed to be unavailable after the SSE, thus necessitating reliance on the RHR system for removal of heat from the spent fuel pool following an SSE.18 In addition, since the SSE is assumed under conventional accident analysis practice i 18 There is no question that the SFPCS is not seismically qualified. See letter from War-ren Murphy, VY, to NRC, dated 7 June 1988, Attachment A at A-5, referring to the "nonseismic Normal Fuel Pool Cooling Subsystem"; NRC Safety Evaluation Reports dated May 1988 [ enclosure to letter dated 20 May 1988 from Vernon L. Rooney, NRC to R.W. Captstick VY, at 2, Attachment 7], and October 1988 [ enclosure to letter dated 14 October 1988 from Vernon L. Rooney, NRC to R.W. Captstick VY, at 2, Attachment 8], which refer to the existing system as non-Seismic Category I. 1

                       . to be accompanied by a single failure, as demonstrated above what would be left would be a single train of RHR to remove decay heat from both the reactor coolant system and the      l spent fuel pool. The NRC staff has already determined that using one train of RHR to 4

cool both the spent fuel pool and the reactor coolant system is not appropriate since too many operator actions and RHR pump starts are required to operate in this mode." The l difficulties inherent in such a mode of operation (switching the single available RHR between the reactor and the spent fuel pool) are exacerbated by the proposed amendment, i since the heat removal needs of the spent fuel pool will be greater with 2870 assemblies than with 2000. Thus, the RHR would have to be devoted to spent fuel pool cooling a greater proportion of the time, and would be correspondingly less available to serve its reactor cooling function. l

21. Moreover, spent fuel pool makeup is provided by the condensate transfer sys-l tem. This system is not Seismic Category I and, therefore, cannot be relied upon to provide l

makeup to the spent fuel pool following a safe shutdown earthquake (SSE). Indeed, under contemporary NRC staff technical review practice, the condensate transfer system would l be assumed to be unavailable following an SSE. Thus, the primary means of providing makeup coolant to the spent fuel pool does not meet the single failure criterion either. The licensee has identified other means of providing makeup to the spent fuel pool, but none of these means meet SSE seismic design criteria, and therefore they do not meet 1 1

                        "     NRC Staff Status Report on Review of Expansion of the Spent Fuel Pool Storage Capacity, Vermont Yankee Nuclear Power Station, at 8 (undated enclosure to letter       ;

dated January 21,1988 from Steven A. Varga (NRC) to Warren P. Murphy (VYNPC)). l l-Attachment 6.

i l l single failure criterion and must be assumed to be unavailable following an SSE.20 2i I 2 The licensee, in its 25 April 1986 submittal to the NRC requesting an increase in the authorized spent fuel capacity to 2,870 fuel assemblies, asserted that there were three  ! seismic Category I sources for providing makeup water to the spent fuel pool. These  ! systems are the torus (suppression pool), the fire water system in conjunction with the service water system from the Connecticut River, and the cooling tower deep basin alternate cooling cell. Subsequently, however,in an 11 December 1987 submittal, the ' licensee indicated that the fire water pumps were not Seismic Category I. Thus, these pumps cannot be relied upon following an SSE. In addition, the NRC Staff has determined that the torus is an acceptable source of water makeup to the spent fuel pool only when the reactor is in the refueling mode with the vessel head removed and the gate between the open reactor vessel and the spent fuel pool is open. Use of the torus during power operation has been concluded by the NRC Staff to be unacceptable since the torus safety function would be compromised by lowering the torus water level to provide spent fuel pool makeup. See, U.S. Nuclear Regulatory Commission, Status Repon on Review of Expansion of the Spent Fuel Pool Storage Capacity, Vennont Yankee Nuclear Power Station, Docket Number 50-271, at 6-7, undated enclosure to letter dated 21 January 1988 from Steven A. Varga (NRC) to Warren P. Murphy (VYNPC). Attachment 6. In order for the service water system to be considered a Seismic Category I makeup source of water for the spent fuel pool, at least one service water pump must be Seismic Category I and powered from an onsite power source, and the fire water pipe must be Seismic Category I up to the suction of the Service Water pump, including appropriate isolation valves at connections between Seismic Category I and non-Seismic Category Ilines. See,Id., at 7. The NRC Staff has concluded that the service water pump meets the seismic requirements, but the fire protection system has not been shown to have the requisite features. The Staff concluded that the service water system is therefore not an acceptable Seismic Category I makeup source of water for the spent fuel pool. Ibid. The cooling tower deep basin alternate cooling cell's seis-mic classification is not addressed in the updated FSAR. In addition, piping from the alternate cooling cell to the service water system pumps w'ould also have to be Seismic Category I, and appropriate isolation valves would have to be provided. The NRC Staff has concluded that the alternate cooling cell is not an acceptable Seismic Category I makeup source for the spent fuel pool. Ibid. 21 More recently, the NRC staff stated that makeup to the spent fuel pool can be pro-vided by the " seismic Category I service water system" (see, NRC SER, attachment to letter dated 14 October 1988 from Vernon L Rooney (NRC) to R.W. Capstick (VYNPC), at 34). Attachment 8. The staff has not demonstrated, however, that the conditions required to render this makeup pathway fully Seismic Category I, as set forth in the preceeding footnote, have been met for the existing SPFCS or for the pro-posed enhanced system. Accordingly, the Board should not draw from the aforemen-tioned NRC staff statement a conclusion that a Seismic Category I-qualified means of providing spent fuel pool makeup exists at Vermont Yankee.

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22. Accordingly, it appears that there does not now exist (and has not ever existed) at Vermont Yankee a Seismic Category I makeup water source for spent fuel pool makeup. This violates 10 CFR 50, Appendix A, General Design Criterion 2 (GDC-2) which requires protection against earthquakes for systems performing functions which are important to safety (as noted above, the Vermont Yankee Updated FSAR makes it clear that the spent fuel pool cooling function is a safety-related function). This also violates GDC-61, since under postulated accident conditions (i.e., the Safe Shutdown Earthquake or SSE) the SFPCS is not designed to " prevent significant reduction in fuel storage coolant i inventory". l

23. Accordingly, the licensee is incorrect that "Given the most critical single active failure, therefore, the configuration of the system is one pump and two heat exchangers."22 In fact, given the most critical single active failure, the configuration of the Vermont Yankee SFPCS is no pumps or heat exchangers for SSE conditions, and one pump and one heat exchanger otherwise.

24. In addi; ion, since the SFPCS itself is not designed to SSE seismic standards, it does not meet the single failure criterion (indeed,it would have to be assumed to fail as a i

result of an SSE), and this also violates GDC-2.23 Further, this also violates GDC-61 m that the design of the SFPCS does not have "a residual heat removal capability having 22 " Memorandum of Vermont Yankee Nuclear Power Corporation in Response to Memorandum anr1 Order of 10/24/88 and Motion for Leave to File the Same",10 November 1988, at 3. 23 In addition, as discussed in paragraph 11 and note 12, the SFPCS is not Class IE. The NRC staff has not identified in what respects the existing system is not Class IE, but the failure of the system to meet Class IE criteria represents an additional basis for concluding that the system does not meet the single failure criterion.

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                              .           r-reliability and testability that reflects the importance to safety of decay heat and other residual heat removal".

25. Neither the NRC staff nor the licensee (nor any other party on the public i record) has determined the failure probability of one RHR train, both RHR trains, one train of service water, both trains of service water, the RBCCW system, or one or both trains of the SFPCS. Thus, a probabilistic argument cannot be made that an acceptable level of safety is achieved should the proposed amendment be approved.

26. One other. matter raised by the licensee requires a response. In footnote 4 of its November 10,1988, filing, the licensee states that prior to restart of the reactor there is -

no requirement of" redundant RHR trains on the reactor". This is quite likely incorrect, since operability of redundant RHR trains (and,in fact, operability of redundant diesel generators) is typically required for a BWR before the plant can proceed from cold shut-down to refueling. D. EVALUATION OF LICENSEE SUBMITI'ALS ON THE PROPOSED ENHANCED SFPCS

                                             '7. It is standard NRC practice to require final design details for a safety-related plant system prior to issuing an operating license. Such details include piping and instrumentation diagrams (P& ids), one-line electrical diagrams (showing power sources for pumps and valves and instrumentation), piping isometric drawings (showing a three-dimensional depiction of the arrangement of the piping and compon'ents in the system),

and design criteria for all structures, systems, and components necessary for operation of the system.

28. In the case of the licensee's proposed enhanced spent fuel pool cooling sys-

_-_--_______-___a

tem, no such level of detail has been provided in the March 2,198824 and June 7,108825 submittals on which the staff purportedly relied for its safety evaluation of the enhance-ment.26 Only two one-line piping diagrams have been provided, one for the combined existing SFPCS and the enhanced sytem, and one for the filtration system. This provides no information on electrical design for the enhanced system's pumps, leaving critical design questions unanswered (e.g., Are the pumps powered from separate AC and DC buses? Are there any single-failure points among the AC and DC power supplies for valves in the system?). This information is necessary for an evaluation of whether the proposed enhanced system satisfies the single failure criterion. In effect, the NRC staff proposes to issue an amendment to the license based upon a level of design detail for which the staff would otherwise only issue a construction permit. This is unacceptable practice. E. CONCLUSIONS

29. Based on the foregoing, we conclude that the existing Spent Fuel Pool Cool-ing System at Vermont Yankee violates GDC-2, GDC-44, and GDC-61, and violates the single failure criterion. Further, we conclude that none of the spent fuel pool makeup sys-tems identified by the licensee meet the single failure criterion, and this also violates GDC-2, GDC-44, and GDC-61. These conditions have existed since the first refueling outage i

1 24 Letter from Warren P. Murphy to NRC, FVY88-17, Attachment 4. l 25 Letter from Warren P. Murphy to NRC, FVY 88-47. Attachment 9. 2c see NRC Staff Response to NECNP's First Set ofInterrogatories and Request for Pro-duction of Documents to the NRC Staff on the Staff's Spent Fuel Pool Expansion Safety Evaluation, dated December 27,1988, at 4-6 (Responses to NECNP Inter-rogatory 4), and NRC Staff Response to NECNP's Second Set of Interrogatories and Request for Production of Documents to the NRC Staff on the Staff's Spent Fuel Pool Expansion Safety Evaluation, dated January 31,1989, at 2-4 (Responses to NECNP Interrogatory 2). l l

16 - - and will continue until such time as a seismically qualified spent fuel pool cooling system l enhancement and a seismically qualified spent fuel pool makeup system are provided for l Vermont Yankee.

30. Insufficient design details have been provided by the licensee to support a finding that the proposed enhancement to the Spent Fuel Pool Cooling System will meet the necessary requirements for a safety-related system, including compliance with GDC 2, GDC 44, GDC-61, and the single failure criterion.

31. Given the above conclusions, there is no basis for the Beard to approve the license amendment request permitting storage of an additional 870 fuel assemblies in the j spent fuel pool at vermont Yankee.

We, Dale G. Bridenbaugh and Steven C. Shelly, certify that the foregoing is true and correct to the best of our information, knowledge, and belief. Dale G. Bridenbaugh AC Steven C. Sholly y Subscribed to and sworn before me this First day of March,1989.

                                                                  !            $.           -    /b o r r7c7A t 2UY'"                                                                                         ,

MARY D. M0KINNEY . . g  : wenn evauc- uurm My Commission expires: i 4 SANTA CLARA COUNTY udMESN".'_, hAftW /f,,/W/ 408 266 7149 P.02 l MAR- 1-89 WED 17:15 )I

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• i PROFESSIONAL QUALIFICATIONS OF DALE G. BRIDENBAUGH 'l f

I I DALE G. BRIDENBAUGH MHB Technical Associates 1723 Hamilton Avenue Suite K SanJose, California 95125 (408) 266-2716-EXPERIENCE: 1976 - PRESENT - President - MHB Technical Associates. San Jose. California Co-founder and partner of technical consulting firm. Specialists in energy consulting to governmental and other groups interested in evaluation of nuclear plant safety and licensing. . Consultant in this capacity to state agencies in California, New York, Illinois, New Jersey, Pennsylvania, Oklahoma and Minnesota and to the Norwegian Nuclear Power Coc mittee, Swedish Nuclear Inspectorate, and various other organizations and environmental groups. Per'7rmed extensive safety analysis for Swedish Energy Commission and contributed to the Union of Concerned Scientists Review of WASH-1400. Consultant to the U.S. NRC - LWR Safety Improvement Program, performed Cost Analysis of Spent Fuel Disposal for the Natural Resources Defense Council, and contributed to the Department of Energy LWR Safety Improvement Program for Sandia Laboratories. Served as expert witness in NltC and state utility commission hearings. 1 1976-(FEBRUARY- AUGUST) Consultant. Project Survival. Palo Alto. California Volunteer work on Nuclear Safeguards Initiative campaigns in California, Oregon, Washington, Arizona, and Colorado. Numerous presentations on nuclear power and alternative energy options to civic, government, and college groups. Also resource person for public service presentations on radio and television. 1973 - 1976 Manacer. Performance Evaluation and Improvement. General Electric Company - Nuclear Enercy DMsion. San Jose. California Managed seventeen technical and seven clerical personnel with responsibility for establishmtmt and management of systems to monitor and measure Boiling Water Reactor equipment and system operational performance. Integrated General Electric resources in customer plant modifications, coordinated correction of causes of forced outages and of efforts to improve reliability and performance of BWR systems. Also responsible for development of Division Master Performance Improvement Plan as well as for numerous Staff special assignments on long-range studies. Was on special assignment for the management of two different ad hoc projects formed to resolve unique technical problems.

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1972 1973 Manacer. Product Service. General Electric Comnany - Nuclear Enerev Division. San Jose. California Managed group of twenty-one technical and four clerical personnel. Prime responsibility was to direct , l interface and liairon personnel involved in' corrective actions required under contract warranties. Also in charge of refueling and service planning, performance analysis, and service communication functions supporting all compkted commercial nuclear power reactors supplied by General Electric, both domestic and overseas (Spain, Germany, Italy, Japan, India, and Switzerland). 1968 - 1972 Manacer. Product Service. General Electric Comnany - Nuclear Enercy Division. San Jose. California Managed sixteen technical and six clerical personnel with the responsibility for all customer contact, planning and execution of work required after the customer acceptance of department-supplied plants and/or equipment. This included quotation, sale and delivery of spare and renewal parts. Sales volume of parts increased from $1,000,000 in 1968 to over $3,000,000 in 1972. 1966 - 1968 l Manacer. Comolaint and Warranty Service. General Electric Comnanv - Nuclear Enerev Division. San - Jose. Califors Managed group of six persons with the responsibility for customer contacts, planning and execution of  ; work required after customer acceptance of department-supplied plants and/or equipment--both domestic and overseas. I 1963 1966 Field Encineerine Sunervisor. General Electrie Comnanv. Installation and Service Encineerine Denartment. Los Ance!es. California Supervised approximately eight field representatives with responsibility fcr General Electric steam and gas turbine installation and maintenance work in Southern California, Arizona, and Southern Nevada. During this period was responsible for the installation of eight different. central station steam turbine-generator units, plus much maintenance activity. Work included customer contact, preparation of quotations, and contract negotiations. 1956 - 1 % 3 Field Er:cineer. GeneraLElectric Comnany. Installation and Service Encineerine Denartment. Chicagg., Illinois Supervised installation and maintenance of steam turbines of all sizes. Supervised crews of from ten to more than one hundred men, depending on the job. Worked primarily with large utilities but had significant work with steel, petroleum and other process industries. Had four years of experience at construction, startup, trouble-shooting and refueling of the first large-scale commercial nuclear power l unit.

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1955 - 1956 Encineerine Trainine Procram. General Electric Comnanv. Erie. Pennsvivania. and Schenectadv. New York Training assignments in plant facilities design and in steam turbine testing at two General Electric factory locations. , 1953 - 1955 United States Armv- Ordnance School. Aberdeen. Marviand Instructor - Heavy Artillery Repair. Taught classroom and shop disassembly of artillery pieces. 1953 Encineerine Traininc Procram. General Electric Comnanv. Evendale. Ohio Training assignment with Aircraft Gas Turbine Department. EDUCATION & AFFILIATIONS: BSME - 1953, South Dakota School of Mines and Technology, Rapid City, South Dakota, Upper 1/4 of class. Professional Nuclear Engineer - California. Certificate No. 0973. Member - American Nuclear Society Various Company Training Courses during career including Professional Business Management, Kepner Tregoe Decision Making, Effective Presentation, and numerous technical seminars. HONORS & AWARDS: Sigma Tau - Honorary Engineering Fraternity. General Managers Award, General Electric Company. PUBLICATIONS & TESTIMONY OF DALE G. BRIDENBAUGH:

1. Oneratinc and Maintenance Exnerience. presented at Twelfth Annual Seminar for Electrie Utility Executives, Pebble Beach, California, October 1972, published in General Electric NEDC-10697, December 1972.

2. Maintenance and In-Service Inspection. presented at IAEA Symposium on Experience From Operating and Fueling of Nuclear Power Plants, Bridenbaugh, Lloyd & Turner, Vienna, Austria, October,1973.

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3. Oneratinc and Maintenance Experience. presented at Thirteenth Annual Seminar for Electric Utility Executives, Pebble Beach, California, November 1973, published in General Electric NEDO-20222, January 1974.

4. Imnrovine Plant Availability. presented at Thirteenth Annual Seminar for Electric Utility Executives, Pebble Beach, California, November 1973, published in General Electric NEDO-20222, January,1974.

5. Annlication of Plant Outace Exnerience to improve Plant Performance. Bridenbaugh and Burdsall, American Power Conference, Chicago, Illinois, April 14,1974.

6. Nuclear Valve Testine Cuts Cost. Time. Electrical World, October 15,1974

7. Testimony of D. G. Bridenbaugh, R. B. Hubbard, and G. C. Minor before the United States Congress, Joint Committee on Atomic Energy, February 18,1976, Washington, D.C. (Published by the Union of Concerned Scientists, Cambridge, Massachusetts.)

8. Testimony of D. G. Bridenbaugh, R. B. Hubbard, and G. C. Minor w the California State Assembly Committee on Resources, Land Use, and Energy, March 8,1976.

9. Testimony by D. G. Bridenbaugh before the California Energy commission, entitled, Initiation of Catastronhie Accidents at Diablo Canvan, Hearings on Emergency Planning, Avila Beach, California, November 4,1976.

10. Testimony by D. G. Bridenbaugh before the U. S. Nuclear Regulatory Commission, subject: Diablo Canyon N.qglear Plant Performance. Atomic Safety and Licensing Board Hearings, in the matter of Pacific Gas and Electric Company, (Diablo Canyon Nuclear Power Plant, Units 1 and 2 ), Docket Nos.

50-275-OL, 50-323-O L,Decembe r,1976.  ; IL Testimony by D. G. Bridenbaugh before the California Energy Commission, subject: Interim Snent Fuel Storace Considerations. March 10,1977.

12. Testimony of D. G. Bridenbaugh before the New York State Public Service Commission Siting Board Hearings concerning the Jamesport Nuclear Power Station, subject: Effect of Technical and Safety Deficiencies on Nuclear Plant Cost and Reliability. in the matter of Long Island Lighting Company (Jamesport Nuclear Power Station, Units 1 and 2), Case No. 80003, April,1977.

13. Testimony by D. G. Bridenbaugh before the California State Energy Commission, subject:

Decommissioning of Pressurized Water Reactors. Sundesert Nuclear Plant Hearings,in the matter of San Diego Gas and Electric Company (Notice ofIntention to File Application for Certification of Site and Related Facilities), Docket No. 76-NOI-2, June 9,1977.

14. Testimony by D. G. Bridenbaugh before the California State Energy Commission, subject: Economic Relationships of Decommissioning. Sundesert Nuclear Plant, for the Natural Resources Defense Council,in the matter of San Diego Gas and Electric Company; Notice ofIntention to File Application for Certification of Site and Related Facilities, Docket No. 76-NOI-2, July 15,1977.

15. The Risks of Nuclear Power Reactors- A Review of the NRC Reactor Safety Study WASH-1400.

Kendall, Hubbard, Minor & Bridenbaugh, et. al., for the Union of Concerned Scientists, August,1977.

16. Testimony by D. G. Bridenbaugh before the Vermont State Board of Health, subject: Oneration of y_crmont Yankee Nuclear Plant and Its impact on Public Health and Safety. October 6,1977.

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17. Testimony by D. G. Bridenbaugh before the U.S. Nuclear Regulatory Commission, Atomic Safety and Licensing Board, subject: Deficiencies in Safety Evaluation of Non.Seisme fyues. Lack of a Definitive Findinc of Safety. Diablo Canyon Nuclear Units, October 18,1977, Avila Beach, California.

18. Testimony by D. G. Bridenbaugh before the Norwegian Commission on Nuclear Power, subject:

Reactor Safetv/ Risk, October 26,1977.

19. Swedish Reactor Safety Study Barseback Risk Assessment. MHB Technical Associates, January,1978. j (Published by the Swedish Department ofIndustry as Document Dsl 1978:1)

20. Testimony by D. G. Bridenbaugh before the Louisiana State Legislature Committee on Natural l Resources, subject: Nuclear Power Plant Deficiencies Imnactine on Safety & Reliability. Baton Rouge,  ;

Louisiana, February 13,1978. l

21. Soent Fuel Disnosal Costs. report prepared by D. G. Bridenbaugh for the Natural Resources Defense Council (NRDC), August 31,1978. {

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22. Testimony of D. G. Bridenbaugh, G. C. Minor, and R. B. Hubbard before the Atomic Safety and Licensing Board, in the matter of the Black Fox Nuclear Power Station Construction Permit Hearings, September 25,1978, Tulsa, Oklahoma.

23. Testimony of D. G. Bridenbaugh and R. B. Hubbard before the Louisiana Public Service Commission, Nucler.r Plant and Power Generation Costs. November 16,1978, Baton Rouge, Louisiana.  ;

24. Testimony by D. G. Briden'oaugh before the City Council and Electric Utility Commission of Austin, Texas, Desien. Construction. and Oneratine Exnerience of Nuclear Generatine Facilities. December 5, l 1978, Austin, Texas. 3

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25. Testimony by D. G. Bridenbaugh for the Commonwealth of Massachusetts, Department of Public Utilities, Imnact of Unresolved Safety Issues. General Deficiencies. and Three Mile Island-Initiated Modifications on Power Generation Cost at the Pronosed Pilcrim-2 Nuclear Plant. June 8,1979. i

26. Imorovine the Safety of LWR Power Plants. MHB Technical Associates, prepared for U.S. Dept. of

[ Energy, Sandia Laboratories, September 28,1979. e f l j 27. BWR Pine and Nozzle Cracks. MHB Technical Associates, for the Swedish Nuclear Power Inspectorate (SKI), October,1979. ) } l l 28. Uncertainty in Nuclear Risk Assessment Methodolocv. MHB Teehnical Associates, for the Swedish i Nuclear Power Inspectorate (SKI), January 1980. l 29. Testimony of D. G. Bridenbaugh and G. C. Minor before the Atomic Safety and Licensing Board, in the natter of Sacramento Municipal Utility District, Rancho Seco Nuclear Generating Station following TMI-2 accident, subject: Onerator Trainine and Human Factors Eneineerinc. for the Cali- ) fornia Energy Commission, Docket No. 50-312-SP, February 11,1980. f l

30. Italian Reactor Safety Study: Caorso Risk Assessment. MHB Technical Associates, for Friends of the Earth, Italy, March,1980.

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31. Decontamination of Krypton-85 from Three Mile island Nuclear Plant. H. Kendall, R. Pollard, and D.

G. Bridenbaugh, et al, The Union of Concerned Scientists, delivered to the Governor of Pennsylvania, May 15,1980.

32. Testimony by D. G. Bridenbaugh before the New Jersey Board of Public Utilities, on behalf of New Jersey Public Advocate's Office, Division of Rate Counsel, Analvsis of lo70 Salem-1 Refueline Outace.

in the matter of the Petition of Public Service Electric and Gas Company for approval of an increase in Electric and Gas rates and for changes in the tariffs for Electric and Gas service., P.U.C. N.J. No. 7, Electric, and P.U.C. NJ. No. 5, Gas, Pursuant to M 48:2-21, August 1980.

33. Minnesota Nuclear Plants Gaseous Emissions Study. MHB Technical Associates, for Minnesota Pollution Control Agency, September,1980.

34. Position Statement, Prooosed Ruimakinc on the Storace and Disoosal of Nuclear Waste. Joint Cross-Statement of Position of the New England Coalition on Nuclear Pollution and the Natural Resources Defense Council, September,1980.

35. Testimony by D. G. Bridenbaugh and G. C. Minor, before the New York State Public Service Commission, in the matter of Long Island Light Company Temporary Rate Case, prepared for the Shoreham Opponents Coalition, September 22, 1980, Case No. 27774, Soreham Nuclear P! ant Construction Schedule. ,

36. Supplemental Testimony by D. C. Bridenbaugh before the New Jersey Board of Public Utilities, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, Analysis of 1979 Salem-1 Refuelinc Outace. in the matter of the Petition of Public Service Electric and Gas Company for approval of an increase in Electric and Gas rates and for changes in the tariffs for Electric and Gas Service, P.U.C. NJ. No. 7, Electric, and P.U.C. NJ. No. 5, Gas, Pursuant to M 48:2-21, Docket No.

794-310, OAL Docket No. PUL-877-79, December,1980.

37. Testimony by D. G. Bridenbaugh and G. C. Minor, before the New Jersey Board of Public Utilities, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, Ovster Creek 12.80 Refueline Outace Investigation. in the matter of the Petition of Jersey Central Power and Light Company for approval of an increase in the rates for electrical service and adjustment clause and factors for such service, OAL Docket No. PUC-3518-80, BPU Docket Nos. 804-285,807-488, February 1981.

38. Economic Assessment: Ownershin Interest in Palo Verde Nuclear Station. MHB Technical Associates, for the City of Riverside, September 11,1981.

39. Testimony of D. G. Bridenbaugh before the Public Utilities Commission of Ohio, in the Matter of the l

Regulation of the Electric Fuel Component Contained Within the Rate Schedules of the Toledo l Edison Company and Related Matters, subject: Davis-Besse Nuclear Power Station 1980-81 Outace l Review. Case No. 81-306-EL-EFL, November,1981.

40. Supplemental Testimony of D. G. Bridenbaugh before the Public Utilities Commission of Ohio, in the matter of the .Regulation of the Electric Fuel Component Contained within the Rate Schedules of the Toledo Edison Company and Related Matters, subject: Davis-Besse Nuclear Power Station 1980-81 Outace Review. Case No. 81-306-EL-EFL, November 198L 4L Systems Interaction and Sincie Failure Criterion. Phase 2 Report. MHB Technical Associates for the Swedish Nuclear Power Inspectorate (SKI), January,1982.

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42. : Testimony of D. G. Bridenbaugh and G. C. Minor on behalf of Governor Edmund G. Brown Jr. .

before the Atomic Safety and Licensing Board, regarding Contention 10. Pressuri7er Heaters. in the c ;;. ' matter of Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), . Docket Nos. 50-275-0L,50-323-OL, January 11,1982.

43. Testimony of D. G. Bridenbaugh and G. C. Minor on behalf of Governor Edmund G.- Brown Jr.,

before the Atomic Safety and Licensing Board, regarding Contention 12 Block and Pilot Onerated Relief Valver, in the matter of Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Docket Nos. 50-275-OL,50-323-OL, January 11,1982.

44. Testimony of D. G. Bridenbaugh before the Commonwealth of Massachusetts, Department of Public Utilities, on behalf of the Massachusetts' Attorney General, Pilcrim Nuclear Power Station. 1981-82 Outace Investigation, in the matter of Boston Edison Company, DPU Docket No.1009-F, March 11, 1982.

45. Testimony of D. G. Bridenbaugh before the Pennsylvania Public Utility Commission, on behalf of the i Pennsylvania Office of Consumer Advocate, Beaver Vallev Outace. March,1982.

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                           - 46. Interim testimony of D. G. Bridenbaugh on Emected Lifetimes and Performance of Nuclear Power Plants. State of Illinois Commerce Commission,82-0026, AG Exhibit 6, March,1982.

47. Testimony of D. G. Bridenbaugh and G. C. Minor before the Atomic Safety and Licensing Board, on behalf of Suffolk County, in the matter of Long Island Lighting Company, Shoreham Nuclear Power Station, Unit 1, regarding Suffolk County Contention 11. Passive Mechanical Valve Failures. Docket-No.50-322-OL, April 13,1982.

48. Testimony of D. G. Bridenbaugh and R. B. Hubbard, in the Matter of Jersey Central Power and Light Company For an Increase in Rates for Electrical Service, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, Three Mile Ishnd Units 1 & 2. Cleanun and Modification ,

Procrams. DPU Docket Nos. 818-726,818-736, May,1982.

49. Testimony of D. G. Bridenbaugh and G. C. Minor on behalf of Suffolk County, before the Atomic Safety and Licensing Board, in the matter of LongIsland Lighting Company, Shoreham Nuclear Power Station, Unit 1, regarding Suffolk County Contention 22. SRV Test Procram. Docket No. 50-322-OL, May 25,1982. .

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50. Testimony of D. G. Bridenbaugh and G. C. Minor on behalf of Suffolk County, before the ' Atomic Safety and Licensing Board, in the matter of 1.ong Island Lighting Company, Shoreham Nuclear Power Station, Unit 1, regarding Suffolk County Contention 28fa)(vil and SOC Contention 7A(61 Reduction of SRV Challences. Docket No. 50-322-OL, June 14,1982.

51. Testimony of D. G. Bridenbaugh before the Illinois Commerce Commission, on l$ehalf of the Illinois Attorney General's Office, Emected Lifetimes and Performance of Nuclear Power Plants. in the matter of Commonwealth Edison (Proposed general increase in electric rates), ICC Docket No. 82-0026, June 18,1982.

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52. Testimony of D. G. Bridenbaugh and R. B. Hubbard on behalf of the Ohio Consumers Counsel, before the Public Utilities Commission of Ohio, regarding Construction of Perry Nuc! ear Generatine Unit No.

1, in the matter of the application of the Cleveland Electric Illuminating Company for authority to amend and increase certain of its filed schedules fixing rates and charges for electric service, Case No. i I 81-1378-EL-AIR, October 7,1982.

53. Issues Affectine the Viability and Acceptability of Nuclear Power Usace in the United States. prepared by MHB Technical Associates for Congress of the United States, Office of Technology Assessment for use in conjunction with Workshop on Tec.%ological and Regulatory Changes in Nuclear Power, December 8 & 9,1982.

54 Testimony of D. G. Bridenbaugh on behalf of Rockford League of Women Voters, before the Atomic Safety and Licensing Board, in the matter of Commonwealth Edison Company, Byron Station, Units 1-and 2, regarding Contention 22. Steam Generators. Docket Nos. 50-454,50-455, March 1,1983.

55. Testimony of G. C. Minor and D. G. Bridenbaugh before the Pennsylvania Public Utility Commission, on behalf of the Office of Consumer Advocate, Recardine the Cost of Consttuctine the Susouchanna Steam Electric Station. Uni.! L Re: Pennsylvania Power and Light, Docket No. R-822169, March 18, 1983.

56. Surrebuttal Testimony of D. G. Bridenbaugh before the Pennsylvania Public Utility Commission, on behalf of the Office of Consumer Advocate, Recardinc the Cost of Constructive the Susouchanna Steam Electric Station. Unit L Re: Pennsylvania Power and Light, Docket No. R-822169, April 20, 1983.

57. Testimony of D. G. Bridenbaugh In the Matter of Public Service Gas & Electric, Base Rate Case.

Nuclear Construction Emenditures. on behalf of New Jersey Department of the Public Advocate, Dhision of Rate Counsel, Docket No. 836-620, OAL Docket No. PUC@930-83, October 13,1983.

58. Affidavit of D. G. Bridenbaugh, in the Matter of Jersey Central Power and Light, on behalf of New Jmey Department of the Public Advocate, Dhision of Rate Counsel, TMI Fault Investigation. DPU Do .ket No. 836-500, November 23,1983.

59. Tes:imony of D. G. Bridenbaugh, in the Matter of Public Service Electric & Gas, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, LEAC Investigation. Salem-1 hingn, DPU Docket No. 831-25, December 1,1983.

l 60. Rebuttal Testimony of D. G. Bridenbaugh, in the Matter of Public Service Electric & Gas, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, LEAC Investigation. Salem. 1 Outaces. DPU Docket No. 831-25, January 18,1984. l l 61. Testimony of D. G. Bridenbaugh, L. M. Danielson, R. B. Hubbard and G. C. Minor before the State of l New York Public Service Commission, PSC Case No. 27563, in the matter of Long Island Lighting Company Proceeding to Investigate the Cost of the Shoreham Nuclear Generating Facility -- Phase II, ' on behalf of County of Suffolk, February 10,1984.

62. Testimony of D. G. Bridenbaugh, in the Matter of Jersey Central Power & Light Company, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, Base Rate Case. Ovster Creek 1983.d4 Outace and O&M and Capital Emenditures. OAL Docket No. PUL-00797-84, BPU Docket No. 841-55, May 23,1984.

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63. Direct Testimony of Dale G. Bridenbaugh and Richard B. Hubbard, Before the Illinois Commerce Commission, Illinois Power Company, Clinton Nuclear Station, on its own motion, an investigation to consider a plan for moderating the initial rate increase associated with placing Illinois Power Company's Clinton Unit No.1 generating station in service, Docket No. 84-0055, available from Illinois Governor's Office of Consumer Services, July 30,1984.

64. Joint Direct Testimony of Dr. Robert N. Anderson, Professor Stanley G. Christensen, G. Dennis Eley, Dale G. Bridenbaugh and Richard B. Hubbard Regarding Suffolk County's Emergency Diesel Generator Contentions, Before the Atomic Safety and Licensing Board, in the matter of Long Island Lighting Company, Shoreham Nuclear Plant Unit 1, NRC Docket No. 50-322-OL, Julv 31,1984.

65. Surrebuttal Testimony of Dale G. Bridenbaugh, Lynn M. Danielson, Richard B. Hubbard, and Gregory C. Minor, Before the New York State Public Service Commission, PSC Case No. 27563, Shoreham Nuclear Station, Long Island Lighting Company, on behalf of Suffolk County and New York State Consumer Protection Board,in the matter of Long Island Lighting Company Proceeding to Investigate the cost of the Shoreham Nuclear Generating Facility - Phase II, October 4,1984

66. Direct Testimony of Dale G. Bridenbaugh, Lynn M. Danielson and Gregory C. Minor on Behalf of Massachusetts Attorney General, DPU 84-145, Before the Massachusetts Department of Public Utilities, regarding the prudency of expenditures by Fitchburg Gas and Electric Light Company on Seabrook Unit 2, November 23,1984,84 pgs.

67. Direct Testimony of Dale G. Bridenbaugh, Richard B. Hubbard and Lynn K. Price on Behalf of Massachusetts Attorney General, DPU 84-152, Before the Massachusetts Department of Public Utilities, regarding the investigation by the Department of the Cost and Schedule of Seabrook Unit 1, Dcccmber 12,1984.

68. Direct Testimony of Dale G. Bridenbaugh, Lynn M. Danielson and' Gregory C. Minor on Behalf of l Maine Public Utilities Commission Staff regarding Seabrook Unit 2, Docket No. 84-113, December 21, 1984.

69. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor Regarding Suffolk County's Emergency Diesel Generator Load Contention, Docket No. 50-322-OL, January 25,1985.

70. Direct Testimony of Dale G. BridenbAugh,in the Matter of the Motion of Public Service Electric &

Gas, on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, Motion To increase The Level of the Leveli7ed Enercy Adiustment Clause. Docket No. ER 8501166 and Docket No. 837-620, April 24,1985.

71. Direct Testimony of Dale G. Bridenbaugh on behalf of the Attorney General of the Commonwealth of Massachusetts, in the Matter of Boston Edison Company DPU 851B, A Hearing to Determine Whether Fuel and Purchased Power Costs Associated with the Outage at Pilgrim Nuclear Power Station Which Began on December 10,1983 and Ended on December 30,1984 Were Reasonably and

               , Prudently Incurred. May 13,1985.

72. Direct Testimony of Dale G. Bridenbaugh on behalf of the Residential Ratepayer Consortium,in the Matter of the Application of Consumers Power Company for a Power Supply Cost Ecconciliation proceeding for the 12-month period ended December 13, 1984, regarding Palisades Outage Review, Case No. U-7785 R, August 28,1985.

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73. Direct Testimony of Dale G. Bridenbaugh, Lynn M. Danielson, and Gregory C. Minor on behalf of the Department of Public_ Service, State of Vermont Public Service Board Docket No. 5030, Central Vermont Public Service Corporation, November 11,1985.

74 Direct Testimony of Dale G. Bridenbaugh on behalf of New Jersey Department of the Public Advocate,in the matter of JCP&L for an increase in rates, Base Rate Case, Oyster Creek O&M and Capital Expenditures, OAL Docket No. 4929-85, BPU Docket No. 8507 698, November 25,1985.

75. Direct Testimony of Dale G. Bridenbaugh on behalf of New Jersey Department of the Public Advocate, in the matter of JCP&L, TMI-Restart - LEAC, Re: TMI Restart Commercial Operation Standards & Reliability of Service, January 31,1986.

76. Direct Testimony of Dale G. Bridenbaugh, Gregory C. Minor, Lynn K. Price, and Steven C. Sholly on behalf of State of Connecticut Department of the Public Utility Control Prosecutorial Division and Division of Consumer Counselin the matter of Connecticut Light and Power Company Retrospective Audit of the Prudence of the Management and Financing of the Construction of Millstone Unit 3, February 18,1986.

77. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of Massachusetts Attorney General regarding the prudence of expenditures by New England Power Co. on Seabrook Unit 2, Docket Nos. ER 85-M6-000, ER 85-647-000, February 21,1986.

78. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of Massachusetts Attorney General regarding WMECo Construction Prudence for Millstone Unit 3, in tha matter ofinvestigation by the department on it own motion as to the priority of the rates and charges set forth in schedules

     ' filed with the department Dec. 17, 1985 by Western Massachusetts Electric Co. to become effective Jan.1,1986, Docket No. 85 270, March 19,1986.

79. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of Massachusetts Attorney General regarding WMECo's Commercial Operating Dates and Deferred Capital Additions on Millstone Unit 3, Docket No. 85-270, March 19,1986.

80. Rebuttal Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of Massachusetts Attorney General regarding New England Power Company's Seabrook 2 Rebuttal, Docket Nos. ER 85-646-001, ER 85-647-001, April 2,1986.

81. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of State of Maine Staff of Public Utilities Commission regarding Construction Prudence of Millstone Unit 3, in the matter of Maine Power Company Proposed Increase in Rates, Docket No. 85 212, April 21,1986.

82. Direct Testimony of Dale G. Brider.baugh and Peter M. Strauss on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, regarding Base Rate Case:In Service Criteria for Hope Creek, Hope Creek O&M and Decommissioning Costs, and Operating Plant O&M Costs, OAL Docket No. PUL 023186, BPU Docket No. ER 85121163, May 19,1986,107 pp.

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83. Direct Testimony of Dale G. Bridenbaugh on behalf of New Jersey Department of the Public Advocate, Division of Rate Counsel, regarding Base Rate Case: Hope Creek Commercial Operating Date and Criteria, Hope Creek O&M Costs, Operating Life, Capital Additions, and Decommissioning Costs,in the matter of Atlantic City Electric Company increasing its rates for electric service Phase II, OAL Docket No. PUL 3290-85, BPU Docket No. ER 8504-434, May 27,1986,85 pp.

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84. Direct Testimony of Dale G. Bridenbaugh, Richard B. Hubbard, and Lynn K. Price on behalf of State of Illinois Office of the Attorney General and Office of Public Counsel, in the matter of Illinois Commerce Commission on its own motion, an investigation to consider a plan for moderating the initial rate increase associated with placing Illinois Power Company's Clinton Unit 1 generating station in service, Docket No. 84-0055, July 9,1986.

85. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor on behalf of the Vermont Department of Public Service, regarding Tariff Filing of Central Vermont Public Service Corporation Requesting a 12% Increase in Rates, Docket No. 5132, August 25,1986.

86. Direct Testimony of Dale G. Bridenbaugh and Richard B. Hubbard on behalf of the Pennsylvania Office of Consumer Advocate, regarding Pennsylvania Public Utility Commission vs. Duquesne Light Company and Pennsylvania Power Company, Docket Nos. R-860378 and R-850267, September 22, 1986.

87. Direct Testimony'of Dale G. Bridenbaugh and Richard B. Hubbard on behalf of The Public Parties Committee, Public Utility Commission of Texas, regarding the Evaluation of Costs of River Bend Nuclear Generating Station,in the matter of application of Gulf States Utilities for authority to change rates, Docket Nos. 7195 and 6755, February 23,1987.

88. Direct Testimony of Dale G. Bridenbaugh on behalf of Maryland People's Counsel,in the matter of the Application of the Baltimore Gas and Electric Company to Adjust Its Electric Fuel Rate Charges, Pursuant to Section 54F of Article 78 of the Annotated Code of Maryland, Case No. 8520-D, April 29, 1987.

89. Direct Testimony of Dale G. Bridenbaugh on behalf of Florida Office of Public Counsel, in regard to Fuel and Purchased Power Cost Recovery Clause with Generating Performance Incentive Factor (Florida Power Corporation - Crystal River 3), Docket No. 860001-EI-B, June 12,1987.

90. Direct Testimony of Dale G. Bridenbaugh on behalf of the Residential Ratepayer Consortium, before the Michigan Public Service Commission, in the matter of the Application of Consumers Power Company for a Reconciliation of Power Supply Cost Recovery Costs and Revenues for Calendar Year 1986, Palisades Nuclear Power Plant, Case No. U-8286-R, July 13,1987.

91. Direct Testimony of Dale G. Bridenbaugh on behalf of the City of El Paso, before the Public Utility Board, in the matter of the Application of the El Paso Electric Company for a Rate Increase in the City of El Paso, Evaluation of Costs of Palo Verde Units 1 and 2, July 15,1987.

92. Direct Testimony of Dale G. Bridenbaugh on behalf of the City of El Paso, before the Public Utility Commission of Texas,in the matter of the Application of the El Paso Electric Company for Authority to Increase Electric Rates, Evaluation of Operational and Decommissioning Costs of Palo Verde Units 1 and 2, Docket No. 7460, July 29,1987.

93. Direct Testimony of Daic G. Bridenbaugh and Gregory C. Minor on behalf of Massachusetts Attorney General, before the Federal Energy Regulatory Commission, regarding Canal Electric Company Prudence Related to Seabrook Unit 2 Construction Expenditures, Docket No. ER86-704-001, July 31, j

1987. l

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94. Direct Testimony of Dale G. Bridenbaugh on behalf of Maryland Peop!c's Counsel, before the Public Service Ccmmission of Maryland, in the matter of the Application of Delmarva Power & Light Company for Electric Fuel Rate Adjustment, Pursuant to Section 54F of Article 78, of the Annotated .

Code of Maryland, Case No 8521, Phase II, August 10,1987, PROPRIETARY. i

95. Direct Testimony of Dale G. Bridenbaugh and Gregory C. Minor before the Pennsylvania Public Utility Commission, Regarding Beaver Valley Unit 1, Docket No.1-79070318, OCA Statement No. 2, August 31,1987.

96. Direct Testimony of Dale G. Bridenbaugh on behalf of Maryland People's Counsel, Case No. 8520-C, in the Matter of the Application of the Baltimore Gas & Electric Company to Adjust its Electric Fuel Rate Charges, Pursuant to Section 54F of Article 78 of the Annotated Code of Maryland, before the Public Service Commission of Maryland, October 20,1987.

97. Direct Testimony of Dale G. Bridenbaugh on behalf of the Pennsylvania Office of Consumer Advocate, before the Pennsylvania Public Utility Commission, regarding Evaluation of Perry Plant Power Ascension Program, Docket Nos. R-870651 and R-870732, OCA Statement No. 7, October 1987. .

98. Surrebuttal Testimony of Dale G. Bridenbaugh before the Pennsylvania Public Utility Commission, ren &g Beaver Valley Unit 1, Docket No. I-79070318, OCA Statement No. 2A, October 30,1987.

99. Surret,cN Testmony of Dale G. Bridenbaugh before the Pennsylvania Public Utility Commission, on behalf of the n.asylvania Office of Consumer Advocate, regarding Evaluation of Perry Plant Power Ascension Program, Docket Nos. R-870651 and R-870732, December 1987, 100. SurrebuttalTestimony of Dale G. Bridenbaugh before the Public Service Commission of Maryland, on behalf of Maryland People's Counsel, in the matter of the Application of the Baltimore Gas and Electric Company to Adjust Its Electric Fuel Rate Charges, Pursuant to Section 54F of Article 78 of The Annotated Code of Maryland, C.ae No. 8520-C, December 17,1987.

101. Direct Testimony of Dale G. Bridenbaugh before the Michigan Public Service Commission on behalf of the Michigan Attorney General Concerning Evaluation of the Investment - MidIr.ad Nuclear Power Station,in the Matter of the Application of Consumers Power Company for Authority to Increase Its Rates for the Sale of Electricity, Case No. U-7830 (Midland), Step 3B, December 29,1987. 102. Direct Testimony of Dale G. Bridenbaugh before the Minnesota Public Utilities Commission on behalf of the Minnesota Department of Public Service Concerning Northern States Power's Decommissioning Cost Estimates for the Monticello and the Prairie Island Nuclear Power Station, in the Matter of the Application of Northern States Power Company for Authority To Increase Its Rates for Electric Services in Minnesota, Docket No. E-OOZ/GR-87-670, February 19,1988. 103. Direct Testimony and Exhibits of Dale G. Bridenbaugh on behalf of Residential Ratepayers Consortium before the Michigan Public Service Commission, State of Michigan, in the Matter of the application of Consumers Power Company for a Reconciliation of Power Supply Cost Recovery (PSCR) Costs and Revenues for Calendar Year 1987, Case No. U-8545R (1987 PSCR Reconciliation), August 17,1988.

c , . e i PROFESSIONAL OU ALIFICATIONS OF STEVEN C. SHOLLY i STEVEN C. SHOLLY MHB Technical Associates 1723 Hamilton Avenue Suite K San Jose, California 95125 (408) 266-2716-

              . EXPERIENCE:

September 1985 - PRESENT. Associate - MHB Technical Associates. San Jose. California Associate in energy consulting firm that specializes in technical and economic assessments of i energy production facilities, especially nuclear, for local, state, and federal governments and private organizations. MHB is extensively involved in regulatory proceedings and the preparation of studies and reports. Conduct research, write reports, participate in discovery l ) process in regulatory proceedings, develop testimony and other documents for regulatory proceedings, and respond to client inquiries. Clients have included: State of California, State of l New York, State ofIllinois. l February 1981 - September 1985 l Technical Research Associate and Risk Analyst - Union of Concerned Scientists. Washincton. DS. Research associate and risk analyst for public interest group based in Cambridge, [ Massachusetts, that specializes in ernmining the impact of advanced technologies on society, l l principally in the areas of arms control and energy. Technical work focused on nuclear power I plant safety, with emphasis on probabilistic risk assessment, radiological emergency planning and preparedness, and generic safetyissues. Conducted research, prepared reports and studies, participated in administrative proceedings before the U.S. Nuclear Regulatory Commission, developed testimony, analyzed NRC rule-making proposals and draft reports and prepared comments thereon, and responded to inquiries from sponsors, the general public, and the j I media. Participated as a member of the Panel on ACRS Effectiveness (1985), the Panel on Regulatory Uses of Probabilistic Risk Assessment (Peer Review of NUREG-1050; 1984), Invited Observer to NRC Peer Review meetings on the source term reassessment (BMI-2104; 1983-1984), and the Independent Advisory Committee on Nuclear Risk for the Nuclear Risk Task Force of the National Association ofInsurance Commissioners (1984). January 1980 - January 1981 Proiect Director and Research Coordinator - Three Mile Island Public Interest Resource Center. Harrisburc. Pennsvivania Provided administrative direction and coordinated research projects for a public interest group based in Harrisburg, Pennsylvania, centered around issues related to the Three Mile Island Nuclear Power Plant. Prepared fundraising proposals, tracked progress of U.S. Nuclear Regulatory Commission, U.S. Department of Energy, and General Public Utilities activities concerning cleanup of Three Mile Island Unit 2 and preparation for restart of Three Mile Island Unit 1, and monitored developments related to emergency planning, the financial health of General Public Utilities, and NRC rulemaking actions related to Three Mile Island. July 1978 - January 1980 Chief Biolocical Process Operator - Wastewater Treatment Plant. Derry Townshin Municinal Authority. Hershev. Pennsylvania 1 Chief Biological Process Operator at a 2.5 million gallon per day tertiary, activated sludge, wastewater treatment plant. Responsible for biological process monitoring and control, including analysis of physical, chemical, and biological test results, process fluid and mass flow management, micro-biological analysis of activated sludge, and maintenance of detailed process logs for input into state and federal reports on treatment process and effluent quality. Received certification from the Commonwealth of Pennsylvania as a wastewater treatment plant operator. Member cf Water Pollution Control Association of Pennsylvania, Central Section,1980. July 1977 - July 1978 Wastewater Treatment Plant Onerator - Borouch of Lemovne. Lemovne. Pennsylvania Wastewater treatment plant operator at 2.0 million gallon per day secondary, activated sludge, wastewater treatment plant. ' Performed tasks as assignea by supervisors, including simple physical and chemical tests on wastewater streams, maintenance and operation of plant equipment, and maintenance of the collection system. September 1976 - June 1977 Science Teacher - West Shore School District. Camn Hill. Pennsylvania Taught Earth and Space Science at ninth grade level. Developed and implemented new course l materials on plate tectonics, environmental geology, and space science. Served as Assistant Coach of the district gymnastics team. September 1975 - June 1976 Science Teacher - Carlisle Area School District. Carlisle. Pennsylvania Taught Earth and Space Science and Environmental Science at ninth grade level. Developed and implemented new course materials on plate tectonics, environmental geology, noise pollution, water pollution, and energy. Served as Advisor to the Science Projects Club.

l f l EDUCATION. B.S., Education, majors in Earth and Space Science and General Science, minor in Environmental Education, Shippensburg State College, Shippensburg, Pennsylvania,1975. Graduate coursework in Land Use Planning, Shippensburg State College, Shippensburg, Pennsylvania,1977-1978. PUBLICATIONS:

1. " Determining Mercalli Intensities from Newspaper Reports," Journal of Geolocical Education.

Vol. 25,1977.

2. A Criticue of: An Independent Assessment of Evacuatica Times for Three Mile Island Nuclear Power Plant. Three Mile Island Public Interest Resource Center, Harrisburg, Pennsylvania, January 1981.

3. A Brief Review and Critiaue of the Rockland County Radiological Emercency Frenaredness Eign, Union of Concerned Scientists, prepared for Rockland County Emergency Planninr; Personnel and the Chairman of the County Legislature, Washington, D.C., August 17,1981.

4. The Necessity for a Promnt Public Alertine Canability in the Plume Exposure Pathway EPZ at Nuclear Power Plant Sites. Union of Concerned Scientists, Critical Mass Energy Project, Nuclear Information and Resource Service, Environmental Action, and New York Public In-terest Research Group, Washington, D.C., August 27;1981. *

5. " Union of Concerned Scientists, Inc., Comments on Notice of Proposed Rulemaking, Amendment to 10 CFR 50, Appendix E, Section IV.D3," Union of Concerned Scientists, Washington, D.C., October 21,1981. *

6. "The Evolution of Emergency Planning Rules,"in The Indian Point Book: A Briefinc on the Safety Investigation of the Indian Point Nuclear Power Plants. Anne Witte, editor, Union of Concerned Scientists (Washington, D.C.) and New York Public Interest Research Group (New York, NY),1982.

7. " Union of Concerned Scien:ists Comments, Proposed Rule,10 CFR' Part 50, Emergency Planning and Preparedness: Exercises, Clarification of Regulations,46 F.R. 61134," Union of Concerned Scientists, Washington, D.C., January 15,1982. *

8. Testimony of Robert D. Pollard and Steven C. Sholly before the Subcommittee on Energy and the Environment, Committee on Interior and Insular Affairs, U.S. House of Representatives, )

Middletown, Pennsylvania, March 29,1982, available from the Union of Concerned Scientists.

9. " Union of Concerned Scientists Detailed Comments on Petition for Rulemaking by Citizen's Task Force, Emergency Planning,10 CFR Parts 50 and 70, Docket No. PRM-50-31,47 F.R.

12639," Union of Concerned Scientists, Washington, D.C., May 24,1982.

10. Supplements to the Testimony of Ellyn R. Weiss, Esq., General Counsel, Union of Concerned >

Scientists, before the Subcommittee on Energy Conservation and Power, Committee on Energy l

and Commerce, U.S. House of Representatives, Union of Concerned Scientists, Washington, D.C., August 16,1982.

11. Testimony of Steven C. Sholly, Union of Concerned Scientists, Washington, D.C., on behalf of the New York Public Interest Research Group, Inc., before the Special Committee on Nuclear Power Safety of the Assembly of the State of New York, hearings on Legislative Oversight of the Emergency Radiologic Preparedness Act, Chapter 708, Laws of 1981, September 2,1982.

12. " Comments on ' Draft Supplement to Final Environmental Statement Related to Construction and Operation of Clinch River Breeder Reacto.- Plant'," Docket No. 50-537, Union of Concerned Scientists, Washington, D.C., September 13,1982. *

13. " Union of Concerned Scientists Comments on ' Report to the County Commissioners', by the Advisory Committee on Radiological Emergency Plan for Columbia County, Pennsylvania,"

Union cf Concerne.d Scientists, Washington, D.C., September 15,1982.

14. " Radiological Emergency Planning for Nuclear Reactor Accidents." presented to Kernenergie Ontmanteld Congress, Rotterdam, The Netherlands, Union of Concerned Scientists, Washington, D.C., October 8,1982.

15. " Nuclear Reactor Accident Consequences: Implications for Radiological Emergency Planning,"

presen.ed to the Citizen's Advisory Committee to Review Rockland Cour.ty's Own Nuclear Evacuation and Preparedness Plan and General Disaster Preparedness Plan, Union of Concerned Scientists, Washington, D.C., November 19,1982. i

16. Testimony of Steven C. Sholly before the Subcommittee on Oversight and Investigations, '

Committee on Interior and Insular Affairs, U.S. House of Representatives, Washington, D.C, Union of Concerned Scientists, December 13,1982,

17. Testimony of Gordon R. Thompson and Steven C. Sholly on Commission Question Two, Contentions 2.1(a) and 2.1(d), Union of Concerned Scientists and New York Public Interest Research Group, before the U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board,in the Matter of Consolidated Edison Company of New York (Indian Point Unit 2) and the Power Authority of the State of New York (Indian Point Unit 3), Docket Nos. 50-247-SP and 50-286-SP, December 28,1982. *

18. Testimony of Steven C. Sholly on the Consequences of Accidents at Indian Point (Commission Question One and Board Question 1.1, Union of Concerned Scientists and New York Public Interest Research Group, before the U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board, in the Matter of Consolidated Edison Company of New York (Indian Point Unit 2) and the Power Authority of the State of New York (Indian Point Unit 3), Docket Nos.

50-247-SP and 50-286-SP, February 7,1983, as corrected February 16,1983. *

19. Testimony of Steven C. Sholly on Commission Question Five, Union of Concerned Scientists and New York Public Interest Research Group, before the U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board, in the Matter of Consolidated Edison Company of New York (Indian Point Unit 2) and the Power Authority of the State of New York (Indian Point Unit 3), Docket Nos. 50-247-SP and 50-286-SP, March 22,1983. *

20. " Nuclear Reactor Accidents and Accident Consequences: Planning for the Worst," Union of Concerned Scientists, Washington, D.C., presented at Critical Mass '83, March 26,1983.

4

l l l

21. Testimony of Steven C. Sholly on Emergency Planning and Preparedness at Commercial

                    .                   Nuclear Power Plants, Union of Concerned Scientists, Washington, D.C., before the Subcommittee on Nuclear Regulation, Committee on Environment and Public Works, U.S.

Senate, April 15,1983, (with "Unics of Concerned Scientists' Response to Questions for the Record from Senator Alan K. Simpson," Steven C. Sholly and Michael E. Faden).

22. "PRA: What Can it Really Tell Us- About Public Risk from Nuclear Accidents?," Union of Concerned Scientists, Washington, D.C., presentation to the 14th Annual Meeting, Seacoast Anti-Pollution League, May 4,1983.

23. "Probabilistic Risk Assessment: The Impact of Uncertainties on Radiological Emergency Planning and Preparedness Considerations," Union of Concerned Scientists, Washington, D.C.,

June 28,1983. 24, " Response to GAO Caesdons on NRC's Use of PRA," Union of Concerned Scientists, Washington, D.C., October 6,1983, attachment to letter dated October 6,1983, from Steven C. Sholly to John E. Bagnulo (GAO, Washington, D.C.).

25. The Imnact of " External Events" on Radiological Emercency Resnonse Plannine Considerations. Union of Concerned Scientists, Washington, D.C., December 22, 1983, attachment to letter dated December 22,1983, from Steven C. Sholly to NRC Commissioner James K. Asselstine.

26. Sizewell'B' Public Inquiry, Proof of Evidence on: Safety and Waste Manacement Implications of the Si7ewell PWR. Gordon Thompson, with supporting evidence by Steven Sholly, on behalf of the Town and Country Planning Association, February 1984, including Annex G, "A review of Probabilistic Risk Analysis and its Application to the Sizewell PWR," Steven Sholly and Gordon Thompson, (August 11, 1933), and Annex O, " Emergency Planning in the UK and the US: A Comparison," Steven Sholly and Gordon Thompson (October 24,1983).

27. Testimony of Steven C. Sholly on Emergency Planning Contention Number Eleven, Union of Concerned Scientists, Washington, D.C., on behalf of the Palmetto Alliance and the Carolina Environmental Study Group, before the U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board, in the Matter of Duke Power Company, et. al. (Catawba Nuclear Station, Units 1 and 2), Docket Nos. 50-413 and 50-414, April 16,1984. *

28. " Risk Indicators Relevant to Assessing, Nuclear Accident 1.iability Premiums," in Preliminary Report to the Independent Advisory Committee to the NAIC Nuclear Risk Task Force.

December 11,1984, Steven C. Sholly, Union of Concerned Scientists, Washington, D.C.

29. " Union of Concerned Scientists' and Nuclear Information and Resource Service's Joint Comments on NRC's Proposal to Bar from Licensing Proceedings the Consideration of Earthquake Effects on Emergency Planning," Union of Concerned Scientists and Nuclear Information and Resource Service, Washington, D.C., Diaac Curran and Ellyn R. Weiss (with input from Steven C. Sholly), February 28,1985.

• l I

l

30. " Severe Accident Source Terms: A Presentation to the Commissioners on the Status of a Resiew of the NRC's Source Term Reassessment Study by the Union of Concerned Scientists," Union of Concerned Scientists, Washington, D.C., April 3,1985 *

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31. " Severe Accident Source Terms for Light Water Nuclear Power Plants: A Presentation to the

- Illinois Department of Nuclear Safety on the Status of a Rcview of the NRC's Source Term Reassessment Study (STRS) by the Union of Concerned Scientists," Union of Concerned Scientists, Washington, D.C., May 13,1985.

32. The Source Term Debate: A Review of the Current Basis for Predictine Severe Accident Source Terms with Soecial Emohasis on the NRC Source Term Reassessraent Procram (NUREG-0956L Union of Concerned Scientists, Cambridge, Massachusetts, Steven C. Sholly and Gordon Thompson, January 1986.

33. Direct Testimony of Dale G. Bridenbaugh, Gregory C. Minor, Lynn K. Price, and Steven C.

Sholly on behalf of State of Connecticut Department of Public Utility Control, Prosecutorial Division and Division of Consumer Counsel, regarding the prudence of expenditures on Millstone Unit III, February 18,1986. l

34. Implications of the Chernobyl-4 Accident for Nuclear Emergency Planning for the State of New l I

York, prepared for the State of New York Consumer Protection Board, by MHB Technical Associates, June 1986.

35. Review of Vermont Yankee Containment Safety Study and Analysis of Containment Ventine Issues for the Vermont Yankee Nuclear Power Plant. prepared for New England Coalition on Nuclear Pollution, Inc., December 16,1986. j

36. Affidavit of Steven C. Sholly before the Atomic Safety and Licensing Board, in the matter of Public Service Company of New Hampshire, et al., regarding Seabrook Station Units 1 and 2 Off-site Emergency Planning Issues, Docket Nos. 50-443-OL & 50-444-OL, January 23,1987.

37. Direct Testimony of Richard B. Hubbard and Steven C. Sholly on behalf of California Public Utilities Commission, regarding Diablo Canyon Rate Case, PG&E's Failure to Establish Its Committed Design OA Program, Application Nos. 84-06-014 and 85-08-025, Exhibit No.10,935, March,1987.

38. Testimony of Gregory C. Minor, Steven C. Sholly et. al. on behalf of Suffolk County, regarding LILCO's Reception Centers (Planning Basis), before the Atomic Safety and Licensing Board, in the matter of Long Island Lighting Company, Shoreham Nuclear Power Station Unit 1, Docket No. 50-322-OL-3, April 13,1987.

39. Rebuttal Testimony of Gregory C. Minor and Steven C. Sholly on behalf of Suffolk County regarding LILCO's Reception Centers (Addressing Testimony of Lewis G. Hulman), Docket No. 50-322-OL-3, May 27,1987.

40. Review of Selected Aspects of NUREG-1150, " Reactor Risk Reference Document," prepared for the Illinois Department of Nuclear Safety by MHB Technical Associates, September 1987.

41. Direct Testimony of Richard B. Hubbard and Steven C. Sholly on behalf of the Pennsylvania I

Office of Consumer Advocate, before the Pennsylvania Public Utility Commission, Evaluation of Beaver Valley Unit 2 Plant Costs, OCA Statement 6, Docket No. R-870651, October 23,1987.

42. Final Report: SitniGcant Factors Affectinc the Cost of Beaver Vallev Power Station. Unit 2.

prepared for Pennsylvania Office of Consumer Advocate, by MHB Technical Associates, OCA Exhibit 6A, October 1987.

         *         . 43. Surrebuttal Testimony of Richard B. Hubbard and Steven C. Sholly before the Pennsylvania
         '               Public Utility Commission, on behalf of the Pennsylvania Office of Consumer Advocate, regarding Evaluation of Beaver Valley Unit 2 Plant Costs, OCA Statement 6-1, Docket No. R-870651, December 7,1987.

44. Testimony on Diablo Canyon Rate Case, Desien Ouality Assurance. Supplemental and Rebuttal Testimony of Richard B. Hubbard and Steven C. Sholly, on behalf of the California Public Utilities Commission, Division of Ratepayer Advocates, Application Nos. 84-06-014 and 85 025, Exhibit No.16,690, September 1988.

45. Testimony on Diablo Canyon Rate Case, Evolution of OA Requirements And Their Understanding By The Nuclear Industrv Ouality Assurance As A Manacement Tool. Volumes I and II, Supplemental and Rebuttal Testimony of Richard B. Hubbard and Steven C. Sholly on be"nalf of the California Public Utilities Commission, Division of Ratepayer Advocate, Application Nos. 84-06-014 and 85-08-025, Exhibit No.16,650, September 1988.

• Available from the U.S. Nuclear Regulatory Commission, Public Document Room, Lobby,1717 H Street, N.W., Washington, D.C.

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VERMONT. YANKEE - 9 Proposed Change No. 133-NUCLEAR POWER CORPORATION . I I

                                                         ' RO 5. Box 169. Ferry Road, Brattleboro. VT 05301'                  y,g,                       .

ENGINEERING OFFICE L ] 1671 WORCESTER ACAD' j v F R AMINGH AM. M ASS ACHUSETT S 01701 {

                                                                                                                      ' TELE PMoNE617 872 4'00 -           1 April 25, 1986                                                          .f TVY 66-34 F

United States Nuclear Regulatory Commission Washington, DC 20555 t, Attention: Office of Nuclear Reactor Regulation Mr. H. R. Denton, Director

References:

(a) License No. DPR-28 (Docket No. 50-271) (b) Letter, USNRC to VYNPC, Amendment No. 37, dated 3 September 15, 1977

Subject:

Proposed Technical Specification Change for Spent and New Fuel

                                                 ' Storage

Dear Sir:

Pursuant to Section 50.59 of the Commission's Rules and Regulations, vennont Yankee Nuclear Power Corporation hereby proposes the following change to Appendix A of the operating license. Proposed Change Replace Page 189 of the Vermont Yankee Technical Specifications with the enclosed revised Page 189. This proposed change will revlse Section 5.5,

                -              " Spent and New Fuel' Storage," of the Vermont' Yankee Technical Specifications to increase the number of spent fuel assemblies allowed to be stored in the spent fuel pool.

• Reason for Change l

Vermont Yankee's (VY) spent fuel storage pool was originally designed and l licensed on the basis that a fuel cycle would be in existence that would only require storage of spent fuel for a year or two prior to shipment to a reprocessing facility. As the reactor core for VY contains 368 fuel assemb)les with approximately 92 being replaced on an annual refueling schedule, a fuel storage capacity of 600 assemblies was considered adequate. In September 1977 (Reference (b)), VY received a license amendment allowing for the phased increase of its spent fuel storage pool capacity from 600 to 2,000 assemblies. This would have permitted VY to operate and maintain full core reserve discharge capability until 1990. At the time this license

• United States Nuclear Regulatory Commission April 25, 1986 page.2 Attention: Mr. H. R. Denton, Director amendment was granted, it was fully anticipated that away-from-reactor storage would. be available during the 1980's to compliment reactor pool storage. ,

Thus, VY anticipated shipping spent fuel of f-site to maintain full core l reserve discharge capability. However, in 1981, the federal government announced that it intended to discontinue funding the away-from-reactor storage program, and utilities were given a clear mandate by the Department of j Energy to develop their own storage programs.  ! This policy was not affected by the passage of the Nuclear Waste policy Act of 1982. Although the Act provides for limited away-from-reactor storage, it states that all other alternatives must be exploited before federal storage will be made available. However, the Act did stipulate that a spent fuel repository will be available by 1998. Since the Act does not require a 2 repository before this date, it is very doubtful that thereTherefore, will be any place VY has

    '      to ship spent fuel in the 1980's or early to mid-1990's.                                                        )

decided to further expand its existing. spent fuel storage capacity in order to maintain full core reserve discharge capability eeyond 1990. Because Section 5.5 of VY's Technical Specifications currently limits the number of spent fuel assemblies allowed to be stored in the spent fuel pool, an amendment to this licensed storage capacity is required. Basis for chant.e VY evaluated the available alternatives to augment its current storageIn capacity within the context of the Nuclear Waste policy Act of 1982. pertinent part, Section 131 of the Act states that, "The persons owning and operating civilian nuclear power reactors have the primary responsibility for providing interim storage of spent nuclear fuel from such reactors, by maximizing, to the extent practical, the effective use of existing storage f facilities at the site of each civilian nuclear power reactor, and byFurther, adding new on-site storage capacity in a timely manner where practical." Section 132 of the Act states that, "The Secretary (USDOE), the Commission (USNRC) and other authorized Federal officials shall each take such actions as such official considers necessary to encourage and expedite the effective use of available storage, at the site of each civilian nuclear power reactor..." i The following alternatives to increasing spent fuel storage capacity at VY were considered: storage

1. Shipment to another reactor site or Away-From-Reactor (AFR) facility;

                                                                                                                           )

Modifying the plant fuel management plan to reduce the spent fuel

2. '

generation rate; and

3. Increasing on-site storage.

The option of off-site shipment of fuel to another reactor site or an AFR i ' storage facility was considered, but determined not to beFurther, feasible the due to the unavailability of an off-site storage site or facility. 1 provision of the Nuclear Waste policy Act of 1982 which sets a target date of 1998 for operation of a waste repository pree'uies any considet. tion of shipping VY spent fuel off-site to a repository prior to maximizing on-site

     .      storage.

4 United States Nuclear Regulatory Commission April 25, 1986 Mr. H. R. Denton, Director Page 3

      '          Attention:

I The currently proposed fuel management plan at VY is to increase design fuel burnup beginning in 1988, thereby slightly decreasing the number of spent fuel assemblies discharged per year to the SFP. However, this plan will not alleviate the need for additional storage capacity. The following methods for increasing on-site storage were considered:

1. Pin Consolidation;

2. Independent Dry (Cask, Drywell and Concrete Silo) Storage;

3. Independent Wet (Water Pool) Storage Pool;

4. Independent Air-Cooled Vault Storage;.and

5. Reracking With High Density Storage Racks.

6, With the exception of Reracking, the above alternatives have not Since i previously been fully licensed for commercial power plants by the 'NRC. additional spent fuel storage has to be in place at VY by 1987,'it is not considered prudent to select a storage option that has not been previously licensed due to uncertainties in the ability to license such methods and uncertainties concerning the licensing schedule. In addition, the above unlicensed options have, in general, not been demonstrated on other than a theoretical or prototype basis, adding to the uncertainty concerning the schedule for design and construction. Also, the Act requires that reactor t licensees utilize previously licensed technologies for maximization of on-sii

        .'       storage.

In view of the above considerations and schedular constraints, increasing on-site storage capacity by replacing existing freestanding racks with a similar proven design to allow closer spacing of the fuel Therefore, assembliesinwas order to concluded to be the only practical alternative for VY. maintain full core reserve discharge capability until the federally mandated repository is available in 1998, VY chose to replace all existing fuel racks i with high density racks. The r.ew racks are capable of storing 2,870 f ' assemblies which is sufficient capacity 'to maintain full core reserve l discharge until 1999. Safety considerations l This proposed change does not present an unreviewed safety question, as defined in 10CFR50.59. VY's spent fuel pool storage expansion method consists of replacing  ; existing racks with a proven design to allow closer spacing of the fuel assemblies. Only proven, well developed,' and demonstrated technology is utilized in both the construction process and in the analytical techniques applied to the expansion. We have evaluated the physical and mechanical processes which may create potential hazards such as criticality long-term considerations, seismic and mechanical loading, pool cooling, l corrosion and oxidation of fuel cladding, and probabilities and consequences Also, the of various postulated accidents and f ailures of decayed spent fuel. neutron poison and rack structural materials were evaluated.

                                                                                                - - - - - - . . _ _ _ _ _ _ - _ _ _ _ )

                                                                                                                                      ~
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                                                        <,y      .,                                                                              a 4            '

N I e April 25. 1986 - United' States Nuclear Regulatory Commission-

                     ~

L in , Mr..Hi R. Denton, Director -Page 4c l

                                          - Attention:

1 '4

                                                                                                     -                                          1
                                                                                                                                            . j 1
                                                    'No facility modifications other than the replacement of the spent: fuel-storage racks and the shortening'of the two cooling water return spargerflines. wn were determined to be necessary.         The design'of" Vermont-Yankee's spent fuel   t pool.is such that no fuel in.the spent fuel; storage racks can be uncovered:in the event of a" failure of the reactor cavity. seal or:the failure of piping associated with the spent fuel storage system or the reactor vessel.

In general,' potential safety hazards associated with spent' fuel pool-expansions are.not as large as those associated with. reactor operation because the purpose of the expansion is to' allow longer term storage of aged spent . fuel. . The VY expansion request is to allow continued storage of spent fuel that has decayed over a decade along with the normal discharge Af terofarelatively year of ' new spent fuel for which:the' pool was originally designed. storage, the majority of both' the initial radioactivity. and heat load. have -

                              .               decayed.
                                -                  . The design of the spent fuel storage racks provides for's suberitical                    ;4 multiplication factor-(k.gg) which was analytically demonstrated to be less                        I than the criticality criterion cf 0.95 for both normal and abnormal storage .

conditions. Normal conditions exist when the fuel storage racks are. located-at the bottom of the pool covered with a normal depth of water for radiation shielding and with'the maximum number of* fuel assemblies in their design ~ storage position. ~ Abnormal conditions may result from external events (such-as an earthquake) or failure of an engineered system (such as the accidental

                             '                dropping of an assembly).
                                                     . criticality calculations were, performed using a two-dimensional,-

0 Water.

' '                                           four-group, dif fusion theory code with a water temperature of. 68 F.

temperatures of 1500F.and 2000F were analyzed to assure that 680F was the more reactive temperature under norma 1' conditions. Monte Carlo calculations demonstrated the adequacy'of the diffusion theory representation. i

                                                  . Analyses were performed to verify that the existing spent fuel pool cooling system can maintain fuel pool temperatures within the- required range under all postulated fuel pool' loading conditions and that' natural circulation is sufficient to remove decay heat and prevent local boiling in the high density racks.

Calculated stress in a fully loaded rack will' not exceed 'the specified requirements of the Standard Review plan Section 3.8.4 when subjected to seismic loadings. Each rack module is a free-standing module that satisfies 1 the seismic design requirements without mechanical dependence on neighboring modules or fuel pool walls for support. The rack modules are classified as j Seismic Category I equipment. Racks of similar design have been licensed for other nuclear faci Fcies. The capacity of the existing fuel pool structure is well abdve the load Both the design condition specified by imposed by the fully loaded racks. Standard Review planning Section 3.8.4 and the design requirements of ACI 349-80 were used in.the calculation of fuel pool capacity. 1

1 United States Nuclear Regulatory Commission April 25, 1986 Mr. H. R. Denton, Director Page 5 Attention: All materials used in the construction of the racks are specified in accordance with the applicable ASME or equivalent ASTM specifications, and all welds in accordance with written procedures which meet the requirements of Section II of the ASME code. Materials selected are corrosion-resistant. This proposed change was previously reviewed and approved by the VY Nuclear Safety Audit and Review Committee. Significant Hazards Consideration The standards used to arrive at a determination + hat a request for amendment involves no si6nificant hazards consideration are included in the Commission's regulations. Specifically, 10CFR50.92 states that a proposed i' amendment will involve a no-significant hazards consideration if the proposed , I amendment does not: (1) involve a significant increase in the probability or create the possibility consequences of an accident previously evaluated; (2) of a new or different kind of accident from any accident previously evaluated; In addition, or (3) involve a significant reduction in a margin of safety. the Commission has provided guidance concerning the application of standards l l for determining whether a significant hazards consideration exists by providing certain examples of amendments that are considered likely, and These examples not were } likely, to involve significant hazards considerations. republished in the Federal Register on, March 6, 1986 (51 FR 7744, " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule), and included the following new example of an amendmenc which the Commission considered not likely to involve significant hazards considerations, "an expansion of the storage capacity of a spent fuel pool when all of the following are satisfied: (1) the storage expansion method consists of either replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies or placing additional racks of j the original design on the pool floor if space permits; (2) the storage (3) the expansion method does not involve rod consolidation or double tiering; l Ke rr fothe pool is maintained less than or equal to 0.95 and (4) no new

          *~                  technology or unproven technology is utilized in either the construction
              -               process or the analytical techniques necessary to justify the expansion."

The discussion below addresses the three 10CFR50.92 standards and summarizes VY's technical evaluation of the proposed increase in spent fuel storage capacity in relation to each. Our evaluation of the proposed plant modifications and operations in support of the amendment request is contained in the enclosed Vermont Yankee Spent Fuel Storage Rack Replacement Report. First Standard Vermont Yankee has determined that the proposed change to increase the spent fuel pool capacity does not involve a significant increase inVY's the safety probabilltf or consequences of an accident previously evaluated. analysis of the proposed reracking has been accomplished using current NRC Staff-accepted codes and standards. The results of the safety analysis demonstrated that the proposal meets the specified acceptance criteria set forth in these standards. In addition, VY has reviewed NRC Staff Safety Evaluation Reports for prior spent fuel pool rerackings involving spent fuel _--_--____-__-_____a

April 25, 1986

                                 ~

United States Nuclear Regulatory Commission page 6 Attention: Mr. H. R. Denton, Director l I pool rack replacements to ensure that there are no identified concerns not fully addressed in this schmittal. VY has identified no such concerns. VY's proposed storage expansion method consists of replacing existing freestanding racks with a similar proven design to allow closer spacing of fuel assemblies within the existing pool. No new technology or unproven technology is utilized in either the construction process or in the analytical techniques applied to the expansion. Vermont Yankee has performed nuclear, thermal-hydraulic, mechanical, structural and radiological analysesTheseofinclude normal and abnormal conditions which could create potential hazards. criticality considerations, seismic and mechanical loading, pool cooling, long-term corrosion and oxidation of fuel cladding, and the probabilities and consequences of postulated accidents and failures of decayed spent fuel. Additionally, the neutron poison and rack structural materials were evaluated The probability and and shown to be conpatible with the pool environment. occurrence of potential abnormal conditions and accident scenarios initiated either by external events (such as a seismic event) oc by failure of an engineered system (such as dropping a fuel assembly) are not affected by the racks themselves; thus, the reracking does not increase the probability of The radiological consequences of these these conditions and accidents. events, as well as the probability and radiological consequences of criticality or installation accidents, were evaluated and all previously analyzed accidents and consequences were found to be conservatively bounded.

• Second Standard VY has determined that the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated. VY has evaluated the proposed rack replacement in accordance with the NRC position paper, "NRC position for Review and Acceptance of Spent Fuel Storage and Handling Application," as well as appropriate NRC Regulatory Guides, appropriate NRC Standard Review plan sections and appropriate industry codes and standards. In addition VY has reviewed the NRC Safety Evaluation Report for the previous VY spent fuel 'The rack replacement application and for other prior spent fuel pool rerackings.

proposed storage eKpansion method consists of replacing existing racks with a i previously approved and proven design which allows closer spacing between stored spent fuel assemblies. Additionally, the storage expansion method does not involve rod consolidation or double tiering and no new technology or unproven technology is utilized in either the construction process Further,orthethe basic analytical techniques necessary to justify the expansion. reracking technology to be used has been developed and demonstrated in numerous applications for fuel pool capacity increases which have previously received NRC staff approval. All credible accidents and consequences evaluated have been found to be conservatively bounded and no new categories or types of accidents have been identified. Third Standard VY has determined that the pecposed change does not involve a significant reduction in a margin of safety. The issue of " margin of safety" when applied to a reracking modification, includes the following considerations:

O United States Nuclear Regulatory Commission April 25, 1986 page 7 Attention: Mr. H. R. Denton, Director

a. Nuclear criticality cotisiderations,

b. Thermal-hydraulic considerations,

c. Mechanical, material and structural considerations.

The margin of safety that has been established for nuclear criticalityin considerations is that the ef fective neutron multiplication factor (k gg) e the spent fuel pool is to be less than or equal to 0.95, including all The criticality reasonable uncertainties and under all postulated conditions. analysis for the proposed modification, described in the enclosed Reracking Report, concluded that for all bounding normal and abnormal storage conditions analyzed, the suberitical multiplication factorThe(k gg) e was verified techniques used to tocalculate be

                           less than the criticality criterion of 0.95.

have been benchmarked against experimental data and are considered very keff The NRC Staff determined in 1976 that as long as the maximum value reliable. of the effective neutron multiplication factor k,gg was less than or equal to 0.95, then any change in pool reactivity would not significantly reduce The methode the i margin of safety, regardless of the storage capacity of the pool. used in the criticality analysis for the reracking conform to the applicable portions of Codes, Standards and Specifications listed in the Reracking I Report including ANSI N210-1976 " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear power Stations," ANSI N16.9-1975, " Validation of ,

               '             Calculation Methods for Nuclear Criticality Safety " the NRC guidance

• l document, "NRC position for Review and Acceptance of Spent Fuel Storago and Handling Applications," and Regulatory Guide 1.13, " Spent Fuel Facility Design

               .             Basis." proposed Revision 2. The computer programs, data libraries and benchmarking data used in the evaluation have been used in previous spent                      Thefuel reracking applications and have been reviewed and approved by the NRC.

criticality analysis for the reracking ascumed operation of the spent fuel The storage facilities consistent with the proposed Technical Specifications. results of these analyses indicate that k,gg is less than 0.?S at 95/95 probability / confidence level under all postulated conditions, including a

                           margin for uncertainties in reactivity calculations and mechanical tolerances. Thus, in meeting the acceptance criteria for criticality, the proposed reracking does not involve a significant reduction in the margin of safety for nuclear criticality.

The margin of safety that has been established for the thermal-hydraulic considerations is that fuel pool cooling be capable of maintaining spent fuel pool water temperatures below the boiling point for any postulated pool heat load. The thermal-hydraulic evaluation is described in the enclosed report. Analyses performed verify that the installed fuel pool cooling The maximum can maintain heat load spent fuel pool temperatures within the design limit. predicted for a full pool with the proposed reracking remains within theIt has also design capacity of existing equipment. the Spent Fuel pool Cooling System is lost for any reason, there isThus, sufficient the time and make-up capacity available to maintain pool water level. ' proposed reracking does not involve a significant reduction in any thermal-hydraulic margins of safety.

United States Nuclear Regulatory Commission April 25, 1986 page 8 Attention: Mr. H. R. Denton, Director The mechanical, material, and structural considerations of the proposed rack replacement are also analyzed in the enclosed report. The racks are designed in .secordance with applicable IIRC Regulatory Guides, Standard Review plans, position papers and appropriate industry codes and standards, as well as to Celsmic Category I requirements. All materials selected are corrosion-resistant. The materials utilized are compatable with the spent fuel pool and the spent fuel assemblies. The conclusion of the analysis is that the margin of safety is not significantly reduced by the proposed reracking. The main function 'of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a stable configuration through all normal and abnormal loadings, such as an earthquake and under accident conditions, Nuclear criticality, thermal-hydraulic, material and structural considerations of the proposed new racks are described in the enclosed report. The neutron poison and rack materials are compatible with materials used for the spent fuel pool liner and the spent fuel assemblies. The rack , j structural considerations address adequate margins of safety ofFurther, critical items the during seismic motion and the racks are seismically qualified. load of the fully leaded racks has been analytically demonstrated to be well within the fuel pool's structural capacity. Thus, the proposed increase in the storage capacity of the spent fuel pool does not involve any Therefore, VY's significant reductions in existing design limitations or safety margins. existing margins of safety are not significantly reduced by the proposed l expansion of pool storage capacity. In summary, VY's request to expand the spent fuel storage pool capacity satisfies the following conditions:

1. The storage expansion method consists of replacing existing racks with a design which allows closer spacing between stored spent fuel assemblies. l

2. The storage expansion method does not involve rod consolidation or double tiering.

3. The ke gf of the pool is maintained less than or equal to 0.95.

f 4. No new technology or unproven technology is utilized in either the construction process or the analytical techniques necessary to justify the expansion. On the basis of the above, VY has determined that operation of the facility in accordance with the proposed amendment does not involve in that it: a significant hazards consideration as defined in 10CFR50.92(c), (1) does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) does not create the possibility of a 1 new or different kind of accident from any accident previously evaluated; and (3) does not involve a significant reduction in a margin of safety. l Fee Determination In accordance with the provisions of 10CFR170.12, an application fee of

                  $150.00 is enclosed.                                                                                                                                I

_ _ - - _ _ _ _ _ - _ _ _ _ _ _ _ - _ ________L

April 25, 1986 United States Nuclear, Regulatory Commission Page 9 Attention: Mr. H. E. Denton, Director Schedule of Chante We request that your review and approval of this proposed change be completed no later than November 15, 1986 in order to insure that the change

       '                     is incorporated in the Vermont Yankee Technical Specifications prior to loss-of full core reserve.

This change will be incorporated into the Vermont Yankee Technical Specifications as soon as practicable following receipt of your approval. We trust that the information above adequately supports our request; however, should you have any questions in this matter, please contact us. Very truly yours, 4f VERMOWT YANKEE NUCLEAR POWER CORPORATION W f**1 Warren P Murphy Vice President and nager of Operations WPM /dps Enclosure ec: U. S. Nuclear Regutatory Commission *

• Document Control Desk (40 copies)

Vermont Department of Public Services 120 State Street I Montpelier, Vermont 05602 f Attention: Mr. Gerald Tarrant, Chairman ) STATE OF VERMONT)

                                                  )ss WINDHAM COUNTY )

Then personally appeared before me, Warren P. Murphy, who, being duly sworn, did state that he is a Vice President and Manager of Operatiurs of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing document in the name and on the behalf of

• Vermont Yankee Nuclear Power Corporation and that the statements therein are true to the best of his knowledge and belief.

qY  ? t & {\lj p ._ Notary Public Diane McCue My Commission Expires ) I I PUSU; bis 2 o s Y $UNT

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Pf esm> VERMONT YANKEE NUCLEAR POWER CORPORATION

                                     .                                                                               FVY 88-17 RD 5, Box 169, Ferry Road, Brattleboro. VT c53o1                y, p'                                                                             ENGINEERING OFFICE 1671 WORCESTER 60 AD FRAMINGHAM, MASsACHUS1 TTs 01701 TEi,EPHoNE 617 872 8100 March 2, 1988                ,
                        'O.S. Nuclear Regulatory Commission kishington, D.C.      20555 Attn:        Document Control Desk

References:

a) License No. DPR-28 (Docket No. 50-271) b) Letter, VYNPC to USNRC, FVY 86-34, " Proposed Technical Specification Change for New and Spent Fuel Storage", dated 4/25/86 c) Letter, VYNPC to USNRC, FVY 87-87, " Vermont Yankee Proposed Ch'ange No.133 - Spent Fuel Pool Expansion", dated 9/1/87. d) Letter, USNRC to VYNPC, NVY 88-05, " Forthcoming Meeting with Vermont Yankee Nuclear Power Station", dated 1/21/88 .,

Dear Sir:

l

Subject:

Vermont Yankee Proposed Change No. 133 - Spent Fuel Pool Expansion Pursuant to the NRC staff's letter o' January 21, 1988 [ Reference d)], a meeting was held on February 9,1988 during which Vermont Yankee responded to 3 the remaining NRC staff technical information requirements associated with the subject spent fuel pool expansion amendment request [ Reference b)]. In accor- . dance with the NRC staff's request, Attachments 1 and 2 to this letter pro- l vide the documentation and information presented by Vermont Yankee and requested by the NRC staff at the February 9,1988 meeting. In order to expedite the NRC staff's review of the subject license amend-ment request and definitively resolve all remaining staff review issues, and in an attempt to resolve the issues pending before the Atomic Safety and Licensing

    . jg'                ' Board, Vermont Yankee has committed to design, install, test and make opera-                                   l tional, a redundant seismically designed Spent Fuel Pool Cooling System prior to                               l
    .pfY                   the time Vermont Yankee exceeds the existing 2,000 spent fuel assembly storage                                 i limit in the Vermont Yankee spent fuel pool. This system will be operational no later than the end of Cycle 16 (Projected to be 1993). This commitment is reflected in Attachment 1 and 2. Attachment 1 specifies the design and perfor-mance criteria for the enhanced system. The design, installation and testing of                                l l

the enhanced system will be in accordance with 10 CFR 50.59 and the NRC's normal " inspection program. l L-____________________._____________________

L , VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Comission March 2, 1988 Page 2 i Attachment 2 to this letter documents the information presented by Vermont  ! Yankee at the February 9,1988 meeting which directly addressed each of the NRC l staff's remaining open technical issues as described in the' January 21, 1988 1 status. report [ Reference d)] of the staff's review of Vermont Yankee's spent I fuel. pool expansion amendment request. As documented in Attachment 2, each of the remaining. open technical issues is addressed for both the ex. sting Vermont Yankee . Spent Fuel Pool System and the proposed enhanced Spent Fuel Pool Cooling System. On the basis of the information' submitted in support of the subject amend-ment request since April 1986 and the comitments and information presented herein, Vermont Yankee requests that you expeditiously complete your review of J i the spent fuel pool expansion application allowing Vermont Yankee to rerack the spent fuel pool to 2,870 assemblies. I Very truly yours,  ! I l VERMONT YANKEE NUCLEAR POWER CORPORATION l Is [Y Warren P. Murphy l j Vice President and 1 Manager of Operations. 1 I l'

          /dm cc: Office of Nuclear Reactor Regulation Mr. Steven A. Varga, Director                                                                       ,

Division of Reactor Projects I/II l 1 l U.S.N.R.C. Region I , Mr. William T. Russell, Regional Administrator l U.S.N.R.C. Resident Inspector Vermont Yankee Nuclear Power Corporation j ASLB Service List I l 1

                                                                                                                    'l

___-____ _ _ a

ATTACHMENT 1 i Design and Performance Criteria for the Enhanced Vermont Yankee Spent Fuel Pool Cooling System Vermont Yankee has committed to providing spent fuel pool cooling capacity-via an Enhanced Fuel Pool Cooling System. The Enhanced Fuel Pool Cooling System will be designed and installed 'in accordance with Vermont Yankee's Operational Quality Assurance Program. I The functional and performance criteria for the system are as follows:

1. Cooling from spent fuel pool to ultimate heat sink will be available from Seismic Category 1 equipme.nt, independent of the RHR System.

2. System will be Safety Clus 3 and single active. failure proof.

o System will be designed to ensure that heat removal capacity assuming the maximum normal heat' load and a single active failure, will be suf-ficient to preclude any restriction on plant operation. The system-will also address the following in accordance with FSAR criteria and Technical Specifications:  ;

                                          -     Detection and isolation of leaks
                                          -     Flooding
                                         -      Missiles
                                         -      Inservice testing capability
                                          -     Fire Protection o            The spent fuel pool cooling pumps and all other essential electrical equipment will be environmentally qualified per the Vermont Yankee EQ Program, seismically qualified per the FSAR criteria and powered from safety class electrical emergency power sources.                                           ,

I o System layout and installation will take into account ALARA con-siderations in accordance with the Vermont Yankee ALARA Program. o Fuel pool temperature monitoring will be provided for all plant operating modes. . i The structural and mechanical design of the piping will be in accordance with ASME/ ANSI B31.1-1977, which is consistent with the Seismic Reanalysis Program and Recirculation System replacement designs. Seismic input will be the appropriate Reactor Building spectra, based on USNRC Regulatory Guide 1.60 and ASME Code Case N-411 criteria, as was approved by the NRC for the Seismic Reanalysis Program and the Recirculation System replacement. Material selection and processing will use NUREG 0313, Rev. 2, as guidance. However, the maximum operating temperature for the system is only 150'F, which l is below the temperature at which IGSCC is a concern. It is Vermont Yankee's policy to use IGSCC resistant material unless significant cost or schedule  ! penalties would result. l

[*l '

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                                                                                                                                         - Attaachment l'        "
      '                                                                                                                                               Page 2 I

fe i , 1

                       . The installation will be performed under ASME Section XI repair program using the Engineering Design Change Request (EDCR) process, as was used in the
             - Seismic Reanalysis Program and Recirculation System replacement project. -
                                                                                        '                                                                              t:".

Post 'nsta11ation. pressure testing.will be in accordance with.ASME/ ANSI q B31.1 for> isolable portions and ASME'Section-XI for portions of new. piping-un- - isolable from existing piping or . components. ~l

                      , Starti-up testing will be performed to ensure that the system meets spe-:                                                             

l cified performance criteria.

                    , . Vermont Yankee will continue. to comply with the Adminis'trative Guidelines -                                                                     4 as described in our September 1, 1987' submittal [ Reference c)] until such time                                                                           j that the Enhanced Fuel' Pool Cooling System'is operable.1                                                                                                      ,

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• A'.'TACHMENT 2 Vermont Yankee Response to NRC Staff Technical Issues Identified in the January 21, 1988 Status Report INTRODUCTION This attachment documents and expands upon the information presented by Vermont Yankee at the February 9,1988 public meeting and directly addresses each of the NRC staff's remaining open technical issues as described in the January 21, 1988 status report of the staff's review of Vermont Yankee's spent fuel pool expansion amendment request. Additionally, Vermont Yankee wishes to clarify two points with respect to the information contained in your letter of January 21, 1988 (NVY 88-05). Specifically, two items . discussed in the attach-ment under Section A, Background, should be corrected.

First, to date, Vermont Yankee has installed racks of the current design sufficient to store 1,690 fuel bundles, not 1,680 as stated. This discrepancy is due to a typographical error contained in Vermont Yankee's original amendment request submittal (FVY 86-34, dated April 25, 1986). Second, the proposed enhanced high density storage racks would increase the storage capacity of the spent fuel pool to 2,870 fuel bundles and are projected to provide storage capacity while maintaining full core reserve discharge capa-The date of 2001 was documented in a bility until 2001, not 1999 as stated.24, 1986 (FVY 86-107) in response to Question No. 10 letter dated November is a realistic projection based on Vermont Yankee's extended fuel cycle manage-ment plan (i.e., eighteen-month fuel cycles).

x ,< d 4 RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION OPEN ISSUES: No.1 HEAT REMOVAL CAPABILITY ITEMS: 1A. 1971 ANS DRAFT STANDARD USED 1B. 9.1.MBTU/HR USED AS HEAT LOAD-1C. FSAR LISTS 2.23 MBTU/HR 1D. SINGLE FAILURE

RESPONSE

1A. NOT USED FOR SFP DECAY HEAT; USED FOR REACTOR VESSEL DECAY HEAT 1B. CONSERVATIVE VALUE FOR A SPECIFIC . SCENARIO  ! 1C. DESIGN VALUE NOT ACTUAL PERFORMANCE CAPABILITIES l 1D. VY IS SINGLE ACT_IVE FAILURE PROOF (VY HEAT EXCHANGERS CAN BE CROSS CONNECTED) CONCLUSION: ITEMS A,B,C AND D FULLY ADDRESSED BY EXISTING 7 SFP SYSTEM HEAT REMOVAL CAPABILITY. PROPOSED UPGRADE ALSO FULLY ADDRESSES ITEMS B,C AND D. 1

                                                                               .._____ _ ___ _ _ __._____ _ ____ _ _ _ E

su""'--~- - - - - - - _ . - _ - _ . _ _ _ _ _ _ . , _ _ _ _ . _ __

                              /

6 DTSCUSSION 1A. Vermont Yankee has performed heat load calculations for fuel stored in i the spent fuel pool in accordance with the guidance of Standard Review Plan, Section 9.1.3. Reference to the 1971 ANS draft standard and 9.1 l MBtu/hr are specific only to the scenario described in Attachment 1 of the September 1, 1987 submittal. This scenario describes how torus cooling and spent fuel pool cooling can be accomplished by the RHR System only. The initial conditions established an operating reactor and recently discharged spent fuel (ten days). These conditions are essentially impossible to achieve since an actual refueling at Vermont Yankee could not be done in less than ten days, thus, these conditions establish a very con-servative analysis. The 1971 ANS draft standard was used in determining the heat load from the reactor vessel just after scram, not the SFP heat load. The heat load in the SFP was determined by using the SRP methodo-logy.

18. The 9.1 MBtu/hr is a conservative value picked because it is the hest transfer capability of one pump and two heat exchangers. It is not the maximum normal heat generation rate. Using SRP guidance, the analysis per-formed by Vermont Yankee shows the heat load at 150 hours (six days) is approximately 10.3 MBtu/hr, which is in agreement with the analysis done by the NRC staff.

1C. Table 10.5.1 of the Vermont Yankee FSAR lists the original design heat transfer rate of the heat exchanger. The original design heat transfer rate was used to purchase the equipment but does not limit the actual heat transfer rate of 2.23 MBtu/hr in the FSAR does not limit the heat exchangers performance to just 2.23 MBtu/hr. Based on the conservation of In energy, as the inlet parameters change so does the heat transfer rate. the Vermont Yankee SFP cooling analysis, the original heat exchanger data 1 sheet inlet parameters were analyzed as a bench mark. The analysis yielded the same cutlet parameters as listed on the heat exchanger data sheet, showing that the original design is just another point within the heat exchanger performance capabilities.

10. Single failure for Vermont Yankee is defined as " single active failure."

The definition is contained in SRP 9.1.3 and the response to Interrogatory 26; "NRC Staff Response to NECNP's First Set of Interrogatories andon this, Based Document Request to the NRC Staff", dated August 5, 1987. Vermont Yankee is single active failure proof with one pump in standby The and one pump operating with two heat exchangers operating in parallel. Vermont Yankee SPFCS piping is arranged in such a way to provide easy pump discharge cross connnection allowing the two heat exchangers to be operated in parallel (refer to FSAR Figure 10.5-1). Considering only one pump to be in operation, and the cross ccanection valve open, the piping to each heat exhanger is routed in such a manner to provide a relatively equal flow resistance. This provides a fairly equal division of flow from the

running pump to each heat exchanger. Figure No.1 sunnarizes the Vermont Yankee pool temperature capability for all combinations of existing SFPCS equipment. As can be seen from the one pump and two heat exchanger curve (middle curve), the pool temperature can be held to less than 150'F after 11 days of fuel decay. This figure is based on SRP heat load analysis methods that yield results comparable to heat loads calculated by NRC staff and attached to the "NRC Staff Response To NECNP's First Set Of Interrogatories And Document Request To The NRC Staff", dated August 5, 1987. Figure No. 2 makes a comparison of NRC staff and Vermont Yankee calculated heat loads at several days of decay. The heat loads at these points compare very well with each other, so it.can be concluded that Vermont Yankee calculated heat loads are not in disagreement with the NRC calculated heat loads. CONCLUSION The design of the existing Spent Fuel Pool Cooling System heat generation calculation methods, heat removal requirements, and single failure requirements comply with Standard Review Plan 9.1.3. The Vermont Yankee commitment of February 9, 1738 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore,.be qualified for a seismic event, be single active failure proof, and powered by a safety class electrical emergency power source. As such, reliance on the RHR System to provide seismic spent fuel pool cooling would not be necessary. The enhanced system would acceptably close Open Issue No.1 also, since it would meet the applicable SRP 9.1.3 requirements for single active failure and heat removal capabilities. l

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    'N HEAT LOAD COMPARISON NRC                          VY o            DAYS       HEAT LOAD         DAYS     HEAT LOAD 6.25'      10.17 MBTU/HR     6         10.35 MBTU/HR 6.92      9.91               7        9.93 r

7.92 9.58 8 9.59 8.92 9.31 9 9.32 l

            - 9.92      9.09               10       9.1 L

L 1 FIGURE #2 l  :

RESPONSE TO NRC QUESTIONS VY SPENT FUEL POOL EXPANSION 4 l

                                                                                                                                            ]

l OPEN ISSUES: < N o . a. SPENTFjJELPOOLTEMPERATURELIMIT ITEM 1 NO FUEL POOL TEMPERATURE MONITOR WITH SFPCS NOT OPERATING

                 ' RESPONSE:

POOL TEMPERATURE MONITORING 19 PROVIDED FOR ALL PLANT CONDITIONS AS DOCUMENTED IN VY LETTER 9-1-87 CONCLUSION: EXISTING VY POOL. TEMPERATURE MONITORING SYSTEM ACCEPTABLE AND ADEQUATE. , i PROPOSED SYSTEM WILL ADDITIONALLY PROVIDE SFP TEMPERATURE MONITORING-FOR ALL PLANT CONDITIONS. l l L- _ _ _ _ .__ _ _ . _ _ _ _ _

i l i l DISCUSSION As detailed in Vermont Yankee's letter of September 1,1987, spent fuel A pool temperature is continuously monitored when the system is in operation. Control Room alarm will sound when temperature exceeds an administrative limit of  ; 125*F. ' In recognition of the fact that the temperature monitors would not provide. accurate temperature indication of the Fuel Pool if the Fuel Pool Cooling System. was inoperable, Vermont Yankee has committed to directly monitor fuel pool tem-perature every four hours if one or both fuel pool cooling trains were ino-  ! perable (see Vermont Yankee letter, dated September 1,1987, Attachment 2) until  ! the enhanced Fuel Pool Cooling System is operable. ' Even at the maximum heat-up rate of 3*F/hr ample time would exist for operator action to secure the demi-neralizer before the inlet temperature exceeds the NRC imposed limit of 140*F. In the refueling mode, when the Spent Fuel Pool Cooling System components could be out of service for maintenance, the spent fuel pool and refueling cavity temperature is monitored by the Residual Heat Removal System temperature indicators. CONCLUSION Based on the above, Vermont Yankee concludes that appropriate temperature monitoring exists for all operating modes, satisfying the requirements of Standard Review Plan 9.1.3. The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SPFCS that meets the applicable requirements of SRP 9.1.3 would, therefore. pro-vide temperature monitoring under all plant conditions thus the enhanced 1 tem would acceptably close Open Issue No. 2 also, since pool temperature monitoring would be provided under all plant conditions.

                                                               - - - - - - - - - - --_-_-_m__ ___ _ _ _. ___           __ _

o 1 I n' [j j 1

o. q RESPONSE TO NRC QUESTIONS 1 l

VY SPENT FUEL POOL EXPANSION 4 OPEN ISSUES: g s o__ s. 0 P00L100LINGFOLLOWINGSEISMICEVENTS ITEM: 3A. FIRE WATER SYSTEM CONNECTION TO SERVICE WATER NOT SEISMIC C A T .'I

                 -3B. ALT. COOLING CELL PIPING SEISMIC CAT.I NOT DEMONSTRATED 3 C .' RHRSW TO RHR CROSS-CONNECT SEISMIC CAT.I                         '

NOT DEMONSTRATED

RESPONSE

SA. SWS PIPING IS SEISMIC CAT.I PER FSAR App.A THROUGH A NORMALLY CLOSED, MANUAL FIRE WATER SYSTEM ISO. VALVE 3B. PIPING IS SEISMIC CAT.I PER FSAR App.A 3C. PIPING IS SEISMIC CAT.I PER FSAR App.A CONCLUSION:

                 -EXISTING VY SW PIPING AND CROSS-CONNECTION ARE SEISMIC CATEGORY      I. PROPOSED FPC SYSTEM UPGRADE WILL BE SEISMIC CATEGORY                                    I.

d DISCUSSION 3A. The Fire Water System is not a seismically qualified system and is isolated from the Service Water Syrtem by a normally closed manual valve. The fire water piping and valve making the connection to the Service Water System is seismically qualified Category I in accordance with the Vermont Yankee FSAR, Appendix A to prevent degradation of the Service Water System in a seismic event. 3B. The cooling tower deep basin alternate cooling cell piping connecting to the Service Water System is seismically qualified as noted in the Vermont Yankee FSAR, Appendix A. 3C. The Vermont Yankee service water pumps are powered from an on-site The emergency electrical source and meet Seismic Category I requirements. Vermont Yankee service water path to the fuel pool meets Seismic Category I requirements from the service water pump suction in the river through the RHR service water connection into the RHR System and through the FPC con-nection into the spent fuel pool. CONCLUSION Based on the above, Vermont Yankee concludes that the existing service water piping and cross connections are Seismic Category I as described in Appendix A to the Vermont Yankee FSAR, Standard Review Plan, Section 9.1.3 is satisfied. The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event.

                                                   =            - _ - - - - -                         - _ _ -

i RESPONSE TO NRC QUESTIONS-

                             -VY SPENT FUEL POOL EXPANSION

{ OPEN ISSUES: N o . t+ RADIOLOGICAL CONSEQUENCES OF BOILING ITEM: 4A. PROVIDE ASSUMPTIONS FOR OFF-SITE DOSE 4B. PROVIDE ASSUMPTIONS FOR ON-SITE DOSE 1 4C; PROVIDE ON-SITE DOSE

RESPONSE

4A. ASSUMPTIONS CAN BE PROVIDED 4B. ASSUMPTIONS CAN BE PROVIDED 4C. ON-SITE DOSE CAN BE PROVIDED CONCLUSION: 10CFR2O REQUIREMENTS MET DURING POOL BOILING FOR OFF-SIT.E AND ON-SITE DOSES. PROPOSED NEW SYSTEM WILL PRECLUDE POOL BOILING FOR ALL PLANT CONDITIONS.

                                                                                                                                              )

I l DISCUSSION J In order to assess the on-site and off-site radiological impact of a postu- -) l lated boiling spent fuel pool, a scenario was developed to maximize the release of fission products through boiling.. The scenario assumes that the plant is i shut down for refueling with a normal 136 bundle' fuel load. The 136 spent fuel bundles discharged completely fill the spent fuel pool to its capacity of 2,870. f Just before the start of the outage, maximum Technical Specification. activity leve'Is; are assumed to be.present in the Reactor Coolant System while normal activity levels (as determined from Plant Chemistry data) are assumed to be pre-l l sent in the spent fuel pool water. The two volumes and their activities are then mixed at the start of the outage when the refucling gates are removed. The length of the outage is assumed to be 21 days. This is consistent with the shortest refueling outage in Vermont Yankee's history. At 21 days, the refueling gates are reinstalled, the fuel pool volume is segregated from the Reactor Coolant System volume, and the fuel pool is on spent fuel pool cooling. All spent fuel pool cooling is then assumed to be lost and the pool is allowed to heat up. The heat load in the pool is then determined based upon Standard i Review Plan methodology and uncertainties assuming the spent fuel operated at i 1,665 MWt with a 100% capacity factor. The rate of heat up and subsequent boiling were then determined assuming a spent fuel pool bulk temperature of 150*F when all cooling is lost. The resulting boil-off rate is then calculated as a function of time. Only the maximum boil-off rate is used in the radiologi-cal calculations. The following assumptions were used to evaluate the on-site and off-site radiological conditions resulting from a spent fuel pool boiling incident. ASSUMPTIams FOR SOURCE TERM

1. Constant maximum boil-off rate if 16.6 gpm.

2. The volatile elements in the spent fuel pool (iodine, tritium) are released during boiling.

3. Tritium concentration in water equals 2 x 10-2 micro Ci/ml.

4. Initial concentration of I-131 dose equivalent (DE) at minimum detectable level 00L) (i.e., 4 x 10-7 micro Ci/ml) in spent fuel pool.

5. At t=0, reactor coolant at long-term Technical Specification limit of 1.1 micro Ci/mi 1-131 DE.

6. Partition Factor (PF) of 100 for iodine during boiling. Based on SRP 15.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR),

establishes a partition factor of 100 between the steam generator water and ' steam phases. l l

l f 1

7. No credit taken for fuel pool cleanup of iodine via the spent fuel pool demineralizers prior to initiation of pool boiling.

8. Recovery operations restore spent fuel pool cooling viithin thirty days.

ASSUMPTIONS FOR OFF-SITE CONDITIONS

1. Blow out panels are not present; ground level unfiltered release assumed.

2. Maximum off-site accident X/Q value for release from Reactor Building equals 6.83 x 10-4 sec/m3

3. For tritium, adult dose conversion factor and breathing rate which combine for most conservative dose rate (from Regulatory Guide 1.109).

Adult Inhalation Oose Factor = 1.58 x 10-7 mrem /pCi inhaled 0 8,000 m3 s year. /

4. For 1-131, infant thyroid dose conversion facter is most conservative 1.62 R/hr uCi/m3 (EPA-520/1-75-001)

(Infant Breathing Rate of 2.5 x 10-5 m3/sec) , RESULTS OFF-SITE I0 DINE Maximum I-131 DE off-site concentration = 1.5 x 10-10 micro Ci/cc. Maximum dose rate thyroid = 2.5 x 10-1 mrem /hr. 30 days dose at maximum rate = 1.8 x 102 mrem or 0.06% of Part 100 Limit (300 rem) and 6% uf Part 20 Limit (3 rem /yr implied). RESULTS OFF-SITE TRITIUM Maximum tritium concentration = 1.4 x 10-8 micro Ci/cc. Maximum adult whole body dose rate = 2.1 x 10-3 mrem /hr. 30 day dose at maximum rate = 1.5 mrein or 0.01% of Part 100 Limit (25 rem) and 0.3% of Part 20 Limit (0.5 rem /yr). ASSUMPTIONS FOR ON-SITE CONDITIONS

1. For on-site evaluation, assume the activity is released into a closed volume (blow-oat panels remain intact) equivalent to the top floor of the Reactor Building.

2. Assume 90'F and 100% relative humidity in the SFP aree as a result of boiling (for tritium concentration).

p p. E

3. Assume the concentration of tritium in the water vapor in the SFP area is the same as the concentration in the SFP water (2 x 10-2 micro Ci/ml).

4. Allow the iodine to be released into a closed volume equivalent to the Reactor Building top. floor and compute the time it takes to reach 10,000 x MPC. This is based on an assumed protection factor of 10,000 for supplied air to a worker in the building.

RESULTS FOR ON-SITE For I-131 DE with very conservative bounding assumptions and credit for supplied air, recovery operations could take place for a 30-day period without exceeding the limits of_10 CFR, Part 20. The iodine concentration never reaches 10,000 MPC in the Reactor. Building. H-3 concentration will remain below the limits of Part 20. 7.2 x 10-7 l l: !< micro c1/cc calculated tritium concentration as compared to MPC for restricted ( area = 5 x 10-6 micro Ci/ml. CONCLUSION This-calculation has shown that boiling of the SFP at Vermont Yankee could occur without exceeding the off-site dose limits of 10 CFR, Part 20 (0.5 rem whole body and 3 rem thyroid). The airborne tritium concentrations in the SFP area should not exceed the limits of Part 20. Using conservative assumptions and taking credit for supplied air, the airborne I-131 concentrations should not exceed the limits of 10 CFR, Part 20. Therefore, considering a complete loss of spent fuel pool cooling by both the SFPCS and the RHR System the radiological releases associated with postu-lated spent' fuel pool boiling are below the limits established for an operating plant by 10 CFR, Part 20. The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified for a seismic event, be single active' failure proof, and powered by a i safety class electrical emergency power source. As such, spent fuel pool cooling would be available under all plant conditions and spent fuel pool boiling would not occur. The enhanced system would acceptably close Open Issue No. 4 also, since . l it would provide spent fuel pool heat removal under all plant conditions and ' prevent pool boiling.

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                                                                                                                                           }

H:; ' ,;l: . p

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                  ,t.

l'.n1 RESPONSE TO NRC QUESTIONS VY SPENT. FUEL POOL EXPANSION. d i OPEN ISSUES:

N o . 25 SUPPLEMENTAL' COOLING ITEM:

5A. PROVIDE' ADDITIONAL INFORMATION'ON PARALLEL HEAT EXCHANGER OPERATION

                               '5B. SWITCHING'RHR BETWEEN TORUS COOLING AND SFP COOLING IS UNACCEPTABLE i                          .   ' RESPONSE:

5A. ADDITIONAL' INFORMATION CAN BEsPROVIDED

59. SWITCHING-RHR BACK AND FORTH FROM TORUS COOLING TO SFP COOLING IS WITHIN THE CAPABILITIES OF THE PLANT CONCLUSION:

EXISTING VY SFP SUPPLEMENTAL COOLING SYSTEM IS l, ADEQUATE AND ACCEPTABLE. PROPOSED'NEW i SYSTEM ALSO FULLY ADDRESSES THESE ISSUES. l l _ _ - - - _ _ _ -.__-_-__-___-____.-_._-._m- _ _ - - _ - - _ -

r' - ~ DISCUSSION SA. Additional information concerning SFPCS operation using one pump and two heat exchangers was previously supplied within the Vermont Yankee response to Open Issue ~No. 1.-

58. For 'a seismic event during power operation,. Vermont Yankee's method, pre-  !

sented within the September 1,1987 submittal, of using one train of RHR to { cool both the spent fuel in the fuel pool and the residual heat in the z reactor is not considered appropriate by NRC staff since too many operator actions and RHR pump starts are involved. This scenario describes how torus cooling and spent fuel pool cooling can be accomplished by the RHR System only. The initial conditions assumed the reactor was operating and recently discharged (ten days) spent fuel. These conditions are essentially impossible to achieve since an actual refueling at Vermont Yankee could not be done in less than ten days; thus, these con-ditions establish a very conservative analysis. The RHR cycle-involves six hours of torus cooling and one hour of Augmented Spent Fuel Pool Cooling (AFPC), with 20 minutes allowed for valve realign- j ment between modes. This conservative scenario is within system capability. and is well within the RHR pump's starting limitations listed in plant. Operating Procedure OP 2124, Rev. 19, " Limit RHR pump starts to 3 in 5 minu-tes followed by a 20 minute run or a 45 minute shutdown for cooling." If realistic spent fuel pool heat loads were used (i.e., less conservative than those required by SRP 9.1.3), the spent fuel pool heat up would be slower, which would allow a longer duration on torus cooling, thus limiting the cycle frequency and reduce ope'rator actions. CONCLUSION It is Vermont Yankee's conclusion that using RHR to ensure cooling of the spent fuel pool considering a seismic event.is within the capabilities of the plant, even if conservative scenarios and heat loads are used. The Vermont Yankee commitment of February 9, 1988 to provide an enhanced SFPCS that meets the applicable requirements of SRP 9.1.3 would, therefore, be qualified.for a seismic event, be single active failure proof, and powered by a safety class electrical emergency power source. As such, reliance on the RHR System to provide seismic spent fuel pool cooling would not be necessary. The enhanced system would acceptably close Open Issue No. 5 also, since it would meet the applicable SRP 9.1.3 requirements, operate under all plant conditions and, therefore, eliminate switching of one RHR train between the fuel pool and the reactor for heat removal.

 +

h ..

                                                                        ~

NU REG-0800 t (Formerly NUREG.75/0871 f-1 e ,,- i U.S. NUCLEAR REGULATORY COMMISSION M '%i. STANDARD. OFFICE CJ NUCLEAR REACTOR REGULATION REVIEW PLAN i.3 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM REVIEW RESPONSIBILITIES

                 ...a   ,

tu.,.:lia , Syste s Branch (ASB) M ena , - Chem % . engineering Branch (CMEB) AREM 0F REVIEW - All nucles- reacto'r plants include a spent fuel pool for the wet storage of spent feel assent lies. The methods used to provide cooling for the removal of decay heat f rom t.:e stured assemblies vary from plant to plant depending upon the individual desios 'he sbfety function to be performed by the system in all cases remains the sas., that is, the spent fuel assemblies must be cooled and must remain covered with wate' during all storage conditions. Other functions performed by the system, not relateu tc safety, include water cleanup for the spent fuel pool, refueling canal. refueling water storage tank and other equipment storage pools; means for filling and draining the refueling canal and other storage pools; and surface skim-mir.g 10 provide clear water in the storage pool. The / St rme of the spent fuel pool cooling and cleanup system covers the system from irlet to ar.d exit f rom the storage pool and pits, the seismic Category I water soerce au piping used for fuel pool makeup, the cleanup system filter-demineralizers en" tha regenerative process to the point of discharge to the radwaste system.

             '         Tne capability of the spent' fuel pool cooling and cleanup system to provide adequate cooling to the spent fuel during all operating conditions is reviewed on one of two bases.                The first basis requires the cooling portion of the sys-The ten,to be designed to seismic Category I, Quality Group C requirements.

second basis allows a non-seismic Category I, Quality Group C, spent fuel pool coo'1r;, system provided that the following systems are designed thetofuel seismic pool Category I requirements and are protected against tornadoes: make up water system and its source; and, the fuel pool building and its ven-tila'.ien and filtration system. The makeup, ventilation and filtration sys-tems must also withstand a single active failure. In addition, the transient tcg etatt.rc (T ) used in evaluating combined load on structures a Rev. 1 - July 1981 USNRC STANDARD REVIEW PLAN Star.dard ie<. . plan, a" tepared for the guidance of theThese office documents of Nuclearare Resctor Regulation staff responsible for the review of made available to the pubhc as part of the apphcabc' ' tr cona' twt' and cperate nuclear power plants commiss;oi poheg to sn'nrrn the nuclear indwst'y and the general pubhc of regulatory procedures and policies Standard re plans are fW sulatitute* fo' regulatory guides or the Commission's regulations and compliance with them is not requ s'ander- rev* plan sections are keyec to the Standard Forrr at and Cantent of Safet Analysis Reports for Nuclear Power Plan t, t sh si.urr s ut the stardaad Forma. have a correspond.rg review plan. as appropriate to accommodate commenth ano to ref 6ect new informa r,bbsher standard rey;ew plat.s will be revised periodically tion and esperience Comments and suggestions for improvement will be considered and should be sent to the U.S. Nuclear Regulatory Co Office of Nuclear Reactor Regulation. Washmgton, D C. 2055s

7 a

                                                                                                                                                                       )

(I' 4 . shall'be'the boiling temperature of water'when the cooling system is not oesignec to seismic Category I requirements. 2 .- The ASB reviews the capability of the spent fuel pool cooling, makeup, .- j and cleanup systems to provide adequate cooling to the spent fuel during .- all' operating and accident conditions. The review includes the following. considerations:

                                                                       -a.       .The quantity of fuel to be cooled, including the corresponding require-ments for continuous cooling during normal, abnormal, and accident conditions.                                                                          1 The ability of the systen to maintain pool water levels,                           q b.

c. The ability to provide alternate cooling capability and the associated time required for operation.

d. Provisions to provide adequate makeup to the pool.

e. Provisions to preclude loss of function resulting from single active failures or failures of nonsafety-related components or systems.

3

f. The means provided for the detection and isolation of system components that could develop leaks or failures.

g. The instrumentation provided for initiating appropriate safety actions. .I

h. The ability of the system to maintain uniform pool water temperature conditions.

3. ASB also performs the following reviews under the SRP sections indicated:

I

a. Review for flood protection is performed under SRP Section 3.4.1.

b. Review of the protection against internally generated missiles is performed under SRP Section 3.5.1.1.

c. Review of the structures, systems and components to be protected against externally generated missiles is performed under SRP Section 3.5.2.

d. Review of high- and moderate-energy pipe breaks is performed under SRP Section 3.6.1.

A A secondary review is performed by CHEB and the results used by the ASB to complete the overall evaluation. CMEB provides an SER input to ASB on a routine basis that includes an evaluation of the capability and capacity of the spent fuel pool cleanup system to remove corrosion products, radio-active' materials and impurities from the pool water. Also upon request the CMEB will provide ASB with an evaluation of the spent fuel pool and the spent fuel pool cooling system materials--fluid compatibility and

                                                                         -potential for metal corrosion degradation. ASB will request such input if the materials used in the design differs significant?y from preiiously           .

approveJ designs. 9.1.3-2 Rev.1 - July 1981

f ,' !dJ .: Li Ct,orditated reviews are performed by other. branches and'the results used Dy M in the overall evaluation'of the SFPCCS. -The coordinated reviews are an.follows: The Structural Engineering Branch (SEB) determines the ~

          }.           acce;Aatility of. the design analyses, procedures, and criteria used to estab'ish the ability of seismic Category I structures housing the system' l

i and scpporting systems to withstand the' effects of natural phenomena such as the safe shutdown earthquake (SSE), the probable maximum flood (PMF),, and tornado mis'siles as part of its' primary review responsibility,for.SRP.~ Sections 3.3.1, 3.3.2, 3.5.3, 3.7.1 through 3.7.4, 3.8.4, and 3.8.5. The-

                     . Mechanical Engineering Branch (MEB), determines that.the components piping                                                     ~

i and structures are designed'in accordance with' applicable codes and standards' , i as.part'of its primary review responsibility for SRP Sections 3.9.1 through j

3. 9.1 The MEE, also, determines the acceptability of the seismic and qualitj group classifications for system components as part of its primary j review responsibility for.SRP Sections 3.2.1 and 3.2;2. The MEB also  :

review; the adequacy of the inservice testing program of pumps and valves  ! as part 'of its primary review responsibility for SRP Section 3.9.6. The Materials Engineering Branch (MTEB) verifies that-inservice inspection requirements are met for system components as part of its primary review l responsibilitytfor SRP Section 6.6, and, upon request, verifies the compati-  ; bility of.the materials of construction with, services conditions. The-  : review'for Fire. Protection, Technical Specifications, and Quality Assurance. i are coordinated and performed by the Chemical Engineering Branch, Licensing Guidance Branch, and Quality Assurance Branch'as part of their primary review responsibility for'SRP. Sections 9.5.1, 16.0 and 17.0, respectively. .The EQB reviews the seismic qualifications of Category I instrumentation and electrical equipment and the environmental qualification.of mechanical and electrical equipment as part of its primary review responsibility for SRP. Sections 3.10 and 3.11, respectively. The Instrumentation and Control (. Systems Branch (ICSB) and'the Power Systems' Branch (PSB) will verify the 3 adequacy of the design, installation, inspection and testing all electrical l system (sensing, control and power) required for proper operation of the SFPCCS as part of their primary review responsibility for SRP Section 7.1 and Appendix 7-A for ICSB and SRP Section 8.3.1 for PSB. The Effluent Treatment Systems Branch (ETSB) will verify that the limits for radio- l activity concentrations are not exceeded as part of its primary review responsibility for SRP Sections 11.1 and 11.2. For those areas of review identified above as being the responsibility of othir branches, the acceptance criteria and methods of review are , contained in the SRP sections corresponding to those branches. II. ACCEPTANCE CRITERIA AcaptabiH yt of the design of the spent fuel pool cooling and cleanup system, as describeo in the applicant's safety analysis report (SAR), including related 4 sections of Chapters 2 and 3 of the SAR is based on specific general design j ' criteria and regulatory guides, and on independent calculations and staff judgments with respect to system functions and component selection. J. The design of the spent fuel pool cooling and cleanup system and its makeep system is accepteble if the integrated design is in accordance i I e th t % following criteria: dereral Design Criterion 2, as related to structures housing the j ( t system and the system itself being capable of withstanding'the  ! i 9.1.3-3 Rev. 1 - July 1981 l

l J l 1 a effects of natural phenomena such as earthquakes, tornadoes, and 5urricanes. Acceptance for meeting this criterion is based on con-formance to positions C.1, C.2, C.6 and C.8'of Regulatory Guide 1.13 and position C.1 of Regulatory Guide 1.29 for safety-related portions ,, I and prsition C.? of Regulatory Guide 1.29 for nonsafety-related portions of the sys*.em. This criterion does not apply to the cleanup portion of the system and need not apply to the cooling system if the fuel pool makeup water system and its source, and the fuel pool building and its ventilation and filtration system meet this criterion, and the ventilation and filtration system meets the guidelines of Regula-tory Guide 1.52. The cooling and makeup system should also be designed to Quality Group C requirements in accordance with Regulatory Guide 1.26. However, when the cooling system is not designated Category I it need not meet the requirements of ASME Section XI for inservice inspection of nuclear plant components.

b. General Design Criterion 4, with respect to structures housing the systems and the system being capable of withstanding the effects of external missiles. Acceptance is based on meeting position C.2 of l

Regulatory Guide 1.13. This criterion does not apply to the cleanup system and need not apply to the cooling water system if the makeup system and its source, and the building and its ventilation and j l filtration system are tornado protected and the ventilation and i filtration system meets the guidelines of Regulatory Guide 1.52. I }_ c. General Design Criterion 5, as related to shared systems and components j important to safety being capable of performing required safety func-tions. l d. General Design Criterion 44, tu include: (1) The capability to transfer heat loads from safety-related structures, systems, and components to a heat sirk under both normal cperating and accident conditions. (2) Suitable redundancy of cor;;onents so that safety functions can be performed assuming a single active failure of a component coincident with the loss of all offsite power. (3) The capability to isolate components, systems, or piping, if required, so that the system safety function will not be l compromised. (4) In meeting this criterion acceptance is based on the recommenda-tions of Branch Technical Position ASB 9-2 for calculating the heat loads and the assumptions set forth in item 1.h of subsection 111 of this SRP section. The temperature limitations of the pool water identified in item 1.d of subsection III of this SRP section is also used as a basis for meeting this criterion. l

a. General Design Criterion 45, as related to the design provisions to permit periodic inspection of safety-related compor,ents and equipment.

f. General Design Criterion 46, as related to the design provisicos to pemit operational functioral testing of saf aty-related systems or j

comt.onents to assure structural integrity and system leak tightness, 9.1.3-4 Rev. 1 - July 1981

L-y operability, and' adequate performance of active system components,

                          'and the capability of the integrated system to perform required functions during normal, shutdown, and accident situations.

(

g. General Design Criterion 61,,as related to the system design for fuel

                          ' storage and handling of radioactive materials, including the following elements:

(1) .The capability for periodic testing of components important to safety. (2) Provisions for containment. (3) Provisions for decay heat removal. (4) .The capability 'to prevent reduction in fuel, storage coolant inventory under. accident conditions in accordance with the L guidelines of-position C.6 of Regulatory Guide 1.13. I I (5) The' capability and capacity to' remove corrosion products, radioactive materials and impurities. from the pool water and reducing occupational exposures to radiation. f

h. Gene'ral Design Criterion 63, as it relates.to monitoring systems.

provided to detect conditions that could result in the loss of decay-heat removal, to detect excessive radiation levels, and to initiate appropriate safety actions, i. 10 CFR Part 20, paragraph 20.1(c) as it relates to radiation doses In meeting i being kept as low as is reasonably achievable (ALARA). this regulation Regulatory Guide 8.8, positions C.2.f(2) and C.2.f(3) will be used as a basis for acceptance. III. REVIEW PROCEDURES The procedures set forth below are _used during the construction permit (CP) application review to determine that the design criteria and bases and the pre-liminary design as set forth in the preliminary safety analysis report meet

              'the acceptance criteria given in subsection II of this SRP section. For the review of operating license (OL) applications, the review procedures and accertance criteria and bases have been appropriately implemented                     The reviewinprocedures the final design as set forth in the. final safety analysis report.

fo- OL applications include a determination that the content and intent of the technical specifications prepared by the applicant are in agreement with'the j j. regt.irements for system testing, minimum performance, and surveillance developed I as a result of the staff's review. I L Upon request from the primary reviewer, the coordinating review branches will

              -provide input for the areas of review stated in subsection I of this SRP section.

The secondary review branch, CMEB, will provide an input on a routine basis for those areas c' review indicated in this SRP section. The primary reviewer i .(ASB) obtains and uses such input as required to assure that this review procedure is complete. a typical system. Any variance of I ~en r eview procedures given belo, are f or t% review, to take account of a proposed unigt.e design, will be such as to f 9.1.3-5 Rev.1 - July 1981

assure that the system meets the criteria of subsection II of this SRP section. Ir. the esin , the spent fuel pool cooling and cleanup system and its makeup system are evaluated with respect to their capability to perform the necessary y safety functions during all conditions, including normal operation, refueling, abnormi, storage conditions, and accident conditions. 1 The ,aiety function of the system for refueling and normal operations is identified by reviewing the information provided in the SAR pertaining to the design bases and criteria and the safety evaluation section. The SAR section on the system functional performance requirements is also reviewed to determine that it describes the minimum system heat transfer and system flow requirements for normal plant operation, component operational degrada-tion requirements (i.e., pump leakage, etc.) and describes the procedures that will be followed to detect and correct these conditions should degrada-tion become excessive. The reviewer, using failure modes and effects analyses, determines that the system is capable of sustaining the loss of any active component and evaluates, on the basis of previously approved systems or independent c.lculations, that the minimum system requirements (cooling load and flow) are met for these failure conditions. The system piping and instrumentation diagrams (P& ids), layout drawings, and component descriptions are then reviewed for the following points:

a. Essential portions of the system are correctly identified and are isolable from the nonessential portions of the system. The P& ids are reviewed to verify that they clearly indicate the physical divi-sion between each portion and indicate required classification changes.  !

System drawings are also reviewed to see that they show the means for accomplishing isolation and the system description is reviewed to identify minimum performance requirements for the isolation valves. For the typical system, the drawings and description are reviewed to verify that adequate isolation valves separate non-essential portions and components from the essential portions, b Heat exchangers, pumps, valves and piping for the cooling portion of the system are constructed to Quality Group C and designed to seismic Category I*. requirements in accordance with the guidance provided in Regulatory Guides 1.26 and 1.29. As an acceptable alternative, the cooling loop may be constructed to Quality Group C and nonseismic Category I requirements provided the spent fuel pool water makeup systerr, and the building ventilation and filtration system are designed to seismic Category I requirements, are protected from the effects of tornadoes and meet the single failure requirements. The ventilation I and filtration system must also meet the guidelines of Regulatory Guide 1.52. The review for seismic design is performed by SEB and the review for seismic and quality group classification is performed b, MEB is indicated in subsection I of this SRP section.

c. The stated quantity of fuel to be cooled by the spent fuel cooling system is consistent with the quantity of fuel stored, as stated in Section 9.1.2 of the SAR. i i

d. For the maximum normal heat load with normal cooling systems in opera-tion, and essuming a single active failure, the temperature of the pool should be kept at or below 140 F and the liquid level in the poci shot.ld be maintained. For the abnormal maximum heat load (full 9.1.3-6 Rev. 1 - July 1981

e i

                  "         . v 'c:-d) the te7 erature of the pool water should be kept below
                  ;o' ling and the liquid level maintained with normal. systems in opera-A single active failure need not be considered for the abnormal ti r.

( case The associated parameters for the decay heat load of the fuel l assemblies, the temperature of the pool water, and the heatup time I or rate of pool temperature rise for the stated storage condition: are reviewed on the basis of independent analyses or comparative analyses of pool conditions that have been previously found acceptable.

e. The spent fuel pool and cooling systems have been designed so that in the event of failure of inlets, outlets, piping, or drains, the pool level will not be inadvertently drained below a point approxi-mately 10 feet above the top of the active fuel. Pipes or external lines extending into the pool that are equippeo with siphon breakers, check valves, or other devices to prevent crainage are acceptable as a means of implementing this requirement.

f. A seismic Category I makeup system and an appropriate backup method to add coolant to the spent fuel pool are provided. The backup system need not be a permanently installed system, nor. Category I, but must take water from a Category I source. Engineering judgment and compari-son with plants of similar design are used to determine that the makeup capacities and the time required to make associated hookups are con-sistent with heatup times or expected leakage from structural damage.

g. Design provisions have been made that permit appropriate inservice inspection and functional testing of system components important to safety. It will be acceptable if the SAR provides a statement that f

the spent fuel pool cooling, makeup, and cleanup system is included in the inservice inspection program per SRP Section 6.6 and the inservice testing program of SRP Section 3.6.6. These SRP sections are reviewed by the MTEB and MEB respectively.

o. The calculation for the maximum amount of thermal energy to be removed by the spent fuel cooling system will be made in accordance with Branch Technical Position ASB 9-2, " Residual Decay Energy for Light-Water Reactors for Long-Term Cooling" (located in SRP Section 9.2.5) under the following assumed conditions. .

i. Tne uncertainty f actor K is set equal to 0.1 for long-term cooling (greater than 10 7 seconds).

ii The normal maximum spent fuel heat load is set at one refueling load at equilibrium conditions after 150 hours decay and one refueling 1 cad to equilibrium conditions after one year decay (Maximum pool temperature 140*F) i'i. The spent fuel pool cooling system should have the capacity to remove the decay heat.from one full core at equilibrium conditions after 150 hours decay and one refueling load at equilibrium condi-tions af ter 36 days decay, without spent fuel pool bulk water boiling. Cooling system single failure need not be considered concurt ent f or t8.;s condition.

    /

s

                          .            For poc h with gre:.ter than 1-1/2 core capacitj, aqe additional refueling batch at equilibrium conditions after 400 days decay should be included in the cooling requirements.

9.1.3-7 Rev. 1 - July 1981

a Tl.e " w eser verifies that the system has been designed so that system fur.ctu ns will be maintained, as required, in the event of adverse natural phenomena such as earthquakes, tornadoes, hurricanes, and floods. The res dewer evaluates the system, using engineering judgment and the results o' failure modes and effects analyses to determine the following: 3 The failure of portions of the system, or of other systems not designed to seismic Category I standards and located close to essential portions of the system, or of non-seismic Category I structures that house, support, or are close to essential portions of the pool and cooling system, will not preclude essential functions. Reference to SAR Chapter 2, describing site features and the general arrangement and laycut drawings, will be necessary as well as to the SAR tabulation of seismic design classifications for structures and systems. State-ments in the SAR to the effect that the above conditions are met are acceptable. (CP)

b. Tne essential portions of the spent fuel pool cooling system are protected from the effects of floods, hurricanes, tornadoes, and internally or externally generated missiles. Flood protection and missile protection criteria are discussed and evaluated in detail under the SRP sections for Chapter 3 of the SAR.

The reviewer utilizes the procedures identified in these plans to

              --                                                                                               assure that the analyses presented are valid. A statement to the effect that the system is located in a seismic Category I structure that is tornado missile and flood protected, or that components of the system will be located in individual cubicles or rooms that will withstand the effects of both flooding and missiles is acceptable.

The location and design of the system, structures, and pump rooms (cubicles) are reviewed to determine that the degree of protection provided is adequate.

3. The system design information and drawings are analyzed to assure that the following features will be incorporated. A statement that these features will be included in the design by some appropriate means is a basis for acceptance. (CP)

a. A leakage detection system is provided to detect component or system leakage. An adequate means for implementing this requirement is to provide sumps or drains with adequate capacity and appropriate alarms in the immediate area of the system,

b. Components and headers of the system are designed to provide individual isolation capabilities to assure system function, control system leakage, and allow system maintenance.

                                                                                                         -      Design provisions are made to assure the capability to detect leakage of radioactivity or chemical contamination from one system to another and to preclude long-term corrosion, organic fouling, or the spreading of radioactivity. Radioactivity monitors and conductivity monitors located in the system discharge lines are acceptable means for imple-menting this requirement.                                                   ,

4. The SU descriptive information, P&lDs, layout drawings, and system analyses are reWewad to assure that essential portions of the system will function 9.1.3-8 Rev. 1 - July 1981

y , I' K

    + +-       followinc design basis accidents, assuming'a concurrent single active component 1 failure. The reviewer evaluates failure mode and ' effects analyses
             . presented in.the SAR to assure function of required components, trace the
   '4          availability'of these components on system drawings; and check that minimum system flow, makeup, and heat transfer requir Ments are met for each.

degraded. situation over the required time spans. :For each case the design w'll'be' acceptable if minimum-system requirements are met.

5. The spent fuel pool cleanup system' and 'various auxiliary. systems are designated as nonsafety-related systems and are designed accordingly.

             .(nonseismic Category-I). 'These systems are evaluated to assure that their. failure cannot affect the. functional performance of any safety ,

related' system or' component. The relationship'and proximity between the nonsafety-related system and' safety-related systems or components-are determined by reviewing the integrated structure and' component, layout diagrams. Independent analyses,' engineering judgement,:and comparisons-with previously approved systems are used to verify that where~a nonsafety-

              'related system interconnects'or interfaces with the cooling system, its failu e by.any event or malfunction will not preclude adequate functional performance:of.the' cooling system.

7. .The cleanup system is also reviewed to assure that it has been designed with the capability to maintain acceptable' pool water conditions. The-P& IDS and associated information provided in the SAR is reviewed .to verify the.following:

a. 'A means has been provided for mixing to produce a uniform temperature:

throughout the pool.

    \'
              .b.      The-cleanup system is reviewed by CMEB to verify they have the capacity anei. capability to remove corrosion products, radioactive materials, and impurities so that water clarity and quality will enable safe-o M ating conditions in the pool. This includes instrumentation and sampling to monitor the water purity and need for demineralized resin replacement including the chemical and radiochemical limits such as conductivity,. gross gamma and iodine activity, demineralized dif-ferential pressure, pH and crud level which are used to initiate corrective action.

c. The capability for processing the refueling canal coolant during refueling operations has been provided.

d. Provisions to preclude the inadvertent transfer of spent filter and demineralized media to any place other than the radwaste facilit/

have been provided. IV. EVALUATION FINDINGS The reviewer verifies that sufficient information has been provided and that his review supports conclusions of the following type, to be included in the staff's safety evaluation report: 3 ? rpent fuel pool cooling and cleanup system includes all components ano piping of the system from inlet to and ex;t from the storage pool and r ['

     .           Pits the seismic Category I water source and piping used for fuel pool maaeuc. the cleanup system filter-demineralizers and the regenerative 9.1.3-9                                Rev. 1 - July 1931

i i proc es to the point of discharge to the radwaste system. The scope of m ic 'f the spent fuel pool cooling and cleanup system included layout i l dra. v . process flow diagrams, piping and instrumentation diagrams, ano y descriptive information for the system and the supporting systems that w are o.sential to safe operation. The cooling portion of the system and the errergency primary makeup system are designed to seismic Category 1, Qual: 1., Group C requirements since they are necessary to remove decay heat from the spent fuel and to prevent fuel damage that could lead to unacceptable releases of radioactivity. The cooling portion of the system , neea not be designed to seismic Category I requirements if the makeup system l

               .and the building ventilation and filtration system are seismic Category 1, and if the ventilation and filtration system meet the guidelines of b g"l e nry Guide 1.52.

h sta4 ccncludes that the design of the spent fuci pool cooling and cleanup systen and its makeup sp tem meets the requirements of General Design Criteria 2,  ; 4 5, 44. 45, 46, 61 and 63. This conclusion is based on the following: 1

1. ~he applicant has met the requirements of General Design Criterion 2 with respect to safety-related portions of the system being protected against

                ;)e tt.r al phenc~ena. Acceptance is based on meeting the guidelines of Regulatory Guide 1.13, position C.1 which recommends a seismic Category I de 'r fer necessary portions of the spent fuel storage facility, pcrition C.2 regarding protection against winds and wind generated missiles, posi-tic- C 6 As it relates to the system being capable of withstanding earth-quakes without loss of coolant that would uncover the fuel, and position C.E eh':h recommendt a seismic Category I makeup system with appropriate reduncancy or a backup from a Category I water source. Acceptance is also based on meeting the seismic design requirements of Ragulatory Guide 1.29.                                         l positicn C.1 for safety-related portions of the system necessary for adecuate cocling to prevent excessive radioactivity releases (position C.I.p f  Pagulatory Guide 1.29) and position C.2 as it relates to the failure M i.c 3saf ety-related portions of the system. If the fuel pool building vent flation and filtration systems are designed to seismic Category I requirements and in accordance with the guidelines of Regulatory Guide 1.52 tti. c: Oing portion of the system need not be seismic Category I.

2 The design meets the requirements of General Design Criterion 4 with regards to protection against the effects of externally generated missiles since it 's in accordance with position C.2 of Regulatory Guide 1.13 since no loss >;f watertight integrity or fuel damage occur in the event of tornado  ; missilee. Tne design meets tre requirements of General Design Criterion 5 regarding the sharing of safety-related structures, systems, and components since no single failure will prevent the system from performing its safety-related i function which is cooling the spent fuel.

           *. The design meets the requirements of General Design Criterion 44 regarding decay heat removal redundancy and power supplies, since the system has the capability to remove decay heat from the spent fuel under both normal nperating and accident conditions. The system has redundancy so that decay heat can be removed ase.uM ng a single active failure coincident with a loss of all of f site power, and is designed with isolation capat 'lity of syster components and piping, if requireo, such that tM ability of the systar to remese deca 3 heat will not be compromised.

9.1.3-10 Rev. 1 - July 1981

                                                                                 - - - _ _ _ _ _ - _ _ _ _ _ _ _ - - __________-__a

E The system meets'the inspection and testing. requirements of General Design Criteria 45 and 46 since the system.is designed and constructed with suitable clearances and-location to allow periodic inspection of major

     - (^         components,.and is designed to permit functional operational testing to assure structural integrity and system leak tightness, operability, and adequate performance:of active system components.

E 6i The system is designed in accordance with the requirements of General Design Criterion 61 as it relates to the system design for fuel. storage since the system has the following design capabilities: 'the system has the capability for periodic testing of components important to safety. The system is' designed to provide suitable shielding by maintaining a minimum wate level above the fa i. There is redundancy and testability

                                                   ~

of the decay heat removal port 4 m of the system, and the system is' designed to prevent reduction in fuel storage coolant inventory;under accident conditions in accordance with position C.6 of Regulatory Guide 1.13. The spent fuel pool cleanup portion of the system (1) provides the capability and capacity of removing radioactive tnaterials, corrosion products, and impurities from the pool water and thus meets the require-ments of Criterion 61 as.it relates to appropriate filtering systems for fuel cooling and storage, (2) reduces occupational exposure to radiation by removing radioactive materials from the pool water and thus meets the requirements'of 10 CFR Part 20, 920.1(c) as it relates to maintaining radiation exposures as low as reasonably achievable (ALARA) and, (3) retains radioactive materials and crud in the pool water in the demineralized and filters and thus meets positions C.2.f(2) and (3).of Regulatory Guide 8.8.'

                                                              ~

f -The system design meets the requirements of General Design Criterion 63 7.

                   ;ince it has' provisions to detect the loss of heat removal function through=

( the use of loss of flow and temperature alarms, and to detect conditions that would result'in' excessive radiation through the use of coolant. low level alarms and radiation monitoring alarms. And tne system has the capability to initiate appropriate safety actions since it has an automatic makeup system and the cooling s3 stem.and ventilation and filtration system can be operated from the control room in the event of high radiation or low level alarms. V. IMF;EMENTA110h N fo.'o ing is intended to provide guidance to applicants and licensees regardinc the NRC staff's plans for using this SRP section. Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, ite methou described herein will be used by the staff in its evaluation of conformance with Commission regulations. Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced Regulatory Guides.

              \1    REFERENCES 10 CFR Part 20, b20. ltc), " General Provisions for Standards for Protection Against Radiation."                                                                           j e

s 9.1.3-11 Rev. 1 - July 1981 ____ -- -______ -_ - _ __n

        ;          Z' CFP Part 50, Appendix A, General Design Criterion 2, "Desigr Bases fe" Sro'e-tior, Against Natural Phenomena."

r L C 9 Part 50, Appendix A, General Design Critorion 4, " Environmental e.nc Missile Design Bases." 4 10 CFR Part 50, Appendix A, General Design Criterion 5, " Sharing of Structures, Systems and Components."

5. 10 CFR Part 5.0, Appendix A, General Design Criterion 44, " Cooling Water."

6. 10 CFR Part 50, Appendix A, General Design Criterion 45, " Inspection of Cocling Water Syste. "

10 CFR Part 50. Appendix A, General Design Criterion 46, " Testing of Cooling Water System."

o. 10 CFR Part 50, Appendix A, Ge'ntral Design Criterion 61, " Fuel Storage

                  .and Handling and Radioactivity Control."

9. 10 CFR Part 50, Appendix A, General Design Criterion 63, " Monitoring fuel ar.d Waste Storage."

30. Regulatory Guide 1.13, " Fuel Storage Facility Design Basis."

11. Regulatory Guide 1.26 " Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants."

u. Regulatory Guide 1.29, " Seismic Design Classification."

l' Regi atory Guide 1.52, " Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants."

14. Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational)

Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Acnievable." l l l I I l 9.1.3-12 Rev. 1 - July 1981 l

6

                                                                      ~~ //.
           ,        ,                                                                            I UNITED STATES
                                .goarcy'A'}

j' NUCLEAR REGULATORY COMMISSION a I '* ;# , j

                             ; 4 W A$mNGTON, D. C. 20555 FEB 01199g
                              \[w ****'
                                           /                                     January 21, 1988 I'eck et No. 50-271 Mr. Warren P. Murphy Vice President and Manager of Operations Vermont Yankee Nuclear Power Corporation                                                                   .

1671 Worcester Road Framinohan, Massachusetts. 01701

SUBJECT:

FORTHCOMING MEETING WITH VERMONT YANKEE NUCLEAR POWER STATION - By letter dated December 16, 1987, you requested a meeting with the NRC 6 technical staff to respond to any remaining informational requirements or outstanding questions associated with your proposed spent fuel pool expansion. We agree that such a meeting is appropriate and reenmmend that you be prepared to discuss the following in this meeting: o An evaluation of and/or justification for the calculational method you used for decay heat generation rate. o Explanation of difference between heat exchanger heat renoval capacity of 2.23 M3tu/Hr (FSAR) and 9.1 MBtu/Hr heat removal capacity of one train of SFPC, as described in your application documents. o Provisions for SFP cooling in case of single failure. o Provisions for SFP temperature monitoring. o Additional information on SFP seismic nakeup capability. o Additional information on radiological consequences of SFP boiling. o information to justify reliance on parallel use of SFP heat exchangers. We have enclosed a status report on our review of your proposal, which further describes these issues.

l :.

 ,.         i.

1

                                                              -7'-                 January 21, 1988.

Mr.' Warren P.. Murphy

                                                            ~
                   '.We propese that t;iis meetino be held at the. new NRC Headquarters at One' WMte'
                   . Flint North in Rnckville in Room 16-B-11 en Tuesday, February 9, at 10:00 AM.

Please contact the' NRR Pro,iect Manacer, V. Rooney (301-497-R344), to' confirm arranoenents for this meeting. Sincerely,. l i QI tr Y to'r Division of Reactor roiects I/II Office of Nuclear ctor Regulation l

Enclosure:

Status Report on Review of ' Explanation of the Spent' Fuel Storage Capacity cc: see next page .

         .t F

                                        .i Mr. R.'W. Capstick                   Verront Yankee tuclear Power Coronratinc Vernant. Yankee Nuclear Power Station CC*

i. Mr. J. Gary Weiaand .

W. P. Murphy, Vice-President President & Chief. Executive Officer and Manaoer of Operations Vermont Yankee Nuclear Power. Corp. Vernnnt Yankee Nucin.r Power Corp. R.D. 5, Boi 169- R.D. 5, Box-169-Ferry Road Ferry Road l Brattleboro, Vernont 05301 Brattleboro, Vermont 05301 Mr. John DeVincentis, Vice President, Mr. Gerald Tarrant, Commissioner Yankee Atomic Electric Company. Vernent Department of Public Service 1671 Vorcester Road 120 State Street Framingham, Massachusetts. 01701 Montpelier, Vermont.05602 New England-Coalition on. Nuclear - Public Service Board Pollutioni State of Vermont Hill and Dale Farm '120' State Street

• R.D. 2, Bex 223' Montpelier, Vernent- 0E602 Putney,. Vermont' 05346 Mr. Walter Zaluzny Vermont Yankee Decommissioning Chairman, Board of Selectman Alliance -

Post Office Box 116 Box 53 Vernon, Verment 05354 Vernon, Vermont 05602-0053

            *            -Raymond H. McCandless.                           Vermont Public Interest Research
                          -Vermont Division of' Occupational                  Group, Inc.

and Radiological Hea'lth 43 State Street.

                        ' Administration Building                          Montpelier, Verrent 05602 Montpelier, Vermont 05607
                          ' Honorable John J. Easton                       William Russell, Regional Administrator Attorney General                                Region I Office State of Vermont                                II.S. Nuclear Regulatory Commission 109. State Street                               631 Park Avenue LMortpelier, Vermont 05602                        King of Prussia, Pennsylvania 19406

• John A. Ritscher, Esquire Mr. R. W. Capstick Ropes & Gray Vermont Yankee Nuclear 225 Franklin Street Power Corporation Boston, Massachusetts 02110 1671 Worcester Road Framingham, Massachusetts 01701

____._-.-.m____-m__m-_--______.______m_m_

g,j a, J , 1 w '

                ,                                                                                      u y                        :;                                            .tp[.                                           ++

y* *' Mr; R W.' Capsticki Vermont Yankee Nuclear Power Corporation b g Vernant Yankee Nuclear' Power Statinn !'

       ~

l:. ,

                           - cc:'
                                                                                   ~
                             *Ellyn R.'Heiss, Esq.                         Resident Inspector            ..   .

Harmon A Weiss . U.S. Nuclear Regulatory Comission -

                           '2001~S Street, N.W.                            P.O. Box 176,

Washington,1D.C.: 20009 Vernon,. Vermont 05354

• David J. Hu11ett, Esq.
• Carol'S. Sneider; Esq.

Special. Assistant Attorney General- -Assistant Attorney General. . .

>        .                   Vermont Depart. of. Public Service            Office of the Attorney General..

120 State Street; One: Ashburton Place,19th Floor Mor.tpelier, VT 05602 Boston,=MA102108 Jay'Gutierrez Geoffrey M. Huntington,-Esouire ' Recinnel Counsel ' Office of the Attorney General USNRC, Recion I

                                                                          ' Environmental Protection Bureau 631 Park Avenuel                              State House Annex i

Kino of Prussia, PA 19406 25 Capitol Street 1

                                                                          . Concord, NH 03301-6397              ,

Charles Bechhoefer, Esq.

                                         ~
                             *G.DanaBisbee,[Esq.-

w Office.of the Attorney General Administrative Judae f Environmental-Protection Bureau Atomic Safety and Licensing Board State House Annex U.S. Nuclear Reaulatory Commission

                            -25 Capitol. Street.                           Washington, DC 20555
      ".                   -Concord, NH- 03301-6397 Dr. James H. Carpenter Administrative Judoe Atonic Safety and Licensino Board             Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission            U.S. Nuclear Regulatory Commission Washington .DC 20555                          Washington,'DC 20555
                           .Pr. Glenn'O.-Bright ~                                    ,

Administrative Judge Atonic-Safety and Licensing Board U.S. Nuclear Regulatory Comission Washington, DC 20555

                             *In response to Ms. Weiss's letter of January 4,1988 to Vernon Rooney,-

Project Manager, asking to be informed by telephone of any meeting scheduled in response to Vermont Yankee's December 16, 1987 reouest for a meeting, Staff counsel has telephoned persons identified with an

• to inform them of the meeting.

                                                                                                                          )

STATUS REPORT ON REVIElf 0F EXPAMS*0N OF ThE SPENT FUEL FOOL SiORAGE CAPACITY VERii0h i r ANKEE NUCLE AR P0HER ST ATIGN DOCKET NUMBER 50-27i A. BACKGROUND The Vernant Yankee Nuclear Power Station is a General Electric Company Boi % c Water Reactor (BWR) and received a full power operatinq license on Februar/ 28, 1973. At the time of licensing, the spent fuel pool contained sufficient st,r-age locations to accomodate 600 fuel bundles. Spent fuel bundles eouivalent to a cuarter of the fuel core were to be placed in the spent fuel pool on an annual cycle. Using the current guidelines of NUREG-0000, Standard Review Plac, Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System," and Branch Tech-nical Position ASB 9-?, " Residual Decay Energy for Light-Water Reactors for Long-Term Cooling," the staff determined that annual offleeding of a quarter core until 600 fuel bundles were in the spent fuel pool would yield a heat cen-erat hn rate of 6.02 MBtu/Hr. Table 10.5.1 of the FSAR states that the spent - fuel pool coolino system consists of two redundant trains with each train con- ~ sistirg of one 450 qpm pump and one heat exchanger. This FSAR table states that the design capability of the each heat exchanger is 2.23 MRtu/Hr with a pool water temperature of 125'F. Also, a maximum spent fuel pool water temperature of 150*F is specified in Technical Specification 3.12(H). With the heat generation rate of 6.02 MBtu/Hr and one train of the spent fuel pool cooling system (the redundant train is considered inoperable due to a single failure), the pool water temperature was calculated by the staff to be 149.3'F, which is less than the Technical Specification limit of 150'F. Additionally, the spent fuel pool cooling system is non-seismic Category I, i non-Class IE. In September 1977 the Vermont Yankee Nuclear Power Corporation (the licensee) received approval to replace the original spent fuel storage racks with new high-density spent fuel storage racks. These high-density racks were to be in-stalled in phases providina a total maximum storage capacity of 2000 fuel bundles. To date, the licensee has installed racks sufficiert to store 1680 fuel bundles. On April 25, 1986 the licensee requested approval to rerack the spent fuel pool for a second time. This second rerack application is the sub.iect of this status report. The new high-density storage racks would increase the storage capacity of the spent fuel pool to 2870 fuel bundles and is pro,iected to provide storage capacity until 1999. The licensee provided additional information on the proposed second rerack request in submittals dated September ?.6, October 21, and November 24, 1986, and February 25, March 19. April 9, April 13, May 22, June 11, September 1, and December 11, 1987. The licensee also incorporated by reference information contained in submittals dated September 11, 1981, November 30, 1983, and May 21, June 27, and December 18, 1984.

l

                                                                                                    )

Information related to the licensee's computer modeling of spent fuel pool cooling was provided at a meeting on January 15, 1987 in Richland, Washington. B. Sumary of Procedures with' Resnect to Review of 'fY's Acolication The licensee's submittals are being reviewed in acenrdance with the requirements of Title 10 of the Code of Federal Regulations, Part 50, Appendix A, General Design Criteria 2, 44, 60, 61, and 62, and the guidelines of Reculator.y Guide; 8.8, NUREG-0800, " Standard Review Plan," (SRP) and NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants." The review identified a number of items that require additional information for resolution, j 1 Heat Generation Rate The staff performed an independent assessment of the heat generation rate of the spent fuel in confo-mance vith the guidelines of NUREG-0800, Section 9.1.3, . (Spent Fuel Pool Cooling and Cleanup Systens) and Branch Technical Position ASB 9-2 " Residual Decay En'ergy for Lighc Water Reactors for Long-term Coolinc". Two conditions were evaluated. The first condition was the maximum normal heat load based on the neat generated by spent fuel when the pool is filled to its

            -   capacity of 2870 burdles with fuel removed during normal refueline. In this condition, the past refueling intervals and the number of fuel bundles offloaded are used in the evaluation along with the licensee's predicted refueling schedule and anticipated number of fuel bundles to be offloaded for each refueling until the poc1 is filled. The second condition was the maxinum abnormal heat load case, which is similar tc the norma'l heat lead case except that the last 368 spaces are filled by a full core offload.

The calculated heat load is conservative in several respects. Generally, plant capacity factors are not considered and no credit is given for the lack of power generation of the reactor during refueling when determining the irradiation time. Further, for the purposes of heat load calculation all offloaded spant fuel bundles are assumed to enter the pool simultaneously when in actuality bundles are placed in the pool in succession thus increasing the decay time before the decay heat load is removed by the spent fuel pool cooling system. These conservatism in addition to the conservatism in the equations in the Branch Technical Positiun provide margin in the overall analysis for items such as the inability to accurately predict actual future refuelinas. Also, the NRC staff's calculations assune design power rather than licensed power.

2. Soent Fuel Peel Coolino System

                                                                              ~

Spent fuel pool cooling systems (SFPCS) are reviewed to verify conformance with the guidelines of NUREG-0800, Section 9.1.3. The SFPCS and the cooling water systems that service the SFPCS are reviewed to assure their ability 9

o-.

     .                                                                            3 to remove the normal heat generated byt 'he spent fuel and maintain'the pool water temperature belnw la0'      ith a single active failure. The sta#f's criteria corsider a facilit\       bility to safely control radioactive materials during an event (e.g., earth w ke, flood, hurricanes, etc.) concurrent with a single active. failure (e.g. failure of a pump, motor, air- or motor-operated valves, etc). -Generally, facilities meet these cuidelines with one less'than the number of SFPCS trains installed and with the SFPCS designed to seismic                                                1 Category I criteria and powered from a Class IE power source, usually an onsite diesel generator. If the facility does not have these features, alternative
                                     . approaches are considered.as indicated in the guidelines of SRP Section 9.1.3.

The systems capable of cooling the spent fuel pool are also reviewed for their adequacy for the abnormal (core offload) heat load case. Since the staff does not censider the offloading of the core to be a normal anticipated occurrence, no failures are assumed to occur concurrent with the core offloed. Further, since there is no fuel in the reactor, systems other than the SFPCS may be , available ard used to provide additional cooling for the spent fuel in the spent' fuel pool during the' period that the core is fully offleeded. These additionel systems could include those systems that normally provide a safety

• function for the reacter and thus would be seismic Cateacry I and Class IE powered. Generally if the SFPCS neets the guidelines for the normal heat load case, the SFPCS with the additional cooling system (s) will meet the guidelines-for the abnormal heat load case. The systems that provide cooling for systems that cool the spent fuel pool (such as the service water system or closed cooline water system) are also reviewed to verify that the additional heat load being transferred to them will not result in water temperatures that could degrade their performance.

Lead Handling 3. The handling of light loads with the new storage racks and potential related f accidents is reviewed in accordance with the guidelines of NUREG-0800, Section 1 9.1.4. A light load is defined as any load that weighs less than a channeled fuel. bundle and its handling tool. For applications to re-rack spent fuel pools, the review of light load handling verifies that for anticipated light lead drop accidents, the consequences are equal to or less than those of the desian basis fuel handling accident. This accident is the oostulated dropping of a fuel bundle within the spent fuel pool and the resulting consequences. If the licensee can demonstrate that damage will not result in the spent fuel Keff exceeding 0.95 (with all uncertainties being included), and the release of radioactive material is equal to or less than a fraction of 10 CFR Part 100 limits as assumed in the previously analyzed fuel handling accident, the staff can conclude that the new rack desicn meets the requirements of General Des?cn r Criterion 62, " Prevention of Criticality in Fuel Storage and Handline." l l l l l ___ . _ _ _ _ _ _ _ _

The handling of heavy loads, and specifically the handling cf the existino and new spent fuel storage racks, is reviewed in accordance with NbREG-0612, " Con-trol of Heavy loads' at NL. clear Power Plants," and NUREG-0554, " Single-Failure - Proof Cranes for Neclear Powe* Plants." While all plants have been reviewed against the cuidelines of NUREG-0612 and NUREG-0554, those reviews were for normal, anticipated, or routine, heavy load handling. The handling of spent fuel storage racks is not routine heavy load handling. Thus, for a licensee to meet the guidelines of SRP Section 9.1.5, compliance with the guidelines.of NUREG-0612 and NUREG-0554 for the specific operations related to re-racking'of

                           'the spent fuel pool should be demonstrated. Compliance is demonstrated, in part, by providing drawings that show the heavy load paths for ensuring that loads are not carried over spent fuel, using sincle-failure-proof lifting rigs, and providing load handling procedures and training. An assessment of the information regarding these matters allot:s the sta#f to determine whether the licensee complies with the guidelines in NUREG-0800, Section 9.1.5, and thus whether the requirements of General Design Criterion 61, " Fuel Storage and                                                -

Handling and Radioactivity Contrc1,"-are met. 4 Spent Fuel Storace Racks The spent fuel storage racks are reviewed to verify ~conformance with the guidelines of SRP Section 9.1.2. The racks are reviewed to confirm that - they can withstand the maximum fuel handling uplift forces without an increase in Keff or a decrease in pool water inventory. Further, t5e design of the racks should ensure that a fuel bundle cannet be inserted in other than

                    '        a desion location. By limitino the maximum uplift forte and limiting the ability to place fuel in non-storage locations, the licensee conforms with the guidelines of NUREG-0800, Section 9.1.2 and thereby meets the requirements of General Design Criterion 62.

J' C. OPEN ISSUES The following are those items that reauire additional information for resolution.

1. Heat Removal Capability The SFPCS at Vernent Yankee consists of two identical trains of eouipment.

Each train consistt of one 450 gpm centrifugal pumo and one ? 23 MBtu/Hr tube-and-shell heat exchanger. The two trains are headered together on'the suction side of the pumps and at the dist.narge of the heat exchanners. The licensee stated that pool temperature monitoring instrumentation sensors are located on the SFPCS piping, which will give meaningful temperature readings whenever the SFPCS pump or. that line is operating. After water from the spent fuel pool is cooled by the heat exchangers, it is purified by the spent fuel pool cleanup ' system. 9 _ _-_- _ _ _ _ _ . - . - _ . _ _ _ _ . _ _ - _ . _ - _ - _ _ _ _ . _ _ - __ A

a= ,

                             'The' licensee stated in the September 1,1987 subnittal that one of the assump-
                            ' tions in its analysis is that the reactor decay heat is calculated in accord-ance with-the 1971 ANS' draft standard. Using the draft standard and pool water' temperature of 150'F, the licensee stated that the heat load in the pool is -

conservatively assumed to be 9.1 MBtu/Hr. However,'it.is not clear that this heat generation rate is meant to be the naximum normal heat generation rate, as

                             ' defined in SRP Section. 9.1.3. Further, the licensee has not provided the cor-responding abnormal heat generation rate (full core offload).

The staff's calculations were performed in accordance with the guidelines , discussed above, using historical 12-month operating cycles and assumina the 18-month operatino cycles anticipated for future plant operation. The results of these calculations indicate that the normal neximum heat ceneration rate is 10.1 M3tu/Hr and the abnormal maximun heat generation

• rate is 21.46.MBtu/Hr.

The licensee's calculational method is not in accordance with the SRP, and provides a less conservative result. The licensee has not provided adequate *

                               , justification for using the 1971 ANS draft standard to calculate decay heat generation. If the'licens'ee wishes'to use this method, further . justification would be required.

The licensee stated in the September 1,1987 submittal that the heat load in the pool is conservatively assumed to be 9.1 MBtu/Hr, which is the heat removal capacity of one train of SFPC. Table 10.5.1 of the FSAR states that the heat removal capacity of a spent fuel pool heat exchanger is 2.23 MBtu/Hr. The

                    '          basis for the licensee's assertion that adequate heat exchanger capacity is available should be provided to the staff for review.

1 Based upon the staff's decay heat calculations, the heat removal capacity of each heat exchanger must be at least 10.1 MBtu/Hr in order to satisfy the single failure' criterion for the maximum normal heat load condition. This heat generation rate, is based on a SFP full of spent fuel that has been discharged on an annual refuelino cycle up to the 1987 refueling outage and then on an 18-month refueling cycle until the spent fuel pool is completely filled with 2870 fuel

                             ' bundles.

General Design Criterion 44 requires that the cooling water system be capable of removing the heat load under normal conditions given a sinole failure. Based upon the staff's calculated decav heat load, under the normal maximum heat load and a single failure, one SFPCS train cannot maintain the SFP water temperature below 150'F for the first 68 days following shutdown for refuelino. The staff calculated a maximum spent fuel pool water temperature for this scenario of 198.1*F. Further, operation of both treins tocether would .not be able to maintain the SFP water temperature below 150'F for the first 10 days into the refueling outage. After 69 days, the SFPCS neets the single failure criteria and heat removal requirements of General Design Criterion 44 because the decay heat generation rate will have been reduced sufficiently to permit one heat exchanger to remove the decay heat. The licensee should provide additional information which addresses removal of normal fuel pool decay heat in the event of a single failure or propose an acceptable alternative method for decay heat renoval. J - - _ _ _ _ ____--_ _ _ - - -

V- <

                                                                                                                            'i 2.-    Spent Fuel Pool Temperature Limit
                        ; Standard Rev'iew Plan Section 9.1.3 identifies an acceptable spent fuel pool temperature linit of la0'F for.the normal-maximun heat load case. Verrent Yankee was or.iqinally: licensed with Technical Specification 3.1200, which limits.the maximum pool-temperature to 150*F. The licensee stated in the submittal dated April 9, 1987, that the SFPCS was qualified for a pool water temperature of 150*F. Specifically, the evalification temperatures for the ma,4cr components are:       140*F'for the deminerclizer,150'F for the SFPC pumps and heat exchangers, .and 175'E for the SFPCS pipino. At water temperatures t                           greater.than 140'F, the demineralizers resins may start to degrade. In order-to-prevent degradation of the demineralized resin, and to.be in conformance with the guidelines of SRP Section 9.1.3, the licensee committed in a submittal dated June 11, 1987, to isolate the denineralizers when the SFPCS inlet temperature is 140'F or hieher. However, there is no control room alarm that would indicate that the spent fuel pool water temperature is 140' or higher. The licensee stated that the control room alarms.that could indicate that the spent fuel pool water temperature limit may be exceeded.are: (1) low spent fuel pool water level (2) SFPCS pump trip, (3) Reactor Building Closed Cooling Water System (RBCCWS) pump trip, (4) RBCCWS tank high and low water levels, (5) low RBCCWS header pressure, (5) Service Water System (SWS) purp' trip, and (7) low SWS
                                    ~

header. pressure. The.RBCCHS heat exchangers are cooled by the SWS. An alarm in either'the RPCCWS or.SWS could be indicative of the potential for increasing spent fuel pool water temperature. None of these alarms is based on a 140'F 6 or 150'F limit in the spent fuel pool. .The licensee should provide information that indicates how the operator will know the pool water temperature.in the event of a loss of the SFFCS.

3. Pool Cooline Following Seismic Events The SFP coolino capability was reviewed with respect to the requirements of General Design Criterion 2. " Design Bases for Protection Against Natural

                        . Phenomena," which includes protection against earthquakes, hurricanes, tornadoes, or other natural events. The Vermont Yankee SFPCS is not seismic Category I and it is not powered by a Class 1E source. Under such. circumstances, SRP Section 9'1.3 identifies an alternative method for cooling tne spent fuel 1

following an earthquake. Specifically, the SRP discusses use of a seisric Category I spent fuel pool makeup water capability and a seismic Category I ventilation system to process the potential steam environment in the SFP-building resulting from pool boiling. In the April 25, 1986 sube'ttal, the licensee identified three sources of makeup water that it asserted were seismic  ; Category I. They are: the torus, fire water systen in conjunction with the l service water system from the Connecticut River, and alternate coolino cell. Subsequently, the licensee stated in the December 11, 1987 submittal that no reliance was being placed on the fire water system but that the nonseismic Category I fire water pumps and related piping could be used if they were ~ available and the seismic Category I service water pumps were not available.

7 l b

                                                               ..71 The staff determined that the torus-is an acceptable source of makeup only when the reactor is in a refueling mode with the vessel head removed and the gate between the open reactor vessel and the spent fuei pool is open. The terus safety furetion during power operation would be compromised.if the level                            l were 10wered to provide SFP makeup.

In order for the service water system fron the river to be considered a seismic Category 1 makeup source of water for the spent fuel pool, at least one service water pump must be. seismic Category I and powered from an ensite source, and the fire water pipe must be seismic Category I up to the suction of the RHR service water (RHRSW) pump, including appropriate isolation valves at connec-tiens between seismic Category I and nonseismic Category I lines. Tna service water pump meets the required seisnic requirements. However, the fire protection system has not been shown to have these features and, therefore, based upon the information'provided by the licensee-to date, the l service water system is not an acceptable seicmic Cateoory I makeup source.

• 1 The licensee identified the coolino tower deep basin alternate cooling cell

as a seismic Catepory I SFP makeup source. Neither the September 1,1987 submittal nor the FSAR addresses the seismic classification of.the associated pipino. Piping from the alterrate cooling cell to the RHRSW pumps would be required to be seismic Catecory I and appropriate isolation in accordance with the single failure criterion would be recuired between the seismic Category I line and ncnseismic Category I lines in order for the alternate cooling cell  ! e to be an acceptable seismic makeup source. The alternate cooling cell has not l been shown to neet these criteria and, therefore, is not an acceptable seismic Category I makeup source. It should also he noted that, according to the information provided in the September 1, 1987 submittal, the service water system, RHR service water systen, and alternate cooling cell require Thethe use n# licensee the RHRSW pump cross connection in order to provide SFP makeup. has identified a connection between the discharge of the RHRSW pump and the RHR cross-tie line. However, the licensee has not indicated that this cross-tie line is seismic Category I. The licensee should provide additional information to demonstrate that any of the above identified sources of SFP makeup is seismic Category I or propose a suitable seismic Category I alternative.

4. Radiological Consequences of Boiling  !

SRP Section 9.1.3 addresses ventilation and filtration to protect aceinst unacceptable radiological consequences due to boiling. In order to confirm the licensee's statements concernino radiological releases during SFP boiling, the staff reviewed the licensee's analysis. The licensee provided the results of

   '                      its offsite dese consequence analysis but has not provided the assumptions related to that analysis. Further, the licensee has provided neither the assumptions nor .the results of its analysis relating to personnel exposure.                           ;

The licensee should provide additional information on the assumptions and ' results of the offsite and onsite radiological consequences of SFP boiling. 1 e 1

l h a; l

5. Sucolenental Coolina.

In the. April 25,-1986 submittal, the licensee indicated that the RHR'systen-would be used to provide supplemental SFP cooling. Supplemental cooling is recuired based on calculated heat. loads, as discussed previously, for the

                    'first 68 days following reactor shutdown. While the licensee noted that SFP cooling'with the SFP heat exchangers in parallel was possible, sufficient
                    .information has not been provided to thoroughly describe the SFPCS cooling capacity in this mode.

The staff notes that in the September ".,1987 submittal, the licensee identi#4ed a scenario involving spent fuel pool accidents i.e., e seismic event with the resulting loss of offsite power and a single f ailure of one train of RHR. The licensee. indicated that manual operator action would'be utilized to transfer the one RHR train between the torus cooling and augmented fuel pool cooling medes as necessary to maintain the fuel pool temperature below 200*F and the torus- . temperatures below 150'F. .The licensee. maintained that reactor and spent fuel pool cooling could be provided in this manner followine an earthoueke. The operator action cycle enteils approximately 6 hours of torus ' cooling followed by 1 heur of augmented fuel pool cooling with 20 minutes allowed for valve realignment at each transfer. This scenario represents approximately six. starts of the RhR' pump per day with a do7en or more operator actions per day. This increase in the number of operator actions represents an increase in the probability of failure of the operable RHR pump and is not appropriate for mit-iaation of design basis accidents. , Based upon.the information reviewed to date, cycling the RHR systez between the torus and the spent fuel pool is not an acceptable method to cool the spent fuel pool. 1

t v M Of0p c9 UNITED STATES

      !"             'o*%            NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 h
                     )                                                  .

M M 2 0 1988 kC$g~,,IAC'2 s ,30g Charles Bechhoefer, Esq. Mr. Glenn O. Bright Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing Board Panel Board Panel U.S. Nuclear Pegulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. James H. Carpenter Administrative Judge , Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 In the Matter of VERMONT YANKEE NUCLEAR POWER CORPORATION (Vermont Yankee Nuclear Power Station) Docket No. 50-271-OLA (Spent Fuel Pool Amendment)

Dear Administrative Judges:

This letter forwards documents that the NRC staff is issuing today relating to its issuance of an amendment authorizing Vermont Yankee Nuclear Power Corporation to place new racks designed to accommodate 2870 spent fuel assem-blies in the spent fuel pool at Vermont Yankee Nuclear Power Station. The amendment does not authorize use of the racks to store spent fuel beyond the 2,000 assemblies previously authorized. This amendment is in partial response to the application dated April 25, 1986 but does not relate to the sole matter at issue before the Licensing Board: whether the spent fuel pool cooling system and the residual heat reinoval system are single failure proof. In any event the Staff has made a final no significant hazards consideration determination with respect to this amendment. Sincerely, p C M 0% - Ann P. Hodgdon Counsel for NRC Staff cc: Service List

s. .

  +       /p3 Cui   o                            UNITED STATES
         !'     #      kE-              NUCLEAR REGULATORY COMMISSION

g. WASHINGTON, D. C. 20555

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                    , /

MAY 2 01388 Docket No. 50-271 Mr. R. W. Capstick Licensing Engineer Vermont Yankee Nuclear Power Corporation 1671 Worcester Road Framingham, Massachusetts 01701 1

Dear Mr. Capstick:

) )-

SUBJECT:

SPENT FUEL POOL EXPANSION RERACKING - AMENDMENT NO. 104 ' (TAC N0.61351) The Commission has issued the enclosed Amendment No.194 to Facility Operating License No. OPR-28 for the Vermont Yankee Nuclear Power Station. The amendment is in partial response to your application dated April 25. 1986, as supplemented on August 15, September 26, October 21, and November 24, 1986, 1 and February 25 March 31, April 9. April 13, May 22, June 11 September 1, . and December 11, 1987, and March 2, 1988. i l This amendment allows the installation of racks of a new design in the spent fuel pool sufficient to accommodate 2870 fuel assemblies, and the storage of fuel assemblies in the new racks up to the present Technical Specification limit of 2000 assemblies in the pool. Use of the remaining 870 storage positions for the storage of fuel assemblies is not authorized by this amendment. The reracking is to be achieved by removing the existing racks and installing new, higher density racks of an improved design. Your request for amendment was noticed in the Federal Reaister on June 18, 1986 (51FR22226) and again on December 31, 1986 (51FR47324) with respect to no significant hazards consideration determination and opportunity for hearing. On January 29, 1987 The New England Coalition on Nuclear Pollution (NECNP)'and the State of Vennont petitioned to intervene and on January 30, 1987 the Commonwealth of Massachusetts petitioned to intervene. Following the ruling on contentions by an Atomic Safety and Licensing Board and a subsequent ruling by an Atomic Safety and Licensing Appeal Board, only one contention remains. That contention concerns the single failure proof characteristics of the spent fuel pool cooling system and the residual heat removal system. This amendment does not change the heat load on the spent fuel or residual heat removal system, and is unrelated to the issue pending before the ASLB. The comments and concerns relevant to this amendment are addressed in the enclosed Safety Evaluation. The Safety Evaluation also includes a final determination of l No Significant Hazards Consideration. In a meeting with you on February 9,1988, the staff learned that you plan to make substantial improvements to your spent fuel pool cooling system. The staff will complete review of the thermal-hydraulic aspects of your proposal and consider a decision regarding your request to increase the Technical Specification  ; limit to 2870 assemblies after learning more about your plans for enhancing the l spent fuel pool cooling system. l 1 _.____ _______ - _ w

i 7 Under-NRC regulations, the Commission may issue and make an amendment . immediately effective, notwithstanding a request for a hearing, in advance of holding the hearing where, as here, it has been determined that the amendment involves no significant hazards consideration. Such issuance is also consistent with Section 132 of the Nuclear Waste Policy Act of 1982, which requires the Commission to encourage and expedite the effective use of available storage at civilian reactor sites. A copy of the Safety Evaluation is enclosed. A copy of Notice of Issuance and i Final Determination of No Significant Hazards Consideration is also enclosed. 1 The Notice of Issuance will also be included in the Commission's bi-weekly Federal Register notice. Sinc ely, Vernon L. Rooney, Project M nager Project Directorate I-3 Division of Reactor Projects I/II

Enclosures:

As Stated  ! cc: See next page og 1 1

e ' s Mr.'R..W. Capstick . Vermont Yankee Nuclear Power Corporation ' Vermont Yankee Nuclear Power Station 1 [ cc:~ Mr. J. Gary Weigand 1 W. P. Murphy, Vice President. President & Chief Executive Officer and Manager of Operations.

        -Vermont Yankee Nuclear Power Corp.          Vermont Yankee Nuclear Power-Corp.

R.D. 5, Box 169 ~R.D. 5, Box'169 .-

         < Ferry Road'                               Ferry Road                                                                                            j Brattleboro, Vermont 05301                 Brattleboro. - Vermont 05301 Mr. John DeVincentis, Vice President      Mr. Gerald Tarrant, Commissioner                                                                         i
         ; Yankee Atomic Electric Company            Vermont Department.of Public Service L        '1671 Worcester Road                         120 State Street-Framingham, Massachusetts 01701          -Montpelier, Vermont 05602 New-England Coalition on Nuclear           Public Service Board-Pollution                                State of Vermont
        . Hill and Dale Farm                         120 State Street
        'R.D. 2, Box 223                            Montpelier, Vermont 05602 Putney, Vermont 05346 Vermont Public Interest Research                                                                                                               .
                                                                                                                                                         ~

Group, Inc. Mr. Walter Zaluzny  !

        .43 State Street                            Chairman, Board of Selectman                                                                             l Montpelier, Vermont 05602                  Post Office Box 116 Vernon, Vermont 05354 William Pussell, Regional Administrator    Raymond N. McCandless Region I Office                            Vennont Division of Occupational U.S. Nuclear Regulatory Commission            and Radiological Health 475'A11endale Road                .

Administration Building

        . King of Prussia, Pennsylvania 19406       Montpelier, Vermont 05602 Mr. R. W. Capstick                         Honorable John J. Easton                                                                                 l Vermont Yankee. Nuclear ~                  Attorney General Power Corporation                          State of Vermont 1671 Worcester Road                        109 State Street Framingham, Massachusetts 01701            Montpelier, Vermont 05602 John A. Ritscher, Esquire                  Mark J. Wetterhahn, Esq.

Popes & Gray Conner & Wetterhahn, P.C. 225 Franklin Street Suite 1050 Boston, Massachusetts 02110 1747 Pennsylvania Avenue, N.W. Washington, D.C. 20006 i

            .-Vermont. Yankee Nuclear Power       Vermont-Yankee Nuclear Power StationL            i
                . Corporation.

l cc: 4 Ellyn,R. Weiss, Esq.- Resident Inspector. Harmon & Weiss- U.S. Nuclear Regulatory Comission -

             '2001 S Street, N.W.                     P.O. Box 176 Washington, D.C. 20009                  Vernon, Vermont 0535A David J. Mulletto: Esq.                 Carol S. Sneider, Esq.
             .Special Assistant Attorney General      Assistant Attorney General Vermont' Depart. of Public Service      Office of the Attorney General 120 State Street                        One Ashburton Place, 19th Floor Montpelier, VT 05602                    Boston, MA 02108 Jay Gutierrez                           Geoffr'ey M. Huntington, Esquire Regional Counsel                        Office of the Attorney General USNRC, Region I                         Environmental Protection Bureau 475 Allendale Road                      State House Annex King'of Prussia -PA: 19406              25 Capitol Street Concord,.NH 03301-6397 G. Dana Bisbee, Esq.                    Charles Bechhoefer,:Esq.                 .

Office of the Attorney General - Administrative Judge Environmental Orotection Bureau Atomic Safety and Licensing Board State House Annex U.S. Nuclear Pegulatory Comission 25 Capitol Street. Washington, DC 20555 Concord, NH- 03301-6397 Dr. James H. Carpenter

                                                     . Administrative Judge Atomic Safety and Licensing Board       Atomic Safety and Licensing Board U.S. Nuclear Regulatory Comission       U.S. Nuclear Regulatory Commission Washington, DC 20555                    Washington, DC 20555 Mr. Glenn 0. Bright -                   Adjudicatory File (2)

Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board Panel Docket U.S. Nuclear Regulatory Comission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, D.C. 20555 1 L_____-____-- _

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                     \ (,,,, #,a                                VERMONT YANKEE NUCLEAR POWER CORPORATION
                                                                             '0OCKET NO. 50-271 VERMONT: YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE.

Amendment No.104 License No. DPR-28

1. The Nuclear Regulatory Comission (the Comission) has found that:

A .- The application for amendment by Vermont < Yankee Nuclear Power Corporation (the licensee) dated April 25, 1986, as supplemented on August 15, September-26, October 21, and November 24, 1986, and-February 25, March'19, March 31, April 9 April 13, May 22, June 11 September.1, and Decer:ber 11,'1987, and, March 2', 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR' Chapter.I; _ B. 'The facility will operate in conformity with the application, the provisions of the Act, and the rules'and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amer.Jment can be conducted without endangering the health and= safety of the public,'nd~(ii) a that such activities will be conducted in ' compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the-common defense and' security or to the health and~ safety of.the public; and

                                             ~E.      The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by authorizing the installation of ,

spent fuel racks as: described in pages 4 through 6 of the licensee's i April 25, 1986 application for amendment. The request to increase the number of fuel assemblies allowed to be stored in the spent fuel pool is . not granted at this time. l

3. This license amendment is effective immediately.

F HE NUC EAR R GULATORY COMMISSION teven . Varga, Director Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Date of Issuance: May 20, 1938

a o UNITED STATES - f ' g J. g 17, j NUCLEAR REGULATORV COMMISSION . WASHING TON, D. C. 20555

          %, * * "

• f SAFETY EVALUATION BY.THE OFFICE OF-NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0.104 TO FACILITY OPERATING LICENSE NO.DPR-25 VERMONT YANKEE NUCLEAR POWER CORPORATION

                                                                    ' VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 ,

1.0~ INTRODUCTION By letter dated April 25, 1986 Vermont Yankee Nuclear Power Corporation (VYNPC,),

             .the licensee, requested a change to Section 5.5.0 of the Technical Specifications-for Vermont Yankee Nuclear Power Station (VY).                                                                 This change would increase the number of fuel assemblies which could be stored in the spent fuel pool from 2,000 to 2670. Other.previously approved specifications of Section 5.5 would remain unchanged.                The change is based on the installation of new fuel racks in the spent fuel pool which can provide a closer packing of fuel assemblies. Required criticality margins are maintained by incorporation of boron containing material in the rack design. This is a commonly used feature for high density rack design
                                                                                                                                                                                      ~

design, and a large number of similar designs have been approved by the NRC. At this time, the staff is granting the proposed amendment in part: i.e., installation of sufficient fuel storage racks of new design in the pool to accommodate 2870 assemblies, and storage of fuel assemblies in the new racks up to tne present Technical Specification limit of 2000 assemblies in the pool. Use of the remaining 870 storage positions for the storage of fuel assemblies is not authorized by this amendment. I

2.0 BACKGROUND

l l l VY is a General Electric Company Boiling Water Reactor (BWR) which received an operating license on March 21, 1972. At the time of licensing, the spent i L

fuel pool contained sufficient storage locations to accommodate 600 fuel assemblies. The spent fuel pool cooling system consists of two redundant trains wit.h each train consisting of one 450 gpm pump and one heat exchanger. The design capability of each heat exchanger is 2.23 MBtu/Hr with a pool water temperature of 125 . The spent fuel pool cooling system is non-seismic Category I, non-Class IE. VYNPC received approval to replace the original spent fuel storage racks with high-density spent fuel storage racks in September 1977. These high-density racks were to be installed in phases,providing a total maximum storage capacity of 2000 fuel assemblies. To date, the licensee has installed racks sufficient to store 1690 fuel assemblies. On April 25, 1986 the licensee requested approval to rerack the spent fuel pool for a second time. This second rerack application is the subject of this safety evaluation report. The new high density storage racks would  : increase the storage capacity of the spent fuel pool to 2870 fuel assemblies and is projected to provide storage capacity until 2001. The licensee provided additional information on the proposed second rerack request in submittals dated August 15, September 26, October 21 and November 24, 1986, and February 25, March 19, March 31, April 9, April 13, May 22, June 11, September 1, and December 11, 1987, and March 2, 1988. The licensee also incorporated by reference information contained in submittals dated September 11, 1981, November 30, 1983, and May 21, June 27, and December 18, 1984. Information related to the licensee's computer modeling of spent fuel pool cooling was provided at a meeting on January 15, 1987 in Richland, Washington. In the April 25, 1986 submittal, in addition to requesting approval to rerack the spent fuel pool, the licensee identified necessary changes involving removal of the spent fuel pool cooling system return line spargers and related piping inside the spent fuel pool. In a submittal dated September 1, 1987, the licensee further defined this request by proposing to cut off the Spent Fuel Pool Cooling System (SFPCS) return line at approximately 15 feet above the top of the racks (which is approximately 8 feet below the fuel pool water level). This modification would provide

o for the storage of an additional 100 fuel assemblies. The licensee stated

      - that the natural circulation developed by the heat generated by the spent fuel will provide adequate cooling for the spent fuel.

The staff issued a status report dated January 21, 1988, which discussed five technical open issues related to the licensee's request to increase the storage car--ity of the spent fuel pool to 2870 fuel assemblies. Some of these open issues involved the fuel pool cooling system and its cooling capacity. These open issues also involved increased heat load due to an increase from the present 2000 fuel assemblies limit to the requested 2870 limit. The staff met with the licensee on February 9, 1988 to discuss these issues. During the meeting, the licensee revealed that it had reached a decision to design, build, and install an enhanced cooling system for the spent fuel pool. This modification was proposed for the purpose of expediting resolution of  : outstanding issues. Subsequently, the licensee in a submittal dated March 2, 1988, documented its commitment to install an enhanced cooling system. Although no details of the modified design were provided, the licensee did provide some design and performance information for the enhanced SFPCS. In order to allow reracking to commence in such a way that personnel radiation exposure is minimized, without awaiting completion of review with respect to enhanced cooling, at this time the staff is considering the portion of the proposed expansion involving reracking and placement of the new racks in the pool but is not considering the storage of more than 2000 assemblies in the pool. Consideration of storage of more than 2000 assemblies will await a determination of the adequacy of spent fuel pool cooling for more than 2000 assemblies, including the yet-to-be-designed enhanced spent fuel pool cooling system. Before completing review of the proposed expansion the staff requires more information than is presently available related to the enhanced spent fuel pool cooling system. Sufficient information is presently available, however, to enable the staff to consider whether or not it is safe to store spent fuel in the new racks up to the present Technical Specification limit of 2000 fuel assemblies, and whether

t 4

                                - the additional new racks.for future storage of 2870 fuel assemblies may be
   ~

safely installed in the fuel pool. Because of the procedure by which the expansion must be accomplished, it is advantageous to consider the reracking process before sufficient information is available to reach a conclusion with respect to storing 2870 assemblies. In order to begin the proposed expansion, the licensee must place a new rack in the pool and transfer fuel presently stored in an old rack to the new rack. The empty old rack is then removed to make room for another new rack, and the process is repeated until all fpel has been transferred to new racks. Additional new racks will than be added to provide space for future storage. It is expected that several months will be required to complete this task for the inventory of irradiated fuel presently in the Vermont Yankee storage pool. If more irradiated fuel were added to the present inventory stored in old racks, the reracking would take even longer and require even more personnel radiation exposure than is required presently. 3.0 EVALUATION 3.1 Criticality Consideration Reauired criticality margins are maintained by incorporation of boron cortaining material in the rack design. The rack design (described in detail in VYNPC's letter of September 25, 1986) is configured so that the bcron associated with the cells, in the form of Boral, is arranged such that there is boron between each pair of fuel assemblies. This incluoes the Boral on the outer edge of racks, which is arranged so that there is boron between assemblies facing each otier across rack gaps. The B-10 loading of the Boral is 0.027 gm/cm2 minimum. The cell pitch is 6.218 inches and the cell inside width is 5.922 or 6.092 inches (fuel assembly with channel is 5.438 inches). The criticality calculations for the new racks were performed by Yankee

v I Atomic Electric Company (YAEC). The calculations were performed with two methodologies. The reference criticality analyses were performed with I the Monte Carlo code KENO-IV using the NITAWL code to provide cross sections based on the XSDRN code cross section library. For sensitivity calculations and trend analyses the diffusion code PDQ-7 was used with cross sections from the CASM0 code. All of these codes and cross  : I sections are well known industry standards, frequently used for analyses I of fuel pools and other complex criticality problems, and have been approved by the NRC. YAEC has benchmarked its KEN 0 methodology against a number of relevant  ; critical experiment results from Babcock and Wilcox and Battelle Northwest Laboratories. These experiments present geometrically representative configurations for fuel racks. YAEC has used these benchmark calculations to develop an analysis methodology uncertainty - factor to be added to rack k eff calculations. YAEC has also determined the potential variation of the rack and fuel parameters that are used in determining the keff f the rack-fuel I system. These parameters include poison thickness, boron concentration, cell pitch, stainless steel thickness and eccentric fuel position. The variation of k eff with these parameters (taken at a 95/95 probability / confidence level) was determined. These (independent I parameters) were statistically combined to provide a AK uncertainty l which, along with the Monte Carlo statistical uncertainty, is added to the base k eff calculation. YAEC has investigated abnormal conditions that might be associated with the spent fuel pool and has determined that potential reactivity variations caused by abnormal pool conditiens and accidents have either negligible or negative effects on keff . These include changes in pool I temperature from the base conditions, cell or rack displacement from seismic incidents, fuel or heavy object drop events, and fuel assembly placement outside of the racks. Thus k eff f r the fuel pool is determined, both for normal and abnormal conditions by adding the previously discussed method and mechanical uncertainties to the base calculation, without the need for additional factors to account for abnormal conditions.

        .t        y 3

r r For the' base case, the YEAC Monte' Carlo calculations assume (1) an l.

                       ' infinite square array of cells' (2)'witii a pitch of 6.218 inchesi each-containing (3)'an unirradiated fuel assembly,of 64 fuel rods (no water-rods) with /4) a uniform enrichment of:3.25 weight percent U-235, (5) no burnable poison and-(6) infinite length. The water temperature is t                       6'8 F. ~This. fuel assembly enrichment bounds present fuel enrichments and the.use of no burnable poison provides conservatism for reactivity calculations.

For the base configuration, the k eff was cal'culated to be 0.9046. The total uncertainty.at a 95/95 level was 0.0221 ok, giving a total k gff of 0.9267. This is to be compared to a required upper limit of 0.95. The fuel assembly' lattice used for the base rack calculations was calculated to have a standard reactor core geometry uncontrolled km valug of.1.35. YAEC proposed, in.the initial submittal, to use a. fuel assembly . km of 1.35 as the design bases for fuel acceptable for. storage in the racks.(rather than a fuel enrichment limit). This is common practice for BWR fuel storage (see for example NEDE-24011-P-A-8, May 1986) and allows credit for the burnable poison in the fuel assembly in the analyses to meet'the Technical Specification requirement of 0.95. As a result of

                    ' discussions' with the staff concerning the nature of ' additional uncertainties involved when using a km design criteria, this proposed limit was reduced to 1.31 by VYNPC by letter dated October 21, 1986.               The possible reactivity effects of (1) nonuniform enrichment variation in the assembly, (2) uncertainty in the calculation of km and (3) uncertainty in I

average assembly enrichment were exaniined and quantified by YAEC, providing the additional correction factor of 0.04 AK. The basic criticality design of the new racks, using boron lined cells to provide the appropriate neutron multiplication level for the closer packed array of high density racks, is a commonly used concept and has been accepted for many spent fuel storage pools. It is an acceptable design concept for maintaining criticality levels for the VY pool. 1 i i i j

     .                                     l-The methodology used by YAEC to analyze the criticality and reactivity change characteristics of the racks is a state of the art methodology,,

commonly used and approved for other utilities for such analyses. The Monte Carl? method using the KEN 0/NITAWL/X50RN packaoe provides 'an acceptable methodology for the base calculations and the P0Q/CASMO is l acceptable for sensitivity calculations. These methods have been L benchmarked againct an appropriate selection of critical experiments, with results falling within expected ranges of deviations from the experiments. The derivation of the uncertainty of the methodology from tnis benchmarking follows normal procedures and also. falls within an. expected range. It is acceptable. ,

       .The examination of uncertainties to be attributed to variances in dimensions and materials in the fuel and racks has covered an acceptable range of. parameters and has used a suitable, standard methodology for determining the reactivity effects and their statistical combination.                                             ,

The examination of the effects of abnormal conditions has covered the standard events relating to changes in temperature, movements, misloadings and dropping of assemblies and other equipment', and the results, giving nonpositive reactivity additions, are reasonable and acceptable. The base calculations and added factors for uncertainties, giving a total k,ff of 0.9267, are thus acceptable for an average enrichment of 3.25 percent. There is a margin of 2.3 percent AK to the staff required

       . Technical Specification limit of 0.95. The transfer to a fuel assembly km design basis criterion has conservatively considered relevant additional uncertainty fcctors, and the resulting design basis km value of 1.31 is acceptable. The approach of using a km design basis has been approved in other applications, and is used in the staff approved General Electric reload analysis approach (as given in GESTAR II, NEDE-24011-P-A-8, May 1986).

The base k,ff criterion of 0.95 given in Technical Specification 5.5.B remains the same. Also unchanged by this request is the average enrichment limit of 16 grams of U-325 per longitudinal centimeter of assembly. This specification is compatible with the 3.25 percent U-235 enrichment used in the base calculations.

}.

                                                                                "O' Therefore,.it is concluded that the required criticality margins are

[ maintaine'd by the new racks. 3.2, Structural Engineering The new high density racks are stainless steel " egg-crate" cellular structures of approximately 6 inches square. Each cell is designed to

          .contain a spent fuel assembly and a typical rack consists of approximately 300 cells whose dimensions are approximately 10-feet long by.8 feet wide and 15. feet high. Weight of'the rack and fuel is transmitted to the floor of the pool through supporting legs. The racks are each free-standing on the pool floor and a gap is provided between the racks and between the racks'and the pool wall so as to preclude impact during earthquake. Such design provides a margin of safety against tilting and deflection movement.

The spent fuel pool is a reinforced concrete structure supported by the Reactor Building walls. The pool is approximately 26 feet wide by 40 feet long by 39 feet high and is completely lined with seam I welded ASTM A240 Type 305 stainless steel. The licensee's load combinations and acceptance. criteria were found to be consistent with those in the " Staff Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and amended January 18, 1979. The existing concrete pool structure was evaluated for the new loads in accordance with the requirements of the appropriate industry codes such as ASME Section III and ACI 349-80 and the NRC staff guidelines and  ! documents such as Standard Review Plan (NUREG-0800) and Regulatory Guide 1.92 " Combining Model Responses and Spatial Components in Seismic Response Analysis." Seismic loads for the rack design are based on the original design floor acceleration response spectra calculated for the plant at the i licensing stage. The seismic loads were applied to the model in three orthogonal directions. The hydrodynamic loads of pool water acting on pool walls are considered. Loads due to a fuel bundle drop accident were considered in a separate analysis. The

4

 ,                                                  postulated loads from these events were found to be acceptable.

The dynamic response and internal stresses and loads of the racks and pool structure are obtained from a time history seismic analysis. Nonlinear time history analysis are performed utilizing the widely-used industry ANSYS code. Friction between rack support pads and pool floor and hydro-dynamic coupling are considered in the analysis. Calculated stresses for the rack components were found to be within allowable limits. The racks were found to have adequate margins against sliding and tipping. An analysis was conducted by the licensee to assess the potential effects of a dropped fuel assembly on the racks. The external kinetic energy will be absorbed by rack strain energy through deformation of the rack cells. The overall integrity of the rack will not be adversely affected. The existing structures were analyzed by the licensee for the modified fuel rack loads. The strength design method for reinforced concrete wa:; used in conjunction with conventional structural analysis procedures to determine capacities. The existing spent fuel pools are determined to safely support the loads generated by the new fuel racks. It is concluded that the proposed rack installation will satisfy the requirements for 10 CFR 50, Appendix A, GDC 2, 4, 61 and 62, as applicable to structures, and is therefore acceptable. 3.3 Compatibility and Chemical Stability of Rack Materials The staff reviewed the compatibility and chemical stability of the new rack materials wetted by the pool water. The licensee supplemented the original submittal dated April 25, 1986, with additional information regarding rack materiais by letter dated March 31, 1987. The proposed spent fuel racks are to be constructed entirely of Type 304L stainless steel, except for threaded rods attached to leveling pads which are 17-4 The 17-4 [ l PH-hardened stainless steel and the neutron absorber material. PH threaded rods are heat treated, chemically cleaned and chrome plated. The neutron absorber material is Boral with a minimum BIO loading of 0.027

1 gms/cm2 Boral is a dispersion of' boron carbide in' an aluminum matrix with an aluminum clad. The spent fuel rack compartments containing the Boral are not watertight. This will allow venting of gas generated by radiolysis of contained water and by Boral off gassing, preventing pressure buildup and possible swelling. The'austenitic stainless steel (304L) used in the rack fabrication has'a maximum carbon. content of 0,03% by weight which minimizes the sensitization in weld heat-affected zones. The stainless steel rack's are compatible with the spent fuel pool water that is processed,by filtration and demineralization to maintain water purity and clarity. The spent fuel pool purity is maintained at < 1p s/cm conductivity at 25 C, < 500' ppb chloride, < 100 ppb total heavy elements, and a pH range of 5.8 to 8.0. Intergranular corrosion tests performed in accordance with ASTM A262, Practice E are required fof the austenitic stainless steel. Dissimilar metal contact corrosion (galvanic attack) between the stainless steel rack assemblies, aluminum in Boral neutron absorption plates and zircaloy in the fuel assemblies will not -* significant because the materials are protected by highly passiviting' oxide films and are, therefore, at similar galvanic potentials. Dral has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron aosorbing material. Boral has been qualified for 10 11 rads of gamma radiation while maintaining its neutron attenuation capability. The annulus space in each cell assembly which contains the Boral is vented to the pool to allow venting of radiologic gases and Boral outgassing. This will prevent swelling and bulging of the stainless steel plates. Tests have shown that Boral does not possess leachable halogens that could be released into the pool environment in the presence of radiation. Similar conclusions have been made regarding the leaching of elemental boron from the Boral.

4 To provide added assurance that no unexpected corrosion or degradation of the materials will compromise the integrity of the racks, the licensee has committed to conduct a long-term poison coupon surveillance program. Surveillance samples in the form of stainless steel retained sheets of Boral (prototypical of the fuel storage cell walls) will be exposed to tne spent fuel pool water. These coupons will be removed and examined periodically over the expected service life. The staff has reviewed the description of the proposed surveillance program for monitoring the Boral in the spent fuel storage pools and concludes that the program can reveal deterioration that might lead to loss of neutron absorbing capability during the life of the spent fuel racks. The staff does not anticipate that such deterioration will occur, but if it would occur, it would be gradual. In the unlikely event of Boral deterioration in the pool environment, the monitoring program will : detect such deterioration and the licensee will have sufficient time to take corrective action, for example, replacement of the Boral sheets. Based on the above discussion, the staff concludes that the corrosion of the spent fuel pool components due to the spent fuel storage pool environment should be of little significance during the life of the facility. Components in the spent fuel storage pool are constructed of alloys that have a low differential galvanic potential between them and i have a high resistance to general corrosion, localized corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in water indicate that the Boral material will not undergo significant degradationduringtheprojectedservicelifeofapproximately40 years for the racks. The staff further concludes that the environmental compatibility and stability of the materials used in the spent fuel storage pool is adequate based on the test data cited above and actual service experience at operating reactor facilities. Finally, the staff finds that implementation of the proposed monitoring program and the selection of appropriate materials of construction by the

s 7 licensee meet'the requirements of 10 CFR 50 Appendix'A, General Design Criterion 61,. regarding the capability to permit appropriate periodic

 ,          . inspection and testing of components and General Design Criterion 62 regarding preventing criticality by maintaining structural integrity of-components and of boron poison, and are, therefore, acceptable.

3.4 Occupational Radiation Exposure The staff has reviewed the licensee's plan for expansion of the spent fuel pool storage capacity with respect to occupational radiation exposure and finds that the ALARA policy, design, and operational considerations are. acceptable. This finding is based on the licensee ~having considered the provisions of 10 CFR Part 20.101,'20.1(c) and 20.103, and.the guidelines of Regulatory Guides 8.8 and 8.10 with respect to the planned expansion. The licensee is currently developing specific work packages for the rerachk l project and has set a dose goal of 20 man-rem. This 20 man rem dose goal is based on_ information gained by reviewing the experience at other operating nuclear-plants that have recently performed similar spent fuel pool modifications. To meet their dose goal (20 man-rem) end the ALARA provision of 10 CFR 20.1(c)', the licensee has assigned an engineer ALARA responsibility to review and approve each work package for the project. The 20 man-rem dose goal represents about 2% of the average annual occupational exposure at the Vermont Yankee plant. The 20 man-rem dose' goal includes all activities necessary for the reracking operation including vacuum cleaning of the SFP walls and floor; shuffling fuel, installation of the new racks; removal of the old racks; cleaning decontamination, and any necessary cutting of old racks; and disposal of waste resulting from the reracking operation, including the old r;tcks. In terms of radiation dose to workers, the spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel. However, one potential source of radiation to workers during the rerack operation is radioactive activation or corrosion products, which are referred to as crud. Crud may be released to the pool water because of fuel movement during the proposed SFP modification. This could increase radiation levels in the

        ,               4 tvicinity'of the pool. The addition'of. crud to the pool water is greater during refuelings, when the spent fuel is moved first into the fuel pool. -It is'at this time that most of the additional crud is introduced into the pool' water from the fuel assembly and from the introduction of primary coolant. However, significant' releases of crud to the pool water during the'rerack operation is'not expected, since the new racks are.            .

cleaned prior to installation. In addition, the purification system for l the pool, which keeps radiation levels in.the' vicinity of the pool at low levels, includes.a filter to remove crud. This filter will be operating e . during the modification of the pool. By letter dated November 24, 1986, the licensee provided information describing actions to be taken during spent fuel pool (SFP)

                                                            . modification. Some of the ALARA ' activities for reducing.th'e occupational radiation dose include:

(a) vacuum cleaning of the SFP floor and walls'as required; l

                                                                   .(b) -hydrolysing and cleaning of old spent fuel racks; (c) use of remote operations for rack removal and replacement operations; and, (d) utilizing the SFP Filtration System to maintain clean water in the pool The licensee also has provided a description of contained and airborne radioactivity sources related to the SFP water which may become airborne as a result of failed fuel and evaporation.       The staff has reviewed these source terms and finds them to be acceptable.

Recently there has been a concern expressed that a severe reactor accident could lead to loss of water from the spent fuel pool. Specifically, if the pool cooling system was disabled as part of the reactor accident sequence, and repa.rs of this system were precluded for several weeks, due to high radiation fields around the plant, then it is possible to postulate a reduction in SFP water inventory. Vermont Yankee, as wcll as other nuclear plants, employ a defense in depth concept for early warning of,

so and subsequent protective actions in response to, any accident or abnormal-occurrence, including a loss of cooling to the spent fuel pool.

                          , Early warning via monitoring systems and precautions called for by the plant's health. physics program assure minimum radiation dose'to workers during both normal and abnormal conditions. The spent fuel pool has
                            ' temperature indicators, water level indicators, vent radiation monitors, an airborne radioactivity monitoring system and an area radiation monitoring.
                                                      ~

system. The water temperature and level indicator provide redundant and diverse means.of detecting loss of cooling to the spent fuel pool even during an accident. They provide an early warning, so that corrective actions can be made to restore cooling or to add water before the water level in the spent fuel pool decreases due to boiling. In addition to the monitoring system and the plant's overall health physics program, the effects of any accident or abnormal condition on , personnel including spent fuel pool boiling can be mitigated by implemen-tation of the licensee's emergency plan, which contains re-entry criteria for entering potentially high radiation areas. On the basis of the above, the staff finds that the projected activities and the dose goal of 20 person-rem for the proposed spent fuel pool expansion is reasonable. Further, we find that the licensee intends to take ALARA considerations into account, to implement reasonable dose reducing activities. Hence, the licensee will be able to maintain individual occupational radiation exposures within the applicable limits of 10 CFR Part 20, and meet the guidelines of Regulatory Guide 8.8. The staff, therefore, finds that the occupational radiation protection aspect of the spent fuel pool modification program is acceptable. 3.5 Radioactive Wastes The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radioactive material. The radioactive waste treatment systems have beer previously evaluated by the staff and found acceptable. There will be no change in the radioactive waste treatment systems or in the conclusions given regarding the evaluation of these systems as a result

  -p:

of the proposed installation of'the new racks. Our evaluation of.the-radiological considerations supports the conclusions that;the proposed

                      '-installation ofinew scent fuel storage racks at Vermont Yankee is acceptable. The basis: for our conclusions 'is that the previous evaluation of the radioactive waste treatment systems are unchanged by-l:                      the-installation of new spent fuel storage racks.

The present spent fuel racks will be removed from'the SFP and will-probably be disposed of as low level waste. If the existing racks are-disposed of as solid waste, the volume will te approximately.2000 cubic feet. The annual average volume of solid wastes shipped'offsite for burial from Vermont Yankee has been approximately 400 cubic meters. Averaged over the lifetime of the plant the addition of the existing spent fuel racks will increase the total waste volume shipped from the facility by less.than 0.4%. This would not'have any significant additional-environmental impact over:that contemplated and discussed in the FES for the operating license application (U.S. Atemic Energy Commission, Environmental Statement Related to the Operation of Vermont Yankee Nuclear Power Station, July 1972). 3.6 Load Handling 3.6.1 Light Loads A light load is a load that weighs less than the combined weight of a fuel bundle, channel and its handling tool. Since there are , no restrictions on the handling of light loads over the spent fuel, a light load could be carried which, if dropped, could have sufficient kinetic energy and impact force on the fuel or rack to potentially resuit in greater. damage than assumed in a fuel handling accident. In the licensee's September 1, 1987 submittal, the licensee stated that an analysis of light loads normally carried over the spent fuel was performed. The licensee identified a light load to be any load which weighs 700 lbs or less. The results of this analysis indicate that the kinetic energy of these loads is less than that of the design basis fuel handling accident and thus the radiological

consequences of a light load drop are bounded by the fuel handling accident. The staff, therefore, considers handling of light loads to be acceptable. 3.6.2 Heavy Loads Spent fuel storage racks weigh more than a fuel assembly, channel and its handling tool. Thus, spent fuel storage racks are considered to be heavy loads. The reactor building crane will be used to move the storage racks within the reactor' building and the spent fuel pool. As part of the review of the Vermont Yankee facility for compliance with guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," the staff concluded in the Safety Evaluation Report dated June 27, 1984, that the reactor building crane was single failure proof by meeting the guidelines of NUREG-0554, " Single : Failure Proof Cranes for Nuclear Power Plants". In the November 24, 1986' submittal, the licensee provided information that showed the movement of spent fuel within the spent fuel pool, the order of rack replacement, and the path of travel for each of the fuel storage racks. The licensee demonstrated that the storage racks will not be carried over spent fuel or over other racks containing spent fuel. In a subsequent submittal Gated February 25, 1987, the licensee provided a drawing that showed the heavy load handling boundaries and laydown areas for the storage racks. The licensee demonstrated that to the extent practical, the paths of travel follow the fuel building structural floor I members and beams. The licensee also stated that the load paths and I laydown areas will be marked with stanchions and ropes prior to performing heavy load lift. Drawings will be provided to the crane operator in the cab and to the tag man directing the lift to assure f adherence to the load paths. The licensee committed to have all l deviations from the established load paths approved by management l personnel prior to being used. The licensee also committed to prepare installation and removal procedures specifically for the reracking of the spent fuel pool, and to provide qualification, training, and testing of crane operators, as described in D.P. 2201, " Reactor Building and Turbine l

                                                            -   - - - - - - - - -- -_ L

17-

                            ' Building Crane Operator Qualifications." This information hasJoeen reviewed by the staff and found to be acceptable.
                            .Two special lifting devices will be'used in the reracking, one.for the existing par racks and.one.for the new NES racks.. By. submittal dated May 22, 1987, the licensee.provided drawings of the' par spent fuel rack lifting rig which show redundanc.y in the lifting rig. The licensee committed to pre-operationally load test the par lifting-' rig
                            .to 150% of the empty spent fuel rack weight. By submittal dated           j April 13, 1987, the licensee provided dt awings of the NES' spent fuel-rack lifting rig, which also show redundancy in the lifting rig.

The licensee. committed to' pre-operationally load test the'.NES lifting rig to 150% of the empty spent fuel rack weight (equivalent to 27h tons) or a total load test equal to 30 tons. In the February 25, 1987 submittal,.the licensee committed to ensure that the special-lifting devices meet the guidelines of. ANSI N14.6-1978, and'to'pe_rform the load. tests and subsequent inspections in accordance with ANSI N14.6-1978. Based.on the.above review, the staff concludes that heavy. loads handling will be performed in accordance with the guidelines of NUREG-0612, and thus the requirements of General Design Criterion 61,

                              " Fuel Storage and Handling and Radioactivity Control", are met as they relate to proper' load handling to ensure against an unacceptable release of radioactivity, a criticality accident, or the inability to cool the spent fuel in the spent fuel pool due to postulated-load drops. The staff has determined that installation of the new high density racks to provide 2870 storage locations in the VY SFP is acceptable.

3.7 Spent Fuel Shipping Cask Drop Accident In the licensee's response to the staff's request for additional information dated November 24, 1986, it was stated that the Reactor Building Crane is considered to be single failure proof. Also, the cask drop height to the refueling building floor is less than 30 feet (Ref. i FSAR Section 12.2). Therefore, in accordance with Standard Review Plan Section 15.7.5, evaluation findings with respect to radiological l' .

consequences for a cask drop accident are not needed. The staff concludes that the proposed expansion meets the applicable criteria with respect to

                                                                                      , the spent fuel cask drop accident analysis.

3.8 Fuel Handling Accident The staff independently evaluated a postulated fuel handling accident following the guidance of Standard Review Plant Section 15.7.4, " Radiological Consequences of Fuel Handling Accident", and using the assumptions set forth in Regulatory Guide 1.25, " Assumptions Used fer Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors." The calculation was performed by using the staff computer code ACTCODE. :The staff conservatively assumed a 24 hour shutdown time for the two damaged fuel assemblies. Credit is given to the Standby Gas Treatment System (SGTS) in the reactor building because the system provides safety grade HEPA filters and charcoal absorbers. Credit is also given for the reactor building, since it maintains a slightly negative pressure during the accident. The radioactivity produced by this accident is processed by the SGTS, which has a 95% removal efficiency for radioactive iodines. The resulting radiological doses at the EAB are 2.58-rem for the thyroid, and 0.337 reu for the whole body. Similarly, at the LPZ, the doses are 0.361 rem for the thyroid and 0/047 rem for the whole body These doses are f.ar below the criteria of 75 rem for the thyroid and 6 rem for the whole body (SRP 15.7.4). Because VY's control room does not have charcoal and absorber filters, the staff also considered control room doses due to a fuel handling accident involving a radioactivity release. However, since this release is from a 300 foot high stack, and the atmospheric dispersion factors are in the order of 10-6and 10-7 s/n3, the effective radiological doses to the control room are estimated to be negligible. The assumptions used for this analysis are listed as in Table 1.

1 TABLE 1. ASSUMPTIONS USED IN FUEL HANDLING ACCIDENT 1 j j Reactor Power Level 1665 MWth Number of fuel assemblies in core 368 Number of fuel rods damaged 126 Standby Gas Treatment System filter efficiency for elemental and organic iodines 95% Cooldown time for impacted spent fuel 24 hrs Effective pool decontamination factor for iodine 100 l GAP ACTIVITY: l Iodine 10% , l Krypton 30% Total noble gas 10% other than Krypton Location Time Period X/Q  : EAB 0-2 hrs 0.25 x 10 -3 s/m2

                                                                                 ~4 LPZ                   0-8   hrs                   0.35 x 10
                                                                                 ~4 8-24 hrs                    0.21 x 10
                                                                                 -5 24-96 hrs                    0.70 x 10
                                                                                 -5 96-720 hrs                   0.15 x 10 l                                                                                                                         i The staff concludes that the proposed spent fuel pool expansion meets the                                             !

I applicable criteria with respect to the fuel handling accident analysis and l is acceptable. l l l

i' 3.9 -Spent Fuel Cooling System Modification l l The license' proposed to remove the return-line spargers and-to terminate the return-line in'a downward pointing direction at approximately 15 feet above the top of the spent fuel pool stordge racks (8 feet below the surface'of the water). 'With the spargers installed as originally licensed in 1972, the water from the SFPCS was returned at the bottom of the spent fuel pool below the spent fuel storage racks. The water generally traveled up through the racks as it passed to the far side of the spent fuel pool, thus providing " forced"' cooling of the spent fuel. With the. propose'd removal of the spargers, the water will enter. and exit 'I the pool at approximately the same elevation above the spent' fuel storage l racks. The mechanism for cooling the spent fuel in this configuration relies on natural circulation. The staff performed an independent spent-fuel cooling analysis to verify the licensee's claim that removal of the = spargers will not affect spent fuel cooling capability. The results demonstrate that because of adequate mixing in the upper plenum, the relatively open flow area below the fuel, and the 2-inch gaps around the periphery of the racks, adequate spent fuel cooling is provided regardless of the inlet flow orientation, or " loading patterns" of the hot assemblies within the pool. The primary factor controlling pool performance is the total pool heating rate to total pool recirculation  ; flow rate. Additional details of the staff's independent analysis are contained in NUREG/CR-5048, " Review of the Natural Circulation Effect in the Vermont Yankee Spent-Fuel Pool," by C. L. Wheeler of Pacific Northwest Laboratory. 3.10 Spent Fuel Pool Temperature Limit I Even though this amendment does not modify the current SFP temperature limit, and does not authorize an increase in storage and thus does not affect heat load, the staff addressed the spent fuel temperature limit in 4 its review. Standard Review Plan Section 9.1.3 identifies en acceptable spent fuel pool temperature limit of 140 F for the normal maximum heat load case. Vermont Yankee was originally licensed with Technical

   ,E."

4 l Specification 3.12(H), which limits the' maximum pool temperature to-150*F. The licensee stated in the submittal dated April 9, 1987, that the SFPCS is qualified for a pool water temperature of 150 F. Specifically, the qualification temperatures for the major components are: 140 F for the demineralizers, 150 F for the SFPC pumps and heat exchangers, and 175 F for the'SFPCS piping. At water temperatures greater than 140 F, the demineralizers resins may start to degrace. In order to prevent degradation 1 of the demineralized resin, and to be in conformance with the guidelines'of SRP Section 9.1.3, the licensee committed in a submittal dated June 11, 1987 to isolate the demineralizers when the SFPCS' inlet temperature. is 140 F or higher. As detailed in Vermont Yankee's letter of September 1, 1987, spent fuel temperature is continuously monitored when the system is in operation. A Control Room alarm will sound when temperature exceeds an administrative limit of 125 F. Additionally, Vermont Yankee has committea to directly monitor fuel pool temperature every four hours if one or both fuel pool - cooling trains are inoperable (see Vermont Yankee letter, dated September 1, 1987, spent fuel temperature is continuously monitored when the system is in operation. A Control Room alarm will sound when temperature exceeds an administrative limit of 125 . Additionally, Vermont Yankee has committed to directly monitor fuel pool temperature every four hours if one or both fuel pool cooling trains are inoperable (see Vermont Yankee letter, dated September 1, 1987, Attachment 2) until the enhanced Fuel Pool Cooling System is operable. Further, the licensee performed a re-evaluation of the remaining SFPCS components and determined that each of the components (pump, valves, heat exchangers, etc.), piping and supports, and structures required are capable of operation at a fluid temperature of 200 F. The FSAR states that one purpose of the SFPCS is to assure the operability of the Reactor Building Ventilation (HVAC) system. The licensee has re-evaluated the performance of the reactor building HVAC with a pool water temperature of 200 F and concluded that there will be negligible degradation of the reactor building HVAC system. The licensee also evaluated the available NPSH for the SFPCS pumps with a pool water temperature of 212 F and concluded that there is a 20 foot margin above the required NPSH of 25 feet and thus adequate pump operation can be provided at elevated pool water temperature of 200 F.

I

                                                                                   -Based upon the information. reviewed as discussed above, including the 125 F-alarm,;the staff finds'the 150*F maximum pool temperature of-

[ Technical Specification 3.12H to be' acceptable. 4.0 SIGNFICANT HAZARDS CONSIDERATIONS COMMENTS The licensee's request for amendment was noticed on June 18, 1986 (51FR22226) and again on December 31, 1986 (51FR47324) with respect to no significant hazards consideration determination and opportunity.for hearing. On January 25, 1987 The New England Coalition on Nuclear Pollution (NECNP) and the State of Vermont petitioned to intervene and on January 30, 1987 the Commonwealth of-Massachusetts petitioned to intervene. Following ruling en contentions by an Atomic Safety and Licensing Board and a subsequent ruling.by an Atomic Safety and Licensing Appeal Board, only one contention remains. That contention concerns the single failure proof characteristics of the spent fuel pool cooling system and the residual heat removal system, and thus is unrelated to this licensing action, because this licensing action does not change the heat i load on the spent fuel or residual heat removal systems. This amendment approves the placement of new racks in the spent fuel pool and storage of fuel in the racks without exceeding the presently authorized 2000 assemblies in the pool. The New England Coalition en Nuclear Pollution (NECNP) provided the only publ{c comments taking issue with the technical basis of the Commission's proposed finding of no significant hazards consideration. In its filing dated July 21, 1986, NECNP expressed the belief that the expansion could significantly increase the risk and consequences of an accident due to the vulnerability of the pools (sic) to failure in the event of a containment failure. This action authorizes only the usage of fuel storage racks of new design and not the storage of additional fuel. The structural capability of the pool has been considered and found to be completely adequate for the additional racks, and the use of racks of new design, rather than old design, has been considered and found not to have a significant impact for reasonably foreseeable design basis events. Therefore, there will not be a significant increase in the risk and conse-quences of an accident due to vulnerability of the pool in the event of con-tainment failure.

NECNP in its filing dated September 19, 1986 expressed concern that the expansion of the fuel pool storage could increase the probability of a zircaloy cladding fire because of the denser packing of the fuel and the suppression of heat transfer by neutron absorbing material, and increase the consequences of a zircaloy cladding fire by the presence of an increased inventory of radioactivity. NECNP attributed the cause of these accidents to either (1) a reactor accident which by some means, such as a hydrogen explosion, caused a loss of pool water, or (2) an accident which, by some means not involving the reactor, caused a loss of pool water. In a filing dated November 19, 1986 NECNP presented informati6n related to Chernobyl purporting to support the previous filings but introducing no new comments. The Staff's response to NECNP's. comment is that the action being authorized does not involve storage of additional fuel; therefore, the comment relating increased consequences to increased inventory does not apply. With respect to the remaining concern related to increased probability. of an accident  : because of the new rack design, the staff, in Section 3.8 of this evaluation, has addressed both the s,afety and environmental aspects of a fuel handling accident, an event which bounds the potential adverse consequences of accidents attributable to operation of a spent fuel pool with high density racks. A fuel handling accident may be viewed as a " reasonably foreseeable" design basis event which the pool and its associated structures, systems and components (including the racks) are designed and constructed to prevent. The environmental impacts of this accident were found not to be significant. The staff has considered events whose consequences might exceed a fuel handling accident, that is, beyond design basis events. Such occurrences include a criticality accident and a zircaloy cladding fire caused by overheating following the loss of spent fuel pool cooling caused by a pool failure. Compliance with General Design Criteria 61, " Fuel Storage and Handling and Radioactivity Control" and 62, " Prevention of Criticality in Fuel Storage and Handling" of 10 CFR Part 50, Appendix A, and adherence to approved industry codes and standards as set forth in the licensee's rerack application (which includes compliance with certain design and construction criteria contained in the Final Safety Analysis Report) I provides assurance that such events are of very low probability by ensuring l that pool and rack integrity and pool cooling capability are maintained. Acceptance criteria for the General Design Criteria consider all reasonably

( E # , i s [% i 1 foreseeable events. For example, in this case, criticality is prevented by ,! providing very strong racks, which will maintain the proper spacing between { fuel assemblies; the spent fuel pool walls are made of reinforced concrete- j four or more feet thick, rendering pool wall failure a very unlikely event. j i The environmental impacts of criticality and pool wall failure could be significant; however, neither of these events is considered to'be reasonably ] foreseeable in light of the design of the spent fuel pool and racks.  ! Therefore, further discussion of their impacts is not warranted and the staff concludes that the' reasonably foreseeable environmental impacts attributable to the proposed action are not significant.

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The licensee's request for amendment to the operating lice'nse for Vermont =

                                                                                                                              )

Yankee including a proposed determination by the staff of no significant j hazards consideration was individually noticed in the Federal Register or. June 18, 1985, followed by a notice on December 31, 1986, pertaining specifically to the hybrid hearing provisions of the Commission's regulations. At this time the staff is considering only the reracking (installation of sufficient fuel storage racks in the pool to accommodate 2870 assemblies, and storage of fuel  ! assemblies in the new racks up to the present Technical Specification limit of l 2000 assemblies in the pool). The Commission's regulations in 10 CFR 50.92 include three standards used by the NRC staff to arrive at a determination that a request for amendment involves no significant hazards considerations. These regulations state that the Commission may make such a final determination if operation of a facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident i eflously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

l' i The proposed spent fuel pool expansion amendment is similar to more than 100 earlier requests from other utilities for spent fuel pool expansions. The. majority of these requests have already been granted by the NRC; l others are under staff review. The knowledge and experience gained by the NRC staff in reviewing and evaluating these similar requests were used in this evaluation. The licensee's request does not use any new or unproven i technology in either the analytical techniques necessary_to support the expansion or in the construction process. The staff has determined that reracking the spent' fuel pool at the Vermont Yankee Nuclear Power Station does not significantly increase the probability or consequences of accidents previously evaluated; does not create new accidents not previously evaluated; and does not result in any significant reduction in the margins of safety with-respect to criticality, cooling or structural considerations.  : The following staff evaluation in relation to the three standards demonstrates that the proposed amendment for the SFP expansion does not involve a significant hazards consideration. First Standard

                                 " Involve a significant increase in the probability or consequences of an acctdent previously evaluated."

The following postulated accidents and events involving spent fuel storage have been identified and evaluated by the licensee. The staff likewise evaluated the same accidents and events.

1. A spent fuel assembly drop in the spent fuel pool.

2. A seismic event.

3. A spent fuel cask drop.

4. A construction accident.

1 The probabilit'y of occurrence of any of the first three accidents is not affected by the racks themselves; thus the modification cannot increase the probability of occurrence of these accidents. As for the construction accident, the licensee will not carry any rack directly over the stored spent fuel assemblies. All work in the spent fuel pool area will be l controlled and performed in strict accordance with specific written procedures. The crane that will be used to move the racks within the reactor building and the spent fuel pool has been evaluated and found acceptable. Section 3.6 of this safety evaluation contains the details of the staff's analysis. Thus, the probability of a construction accident is not significantly increased as a result of reracking. Accordingly, the proposed modification does not involve'a significant increase in the probability of occurrence of an accident previously evaluated. As noted in Section 3.1 of this safety evaluation, the consequences of a = spent fuel assembly drop in the spent fuel pool was evaluated and it was found that the criticality acceptance criterion, keff less than or equal to 0.95, is not violated. The staff also conducted an evaluation of the potential consequences of a fuel handling accident. The staff analysis found that the calculated doses are less than 10 CFR Part 100 guidelines. The results of the analysis show that dropping a spent fuel assembly on the racks will not distort the racks such that they will not perform their safety function. Section 3.6 contains the details of the staff's accident analysis. Thus, the consequences of this type of accident are not significantly changed from the previously evaluated spent fuel assembly drops which have l been found acceptable. The consequences of a seismic event have been evaluated and are acceptable. The new racks will be designed and fabricated to meet the requirements of applicable portions of the NRC Regulatory Guides and published standards. The new free-standing racks are designed, as are the existing free-standing racks, so that the floor loading from racks completely filled with spent fuel assemblies, partially filled, or empty at the time of the incident, does not exceed the structural capability of the spent fuel pool. The Reactor Building and spent fuel pool structure have been evaluated for the increased loading from the spent fuel racks in accordance with the

   ~

i O 27 l4 criteria previously evaluated by the staff and found acceptable. Section 3.2 contains the details of the staff's analysis. Thus, the consequences of a seismic event are not significantly increased from previously evaluated f events. The consequences of a spent fuel cask drop have been evaluated (see Section 3.7 of this safety evaluat' ). Because the Reactor Building Crane is single failure proof and the cask drop height to the refueling floor is less than 30 feet, the radiological consequences of cask drop meets the applicable criteria and are not significantly increased from previous analysis. The consequences of a construction accident are enveloped by the spent fuel cask drop analysis. No rack (old or new) weights more than a single 25 ton cask. In addition, all movements of heavy loads handled during the rerack operation will comply with the NRC guidelines presented in NUREG-0612, " Control of Heavy Loads at Nuclear =

Power Plants." The consequences of a construction accident are not

) increased from previously evaluated accident analyses. Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a significant increase in the probability or consequences of an accident previously' evaluated. Second Standard

                    " Create the possibility of a new or different kind of accident from any accident previously evaluated."

As noted in various sections of this safety evaluation, the staff evaluated the proposed modification in accordance with the guidance of appropriate NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate industry codes and standards. In addition, the staff has reviewed several previous NRC Safety Evaluations for rerack applications similar to this proposal. No unproven techniques and methodologies were utilized in the analysis and design of the proposed high density racks. No unproven technology will be utilized in the fabrication and

1 1: :. . ,, -i insta11ation' process of the new racks. The basic reracking technology in this case has been developed and demonstrated in numerous applications for a fuel pool capacity increase which have already received NRC staff approval.- Therefore it'is concluded that the proposed amendment to replace.the spent fuel 1 1 racks in the spent fuel pool will not create the possibility of a new or  ; different kind of accident from any accident previously evaluated.  ! Third Standard  ! l

          " Involve a significant reduction in a margin of safety."

The staff Safety Evaluation review process has established that the issue of margin of safety, when applied to a reracking modification, should address the following areas:

1. Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material and structural considerations.

The established acceptance criterion for criticality is that the neutron multiplication factor in spent fuel pools shall be less than or equai to 0.95, including all uncertainties, under all conditions. This margin of safety has been adhered to in the criticality analysis methods for the new rack design. The methcds used in the criticality analysis conform with the applicable portions of the appropriate staff guidance and industry codes, standards, and specifications. In meeting the acceptance crit:.ria for criticality in the spent fuel pool, such that k,ff is always less than 0.95, including uncertainties at a 95%/95% probability / confidence level, the proposed amendment to rerack the spent fuel pool does not involve a significant reduction in a margin of safety for nuclear criticality. Section 3.1 contains the details of the staff's analysis.

o Reracking the Vermont Yankee spent fuel storage pool, without approving expanded fuel storage capacity, adds nothing to the pool heat load. Therefore, all thermal-hydraulic considerations related to bulk pool temperature and the spent fuel pool cooling system remain unchanged. Local cooling effects due to removal of' the return line spargers was independently analyzed by the staff, and found to cause no significant reduction in the margin of safety. The main safety function of the spent fuel pool and the racks is to maintain the spent fuel assemblies in a safe configuration through all normal or abnormal loadings, such as an earthquake, impact due to a spent fuel cask drop, drop of a spent fuel assembly, or drop of any other heavy object. The mechanical, material, and structural design of the new spent fuel racks is in accordance with applicable portions of the "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1987, as modified January 18, 1979; Standard Review Plan 3.8.4; and other applicable NRC guidanee and industry codes. The rack materials used are compatible with the spent fuel , pool and the spent fuel assemblies (see Section 3.3 of this safety evaluation). The structural considerations of the new racks address margins of safety against l tilting and deflection movement, such that the racks are not damaged during impact (see Section 3.2 of this safety evaluation). In addition, the spent fuel assemblies remain intact and no criticality concerns exist. Thus, the margins l of safety are not significantly reduced by the proposed rerack. Therefore, it is concluded that the proposed amendment to replace the spent fuel racks in the spent fuel pool will not involve a significant reduction in a margin of safety. 6.0

SUMMARY

Based on the above-described review, the staf concludes that the reracking l of the Vermont Yankee spent fuel pool to accommodate 2870 fuel assemblies using l the new high density racks is acceptable. The present Technical Specification Section 5.5, which limits the number of spent fuel assemblies stored in the spent fuel pool to 2000 assemblies, remains unchanged.

   'o.                                                                                                                                                     ,

( . [ l l The staff's conclusbns are limited to the removal of the spargers and the' use , of the new racks. .The staff is not at this time authorizing the filling of the racks beyond the 2000 assemblies presently authorized. l l

7. 0 ENVIRONMENTAL CONSIDERATIONS )

This-amendment involves a change in the installation or use of a facility l component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involves no significant increase- ) in the amounts, and no significant change in the' types, of any effluents that f may be released offsite, and that there is no significant increase in  ! individual or cumulative occupational radiation exposure. The Com:nission has l previously published a proposed finding that the proposed amendment involves no l significant hazards consideration, and in Section 5.0 of this evaluation the Commission reaches a final conclusion that this amendment involves no significant  ; hazards consideration. Therefore, this amendment meets the eligibility l criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

8.0 CONCLUSION

S j The staff has reviewed and evalua'ted the licensee's request for reracking  ; the Vermont Yankee spent fuel o001. Based on the considerations discussed in this safety evaluation, the staff concludes that: (1) This amendment will not (a) significantly increase the probability or consequences of accidents previously evaluated, (b) create the possibility of a new or different accident from any accident previously evaluated, or (c) significantly reduce a margin of safety; and therefore, the amendment does not involve significant hazards considerations; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and

e l (3) such activities will be conduc'ted in compliance with the ! Ccmmission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Therefore reracking with the new racks is approved. Principal Contributors: S. Kim, H. Richings, A. Chu, M. Lamastra, J. Lee, F. Witt, J. Ridgely, and V. Rooney Date of Issuance; May 20, 1938

 !-   c 7590-01 l                                     '

UNIT,ED STATES NUCLEAR REGULATORY COMMISSION VERMONT YANKEE NUCLEAR POWER CORPORATION DOCKET NO. 50-271 , NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE-AND FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No.104 to Facility Operating License No. DPR-28, issued to the Vermont Yankee Nuclear Power Corporation, (the licensee), which authorizes the use of spent fuel storage racks of new design for the Vermont Yankee Nuclear Poker Station located in Windham County, Vermont. The amendment was effective as of the date of its issuance.

                          -The amendment allows the installation of racks of new design in the spent l

fuel pool sufficient to accommodate 2870 assemblies, and the storage of fue? assemblies in the new racks up to the present Technical Specification limit of 2000 assemblies in the pool. The amendment is in partial response to the licensee's proposed application for amendment dated April 25, 1986, as supplemented on August 15, September 26, October 21, and November 24, 1986, and February 25, March 19, March 31, April 9, April 13, May 22, June 11, September 1, and December 11, 1987, and Mar-S 2, 1988. Notice of Consideration of Issuance of Amendment and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing in connection with this action was published in the FEDERAL REGISTER on June 13, 1986, Requests for a hearing were filed on January 29, 1987 by the New England Coalition on Nuclear Pollution and by the State of Vermont. On January 30, 1987 a request was filed by the Commonwealth of Massachusetts.

              .o o

Under its regulations, the Commission may issue and make an amendment immediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards considerations are involved. The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards considerations. The basis for this determination is contained in the Safety Evaluation related to this action. Accordingly, as described above, the amendment has been issued and made immediately effective and any hearing will be held after issuance. The Commission has determined that this amendment satisfies the criteria for categorical exclusicn in accordance with 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared. For further details with respect to the action, see (1) the application for amendment dated April 25, 1986, as supplemented by letters dated August 15, September 26, October 21, and November 24, 1986, and February 25, March 19, March 31, April 9, April 13, May 22, June 11, September 1, and December 11, 1987 and March 2, 1988; (2) Amendment No.1ga to Facility Operating License No. DPR-28; (3) and the Conmission's related Safety Evaluation. All of these items are available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the Brooks Memorial Library, 224 Main Street, Brattleboro, Vermont 05301. A copy of items (2), I

o f and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, Attention: Director, Division of Reactor Projects I/II. C-L-I h DatedatRockville,Marylandthis,'[Cj.dayof}Jhg',1988. O FOR THE NUCLEAR REGULATORY COMMISSION 1 Vernon L.'Rooney, Project Manag r Project Directorate I-3 Division of Reactor Projects I/II

  .,g.
  .t_.-
                                                                                                                  -and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory.

Comission, Washington, D.C. 20555 Attention: Director Division of Reactor Projects I/II. - t I,-  % DatedatRockville,Marylandthis[C[dayofJh}3y',1988. { FOR THE NUCLEAR REGULATORY COMMISSION h Vernon L. Rooney, Project Manag r. Project Directorate I-3 Division of Reactor Projects I/II I I i l

RECEWED 0:718 gag l

 .e' -

n [Yu taag3o,, UNITED STATES e o NUCLEAR REGULATORY COMMISSION g E WASHINGTON, D. C. 20555

           ...../                                  OCT 141988
         ' Docket No. 50-271 Mr. R. W. Capstick Licensing Engineer Vermont Yankee Nuclear Power Corporation 1671 Worcester Road Framinoham, Massachusetts 01701

SUBJECT:

SPENT FUEL P00L EXPANSION SAFETY EVALUATION (TAC N0. 69179) The Commission has issued the enclosed Safety Evaluation for the proposed expansion of the spent fuel pool at the Vermont Yankee Nuclear Power Station. The Safety Evaluation pertains to your application for license amendment dated April , 25, 1986, as supplemented on Auoust 15, September 26, October 21, and November 24,.1986; and February 25, March 31, April 9, April 13, May 22, June 11, September 1, and December 11, 1987; and March 2 and June 7, 1988. On May 20, 1988 the staff issued License Amendment No. 104, which authorized you to place new racks in the pool to accommodate 2870 assemblies, and to store fuel in the racks, not to exceed the presently authorized 2000 assemblies. On June 7, 1988, you wrote a letter to the NRC forwarding a document relating to your commitment to provide an enhanced spent fuel pool cooling system. The staff has examined the safety considerations of the installation and operation of such a system, with storage in the spent fuel storage pool of the proposed 2870 assemblies and includes a discussion of these matters in the enclosed Safety Evaluation. The Environmental Assessment related to the proposed amendment was transmitted to you on July 25, 1988. The Notice of Environmental Assessment and Finding of No Significant Impact was published in the Federal Register on ugust 1, 1988 (53 FR 28925).

a I N^ l 2 We understand that ycur plans do not call for' installation of the Emergency Standby System (ESS) in the near future. Our Safety Evaluation assumes the installation of the ESS to handle the heat load if more than 2000 assemblies are stored. For this reason, the ESS must be installed and tested to . I demonstrate operability for the conclusions of the Safety Evaluation to be valid.. You will note that we have not issued the license amendment authorizing the storage of moreuthan 2,000 assemblies. Sincerely, dp Vernon L. Rooney, Project Manager V Project Directorate I-3 Division of Reactor Projects I/II

Enclosure:

As Stated cc: See next page i i I

e .D cr p 4 g

                    ;Mr.:R. W. Capstick
                    - Vermont Yankee Nuclear Power Corporation Vermont Yankee Nuclear Power Station
                     .cc:.

I Mr'.'J.l Gary Weigand W. P. Murphy, Vice President President &. Chief Executive Officer and Manager of Operations Vermont Yankee' Nuclear Power Corp. Vermont Yankee Nuclear Power Corp. R.D. 5, Box 169- R.D. 5,. Box 169 Ferry Road. . Ferry Road' Brattleboro, Vermont 05301_ Brattleboro, Vermont 05301

                    . Mr. John'DeVincentis, Vice President     Mr. George Sterzinger, Commissioner Yankee Atomic Electric Company           Vermont Department of Public Service
                    ' 1671' Worcester Road                     120 State Street, 3rd Floor
                    - Framingham, Massachusetts 01701          Montpelier, Vermont 05602 i

New England Coalition on Nuclear Public Service Board-Pollution State of Vermont-

                     ' Hill'and Dale Farm                      120 State Street R.D. 2, Box 223                          Montpelier, Vermont 05602 Putney, Vermont 05346 Vermont Public Interest Research Group, Inc.                           G. Dean Weyman 43 State Street                          Chairman, Board of Selectman Montpelier, Vermont 05602                Post Office Box 116 Vernon, Verment 05354 William Russell, Regional Administrator  Raymond H. McCandless Region I Office                          Vermont Division of Occupational U.S. Nuclear Regulatory Commission         and Radiological Health 475 Allendale Road                       Administration Building King of Prussia, Pennsylvania   19406    Montpelier, Vermont 05602 Mr. R. W. Capstick                       Honorable John J. Easton Vermont Yankee Nuclear                   Attorney General Power Corporation                        State of Vermont 1671 Worcester Road                      109 State Street Framingham, Massachusetts 01701          Montpelier, Vermont 05602 R. K. Gad III                            Conner & Wetterhahn, P.C.

Ropes & Gray 225 Franklin Street Suite 1050 Boston, Massachusetts 02110 1747 Pennsylvania Avenue, N.W. Washington, D.C. 20006 I f 1 I __ -__--_______ _ - _ -

30. .

l Vermont Yar;kee Nuclear Power Vermont Yankee Nuclear Power Station- [ i Corporation cc: Ellyn'R. Weiss, Esq. ' Resident Inspector Harmon & Weiss U.S. Nuclear Regulatory Commission 2001 S Street, N.W. P.O. Box-176 Washington,.D.C. 20009 Vernon, Vermont 05354 Samuel H. Press, Esq. Carol S. Sneider, Esq. c/o Vermont Dept. of Public Service Assistant Attorney General 120 State ~ Street Office of the Attorney General Montpelier, VT 05602 One Ashburton Place,19th Floor Boston, MA 02108 Jay Gutierrez . Geoffrey M. Huntington, Esquire Regional Counsel Office of the Attorney General USNRC, Region I Environmental Protection Bureau 475 Allendale Road State House Annex King of Prussia, PA 19406 25 Capitol Street Concord, NH 03301-6397 G. Dana Bisbee, Esq. Charles.Bechhoefer, Esq. Office of the Attorney General Administrative Judge Environmental Protection Bureau Atomic Safety and Licensing Board State. House Annex U.S. Nuclear Regulatory Commission

25. Capitol Street Washington, DC 20555 Concord, NH 03301-6397 Dr. James H. Carpenter Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Mr. Glenn 0. Bright Adjudicatory File (2)

Administrative Judge Atomic Safety and Licensing Board Atomic Safety and Licensing Board Panel Docket U.S. Nuclear Regulatory Connission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, D.C. 20555

Y SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING THE SPENT FUEL POOL EXPANSION VERMONT YANKEE NUCLEAR POWER CORPORATION I VERMONT YANKEE NUCLEAR POWER STATION DOCKET N0. 50-271

1.0 INTRODUCTION

By letter dated April 25, 1986, Vermont Yankee Nuclear Power Corpora-tion (VYNPC), the licensee, requested a change to Section 5.5.0 of the Technical Specifications for Vermont Yankee Nuclear Power Station (VY). l This change would increase the number of fuel assemblies that could be stored in the spent fuel pool from 2000 to 2870. Other previously approved specifications of Section 5.5 would remain unchanged. The change l is based on the installation of new fuel racks in the spent fuel pool which provide a closer packing of fuel assemblies. Required criticality  ; i margins are maintained by incorporation of boron containing material in the rack design. This is a commonly used feature for high density rack design, and a large number of similar designs have been approved by the NRC. l l On May 20, 1988, the NRC staff issued Amendment No. 104, granting the proposed amendment in part: i.e., authorizing installation of sufficient fuel storage racks of new design in the pool to accommodate 2870 l assemblies and storage of fuel assemblies in the new racks up to the present Technical Specification limit of 2000 assemblies in the pool.

          -J In connection with Amendment No. 104, the staff. issued a Safety Evaluation. The staff has_ incorporated the' discussion in that Safety-Evaluation in the Safety Evaluation being issued today, which addresses the total request, including the storage of 2870 assemblies and a new enhanced spent' fuel pool cooling system.

2.0 BACKGROUND

VY is a General Electric Company Boiling Water Reactor (BWh) wnich received an operating license on March 21, 1972. At the time of licensing, the spent fuel pool contained sufficient storage locations to accommodate 600 fuel assemblies. The spent fuel' pool cooling system consists of two redundant trains with each train consisting of one 450 gpm

                                                                                                                             ~

pump and one heat exchanger. The design capability of reach heat exchanger is 2.23 MBtu/Hr with a pool water temperature of 125 . The spent fuel pool cooling system is non-seismic Category I, non-Class IE. VYNPC received approval to replace the original spent fuel storage racks with high-density spent fuel storage racks in September 1977. These high-density racks were to be installed in phases providing a total maximum storage capacity of 2000 fuel assembifes. As of April 1986, the On licensee had installed racks sufficient to store 1690 fuel assemblies. i April 25, 1986 the 'icensee requested approval to rerack the spent fuel pool for a second time. This second rerack application is the subject of this safety evaluation report. The new high density storage racks would

7---__-_. 1

    .x

[? L \ :- - 3.-

                                                                   ~

increase the storage capacity of the spent fuel poo1 to.2870= fuel assemblies and is projected to provide storage capacity until 2001. The licensee provided add'itional information on the proposed second rerack  ; request in submittals dated August 15, September 26, October 21 and November 24, 1986; February 25, March 19, March 31, April'9,-April'13, May 22, June 11, September 1,.and December 11, 1987; and March 2 and June 7, 1988. The licensee also incorporated by reference information contained, l 1

         .in cubmittals dated September 11, 1981; November 30, 1983; and May 21, June 27, and December 18, 1984.' .Information related to'the licensee's                        ,

computer modeling of spent' fuel pool cooling was provided at a meeting on January 15, 1987, in Richland, Washington. 1 In the April 25, 1986 submittal, in addition to requesting approval to-expand' the capacity of its spent fuel pool by reracking, the licensee j identified necessary changes involving removal of the spent fuel pool cooling system return line spargers and related piping inside the spent fuel pool. In a submittal dated September 1, 1987, the licensee further l defined this request by proposing to cut off the Spent Fuel Pool Cooling System (SFPCS) return line at approximately 15 feet aoove the top of the racks (which is approximately 8 feet below the fuel pool water level). This modification would provide for the storage of an additional 100 fuel assemblies beyond the capacity available if the return line were not cut. The licensee stated that the natural circulation developed by the heat generated by the spent fuel would provide adequate cooling for the spent fuel.

1 .. t. 4

                       -The staff issued a status report dated January 21, 1988, which discussed five technical open issues related to the licensee's request to increase the storage capacity of the spent fuel pool to 2870 fuel assemblies.              Some of these open issues involved the. fuel pool cooling system and its cooling capacity.          These open issues also involved increased heat load due to an increase from the present 2000 fuel assemblies limit to the requested 2870 limit.           The staff met with the licensee on February 9,.1988 to ciscuss these issues.

During the meeting, the licensee revealed that it had reached a decision to design, build, and install an enhanced cooling system for the spent fuel pool. This modification was proposed for the purpose of expediting resolution of outstanding issues. Subsequently, the licensee in a submittal dated March 2, 1988, documented its commitment to install an enhanced cocling system. Although no details of the modified design were provided, the licensee did provide some design and performance information for the enhanced SFPCS. In order to allow reracking to commence in such a way that personnel radiation exposure would be minimized, without awaiting completion of review with respect to enhanced cooling, on May 20, 1988 the staff issued Amendment No. 104, which considered the portion of the proposed expansion involving reracking and placement of the new racks in the pool. The staff stated that consideration of storage of more than 2000 assemblies would await a determination of the adequacy of spent fuel pool cooling for more than

S-2000 assemblies, including the yet-to-be-designed enhanced spent fuel pool cooling system, for which more information was required than was available. at the time. Subsequently, on June 7, 1988, the licensee supplemented its application wich a document describing in more detail its plans .for the new enhanced spent fuel cooling system. The NRC staff's evaluation includes a review of this document. 3.0 EVALUATION 3.1 Criticality Consideration. Required criticality margins are maintained by incorporation of boron containing material in the rack design. The rack design (described in detail in VYNPC's letter of September 25, 1986) is configured so that the baron associated with the cells, in the form of Boral, is arranged such that there is boron between each pair of fuel assemblies. This includes the Boral on the outer edge of racks, which is arranged so that there is boron between assemblies facing each other across rack gaps. The B-10 loading of the Boral is 0.027 gm/cm 2minimum. The cell pitch is 6.218 inches and the cell inside width is 5.922 or 6.092 inches (fuel assembly with channel is 5.438 inches). The criticality calculations for the new racks were performed by Yankee Atomic Electric Company (YAEC). The calculations were

{ -

 ~

performed with two methodologies. The reference criticality analyses , were performed with the Monte Carlo code KENO-IV using the NITAWL code to provide cross sections based on the XSDRN code cross section library. For sensitivity calculations and trend analyses the diffusion code PDQ-7 was used with cross sections from the CASMO code. All of these codes and cross sections are well known industry standards, frequently used for analyses of fuel pools and other complex criticality problems, and have been approved by the NRC. YAEC has benchmarked its KENO methodology against a number of relevant critical experiment results from Babcock and Wilcox and Battelle Northwest Laboratories. These experiments present geometrically representative configurations for fuel racks. YAEC has I used these benchmark calculations to develop an analysis methodology uncertainty factor to be added to rack keff calculations. YAEC has also determined the potential variation of the rack and fuel parameters that are used in determining the keff of the rack-fuel system. These parameters include poison thickness, boron concentration, cell pitch, stainicss steel thickness and eccentric fuel position. The variation of k eff with these parameters (taken at a 95/95 probability / confidence level) was determined. These (independent parameters) were statistically combined to provide a AK uncertainty which, along with the Monte Carlo statistical uncertainty, is added to the base keff calculation.

     =

YAEC has investigated abnormal conditions that might be associated with the spent fuel pool and has determined that potential reactivity variations caused by abnormal pool conditions and accidents have either negligible or negative effects of keff. These include char.ges in pool temperature from the base conditions, cell or rack displacement from seismic incidents, fuel or heavy object drop events, and fuel assembly placement outside of the racks. Thus k eff r for the fuel pool is determined, both for normal and abnormal conditions, by adding the previously discussed method and mechanical uncertainties to the base calculation, without the need for additional factors to account for abnormal conditions. For the base case, the YAEC Monte Carlo calculations assume (1) an infinite square array of cells (2) with a pitch of 6.218 inches, each containing (3) an unirradiated fuel assembly of 64 fuel rods (no water rods) with (4) a uniform enrichment of 3.25 weight percent  ; U-235, (5) no burnable poison and (6) infinite length. The water temperature is 68 F. This fuel assembly enrichment bounds present fuel enrichments and the use of no burnable poison provides conservatism for reactivity calculations. For the base configuration, the keff was calculated to be 0.9046. The total uncertainty at a 95/95 level was 0.0221 ax, giving a total k of 0.9267. This is to be compared to a required upper limit of eff 0.95.

L l- . ..

     .                                                                       .g.

The fuel assembly lattice used for the base ' rack calculations was calculated to have a standard reactor core geometry uncontrolled km value of 1.35. YAEC proposed, in the initial submittal, to use 'a fuel assembly km of 1.35 as the design bases for fuel' acceptable for storage in the racks (rather than fuel enrichment limit). This.is common practice for BWR fuel storage (see, for example, NEDE-24011-P-A-8, May 1986) and allows credit for the burnable poison in the fuel assembly in the analyses to meet the Technical Specification requirement of 0.95. As a result of discussions with the staff concerning the nature of additional uncertainties involved when using AK design criteria, this proposed ifmit was reduced to 1.31 by VYNPC by letter dated October 21, 1986. The possible reactivity effects of (1) nonuniform enrichment variation in the assembly, (2) uncertainty in the calculation of km and 3) uncertainty in average assembly enrichment were examined and quantified by YAEC, providing the additional correction factor of 0.04 ax. The basic criticality design of the new racks, using boron lined cells to provide the appropriate neutron multiplication level for the closer packed array of high density racks, is a commonly used concept and has been accepted for many spent fuel storage pools. It is an ) acceptable design concept for maintaining criticality levels for the VY pool. The methodology used by YAEC to analyze the criticality and  : reactivity change characteristics of the racks is a state of the art

i

     -c                                                  i l

methodology, commonly used and approved for other utilities for such analyses. The Monte Carlo method using the KEN 0/NITAWL/XSDRN package provides an acceptable methodology for the base calculations and the PDQ/CASMO is acceptable for sensitivity calculations. These methods have been benchmarked against an appropriate selection of critical-experiments, with results falling within expected ranges of deviations from the experiments. The derivation of the uncertainty of the methodology from this benchmarking follows normal procedures and also falls within an expected range. It is acceptable. The examination of uncertainties to be attributed to variances in dimensions and materials in the fuel and racks has covered an acceptable range of parameters and has used a suitable, standard methodology for determining the reactivity effects and their statistical combination. The examination of tho effects of abnormal conditions has covered the standard events relating to changes in temperature, movements, misloadings and dropping of assemblies and other equipment, and the results, giving nonpositive reactivity additions, are reasonable and acceptable. The base calculations and added factors for uncertainties, giving a total k,7f of 0.9267, are thus acceptable for an average enrichment of 3.25 percent. There is a margin of 2.3 percent AK to the staff required Technical Specification limit of 0.95. The transfer to a fuel assembly km design basis criterien has conservatively considered relevant additional uncertainty factors, and the resulting design basis km value of 1.31 is acceptable. The approach of using a km

design basis has been approved in other applications, and is used in the staff approved General Electric reload analysis approach (as given in GESTAR II, NEDE-24011-P-A-8, May 1986). The base k eff criterion of 0.95 given in Technical Specification 5.5.B remains the same. Also unchanged by this request is the average enrichment limit of 16 grams of U-325 per longitudinal centimeter of assembly. This specification is compatible with the 3.25 percent U-235 enrichment used in the base calculations. Therefore, it is concluded that the required criticality margins are maintained by the new racks. 3.2 Structural Engineering The new high density racks are stainless steel " egg-crate" cellular structures of approximately 6 inches square. Each cell is designed to contain a spent fuel assembly and a typical rack consists of approximately 300 cells whose dimensions are approximately 10 feet long by 8 feet wide and 15 feet high. Weight of the rack and fuel is transmitted to the floor of the pool through supporting legs. The racks are each free-standing on the pool floor and a gap is provided between the racks and the pool wall so as to preclude impact during earthquake. Such design provides a margin of safety against tilting and deflection movement.

o

   ...                                  11 _

The spent fuel pool is a. reinforced concrete structure supported by the Reactor Building walls. _ The poo? is approximately 26 feet wide by 40 feet long by 39 feet deep and is completely. lined with seam welded ASTM A240 Type 305 stainless steel. The licensee's load combinations and acceptance criteria were'found to be consistent with those in the " Staff Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14,1978 and amended January 18,.1979. The existing concrete pool structure was evaluated for the new. loads in accordance with the requirements of the appropriate industry codes such as ASME Section III and ACI 349-80 and the NRC staff guidelines and documents such as Standard Review Plan (NUREG-0800) and Regulatory Guide 1.92

       " Combining Model Responses and Spatial Components in Seismic Response Analysis."

[. 1 Seismic load.s for the rack design are based on the original design I ' floor acceleration response spectra calculated for the plant at the licensing stage. The seismic loads were applied to the model in three orthogonal directions. The hydrodynamic loads of pool water acting on pool walls are considered. Loads due to a fuel bundle drop accident were considered in a separate analysis. The postulated loads from these events were found to be acceptable. The dynamic response and internal stresses and loads of the racks and pool structure are obtained from a time history seismic analysis. Nonlinear time history analyses are performed utilizing the

c..

l widely-used industry ANSYS code. Friction between rack support pads I and pool floor and hydro-dynamic coupling are considered in the analysis. Calculated stresses for the rack components were found to j be within allowable limits. The racks were found to.have adequate margins against sliding and tipping. An ' analysis was conducted by the licensee to assess the potential effects of a dropped fuel assembly on the racks. The external kinetic energy will be absorbed by rack strain energy through deformation of the rack cells. The overall integrity of the rack

          .will not be adversely affected. The existing structures were analyzed by the licensee for the modified fuel rack loads. The strength design method for reinforced concrete was used in conjunction with conventional structural analysis procedures to determine cepacities.

The existing spent fuel pool is determined to safely support the loads generated by the new fuel racks. It is concluded that the' proposed rack installation will satisfy the requirements for 10 CFR 50, Appendix A, GOC 2, 4, 61 and 62, as applicable to structures and is, therefore, acceptable. 3.3 Compatibility and Chemical Stability of Rack Materials The staff reviewed the compatibility and chemical stability of the new rack materials wetted by the pool water. The licensee supplemented the original-submittal dated April 25, 1986, with

4 9; , L i

       ~~

l additional information regarding rack materials by letter dated March 31, 1987. The proposed spent fuel racks are to be constructed .l entirely of. Type 304L stainless steel, except for threaded rods attached to leveling pads, which are 17-4 PH-hardened stainless steel, 4 and the neutron absorber material. The 17-4 PH threaded rods are ) i

!             heat treated, chemically cleaned and chrome plated.                     The neutron               !

absorber material is Boral with a minimum B10 loading of 0.027 gms/cm2 Boral'is a dispersion of boron carbide in an aluminum inatrix with an aluminum clad.  !

                                                                                                                ?

The spent fuel rack compartments containing the Boral are not j i watertight. This will allow venting of gas generated by radiolysis of contained water and by Boral off gassing, preventing pressure

                                                                                                             'I buildup and possible swelling. The austenitic stainless steel (304L) used in the rack fabrication has a maximum carbon content of 0.03% by weight which minimizes the sensitization in weld heat-affected zones.

The stainless steel racks are compatible with the spent fuel pool water that is processed by filtration and demineralization to j maintain water purity and clarity. The spent fuel pool purity is I maintained at < 1p S/cm conductivity at 25 C, < 500 ppb chloride, ! < 100 ppb total heavy elements, and a pH range of 5.8 to 8.0. Intergranular corrosion tests performed in accordance with ASTM A262, Practice E are required for the sustenitic stainless steel. Dissimilar metal contact corrosion (galvanic attack) between the f stainless steel rack assemblies, aluminum in Boral neutron absorption plates and zircaloy in the fuel assemblies will not be significant _ - - _ - - _ _ _ _ - _ - _ _________ _ _ a

 .[+       ,

l

                                                           - 14 i

because the materials are protected by highly passiviting oxide films and are, therefore, at similar galvanic potentials. i Boral has undergone extensive testing to study the effects of gamma irradiation in various environments, and'to verify its structural integrity and suitability as a neutron absorbing material. Boral has been qualified for 10 11 rads of gamma radiation while maintaining its neutron attenuation capability. 1

                    - The annulus space in each cell' assembly which contains the Boral.is                                                                                         .

vented to the pool to allow venting of radiolitic gases and Boral outgassing. This will prevent swelling and bulging of the stainless y steel plates. I Tests have shown that Boral does not possess leachable halogens that could be released into the pool environment in the presence of radiation. Similar conclusions have been made regarding the leaching of elemental boron from the Boral. To provide added assurance that no unexpected corrosion or , 1 degradation of the materials will compromise the integrity of the l racks, the licensee has committed to conduct.a long-term poison coupon surveillance program. Surveillance samples in the form of  : stainless steel retained sheets of Boral (prototypical of the fuel storage cell walls) will be exposed to the spent fuel pool water. l 1 E-____________________. _ _ _ _ _ _ ___ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

   .                                                             There coupons will be removed and examined periodically over the expected service life.

The staff has reviewed the description of the proposed surveillance l program for monitoring the Boral in the spent fuel storage pools and concludes that the program can reveal deterioration that might lead to loss of neutron absorbing capability during the life of the spent fuel racks. The staff does not anticipate that such deterioration will occur, but if it should occur, it would be gradual. In the unlikely event of Boral deterioration in the pool environment, the monitoring program will detect such deterioration and the licensee will have sufficient time to take corrective action such as, for example, replacement of the Boral sheets. Based on the above discussion, the staff concludes that the corrosion of the spent fuel pool components due to the spent fuel storage pool environment should be of little significance during the life of the facility. Components in the spent fuel storage pool are constructed of alloys that have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion. Tests under irradiation and at elevated temperatures in water indicate that the Boral material will not undergo significant degradation during the projected service life of approximately 40 years for the racks.

r r  :

                                                  - 16  -

The staff further concludes that the environmental compat'ibility and stability of the materials used in the spent fuel storage pool is adequate based on the test data cited above and. actual service-experience at operating reactor facilities. Finally, the staff finds'that implementation of the proposed. , monitoring program and the selection of appropriate materials of construction by the licensee meet the requirements of 10 CFR 50 Appendix A, General Design Criterion 61, regarding the capability to permit appropriate periodic inspection and testing of components, and General Design Criterion 62, regarding preventing criticality by maintaining structural integrity of components and of boron poison, and is, therefore, acceptable. 3.4 Occupational Radiation Exposure The staff has reviewed the licensee's plan for expansion of the spent fuel pool storage capacity with respect to occupational radiation exposure and finds that the ALARA policy, design, and operational considerations are acceptable. This finding is based on the licensee having considered the provisions of 10 CFR Part 20.101, 20.1(c) and 20.103, and the guidelines of Regulatory Guides 8.8 and 8.10 with respect to the planned expansion. The licensee set a dose goal of 23 person-rem for the SFP modification project before committing to add an enhanced fuel pool cooling system. The goal is based on j information gained by reviews of the experience gained with similar l l

l y

• t-17 -

                                      ~

projects at other plants. The redundant, seismically designed spent fuel pool cooling system, which would be. operational prior to the l time Vermont Yankee exceeds the existing 2000 spent fuel assembly-storage limits, was proposed by the licensee to resolve all remaining staff concerns related to increasing the storage limit. By telephone conversations on July 7, 1988 the licensee informed the' staff that i the dose for installation of the enhanced spent fuel pool cooling ] system has been estimated very conservatively to add less than 10 l person-rem to the original dose goal. This results in a dose goal , for the entire SFP modification, including the enhanced SFP cooling system, of 33 person-rem. The staff finds this dose goal will not i affect the licensee's ability to maintain individual occupational doses within the limits of 10 CFR 20, and as low as is reasonably achievable (ALARA). Normal radiation control procedures, in accordance with the guidelines of Regulatory Guide 8.18, should preclude any significant occupational radiation exposures. The 33 person rem dose goal includes all activities necessary for the reracking operation including vacuum cleaning of the SFP walls and floor; shuffling fuel, installation of the new racks; removal of the l old racks; cleaning, decontamination, and any necessary cutting of l old racks; and disposal of waste resulting from the reracking operation, including the old racks. In terms of radiation dose to workers, the spent fuel assemblies J themselves contribute a negligible amount to dose rates in the pool

area because of the depth of water shielding the fuel. However, one potential source of radiation to workers during the~rerack operation is radioactive activation of corrosion products, whichLis referred to as crud. Crud may be released to the pool water because of fuel movement during the proposed SFP modification. This could increase radiation levels in the vicinity of the pool. The addition of crud to the pool water is greater during refuelings, when the spent fuel is first moved into the fuel pool. It is at this time that most of the additional crud is introduced into the pool water from the fuel assembly and from the introduction of primary coolant. However, significant releases of foreign material that might become activat;d is not expected,'since the new racks are cleaned prior to installation. In addition, the purification system for the pool, which keeps radiation levels in the vicinity of the pool at low levels, includes a filter to remove crud. This filter will be operating during the modification of the pool. Thus, we find that the proposed storage of spent fuel in the modified SFP will not result in any significant dose to workers. Recently, there has been a concern exp- ,ed that a severe reactor accident could lead to loss of water from the spent fuel pool. Specifically, if the pool cooling system were disabled as part of a reactor accident sequence and repairs of this system were precluded for several weeks due to high radiation fields around the plant, then it is possible to postulate a reduction in SFP water inventory. Vermont Yankee, like other nuclear plants, employs a defense in depth concept for early warning of, and subsequent protective actions in

          ~

response to, any accident-or abnormal occurrence, including a loss of-cooling to the spent fuel pool. Early warning via monitoring systems and precautions called for by the plant's' health physics program assure minimum radiation dose to. workers during both normal and abnormal conditions. The spent fuel pool has temperature indicators, water level indicators, vent radiation' monitors, an airborne radioactivity monitoring system and an area radiation monitoring system. The water temperature and level-indicator provide redundant and diverse means of detecting loss of cooling to the spent fuel pool even during an accident. They provide an early warning,.so that corrective actions can be taken to restore cooling or to add water before the water level in the spent fuel pool decreases due to boiling. In addition to the monitoring system and the plant's overall health physics program, the effects on personnel of any accident or abnormal condition including spent fuel pool boiling can be mitigated by implementation of the licensee's emergency plan, which contains re-entry criteria for entering potentially high radiation areas. On the basis of the above, the staff finds that the projected activities and the dose goal of 33 person-rem for the proposed spent fuel pool expansion are reasonable. Further, we find that the licensee intends to take ALARA considerations into account to implement reasonable dose reducing activities. Hence, the licensee

             ->                                                20 will be able to maintain individual occupational radiation exposures within the applicable limits of 10 C.F.R.~ Part 20, and meet the guidelines of Regulatory Guide 8.8. The staff, therefore, finds that the occupational radiation protection' aspect of the spent fuel pool modification program is acceptable.

3.5 Radioactive Wastes The plant contains radioactive waste treatment systems designed to collect and process the gaseous, ifquid and solid wastes that might contain radioactive material. The radioactive waste treatment systems have been previously evaluated by the staff and found acceptable. There will be no change in the radioactive waste treatment systems or in the conclusions given regarding the evaluation of these systems as a result of the proposed installation of the new racks. Our evaluation of the radiological considerations supports the conclusions that the proposed installation of new spent fuel storage racks at Vermont Yankee is acceptable. The basis for our conclusions is that the previous evaluation of the radioactive waste treatment systems is unchanged by the installation of new spent fLe1 storage racks. The present spent fuel racks will be removed from the SFP and will probably be disposed of as low level waste. If the existing racks are disposed of as solid waste, the vo1Lme will be approximately 2000 cubic feet. The annual average volume of solid wastes shipped

G

                                                     - 21:-

j . ', E offsite.for burial from Vermont Yankee hasLbeen-approximately'400 cubic meters; Avaraged over the lifetime'of the plant the addition s of the existing spent. fuel racks will increase the' total waste volume shipped from the facility by less than 0.4%. This would not have any significant environmental impact beyond that' contemplated and discussed in the FES for the operating license application.- (U.S. Atomic Energy Commission, Environmental Statement Related to the Operation of Vermont Yankee Nuclear Power Station, July 1972). 3.6 Load Handling 3.6.1' Light Loads A light load is a load that weighs less than the combined weight of a fuel bundle, channel and its handling tool. Since

                                                                                                      ~

there are no restrictions on the handling of light loads over the spent fuel, a light load could be carried which, if dropped, could have sufficient kinetic energy and impact force on the fuel or rack to potentially result in greater damage than assumed in a fuel handling : cident. In the licensee's September 1, 1987 submittal, the licensee stated that an analysis of light loads normally carried over the spent fuel was performed. The licensee identified a light load to be any load which weighs 700 lbs or less. The results of this analysis indicate that the kinetic energy of these loads is less than that of the design basis fuel handling accident and l l l l

j thus the radiological consequences of a light load drop are bounded by the fuel ~ handling accident. The staff, therefore, i considers handling of light loads to be acceptable. 3.6.2 Heavy Loads Spent fuel storage racks weigh more than a fuel assembly, channel and its handling tool. Thus, spent fuel storage racks are considered to be heavy loads. The reactor building crane will be used to move the storage racks within the reactor building and the spent fuel pool. As part of the review of the Vermont Yankee facility for compliance with guidelines of NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," the staff concluded in the Safety Evaluation Report dated June 27, 1984, that the reactor building crane was single failure 1 proof by meeting the guidelines of NUREG-0554, " Single Failure Proof Cranes for Nuclear Power Plants". l In the November 24, 1986 submittal, the licensee provided information that showed the movement of spent fuel within the spent fuel pool, the order of rack replacement, and the path of travel for each of the fuel storage racks. The licensee demonstrated that the storage racks will not be carried over spent fuel or over other racks containing spent fuel. In a subsequent submittal dated February 25, 1987, the licensee provided a drawing that showed the heavy load handling  ! l

O The boundaries and laydown areas for the storage racks. f licer,see demonstrated that to the extent practical, the paths of travel follow the fuel building structural floor members and beams. The licensee also stated that the load paths and laydown areas will be marked with stanchions and ropes prior to performing heavy load lift. Drawings will be provided to the crane operator in the cab and to the tag man directing the lift to assure adherence to the load paths. The licensee committed to have all deviations from the established load paths approved The licensee also by management personnel prior to being used. committed to prepare installation and removal procedures specifically for the reracking of the spent feel pool, and to provide qualification, training, and testing of crane operators, as described in D.P. 2201, " Reactor Building and Turbine Building Crane Operator Qualifications." This information has been reviewed by the staff and found to be acceptable. Two special lifting devices will be used in the reracking, one By for the existing par racks and one for the new NES racks. submittal dated May 22, 1987, the licensee provided drawings of the par spent fuel rack lif ting rig which show redundancy in the lifting rig. The licensee committed to pre-operationally load test the par lifting rig to 150% of the empty spent fuel rack weight. By submittal dated April 13, 1987, the licensee provided drawings of the NES spent fuel rack lifting rig, which also show redundancy in the lifting rig. The licensee committed

y.;

                                                                   - 24 g ?:

to pre-operationally load test the NES lifting rig'to 150% of

                                       'the' empty spent fuel rack weight-(equivalent to 273s tons) or a' total load test equal to 30 tons. In the-February 25, 1987 submittal, the licensee committed to ensure that the special t

lifting devices meet the guidelines of ANSI N14.6-1978,_and to perform the load tests and subsequent inspections 'in accordance [ a with ANSI N14.6-1978. s Based on the above review, the staff concludes that heavy loads , handling will be performed in accordance with the guidelines of NUREG-0612, and thus the requirements of General Design Criterion 61, " Fuel Storage and Handling and Radioactivity Control", are met as they relate to proper load handling to ensure against an unacceptable release of radioactivity, a criticality, accident, or the inability to cool the spent fuel 'in The staff has the spent fuel pool due to postulated load drops. determined that. installation of the new high density racks to provide 2870 storage locations in the VY SFP is acceptable. 3.7 Soent Fuel Shioping Cask Drop Accident L In the licensee's response to the staff's request for additional information dated November 24, 1986, it was stated that the Reactor Also, the Building Crane is considered to be single failure proof. cask drop height to the refueling building floor is less than 30 feet (Ref. FSAR Section 12.2'). Therefore, in accordance with Standard

(

 .                                     - 25 I

Review Plan Section 15.7.5, evaluation findings with respect to radiological consequences for a cask drop accident are not needed. The staff concludes that the proposed expansion -eets the applicable criteria with respect to the spent fuel cask drop accident analysis. 3.8 Fuel Handling Accident The staff independently evaluated a postulated fuel handling accident following the guidance of Standard Review Plan Section 15.7.4,

        " Radiological Consequences of Fuel Handling Accident", and using the assumptions set forth in Regulatory Guide 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors,"

The calculation was performed by using the staff computer code ACTCODE. The staff conservatively assumed a 24 hour shutdown time for the two damaged fuel assemblies. Credit is given for the Standby Gas Treatment System (SGTS) in the reactor building because the system provides safety grade HEPA filters and charcoal absorbers. Credit is also given for the reactor building, since it maintains a The radioactivity slightly negative pressure during the accident. produced by this accident is processed by the SGTS, which has a 9f% removal efficiency for radioactive iodines. The resulting radiological doses at the EAB are 2.58-rem for the thyroid, and 0.337 rem for the whole body. Similarly, at the LPZ, the doses are 0.361 1 l

rem for the thyroid and 0.047 rem for the whole body. These doses are far below the criteria of 75 rem for the thyroid and 6 rem for the whole body (SRP 15.7.4). Because VY's control room does not have charcoal and absorber filters, the staff also considered control room doses due to a-fuel handling accident involving a radioactivity release. However, since this release is from a 300 foot high stack, and the atmospheric

                                                  -6 dispersion factors are in the order of 10 and 10-7 sj,3 , the ef fective radiological doses to the control room are estimated to be negligible.

The assumptions used for this analysis are listed in Table 1. i l l l l t - - - _ . _ _ - . _ - _ _ _ _ _ _ _ _ _ _ _ -

g; .

      .i LTABLE 1. AS'SUMPTIONS USED IN FUEL HANDLING ACCIDENT 1665'MWth Reactor' Power Level        .

368 1 Number'of fuel assemblies in core 126 Number'of fuel rods' damaged L Standby Gas Treatment System filter efficiency for elemental and-organic. 95% iodines .24 hrs

  '                   'Cooldown time for. impacted spent fuel Effective pool decontamination factor.

100 for iodine ~ GAP ACTIVITY:

                               ' Iodine                         10%
                              . Krypton _                       30%
                              -Total noble gas                  10%

other than Krypton

                                               -Time Period                                         X/Q Location
                                                                                                        -3 s/m3 -

EAB '0-2 hrs 0.25 x 10

                                                                                                        ~4 LPZ                     0-8. 'brs                                 0.35 x 10
                                                                                                        ~4 8-24 hrs                                  0.21 x110
                                                                                                        ~0 24-96 hrs                                   0.70 x 10
                                                                                                        -5 96-720 hrs       .                           0.15 x 10 h

1

           'The staff concludes'that the proposed spent fuel pool expansion meets the I

applicable criteria with respect to the fuel handling accident analysis and is acceptable.

l l a 3.9 Sparcer Removal The licensee proposed to remove the return line spargers and to terminate the return line in a downward pointing direction at approximately 15 feet above the top of the spent foal pool storage With the spargers as racks (8 feet below the surface of the water). originally installed, the water from the SFPCS was returned at the bottom of the spent fuel pool below the spent fuel storage racks. The water generally traveled up through the racks as it passed to the far side of the spent fuel pool, thus providing " forced" cooling of the spent fuel. With the spargers removed, the water enters and exits the pool at approximately the same elevation atsove the spent fuel storage racks. The mechanism for cooling the spent fuel in this configuration relies on natural circulation. The staff performed an independent spent fuel cooling analysis to verify the licensee's claim that removal of the spargers will not affect spent fuel cooling capability. The results demonstrate that because of adequate mixing in the upper plenum, the relatively open flow area below the fuel, and the 2-inch gaps around the periphery of the racks, adequate spent fuel cooling - is provided regardless of the inlet flow orientation, or " loading The primary factor patterns" of the hot assemblies within the pool. I controlling pool performance is the total pool heating rate to total pool recirculation flow rate. Additional details of the staff's , independent analysis are containea in NUREG/CR-5048, " Review of the

                                                         - 29           -

Natural Circulation Effect in the Vermont Yankee Spent-Fuel Pool," by C. L'. Wheeler of Pacific Northwest Laboratory. I 3.10 Spent Fuel Pool Temperature Limit Even though this amendment does not modify the current SFP

                                             ~

temperature limit, the staff _ addressed the spent fuel _ temperature limit in its review. Standard Review Plan Section 9.1.3 identifies  ; an acceptable spent fuel pool temperature limit of 140 F for-the-Vermont Yankee was originally normal maximum heat load case.  : licensed with Technical Specificatica 3.12(H), which limits the maximum pool temperature to 150*F. The licensee stated in the , 1 submittal dated April 9,1987, that the SFPCS is qualified for a pool Specifically, the qualification ) water temperature of 150 F. i I temperatures for the major components are: 140 F for the

                    ' demineralized, 150 F for the SFPC pumps and heat exchangers, and 175 F for the SFPCS piping.           At water temperatures greater than                                                                                                              !

140 F, the demineralizers resins may start to_ degrade. In order to prevent degradation of the demineralized resin and to be in I conformance with the guidelines of SRP Section 9.1.3, the licensee committed in a submittal dated June 11, 1987 to isolate the demineralizers when the SFPCS inlet temperature is 140 F or higher. As detailed in Vermont Yankee's letter of September 1,1987, spent l' fuel temperature is continuously monitored when the system is in operation. A Control Room alarm will sound when temperature exceeds Additionally, Vermont Yankee has an administrative limit of 125 F. committed to directly monitor fuel pool temperature every four hours l

L; 4

   'X                                                                                                        ifoneorbothfuel:poolcoolingtrains'areinoperable(seeVermond Yaiikee letter, dated September 1, 1987). Further, the licensee performed'a re-evaluation of the remaining SFPCS components.and-                                    1 determined that each of the components (pump, valves, heat                                          ,

exchangers, etc.), piping and supports', and structures required are-capable of operation at a fluid temperature of 200'F. The'FSAR states that one purpose of the SFPCS is to assure the operability of the Reactor Building Ventilation (HVAC) system. The licensee has re-evaluated the performance of the reactor building HVAC with a pool. water temperature of 200 F and concluded that there will be negligible degradation of the reactor building HVAC system. The-licensee also evaluated the available NPSH for the SFPCS pumps with a I pool water temperature of 212 F and concluded that there is a 20 foot

                                                              '                                                                     .l I
                                 . margin above the required NPSH of 25 feet and thus adequate pump j

operation can be provided at an elevated pool water temperature of 200*F. Based upon the information reviewed as discussed above, including the 125 F alarm, the staff finds that the 150 F maximum pool temperature of Technical Specification 3.12H continues to be acceptable. 3.11 ENHANCED FUEL POOL COOLING SYSTEM By letter dated June 7, 1988, VYNPC described a redundant seismically qualified SFPCS, which they will install, test, and make operational prior to the time Vermont Yankee exceeds the current Technical

o ge --

       .m 6                                                 fi p

below 150 F with. assumed normal. refueling heat loads. It cools.the L4

 >n 4            pool by transferring the spent fuel decay heat through the heat b'           exchangers to'the Service Water. System maintaining pool 1 temperature h;l-               below 150*F. The ESS is normally not .in' use.

3.11.1 Heat Removal Caoability l Based on the guidelines of SRP Section 9.1.3, " Spent Fuel Pool. Cooling and Cleanup Systems" two spent fuF Mol heat generation conditiens were evaluated. The first candition was the normal maximum heat load, which-is the heat generated by spent fuel -l when the pool is filled (2,870 assemblies) with successive  ! regulac refuelings (1/3 of the core for each refueling interval)- assuming a failure of one train of pool cooling. The second condition was the abnormal maxi.um T heat load case, which.is l similar to the normal maximum heat load case except that the

                       'last 368 spaces are filled by t full core' off-load and no failures are assumed.                                                         I I

The decay heat loads from normal refuelings were calculated by the licensee with the assumption of discharging the spent fuel f

                                                                                                     ]

1 I to the storage pool at six cays and ten days following shutdown ) from normal operation. The abnormal decay heat load was calculated assuming that the full core discharge to the storage  ; pool occurs ten days following shutdown from normal operation l l for refueling. l l l^

I i

  'o.

s I' Data in Table A.2 from the licensee's June 7, 1988 submittal indicate that the decay heat generation rate for normal refueling discharge varies from 7.59 MBtu/hr to 10.33 MBtu/hr, and that the rate for the abnormal refueling discharge varies from 16.84 MBtu/hr to 18.26 MBtu/hr (six and ten days respectively). These heat generation rates are comparable to the staff's independent calculated values. Based upon a comparisor) of these calculated. heat loads and the heat removal capability of the cooling systems, the Fuel Pool Cooling and Demineralized System has sufficient capacity to remove the calculated maximum normal and abnormal heat genersted by the spent fuel. l The Normal Fuel Pool' Cooling System (NFPCS) has a heat removal capacity of 2.23 MBtu/hr per train with a pool water temperature of 125 F. Should the NFPCS be unable to maintain fuel pool temperature or if it should lose flow, the Emergency Standby System (ESS) will be used. The ESS has a heat removal capacity l of 11.0 MBtu/hr per train. In the normal maximum heat load 1 condition the maximum calculated heat load - 20.33 MBtu/hr l l as compared with an EES heat removal capability of 11 MBtu/hr per train. In the abnormal maximum heat load condition, the  ; total heat removal capacity is 22.0 MBtu/hr as compared with a l Therefore,  ! maximum calculated heat load of 18.26 MBtu/hr.  ; the ESS has sufficient heat removal capability following the methodology as discussed in SRP 9.1.3. l 1 l

                                                                                -    _ - - - _ _ _ __ a

j'b ,

                                                                                      - 34.-

3.11.2 Water Level,~' Makeup Water and Corrosione  ; p . Leakage of potentially radioactive water Lto the environment from , the.ESS is prevented by providing a. higher pressure in'the )ll 1

                                                                              ~

service waterisystem than the'ESS pressure. Indication of this- l differential pressure is'provided in the Control Room.' Leakage

                                                                                                                                                                                           ]

I from the NFPCS to the service water system is prevented by using I an intermediate closed loop cooling system, Reactor Building j Closed Cooling *0 er (RBCCW), which transfers the decay heat to j i the Service Water system. This closed loop system ensures that. 'l l j fuel pool. water leakage, if any, will be contained within the 1 E RBCCW System and will not be released into the Service Water 'l

                                                                                                                                                                                         -j System.

1 The Fuel Pool Cooling'and Demineralized System has instruments

                                                                                   ~

to monitor the water level and maintain a water level above the . fuel sufficient to provide shielding for normal building

                                                                                                                                                                                           'I occupancy. Makeup water to the SFP is normally provided from the Condensate Transfer System or the Demineralized Water j

System. Makeup to the SFP to account for leakage and j evaporation can also be provided by the seismic Category I service water system. l 1

o The fuel pool system pumps and heat exchangers in contact with the pool water as well as associated piping and valves are corrosion-resistant material. The filter-demineralized maintains total heavy element content in the pool at 0.1 ppm or less. Particulate material is removed by the pressure precoat filter-demineralized unit. A post-strainer is provided in the effluent stream of the filter-demineralized to limit the migration of the filter material. Two small skimmer pumps are provided to remove surface debris by pumping water from the top of the pool through cartridge filters then back to the pool through the service boxes. 3.11.3 Isolation Caoability Isolation of the non-seismic NFPCS from the ESS is achieved by l l l two check valves V-19-18 and V-19-G and two isolation valves of the NFPCS, V-19-H and -I which are nonthrottling MOVs, each I l powered by a different safety-related electrical power supply. Thus, General Design Criterion 2, requiring isolation of seismically qualified systems from non-seismic systems is i satisfied. Each of the two heat exchangers in the ESS has a service water outlet MOV. These two MOVs, V-19-J and -K, are throtting-type valves providing service water flow control, and thereby

         = _ _ _ _ _ -

7 :: , ton

                                                      . 36=-

o > n. controlling'bothithe pool' temperature and the; differential L- pressure between service water and the fuel pool water.

                          ~The' ESS.'is designed'to provide pool cooling under all-licensed                                                       .

plant conditions. This subsystem-.is designed as seismic-Category I using the seismic Category;I Service Water System to n

                          , remove spent: fuel decay heat to the ultimate heat sink-L (Connecticut River). Essential electrical components are also environmentally qualified to ensure operability.under design -

basis accident conditions. Therefore,' pool boiling will not occur, and the Reactor Building environment will not be subject to-the consequences of a boiling spent fuel pool. Instrumentation and controls are provided to detect, control and a A record pump operation, pool temperature, and system flow. e pool leak detention system is provided to monitor leakage 1 I through the pool liner. l-3.11.4 Inspection and Testing L The NFPCS normally has one train .in operation. Redundant units are operated periodically to handle abnormal heat loads or for maintenance. The redundant units of the ESS are periodically operated to ensure that this subsystem can be isolated and provide cooling by remote manual initiation. Also, routine visual inspections for both subsystems' components,

2;

     - lT j.

L instrumentation and alarms will be performed to verify' system. 7

;;                              . operability.

r 3.11.5- Summary: Enhanced Fuel Pool Cooling ~ System i The staff has reviewed and evaluated the enhanced. fuel, pool cooling system as described in Vermont Yankee's submittal dated June .7, 1988. The staff finds that:

• The enhanced spent fuel pool cooling system includes an Emergency' Standby Subsystem (ESS) which has seismic Category I redundant cooling trains with a seismic Category 1 makeup water source.
• The heat removal capacity of one train of the ESS is adecuate'for normal maximum heat load removal. The heat k removal capacity of both trains of the ESS is adequate for the maximum abnormal heat load removal.
• The enhanced systeni has appropriate instruments to monitor the pool water level, and maintains the necessary water level above the spent fuel bundles.
• The filter-demineralized of the enhanced system provides the pool water clarity and purity.

l

• The enhanced system is able to maintain uniform pool water temperature and maintain the temperature at or below the Technical Specification limit of 150*F.

m

6.0 CONCLUSION

S The staff has reviewed and evaluated the licensee's request for expanding

                                      'the capacity of the Vermont Yankee spent fuel pool.        Based on the considerations discussed in this safety evaluation, the staff concludes that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and i (2) such activities will be conducted in compliance with the  ! Commission's regulations and the issuance of this amendment will l not be inimical to the common defense and security or to the 1 health and safety of the public. l l Principal Contributors: S. Kim, H. Richings, A. Chu, M. Lamastra, J. Lee, l F. Witt, J. Ridgely, and V. Rooney. Date of Issuance: l l l l I l l l l l 1 l 1 - - - - - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ \

     ,e e _                 ,

g r ~p~~ /*-" ~' "'] 'l 9 "9 VERMONT YANKEE NUCLEAR POWER CORPORATION b

                                    .                                                                                                               FVY 88-47 RD 5, Box 169. Ferry Road, Brattleboro VT 05301                      ,,,Ly ,o ENGINEERING OFFICE N                                                                                                                 1671 WORCESTER ROAD
                         .                                                                                                           FRAMINGHAM. MASSACHUSETTS 01701
                                  .                                                                                                          TELEPMONE 617-872-8100 June 7, 1988 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn:                 Document Control Desk

References:

a) License No. DPR-28 (Docket No. 50-271) b) Letter, VYNPC to USNRC, FVY 86-34, " Proposed Technical Specification Change for New and Spent Fuel Storage", dated 4/24/86 c) Letter, VYNPC to USNRC, FVY 88-17, " Vermont Yankee Proposed Change No. 133 - Spent Fuel Pool Expansion," dated 3/2/88 d) Letter, USNRC to VYNPC, NVY 88-093, " Spent Fuel Pool Expansion Reracking - Amendment No. 104," dated 5/20/88

Subject:

Vermont Yankee Proposed Technical Specification Change for New and Spent Fuel Storage By letter dated April 24, 1986 [ Reference b)], Vermont Yankee sub-mitted a proposed license amendment request to revise Section 5.5, " Spent and New Fuel Storage" of the Vermont Yankee Technical Specifications to increase the number of spent fuel assemblies allowed to be stored in the spent fuel pool. By letter dated March 2, 1988 [ Reference c)], Vermont Yankee commited to design, install, test, and make operational, a redundant seismically desig..ed Spent Fuel Pool Cooling System prior to the time Vermont Yankee exceeds the existing Technical Specification limit of 2,000 spent fuel assembly storage limit in the Vermont Yankee spent fuel pool. Subsequently, by letter dated May 20, 1988 [ Reference d)], Amendment No. 104 to Vermont Yankee's license was issued allowing the installation of racks of a new desigh in the spent fuel pool sufficient to accommodate 2,870 fuel assemblies, and the storage of fuel assemblies in the new racks up to the present Technical Specification li. nit of 2,000 assemblies in the pool. Use'of the remaining 870 storage positions for the storage of fuel assemblies was not authorized by the license amendment. The NRC letter of May 20, 1988 transmitting Amendment No. 104 to the Vermont Yankee license stated that the staff would complete its review of the thermal-hydraulic aspects of Vermont Yankee's proposed change and con-sider a decision to increase the Technical Specification limit to 2,870

4 VERMONT YANKEE NUCLEAR POWER CORPORATION U.S. Nuclear Regulatory Commission June 7, 1988 Page 2 assemblies after learning more about Vermont Yankee's plans for enhancing the Spent Fuel Pool Cooling System. Accordingly, Vermont Yankee submits as Attachment A to this letter i description of the enhanced Spent Fuel Pool Cooling System in the format of a revised Final Safety Analysis Report (FSAR). The design, installation and testing of the enhanced system will be in accordance with 10 CFR 50.59. On the basis of the information pro-vided in Attchment A, Vermont Yankee requests issuance of the subject license amendment allowing Vermont Yankee use of the full 2,870 storage positions for storage of fuel assemblies in the spent fuel storage pool. Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION

                                                /

A/aa<-- / - Warren P. Mur y Vice Pre ident ap Manager of Ope ns

      /dm cc:   Mr. V. Rooney, USNRC USNRC Regional Administrator, Region I USNRC Resident Inspector,, VYNPC                                         {

ASLB Service List i l ) l 1 }

                    /                                                                                                                                ,

Y a

                                                                                                                                                 .i VYNPS L

FUEL POOL COOLING AND DEMINERALIZED SYSTEM. p. TABLE OF CONTENTS l.

               'Section                     Title                                                             Page
               .A.1   : Power Generation 0bjec tive . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A                                   ,

A.2 ' Safety 0bjective...................................... A-1 A.3 -Power Generation Design Bases......................... A A.4- Safety Design' Basis........................... 1....... A-2  !

               'A.5'    . Description............................................                             A-2                                   l'

!? , ~A.6 Safety Evaluation..................................... A-10 A.7 Inspection'and Testing................................ A-12 1

                                                                                                                                                  .1 l

1 l 1 a i. i J i l 1 1 A-i

                                                                                                                                                     \

A

   -                                          VYNPS FUEL POOL COOLING AND DEMINERALIZED SYSTEM LIST OF FIGURES Figure No.              Title A-1    Fuel Fool Cooling System A-2    Fuel Pool Filter Demineralized System i

i l I i 1 1 l l r f l

                                                                                                                                                .1 l

l l I 1 A-il l

[L. VYNPS FUEL POOL COOLING AND DEMINERALIZED SYSTEM LIST OF TABLES Table No. Title A.1 Fuel Pool Cooling and Demineralized System - System Specifications A.2 Fuel Decay Heat - After Normal Refueling or Full Core Discharged to Pool - Estimated Using SRP 9.1.3 i l l I i i A-lii

VYNPS FUEL POOL COOLING AND DEMINERALIZED SYSTEM j A A.1 Power Generation Objective The objective of the Fuel Pool Cooling and Demineralized System is to remove the decay heat released from the spent fuel elements. The system maintains a specified fuel pool water temperature, purity, water clarity, and water level. j A.2 Safety Objective The safety objective of the Fuel Pool Cooling and Demineralized System is to remove decay heat from the stored fuel and maintain fuel pool water temperature at a level which will help maintain the Reactor Building environment within the bounding limits of the environmental qualification of electrical equipment. A.3 Power Generation Design Bases i

1. The Fuel Pool Cooling and Demineralized System shall minimize corrosion product buildup within the spent fuel pool and shall maintain proper water clarity, so that the fuel assemblies can be efficiently handled underwater.

2. The Fuel Fool Cooling and Demineralized System shall minimize fission product concentration in the spent fuel pool water, thereby minimizing the radioactivity which could be released from the pool to the Reactor Building environment.

3. The Fuel Pool Cooling and Demineralized System shall monitor fuel pool water level and maintain a water level above the fuel sufficient to provide shielding for normal building occupancy.

4 ., The Fuel Pool Cooling System shall be capable of maintaining the spent fuel pool temperature below 150 F. A-1 1

VYNPS-

                                  ~ A.4 Safety Design Basis The Fuel: Pool Cooling and Demineralized System shall be designed to remove the decay heat from the fuel assemblies and maintain fuel pool water temperature at a level which will help maintain the Reactor Building environment within 1            the bounding limits of the environmental qualification of electrical equipment.

A.5 Description General The Fuel Pool Cooling and Demineralized System (FPCDS) consists of four heat

                                  -exchangers,.four pumps, two demineralizers, piping and sufficient valves for coutrol of the design functions and required isolation capability. The Fuel Pool Cooling and Demineralized System pumps and heat exchangers are located in the Reactor Building below the bottom elevation of the fuel pool.

The fuel pool concrete structure, metal liner, spent fuel storage racks, and the Emergency Standby Subsystem of the FFCDS are designed to withstand Seismic Class I earthquake loads. The FPCDS equipment"is arranged"in such a way as to provide a systemwith"two independent means of cooling the-spent fuel pool. Normal' spent fuel pool cooling and cleanup is provided by using the Normal Fuel Pool Cooling Subsystem. This subsystem consist of Pumps P-9-1A and 1B' and Heat Exchangers E-19-1A and 1B which are arranged in two parallel trains with one train normally lined up and operating during plant operation. This ' subsystem of the'FPCDS"is used to provide pool water filtration and demineralization to maintain proper pool water clarity and cleanliness for

                                                                                                                                                                                                                                                      ~ ~                         ~

refueling operations. Thd$'rnEl'Fiiel' Pool'CoolingSubsystem'alsoprovides! l sufficient poo1~ cooling to. maintain posi temperatures within specified limits

                                           ~e-e~ -...

during normal refuelinis'(nominal-one- third-core discharge) and plant'

                                          .m operations.

A-2 l l 1

4 VYNPS However, s..hould an unusually high spen't ' fuel ' decay heat load be placedTin the

           . pool, or a seismic event occur, the Emergency Standby Subsystem can be utilized to maintain pool temperatures within specified limits. The Emergency Standby Subsystem of the FPCDS consists of Pumps P-19-2A and 2B and Heat                                                              ,

Exchangers E-19-2A and 2B which are normally lined up as two parallel trains-in a standby mode to the Normal Fuel' Pool Cooling Subsystem.. Each train of, the Emergency Standby Subsystem can be placed in service remotely. Calculations of expected decay heat loads from normal refuelings and from a full core discharge both with previous cycles of spent fuel in the racks were performed in accordance with the guidance provided in NRC Standard Review Plan 9.1.3, Revision 1, dated July 1981. The normal discharges were assumed discharged to the pool at six days and ten days following shutdown from normal operation. The full core discharge was assumed discharged to the pool ten days following shutdown from normal cperation for refueling. Six days followkg shutdown for a normal refueling is derived from the guidance I provided in NRC Standard Review Plan 9.1.3. Ten days following shutdown for a

 .          normal refueling or a full core discharge is the earliest time at which the s

refueling cavity gates coulu be replaced isolating the reactor vessel from the ~ spent fuel pool. The transfer of the spent fuel assemblies from the reactor vessel to the spent fuel pool is assumed to occur instantly at the six-day or ten-day time period providing a conservative fuel decay heat load in the spent fuel pool. Data from these analyses are provided in Table A.2. Examination of this data shows that while the Normal Fuel Pool Cooling Subsystem heat exchanger capacity may be exceeded for relatively short spent fuel decay times, the backup capability of the Emergency Standby Subsystem of the FPCDS is more than sufficient and can be placed in service until the fuel decay heat load is reduced. l The operating temperature of the fuel pool is permitted to rise up to 25 F above the administrative temperature limit (125 F) when the circulation flow is temporarily interrupted or when larger than normal batches of spent fuel , are placed in the pool. l

    % %+.aw,%.

A-3 l l I

VYNPS Emergency Standby Subsystem The Emergency Standby Subsystem (ESS) of the FPCDS is shown in Figure A-1. The Emergency Standby Subsystem of the FPCDS is a two train, Seismic Class I, Safety Class 3 System designed to prevent a single active failure from disabling both trains. It is designed as a standby system that can remotely be placed in operation from the Control Room. This portion of the system cools the fuel storage pool by transferring the spent fuel decay heat (see Table A.2) to the Service Water System. The pumps circulate the pool water in a closed loop, taking suction from the spent fuel storage pool through the heat exchangers and discharging it back into the fuel pool. The emergency standby heat exchangers are of the shell and tube design, with all parts in contact with the pool water being corrosion resistant material. These heat exchangers are each sized to maintain the fuel pool water temperature below 150 F af ter a normal refueling. Considering one train (one heat exchanger and one pump), this heat removal capability encompasses the normal maximum heat load from completely filling the pool with 2,870 spent fuel assemblies from the last normal discharge. The combined heat removal capability considering both trains (two heat exchangers and two pumps) operating encompasses a full core discharge heat load completely filling the pool with 2,870 spent fuel assemblie's. This provides sufficient heat removal capacity to preclude any impact on plant operation due to insufficient spent fuel pool cooling. The heat exchangers are cooled by the seismically qualified safety-related Service Water System (SWS). The design of the system places the heat exchangers on the suction side of the pumps. In order to protect against fuel pool water leakage into the Service Water System, a positive dif ferential pressure is maintained. The fuel pool water side of the heat exchangers has a maximum operating pressure equivalent to the static pressure head from the pool surface to the heat exchanger. The Service Water System side of the heat exchangers has a minimum operating pressure which is greater than the maximum pressure on the fuel pool side of the heat exchangers. By providing a A-4 l _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

m , F .c

        - +
               .                                                                                                            1 VYNPS positive differential pressure under all conditions of Service Water System-coperation, leakage of fuel pool' water to the environment.is prevented. The differential pressure across each heat' exchanger is monitored by a                                             i i

differential pressure indicator in the Control Room. j n. The Emergency Standby Subsystem of the FPCDS includes two centrifugal pumps each with a: design flow of 700 gpm. All the parts of the pump in contact with the pool water are corrosion-resistant material. The pumps are Seismic Class I and environmentally qualificd to ensure operability after exposure to a harsh environment. The pumps are located within'the FPCDS cubicle in such a manner to prevent common mode failure from fire, flooding, or missiles. A low -] discharge pressure alarm indicates in the Control Room, plus, the pumps are automatically tripped on a low suction pressure condition. One pump alone is designed to provide sufficient flow for the maximum normal heat load from a l normal refueling discharge. For an abnormal heat load, such as full core discharge, two pumps can be running concurrently (one in each train) .) I 1 (reference Table A-1). , Four Motor-Operated Valves (MOVs) provide isolation from the nonseismic Normal Fuel Fool Cooling Subsystem and isolation and throttling of the service water through the heat exchangers. Each heat exchanger service water outlet MOV is powered by the same electrical source as its respective Emergency Standby Subsystem pump. These two MOVs V-19-J and K are throttling-type valves providing service water flow control through its respective heat exchanger, and thereby controlling both pool temperature and service water to fuel pool cooling differential pressure. l The two Normal Fuel Pool Cooling Subsystem Isolation Valves V-19-H and I are I nonthrottling MOVs, each powered by a different safety-related electrical power supply. These isolation valves receive a signal to close on low pool level, providing automatic pool isolation from the Normal Fuel Pool Cooling Subsystem in case of a line break in this nonseismic portion of the FPCDS. In conjunction with the two Normal Fuel Pool Cooling Subsystem isolation MOVs in the supply line, there are two discharge line check valves. These Check Valves V-19-18 and V-19-G provide isolation of the nonseismic Normal Fuel Pool Cooling Subsystem from the A-5

l VYNPS Emergency Star.dby seismic portion of the system. Thus, isolation of the nonseismic portions of the Normal Fuel Fool Cooling Subsystem is assured. Piping associated with the Service Water supply and discharge to the heat exchangers and the fuel pool water piping will be of corrosion resistant ) material. The piping is designed and constructed in accordance with the requirements of ANSI B31.1-77. Valves in the fuel pool water piping are chosen considering their propensity not to collect corrosion products, pressure tight sealing capability, and ease of maintenance. Indication is provided in the Control Room and/or locally near the equipment. Control Room indication for each train includes direct pool temperature, fuel pool water temperature out of the heat exchangers, pump run lights, pump discharge pressures, service water flow, SWS to ESS heat exchanger DP and valve position lights. Local indication includes fuel pool water temperature into the heat exchangers, pump discharge pressures, and heat exchanger DP. Pool tem,arature is provided by redundant thermocouple located within the pool. All other transmitters and sensors are located in or near the Fuel Pool Cooling System cubicle. Control for the two pumps and four MOVs is provided in the Control Room. Control Room controls include pump on/off switches, aervice water throttle valves control switches, and Normal Fuel Fool Cooling Subsystem isolation valves control switches. These remote controls and instrumentation are provided to detect and control pump operation, pool temperature, and system flow, thereby ensuring operability of the Emergency Standby Subsystem of the FPCSD. Normal Fuel Pool Coolisg Subsystem The Normal Fuel Pool Cooling Subsystem (NFPCS) is shown in Figure A-1. The system cools the fuel storage pool by transferring the spent fuel decay heat (see Table A.2) through heat e :-hanger (s) to the Reactor Building Closed Cooling Water System. Water purity and clarity in the storage pool, reactor well, and dryer-separator storage pit are maintained by filtering and A-6

r ,

,  w VYNPS t'

demineralizing the pool water through filter-demineralized (s), which is shown in Figure A-2. The system consists of two circulating pumps connected in parallel, two heat exchangers, two filter-demineralizers, and the required piping, valves and instrumentation. Each pump has a design capacity equal to a filter-demineralized design flow rate (450 gpm) and is capable of simultaneous operation. Two filter-demineralizers are provided. The pumps circulate the pool water in a closed loop, taking suction f rom the spent fuel storage pool, circulating the water through the heat exchanger (s) and filter demineralized (s), and returning it to the fuel pool and reactor well. The fuel pool filter demineralizers are located in the Radwaste Building. The pools (dryer-separator storage pit, reactor well, and fuel storage pool) are filled from the Condensate Transfer System. Make-up to the pools is supplied by the Condensate Transfer System or the Demineralized Water System. Water.is removed from the pools via the fuel pool pumps through the fuel pool filter-demineralized units to the condensate storage tank. Fuel pool water is continuously recirculated except during the temporary periods when the reactor well and dryer-separator pit are being drained. The Normal Fuel Pool Cooling Subsystem is capahle of' removing the decay heat load of the normal discharge batch of spent fue' . tith sufficient decay heat reduction. The Emergency Standby Subsystem can be used in lieu of the Normal Fuel Pool Cooling Subsystem to increase pool cooling in the event that a larger than normal amount of fuel is discharged into the pool or the normal I' fuel pool cooling heat transfer capacity is exceeded. During refueling, when the reactor well is flooded and the gates between the well and the pool are f removed, the RHR System is also available to cool the fuel pool in concert with reactor vessel core cooling. The RHR System has more than enough capacity to cool the reactor vessel core plus the entire inventory in the spent fuel pool. J A-7

VYNPS Two small skimmer pumps are provided which take suction from.the top of the pool.to remove surface debris. These pumps. pump this water through cartridge filters then back to the pool through~the service boxes located around the pools. Pool water clarity and purity are maintained by a combination of filtering and ion exchange processes. The filter-demineralized maintains total heavy element content (Cu, Ni, Fe, Hg, etc.) at 0.1 ppm or less, with a pH range of 5.8 to 8.0 for compatibility with the fuel racks and other equipment. Particulate material is removed from the circulated water by the pressure precoat filter-demineralized unit in which a finely divided disposable filter medium is supported on permanent filter elements. The filter medium is replaced when the pressure drop is excessive or the ion exchange resin is depleted. Backwashing and precoating operations are manually controlled from the Radwaste Building. The spent filter medium is flushed from the elements and transferred to the condensate phase separator tanks by backwashing with air and condensate. The new filter medium is mixed in a precoat tank and transferred as a slurry by a precoat pump to the filter where the solids deposit on the filter elements. The holding pump maintains circulation through the filter in the interval between the precoating operation and the return to normal system operation to hold the precoat on the elements. The pump starts automatically on loss of system flow to maintain sufficient flow through the filter media to retain it on the filter elements. l A post-strainer is provided in the effluent stream of the filter-demineralized l l to limit the migration of the filter material. The filter holding element is capable of withstanding a differential pressure greater than the developed pump head for the system. The maximum pressure drop across the filter and associated process valies and piping should not exceed 25 psid at the time of filter media replacement. The Backwash System is used to completely remove resins and accumulated sludge from the filter demineralizers with a minimum volume of water. Backwash slurry drains to a phase separator. The Precoat System is designed to rapidly apply a uniform precoat of filter media to the holding elements of a filter demineralized. One centrifugal precoat pump and associated piping and valves are provided to precoat either A-8

VYNPS filter-demineralized and recirculate the water to the precoat tank or suction side of the precoat pump. The filter-demineralized units are located separately in shielded rooms. Each room contains only the filter-demineralized and piping. All inlet, outlet, recycle, vent, drain, and other valves are located on the outside of one shielding wall of the room, together with necessary piping and headers, instrument elements, and controls. Penetrations through shielding walls are located so as not to compromise radiation shielding requirements. The fuel pool filter-demineralizers are also used to process liquid radioactive wastes. See Chapt'er 9 of the Vermont Yankee FSAR for details. The system instrumentation is provided for both automatic and remote manual operations. Instrumentation and controls are provided to detect, control and t record pump operation, pool temperature, and system flow. A pool Leak l Detection System has been provided to monitor leakage and thus indicate pool integrity. The pumps can be controlled locally or at Panel 20-22 in the Radwaste Control  ; Room. Pump low suction pressure automatically trips the purps. A pump low discharge pressure alarm indicates in the Radwaste Control Room and a common trouble alarm in the Main Control Room. The flow rate through each of the filter-demineralizers is indicated by a flow indicator on the Pump Room panel and in the Radwaste Control Room. The flow indicators can be seen by the opetators from the vicinity of the Fuel Fool Cooling System control valves. A high rate of leakage lbrough the refueling bellows assembly, drywell to reactor seal, or the fuel pool gates is indicated by lights on the operating floor instrument racks and is alarmed in the Main Control Room. The filter-demineralizers are controlled from a local panel in the Radwaste Building. Differential pressure and conductivity instrumentation are provided for each filter-demineralized unit to indicate when backwash is required. A-9 _ . _ _ _ _ __ ___ ____m

VYNPS Suitable alarms, differential pressure indicators, and flow indicators are provided to monitor the condition of the filter-demineralizers. A.6 Safety Evaluation Maximum normal heat load in the pool will be the sum of the heat from all previous batches plus that just discharged from the current refueling. The Normal Fuel Pool Cooling Subsystem of the Fuel Pool Cooling and Demineralized System is used normally to maintain the pool water temperature below administrative limits during refuelings and plant operation. The Emergency Standby Subsystem is available to provide additional cooling, if needed, to ensure that the pool temperature does not exceed 150 F. Maximum possible heat load would be the sum of the heat from all previous batches plus the heat from a full core discharge. If such a situation arose, the Emergency Standby Subsystem would be used to provide the cooling capacity needed under these conditions, or other high heet load conditions, to maintain the pool water temperature less than 150 F. Also, as an additional means of cooling the spent fuel pool during refueling operations, when the fuel pool and the refueling cavity are connected and filled with water, the Residual Heat Removal (RHR) System can be utilized to provide concurrent cooling to the core and spent fuel poc1 by circulating the water from the core to the pool and back to the core. In this mode, the RER System will be in operation providing cooling to the corn and can be shifted to provide concurrent reactor core and spent fuel pool cooling. The RHR System has more than enough capacity to cool both the reactor core and the entire inventory of stored spent fuel in the spent fuel pool. The Emergency Standby Subsystem is designed to provide pool cooling under all licensed plant conditions. This portion of the system is designed as Seismic Class I using the Seismic Class I Service Water System to remove spent fuel decay heat to the ultimate heat sink (Connecticut River). Essential electrical components in this portion of the system are also environmentally qualified to ensure operability under design basis accident conditions. In , addition, the equipment is located in such a manner as to prevent conson mode i A-10

VYNPS failure from fire, flooding, or missiles. Providing sufficient pool cooling and environmental qualification, assures that the spent fuel will be cooled and boiling _will not occur in the spent fuel pool. Therefore, the Reactor Building environment will'not be subject to the consequences of a boiling spent fuel pool. Leakage of potentially radioactive water from the Emergency Standby Subsystem through the heat exchanger into the Service Water System is prevented by providing a higher service water pressure than the Emergency Standby Subsystem pressure. This differential pressure ensures that leakage, if any, will go into the pool. Indication of this differential pressure is provided in the Control Room along with the controls for initiating the emergency standby portion of the system. Leakage of the potentially radioactive water from the Normal Fuel Pool Cooling Subsystem to the Service Water System is prevented by using an intermediate closed loop cooling system, Reactor Building Closed Cooling Water (RBCCW), which transfers'the heat from the Normal Fuel Pool Cooling Subsystem to the Service Water System. This Closed Loop System arrangement ensures that fuel pool water leakage, if any, is contained within the RBCCW System and not released into the Service Water System. The normal fuel pool cooling flow rato is designed to be larger than that required of two complete water changes per day of the fuel pool, or one change per day of the fuel pool, reactor well, and dryer-separator pit. The Emergency Standby Subsystem flow rate (700 gpm) is approximately 50% greater than the normal fuel pool cooling flow rate (450 gpm). The maximum Normal Fuel Pool Cooling Subsystem flow rate is twice the flow rate needed to

                                                                 ~

maintain the specified water quality. An analysis has been made to determine the consequences of dropping a fully loaded spent fuel shipping cask into the fuel storage pool. The results of that analysis showed that the bottom of the pool would lose its water-tight integrity, thereby making it difficult to maintain adequate shielding and cooling of the stored spent fuel. To prevent any load-drop occurrence, the A-11 ___ _ _ __J

r-VYNPS Reactor Building crane is designed to be single-failureproof. (See Section 12.2.2.2. of the Vermont Yankee FSAR) A.7 Inspection and Testing No special tests are required of the Normal Fuel Pool Cooling Subsystem because at least one pump, heat exchanger, and filter-demineralized are normally in operation while fuel is stored in the pool. Redundant units are operated periodically to handle abnormal heat loads or to replace a unit for servicing. Routine visual inspection of the system components, pumps, heat exchangers, instrumentation, and trouble alarms are adequate to verify system operability. The redundant units of the Emergency Standby Subsystem are periodically operated to ensure that the active components of the subsystem can isolate and provide pool cooling by remote manual initiation. Routine visual inspections of the system components, pumps, heat exchangers, instrumentation, and alarms are adequate to verify system operability. P A-12

VYNPS TABLE A.1 FUEL POOL COOLING AND DEMINERALIZED SYSTEM - SYSTEM SPECIFICATIONS System Function System Specification Normal Fuel Pool Cooling Subsystem Total pool, well, and pit volume 81,500 ft3 Fuel storage pool volume 41,600 ft3 System design flow 450 gpm Maximum flow 900 gpm Pump characteristics 450 gpm, 225 feet TDH, 25 feet NPSH Heat exchanger - Capacity each 2.23 x 106 Btu / hour, FPC temperature 1250F, RBCCW temperature 1000F, RBCCW flow 350 gpm Filter-demineralized 267 square feet, 450 gpm, 25 psi maximum differential pressure (dirty) Emergency Standby Subsystem System design flow 700 gpm Maximum flow 1400 gpm Pump characteristics 700 gpm, 80 feet TDH, 24 feet NPSH Heat exchanger - Capacity each 11.0 x 106 Btu / hour, FPC temperature 1500F, SW temperature 900F, SW flow 700 gpm A-13

VYNPS TABLE A.2 FUEL DECAY HEAT (ESTIMATED) AFTER OPERATION OF 18110NTHS

                               - NORMAL REFUELING, 136 ASSEMBLIES DISCHARGED
                               - FULL CORE DISCHARGE, 368 ASSEMBLIES DISCHARGED Degay Heat (10 Btu /hr)

Normal Refueling Discharge Full Core Discharge Number of 6 Days 10 Days Number of 10 Days Cycle Bundles After After Bundles After Discharged In Pool Shutdown Shutdown In Pool Shutdown 13 (1) 1,586 8.75 7.59 1,818 16.84 14 1,722 9.00 7.79 1,954 17.18 15 1,858 9 18 7.96 , 2,090 17.37 16 1,994 9.35 8.12 2,226 17.53 17 2,130 9.50 8.28 2,362 17.69 18 2,266 9.65 8.42 2,498 17.84 19 2,402 9.80 8.57 2,634 17.99 20 (2) 2,538 9.94 8.71 2,770 18.13 21 2,674 10.07 8.84 2,906 (3) 18.26 22 2,810 10.20 8.97 N/A 23 2,946 (3) 10.33 9.10 N/A NOTE: The decay heat from the previous cycle discharges is included in the above-estimated heat loads.

1. Vermont Yankee is currently in Cycle 13; estimated shutdown is 2/1989.

2. Loss of full core reserve discharge capability.

3. Exceeds capacity of reracked fuel pool.

A-14

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CERTIFICATE OF SERVICE. I' certify that on March 1, 1989, copies of the foregoing pleading were served by first-class mail, or as otherwise.indi-cated, on all parties listed below. E 6$ Charles Bechhoefer, Chairman gs Atomic Safety and Licensing Board Panel Ei- g g U.S. Nuclear Regulatory Commission EE? 'M Ug Washington, D.C. 20555 L. " 4 {"o Gustave A. Linenberger, Jr.  ::1 o Atomic Safety and Licensing Board Panel- ~[Ml $7 U.S. Nuclear Regulatory Commission Dj Washington, D.C. 20555 , Dr. James H. Carpenter Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Secretary of the Commission Attn: Docketing and Service Section U.S. Nuclear-Regulatory Commission Washington,_D.C. 20555 George Dean, Esq. Assistant Attorney General

                ' Commonwealth of Massachusetts Department of the Attorney General One Ashburton Place Boston, MA 02108 Samuel Press, Esq.

George Young, Esq. Vermont Department of Public Service 120 State Street Montpelier, VT 05602 Ann Hodgdon, Esq. 40ffice of the General Counsel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Diana Sidebotham R.F.D. #2 Putney, Vermont 05346 R.K. Gad III Ropes & Gray One International Place Boston, MA 02110 , l t

j;  ; 3 Geoffrey M. Huntington, Esq. Office of the Attorney General Environmental Protection Agency State House Annex 25 Capitol Street Concord, NH. 03301-6397 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dean R. Tousley _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ . . _ _ . _ _ _ _ . _ , _ _ _ _ _ _ _ _ _ _}}