ML20141M613

Forwards Util Response to Re 920715 Meeting. Certificate of Svc Encl.Served on 920827.Reserved on 920828

ML20141M613
Person / Time
Site:	Millstone
Issue date:	08/27/1992
From:	CO-OPERATIVE CITIZEN'S MONITORING NETWORK, INC. (CCMN
To:	NRC OFFICE OF THE SECRETARY (SECY)
References
CON-#392-13191 OLA, NUDOCS 9209030077
Download: ML20141M613 (51)

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Ron e n 5000 August 7, 1992 NL-92-549 Ms. Mary Ellen Marucci 104 Brownell Street New Haven, Connecticut 06511

Dear Ms. Marucci:

This letter is to respond to your questions and requests made during our meeting of July 15, 1992.

We are in receipt of your July 22, 1992, letter which sumnarizes your thoughts on the meeting.

We are pleased that you found the a

meeting informative and helpful, and agree with you that such a non-adversarial approach is more conducive to the sharing of information.

There are two items discussed in your letter which are inconsistent with our recollection.

The first regards direct communication between our respective experts.

As stated in previous correspondence, the Nuclear Licen.;ing department is your point of contact for information related to the license amendment in cuestion.

If your organization requests information which Nuclear Licensing cannot provide, Nuclear Licensing will interf ace with NU technical expe rts, as appropriate, to obtain the inf ormation.

This is the same process used when NU communicates on this issue with other outside organizations, including the NRC.

I will continue to be your primary point of contact on this issue.

In my absence, Mike Wilson can be reached at (203) 665-3684.

For all other technical issues and other matters of concern, your point of contact is our Nuclear Information department (665-5189).

This is our standard company policy for responding to all inquiries from the public.

The Nuclear Information department will work with the appropriate technical experts and will provide you with a response in a timely manner.

Second, although we agreed to reconsider the installation of neutron flux monitors in the Millstone Unit No. 2 Spent Fuel Pool (SFP), we did not agree to bring this issue to the NU Board of Trustees.

Those two issues clarified, the following are the responses to your requests made at the July 15, 1992 meeting.

oS3422 fit.V 4 68

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+

Ms. Mary Ellen Marucci August 5, 1992 NL-92-549/Page 2 i

Request:

Provide data on the activity of each of the re*

muclides i

of a fuel asse:nbly which has undergone 85% of its de.

n burn-up and has decayed for 21 days.

Responses As requested, the calculations cf a Millstone Unit No. 2 fuel assembly which has experienced 85% of its design burn-up and 21 days of decay were performed by members of our Radiological Assessment staff.

The results of these calculations are shown in Attachments 1 and 2 to this letter.

Request:

Provide information concerning the amount of 3 Li 7, produced via neutron interaction with a Boron atom, which gets f rom a Boron coupon to the SFP water.

Response

The amount of 3 LI 7 produced from the neutron-Boron interaction and released into the SFP water is negligible.

Request:

Provide information regarding HU's experience with Boroflex, including which of the units has used it the longest and what the blackness testing results show.

Response

Boroflex is used as a neutron absorbing material in the spent fuel storage racks at all three Millstone units.

Although the portions of each pool which utilize Boroflex are different, Boroflex was installed at all three units around the same time period.

Blackness testing of the Boroflex has been conducted at Millstone Unit Nos. 2 and 3.

The blackness testing most recently conducted at Millstone Unit No.

2 was discussed in the Spent Fuel Pool Criticality Safety Analyses, included as Attachment 2 in the letter which we sent to you dated July 1, 1992. The very conservative l

assumptions utilized in the analyses and test data results are discussed within that attachment.

Blackness testing at Millstone Unit No. 3 was conducted more recently, and we are awaiting the results.

Request:

Evaluate the feasibility of installing an "800" telephone i

number which gives meteorological data at Millstone and the Haddam Neck Plant.

1 i

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4 Ms. Mary Ellen Marucci j

August 5, 1992 i

NL-92-549/Page 3 l

Response

We have evaluated your request end have determined that installing such a service is unnecessary.

NU expends significant resources to assure that our nuclear units are operated in accordance with State and rederal regulations.

Our emergency preparedness measures are in full compliance with NRC, Federal Emergency Management Agency, s

i and State of Connecticut requirements.

We maintain out Emergency Plan Implementing Procedures and our Emergency Organization in a 4

state of readiness in the unlikely event that they are required.

The Emergency Notification System and other communication systems are similarly maintained.

We believe that these measures are comprehensive and fully capable of providing appropriate 4

information to protect public health and safety.

Meteorological data are among the information provided to local and state 1

officials.

Additionally, the appropriate use of meteorological data in determining the effects of radiological releases requires evaluation and interpretation by a trained meteorologist.

This information is provided to State of Connecticut officials who will assess meteorological as well as radiological events.

For the foregoing reasons, we believe that providing such information via i

an "800" number is neither necessary nor appropriate.

Request:

Consider providing to the public the schedule for planned releases in advance of the release.

Responses We have evaluated your request that planned radioactivity releases-be announced in advance of the release.

We have concluded that such announcements are unnecessary and impractical.

i l

l Planned radioactive releases are controlled via station and unit procedures so that they are as low as is reasonably-achievable.

Specifically, the numerical limits prescribed in the technical I

specifications govern these releases.

Accordingly, the levels of l

radioactive material in ef fluents to unrestricted areas are within I

the bounds of federal regulations.

These regulations do not require that planned releases be announced in advance.

Although these releases are planned, the precise schedule for releases is:

l not known far enough in advance to allow for timely public notification.

Hence,. preparing a schedule for such releases in advance would be impractical.

. - ~

o Ms. Mary Ellen Marucci August 5, 1992 NL-92-549/Page 4 In the event of unplanned releases,

federal, state and local officials, the media, and the public are notified in accordance with station and corporate procedures.

Information regarding these releases, should they occur, can be obtained by members of the public from several sources:

Connecticut Department of Environmental Protection, Radiological Controls Division - (203) 566-5134 Connecticut State Office of Emergency Management -

(203) 566-2074 NU Nuclear Information Department - (203) 665-5189 Request:

Provide information which describes the SFP accident scenarios analyzed.

Response

The Millstone Unit No.

2 Final Safety Analysis Report (FSAR) describes the analyzed accident scenarios associated with the STP.

The applicable portions of the FSAR are included as Attachment 3 to this letter.

Request:

Evaluate installing neutron criticality monitors in the d

Unit 2 SFP.

l Responset l

After further consideration, NU's position remains that the installation of neutron sensing criticality monitors in the SFP is an unnecessary, unwarranted expense.

The design of the SFP is such i

I that a significant margin to criticality is maintained during all conditions.

Local monitoring for criticality is obviated by this design.

The SFP criticality analysis previously provided to you describes this in detail.

Request:

Furnish the SFP reactivity calculations performed by HOLTEC, ABB-CE, and NU.

Response

The criticality analysis performed by HOLTEC was submitted to the NRC as part of NU's license amendment request.

This analysis has been provided to you in a previous transmittal.

However, to be responsive to your request for additional detail, Attachment 4-to this letter contains the HOLTEC benchmark calculations.

1

~. -

l Ms. Mary Ellen Marucci l

August 5, 1992 j

14L-92-549/Page 5 Appropriate adjustments to the reactivity calculations were made based on these benchmarks.

Regarding the ABB-CE calculations, NU is not in possession of them.

ABB-CE acknowledged the analysis error and has provided to NU and the NRC a description of the cause of the error.

The reactivity 1

calculations upon which the amendment is based were done by HOLTEC and underwent independent review as part of their Quality Assurance program.

Although NU performed a criticality analysis, it is not the analysis upon which the amendment is based.

It would, therefore, be inappropriate to release any of our calculations.

i Regarding Mr. Sullivan's stated concerns regarding evacuation of the communities surrounding Millstone Station, we recommend that he contact the local Civil Preparedness Directors.

They are:

New London - Mr. Reid Burdick - (203) 447-5200 Waterford

- Ms. Karen Sturgeon - (203) 442-9585 East Lyme

- Mr. Fred Johnson - (203) 739-4434 We will continue to cooperate with your requests for information as we have in the past in the hopes that continuing dialogue will j

resolve your concerns. As previously requested, I would appreciate your directing any further questions you may have on the subject to my attention.

I may be contacted at (203) 665-3298, Sincerely, i

Gum) e.10fe>rnb-h s

Richard M.

Kacich Director, Nuclear Licensing rmk/mjw/lmp Attachments cc:

NL Memo File Nuclear Records

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l Millstone Unit No. 2 Fuel Assembly Fiaal Safety Analysis Report Fuel-Handling Accident Excerpts

. August 1992

. ~

KNPS 2 FSAR 14.7 RADIOACTIVE RELEASES FROM A SUBSYSTEM OR COMPONENT 14.7.1 WASTE GAS SYSTEM FAILURE 14.7.1.1 General The limiting accident considered is the postulated and uncontrolled release to the auxiliary building of the radioactive xenon and krypton gases stored in one vaste gas decay tank.

The credibility of such an occurrence is low since the waste gas system is net subjected to pressures greater than 150 psig, or large stresses.

The result of a rupture of a gas decay tank is analyzed in order that the maximum hazard which would result from a malfunction in the radioactive waste processing system will be defined.

14.7.1.2 Method of Analysis It is assumed that the tank contains the gaseous activity evolved from degassing one system volume of reactor coolant for refueling.

The maximum activity would exist prior to cold shutdown at the end of an operating cycle during which extended operation with one percent defective fuel had occurred.

Based on this and neglecting' decay after degasification, the noble gas activity in the tank is given in Table 14.7.1 1, 14.7.1.3 Results of Analysis Dose (rems)

Site QIga.D Boundary M2 Thyroid Whole body 6.4 x 10'1 6.6 x 10-2 14.7.1.4 Conclusions If a vaste gas decay tank rupture did occur, the dose would be within 10 CFR Part 100 guidelines.

14.7.2 RADIOACTIVE L1 QUID WASTE SYSTEM LEAK OR FAILURE (RELEASE TO ATMOSPHERE)

This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.

16.7.3 POSTU1ATED RADIOACTIVE RELEASES DUE TO LIQUID-CONTAINING TANK FAIUURES This event is not in the current licensing basis for Millstone Unit 2 and therefore is not analyzed.

14.7.4 RADIOLDGICAL CONSEQUENCES OF FUEL HANDLING ACCIDENT 14.7.4.1 General The likelihood of a fuel handling accident is minimized by administrative controls and. physical limitations imposed upon fuel handling operations'.

e214-7. cal 14.7-1 5/17/90

=.

MNPS 2 FSAR All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a qualified supervisor. Also, before any refueling operations begin, verification of complete control element assembly (CEA) insertion is obtained by tripping each CEA individ-ually to obtain indication of assembly drop and disengagement from the drive shaft.

Boron concentration in the coolant is raised to the refueling concentration of 1720 ppm boron, or more and is verified by chemical analysis. At a boron concentration of 1720 ppm, the core will be more than 9.5 percent suberitical, even with all CEA's withdrawn.

After the vessel head is removed, the CEA drive shaf ts are removed from their respective assemblies. A load cell is used to indicate that the drive shaft is free of the CEA as the lifting force is applied.

The maximum elevation to which the fuel assemblies can be raised is limited by the use of hard stops in the fuel handling hoirts and manipulators to ensure that the minimum depth of water above the top of a fuel assembly required for shielding is always present.

This constraint applies in-fuel handling areas inside containment and in the spent fuel pool area.

Supplementing the physical limits on fuel withdrawal, radiation monitors located at the fuel handling areas provide both audible and visual warning of high radiation levels in the event of a low water level in the refueling cavity or fuel pool.

Fuel pool structural integrity is assured by designing the pool and the spent fuel storage racks as Seismic Class I structures.

The design of the spent fuel storage racks and handling facilities in both the containment and fuel storage area is such that fuel will always be in a suberitical geometrical array, assuming zero boron concentration in the fuel pool water. The spent fuel pool and refueling pool water contain a minimum of 800 ppm and 1720 ppm of boron, respectisely.

Natural convection of the surrounding water provides adequate cooling of fuel during handling and storage. Adequate cooling of the water is provided by forced circula-tion in the spent fuel pool cooling system.

At no time during the transfer from the reactor core to the spent fuel storage rack is the spent fuel 7

removed from the water.

Fuel failure during refueling as a result of inadvertent criticality or overheating is not possible.

The possibility of damage to a fuel assembly -

as a consequence of mishandling is minimized by extensive personnel training, detailed procedures, and equipment design. The design precludes the handling of heavy objects such as shipping casks over the spent fuel storage racks with the exception of the consolidated fuel storage box.

Inadvertent disengagement of.a fuel assembly or consolidated fuel storage box from the fuel handling machine is prevented by mechanical interlocks, Consequently, the possibility of dropping either one and damaging of a fuel i

l assembly is remote.

Should a fuel assembly be dropped or otherwise damaged during handling, radioactive release could occur in either the containment or the auxiliary-building. The ventilation exhaust-air from both of these areas is moni-tored before release to the atmosphere (see-Subsection 7.5.6.3).

The radiation monitors immediately indicate the increased activity level and alarm.

The affected area would then be evacuated.

m n *-L ost 14.7-2 5/17/90

MNPS 2 FSAR Release of activity through the containment purge system would be prevented by automatic closure of the containment isolation dampers as described in Subsection 9.9.2.2.

The containment personnel hatches and equipment hatches are closed during fuel handling operations.

Since the auxiliary building cannot be completely isolated, this results in a more limiting activity release to the environment.

Prior to the handling of irradiated fuel, the exhaust air is diverted from the main exhaust system by being manually aligned to the auxiliary exhaust system (AES) and exhausted from the spent fuel pool area through the enclosure building filtration system (EBFS) charcoal filter to remove iodines (see Subsec-tion 9.9.8) prior to release through the Unit i stack.

14.7.4.2 Method of Analysis For the purpose of_ defining the upper limit on fuel damage as the result of a fuel handling accident, it is assumed that the el assembly or consoli-dated fuel storage box is dropped during handling.

Interlocks, procedural and administrative controls make such an event unlikely.

However, if an assembly is damaged to the extent that a number of fuel rods fail, the accumulated fission gases and iodines in the fuel element gap could be released to the surrounding water.

Release of the fission products to the surrounding water is considered negligible as a result of reduced diffusion through the fuel due to the low fuel temperature during refueling, The fuel assemblies and consolidated fuel storage box are stored within the spect fuel rack at the bottom of the spent fuel pool.

The top of the rack extends above the top of the stored fuel.

A dropped fuel assembly or I

consolidated fuel storage box could not strike more then one fuel assembly in the storage rack.

Impact can occur only between the ends of the involved components, the bottom end fitting of the dropped components j

impacting against the top end fitting of the stored fuel assembly.

The-re sults of an analysis on the energy absorption capability of a fuel l

l assembly indicate that a fuel assembly is capabic of absorbing the kinetic l

enargy of the fuel assembly or consolidated fuel storage box drop with no fuel rod failures.

The worst fuel handling incident that could occur in the spent fuel pool is the dropping.of a fuel assembly to the fuel pool floor.

The dropping of a consolidated fuel storage box was evaluated and determined to be bounded by the' fuel assembly drop to the fuel pool floor.

~

After striking the pool floor vertically, the assembly would rotate into a horizontal attitude.

It is postulated that during this rotation the assembly will strike a protruding structure.

The fuel storage pool has been designed without such a protruding structure, hence, the shape and nature of the assumed member is indeterminate.

For this analysis, there-fore, a line load has been assumed.

To obtain an estimate of the number of fuel rods which might fail in the event a fuel assembly is dropped, the energy required-to crush a fuel rod and bend the entire -assembly has been determined.

The point of impact was assumed to be the most effective location for fuel rod damage, the center of percussion.

Resistance to crushing offered by the fuel-pellet is I

considered in the analysis.

Failure of the fuel tube by crushing absorbs the least energy, hence, the model produces a_ conservative upper limit for the number of fuel rod failures. This failure mode is applicable to the outer row of fuel rods only.

Since it is not possible to apply a line load e214-L oel 14.7-3 5/17/90

MNPS-2 FSAR 1

beyond t.he outer row of fuel rods, the failure mode of rods in rows other i

than the outer rows will be by bending rather than by crushing.

3 Approximately 36,000 in. lbs of kinetic energy from rotation must be absorbed.

The energy required to bend the assembly and crush the outer row of fuel rods to failt:re is 4,600 in.-lb.

Failure of.the second row of fuel rods by bending along requires more than 70,000 in.-lbs.

Thus, no more than 14 fuel rods, i.e.,

one outer row of rods, would be expected to fail.

All X/Q values have been chosen in the following manner:

Site meteo-rological data has been examined for the years 1974, 1975, and 1976.

For each release point and dose calculation time period in question, the year with the largest (most conservative) 95% maximum X/Q value has been chosen.

For each accident, the results indicate that for operation of Millstone Unic No. 2 at 2700 MVt. the radiological consequences will not exceed the limitations of 10CFR100, and are in fact significantly below the limits in most cases.

14.7.4.2.1 Fuel Handling Accident in the Spent Fuel Pool This accident has been reanalyzed using the assumptions contained in i

Regulatory Guide 1.25.

A complete list of assumptions is provided in Tabic 14.7.4-1.

The results of this analysis, which are well below the limits of 10CFR100, are summarized in Section 14.7.4.3.1.

14.7.4.2 2 Fuel Handling Accident in Containment i

A complete list of the assumptions used in this calculation is provided in Table 14.7.4-2.

The results of the analysis, which are well within the limits of 10CFR100, are summarized in Section 14.7.4.3.2.

14.7.4.3 Results of Analysis 14.7.4.3.1 Spent Fuel Pool Accident Dose (rems) 11te Boundgy LPZ One One Organ Assembiv 14 Rods Assembly 14 Rods L

Thyroid 3.3 2.7 x 10" 1.1 8.9 x 10-2 Whole Body

7. 9 x 10-2 6.4 x 10-3

2. 6 x 10-2 2.1 x 10'3 a

4 Mr214-7.est 14.7-4 5/17/90 s

MNPS-2'FSAR 14.7.4.3.2 Containment Accident Dose (rems)

Site Boundary LPZ One One Organ _

14 Reds Assembly L4 Pods Assembly Thyroid 1.5 18.1 1.5 x 10'1 1.9 h le Body 5.5 x 10-3

6. 8 x 10-2

5. 7 x 10

7.6 x 10-3 14.7.4'.4 Conclusions The exclusion boundary doses resulting from a fuel handling accident are within the guidelines of 10CFR Part 100. Thus, a dropped fuel assembly will not present any undue hazard to the health and safety of the public.

I' 14.7.5 SPENT FUEL CASK DROP ACCIDENTS As discussed in Section 5.4.3.1.9, dropping a spent fuel cask could :.esult in the rupture of up to 587 intact assemblies.

Per Technical Specifica.

tions, these assemblies must be decayed for a minimum of.120 days.

(Note:

l A larger number of consolidated fuel rods.could rupture, but since these i

assemblies must be decayed at least 5 years, the dose consequences would be less.) A dose calculation was performed for the assumed rupture of 587 assemblies with 120-day decay. This calculation was performed by ratioing MP2 specific parameters to those generic values used in th6 dose assessment section of NUREG 0612. The MP2 specific assumptions, which were different from the NUREG-0612 assumptions, were:

Power Level - 2700 MW, 3

the EAB - 5.4 x 10 sec/m and Number of Assemblies in 0-2 HR y/Q at i

Core - 217. The resulting whole body dose at the EAB was calculated to be 241 m*em.

The thyroid dose is insignificant after 120 days decay.

Therefore, the resulting dose is within the acceptable small fraction of l

10CFR Part 100 limits, l

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KNPS 2 FSAR TABLI 14.7.4 1 Assuretion for Puel Handline Accident in the Scent Puel Pool Assunction Basis (1) Reactor Core Power Level -

2700 MVt Stretch Power (2) Iodine Pool Decontamination Factor - 100 Reg. Guide 1.25 (3) Activity Released from Rods:

Iodines - 10%

Noble Gases (Except KR 85) - 10%

Reg. Guids 1.25 KR 85 - 30%

(4) Chemical Form of Iodines Above the Pool:

Reg. Guide 1.25 25% in Organic Form 75% in Inorganic Form (5) a) One Assembly Assumed to Rupture a) Reg. Guide 1.25 b) 14 Rods Assumed to Rupture b) FSAR i

(6) Decay Time - 72 Hours Tech. Spec f3.9.3 120 Days Tech. Spec. f3.9.16.1 5 Years Tech. Spec. 63.9.20 (7) Number of Assemblies in Core - 217 FSAR (8) EBFS Filter Efficiencies:

Reg. Guide 1.25/MP2 SER Organic Iodine - 70%

Elemental Iodine - 90%

(9) All Activity Released from Fuel Pool Reg. Guide 1.25,

Butiding Instantaneously Through Filters (10) X/Qs (sec/m') (for MP1 Stack Release) 951 Maximum X/Qs during Site Boundary (01 Hour) - (1.03E-4) the years 1974-1976 LPZ (0 1 Hour) - 3.41E 5 (11) Thyroid Dose Conversion Factors See Justification Under from Reg. Guide 1,109 Section V, LOCA (12) Semi-Infinite Cloud Dose Model Reg. Guide 1.25 i

(13) Feaking Factor - 1.65 Reg. Guide 1,25 (14) Breathing Rate - 3.97 x 10** m /sec.

Reg. Guide 1.25 3

czu6x.csi 1 of 1 5/17/90

e a

MNPS 2 FSAR With the exception of the missile impact area, the allowable stresses to resist the effects of tornadoes are 90 percent of the yield strength of the reinforcing steel and 85 percent of the ultimate strength ~of the concrete.

A discussion of the probability of tornado occurrence is presented"in i

Section 2.3.

3.4.3.1.7 Pipe Restraint Loads These are the loads imparted to the structure from the pipe restra'ints i

produced by either a postulated pipe rupture or an earthquake.

5.4. 3.1. 8 Pipe Whipping Loads These are the loads imposed on the sttucture due to whipping from a postulated pipe rupture.

3 5.4.3.1.9 Cask Drop Loads The following design criteria were used in the analysis of the spent fuel J

pool in the event that a cask is accidentally dropped:

l a.

Weight of cask in air (1b) 200,000 b.

Length of cask (ft) 19 c.

Diaceter of cask (ft) 8 d.

Distance of drop (ft)

In air 2-3/4 In water 35-1/2 1

As shown on Figure 5.3-5. the only area of the spent fuel pool into which the cask could be dropped directly is the cask laydown area.

The cask laydown area is isolated from the spent fuel storage area by two foot thick,- permanent, reinforced concrete walls and a temporary gate placed in t

the fuel transfer slot.

The base slab of the cask laydown area is composed of seven feet of reinforced concrete resting on a mass of monolithic concre _ a which in turn rests on bedrock.

Therefore a cask dropped in this area would travel vertically downward as restrained by the surrounding walls.

Any damage would be limited to rupturing of the spent fuel pool liner and local' superficial crushing of concrete in the area of impact of the end of the cask.

Leakage through the ruptured liner would be detected in the control room and woule be stopped by closing the valve that connects the leak collection channel for the ruptured zone (s) to the leak detection instrumentation.

If during handling, the cask is dropped on or near point "A," as shown on j

Figure 5.3-5, there exists a possibility that the cask could fall or tumble into the spent-fuel storage' area.

The fall would provide some local i

concrete crushing in_the spent fuel pool and laydown area walls at eleva-tion (+) 38'-6" The_ cask would then slide into the spent fuel storage area of the pool.

The cask would crush the spent fuel rack module (s) that it landed _on, but the buoyant effect of the water combined with the crushing of the rack would dissipate most of the kinetic energy of the falling. cask.

Therefore the probable damage would be limited to rupture of the spent fuel pool liner and-local crushing of concrete where the cask impacted.

Damage to spent fuel stored in the pool (both intact and l

1 ten-4. c e.

5.4-5 6/13/90

MNPS-2 FSAR consolidated) would be mitigated by administrative 1y controlling the age of the stored fuel in the affected area around the cask laydown area of the Region II spent fuel pool.

Technical Specifications require that all stored fuel within-a specified distance of the cask laydown area sh,all have decayed a minimum of 120 days from suberitical reactor operation whenever a shielded cask is on the refueling floor.

The seven foot thick base and six foot thick walls, of reinforced concrete, would remain intact.

Leakage 4

would be detected and stopped as described above.

Makeup water would be available at a maximum of 300 gpm from the p'rimary makeup water system, at a maximum of 600 gpm from the fire protection system hose stations located in the auxiliary building at elevation (+)

38'-6" and at a maximum of 600 gpm from the auxiliary feedwater system i

(AFWS).

I 5.4.3.1.10 Fuel Transfer Tube Bellows The following loads were used-in the design of the fuel-transfer tube and bellows:

Design pressure, internal (psi) 60 Design temperature (F) 290 Lateral movement (in.)

0.14 Axial movement, expansion or contraction (in.)

C5 Displacements are selected to accommodate an assumed differential settle-ment of one-eighth inch between the buildings.

Since both the containment and auxiliary buildings are founded on rock, this motion is minimal.

5.4.3.2 Design Load Combinations To ensure the structural integrity of the auxiliary building, the working stress method of design is used for the various loading combinations.

For the operating conditions, normal allowable stresses given in the American Institute of Steel Construction (AISC) Manual of Steel Construction 1963, and the American Concrete Institute (ACI) 318-63, " Building Code Require-ments for Reinforced Concrete" are used.

These allowable stresses are increased by 33-1/3 percent for-the 115-mph base wind loads and the,

operating basis earthquake (OBE) loads.

For the tornado wind and the design basis earthquake (DBE), the allowable stresses are 90 percent of the yield strength of the reinforcing, and 85 percent of the ultimate strength of concrete.

The load combinations are listed:

a.

D+L b.

D+L+W, c.

D+L+We d.

D+L+E e.

D + L + E' f.

D + L _ + P. + We + H, g.

D+L+T+E h.

D + L + T + E' i.

D + L + P, + E ten-4. c al 5.4-6 6/13/90

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SPENT FUE! CASK TRWEt LIMITS FIGUkE S.5-6 2

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Millstone Unit No. 2 Spent Fue'l Pool Criticality Analysis Benchmark Calculations-l l

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e-APPENDIX C

BENCHMARK CALCULATIONS 1

by Stanley E. Turner, PhD, PE HOLTEC INTERNATIONAL December, 1991

~

^*V

1.0 INTRODUCTION

AND S1DMARY The objective of this benchmarking study is to verify both the NITAWL-KENO-Sad 'i' methodology with the 27-group SCALE i

cross-section library and the CASMO-3 coded) for use.in criticality safety calculations of high density spent fuel storage racks.

Both calculational. methods are based upon transport theory and have been benchmarked against critical experiments that simulate typical spent fuel storage rack designs as realistically as possible.

Results of these benchmark calculations with both methodologies are consistent with corresponding calculations reported -in the literature.

l Results of the benchmark calculations show that the 27-group (SCALE) NITAWL-KENO-Sa calculations consistently under-predict the critical eigenvalue by 0.0101 1 0.0018 8k (with a 95%

probability at a 95% confidence level) for critical experimentsN that are as representative as possible of realistic spent fuel storage rack configurations and poison worths.

Extensive benchmarking calculations of critical experi-ments with CASMO-3 have also been reportedm, giving a mean k,,, of 1.0004 i O.0011 for 37 cases.

With a K-f actor of 2.14m for 95%

probability at a 95% confidence level, and conservatively neglect-ing the small overprediction, the CASMO-3 bias then becomes 0.0000 1 0.0024.

CASMO-3 and NITAWL-KENO-Sa intercomparison calculations of infinite arrays of poisoned cell configurations (representative of typical spent fuel storage rack designs) show very good agreement, confirring that 0.0000 0.0024 is a reasonable bit.s and uncertainty for CASMO-3 calculations.

Reference 5 also documents good agreement of heavy nuclide concentrations for the Yankee core isotopics, agreeing with the measured values within experimental error.

A-1

,y..,,.

+,

=-

Tho benchmark calculations -roported h3ro confirm that j

either the 27-group (SCALE) NITAWL-KENO or CASMO-3 calculations are acceptable for criticality analysis of high-density spent fuel storage racks.

Where possible, reference calculations for storage rack designs should be performed with both code packages to provide 17idependent verification.

1 2' O NITAWL-KENO Sa BENCHMARK CALCULATIONS Analysis of a

series of Babcock & Wilcox critical 5

experiments"), including some with absorber panels. typical of a

poisoned spent fuel rack, is summarized in Table 1, as calculated with HITAWL-KENO-Sa using the 27-group SCALE cross-section library and the Nordheim resonance integral treatment in NITAWL.

Dancoff.

factors for input to NITAWL were calculated with the Oak Ridge SUPERDAN routine -(from the SCALE (2) system of codes). The mean for

{

these calculations-is 0.9899 i O.0028 (1 o standard deviation of i

the population).

With a one-sided tolerance factor corresponding to 95% probability at a 95% confidence level"),'the calculational bias is + 0.0113 with an uncertainty of the mean of of i O.0018 for the sixteen critical experiments analyzed, a

Similar calculational deviations have been reported by ORNL") for some 54 critical experiments (mostly clean critical without strong absorbers), obtaining a mean bias of 0.0100 i O.0013

)

(95%/95%).

These published results are in good agreement with the f

results obtained in the present analysis and lend further credence to the validity of the 27 group' NITAWL-XENO-Sa calculational model for use in criticality anal' sis of high density spent fuel storage y

racks.

No trends in k with intra-assembly water gap, with g

absorber panel reactivity worth, with enrichment or with poison concentration were identified.

A-2 l

l l

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~. -.. -.

Additional benchmarking calculctiono waro cico made for a series of French critical experiments

• at 4.75% enrichment and for several of the BNWL criticals with 4.26% enriched fuel.

Analysis of the French criticals (Table 2) showed a tendency to overpredict the reactivity, a result also obtained by ORNLUM.

The

--alculated k,,, values showed a trend toward higher values with decreasing core size.

In the absence of a significant enrh. ent effect (see Section 3 below), this trend and the overpredic':

.n is attributed to a small inadequacy in NITAWL-KENO-Sa in calculating neutren leakage from very small assemblies.

Similar overprediction was also observed for the BNWL series of critical experimentsOU, which also are small assemblies (although significantly larger than the French criticals). In this case (Table 2), the overprediction appears to be small, giving a mean k,g of 0.9990 i O.0037 (1 o population standard deviation).

Because of the small size of the BNWL critical experiments and the absence of any significant enrichment effect, the overprediction is also attributed to the failure of NITAWL-KENO-Sa to adequately treat neutron leakage in very small assemblies.

Since the analysis of high-density spent fuel storage racks generally does not entail neutron leakage, the observed inadequacy of NITAWL-KENO-Sa is not significant.

Furthermore, omitting results of the French and BNWL critical experiment analyses from the determination of bias is conservative since any leakage that might enter into the analysis would tend to result in overprediction of the reactivity.

t A-3 I

l

3a CASMO-3 BENCHMARK CAlfULATIONS e

The CASMO-3 code is a multigroup transport tteory code utilizing transmission probabilities to accomplish two-dimensional calculations of reactivity and depletion for BWR and PWR fuel Gsemblies.

As such, CASMO-3 is well-suited to the criticality analysis of spent fuel storage racks, since general practice is to t'reat the racks as an infinite medium of storage cells, neglecting leakage effects.

CASMO-3 is a modification of the CASMO-2E code and has been extensively benchmarked against both mixed oxide and hot and cold critical experiments by Studsvik Energiteknik s)

. Reported ana-t lyses (5) of 37 critical experiments indicate a mean k, of 1.0004 i g

O.0011 (10).

To independently confirm the validity of CASMO-3 (and to investigate any effect of enrichment),

a series of i

calculations were made with CASMO-3 and with NITAWL-KENO-Sa on identical poisoned storage cells representative of high-density spent fuel storage racks.

Results of these intercomparison calculations * (shown in Table 3) are within the normal statistical i

variation of KENO calculations and confirm the bias of 0.0000 1 0.0024 (95%/95%) for CASMO-3.

Since two independent methods of analysis would not be expected to have the same error function with enrichment, results of the intercomparison analyses (Table 3) indicate that there is no significant effect of fuel enrichment over the range of enrich-ments involved in power reactor fuel.

Furthermore, neglecting the French and BNWL critical benchmarking in the determination of bias is a conservative approach.

• Intercomparison between analytical methods is a technique endorsed by Reg. Guide 5.14,

" Validation of Calculational Methods for Nuclear Criticality Safety".

i A-4

I j

i' REFERENCES TO APPENDIX b I

I i

i

. i e

i 1.

Green, Lucious, Patria, Ford, White, and Wright, "PSR l

/NITAWL-1 (code package) NITAWL Modular Code System For Generating coupled Multigroup Neutron-Gamma Libraries from j

ENDF/B", ORNL-TM-3706, oak Ridge National Laboratory, November j

1975.

i 2.

R.M. Westf all et. al., " SCALE: A Modular System for Performing i

Standardized Computer Analysis for Licensing Evaluation",

NUREG/CR-0200, 1979.

7 3.

A.

Ahlin, M.

Edenius, and H.

Haggblom, "CASMO A Fuel i

Assembly Burnup Program", AE-RF-76-4158, Studsvik report.

l f

A. Ahlin and M. Edenius, "CASMO - A Fast Transport Theory Depletion Code for LWR Analysis", ANS Transactions, Vol. 26, l

p. 604, 1977.

l "CASMO-3 A Fuel Assembly Burnup Program, Users Manual",

j Studsvik/NFA-87/7, Studsvik Energitechnik AB, November 1986 4

l 4.

M.N. Baldwin et al., " Critical Experiments Supporting Close

{

Proximity Water Storage of Power Reactor Fuel", BAW-1484-7, l

The Babcock & Wilcox Co., July 1979.

I l

l 5.

M. Edenius and A. Ahlin, "CASMO-3 New Features, Benchmarking, j

and Advancei Applications", Euclear Science and Encineering, j

100, 342-SL, (1988) 6.

M.G. Natrella,_ Experimental Statistics, National Bureau of Standards, Handbook-91, August 1963.

i i

i 7.

R. W. Westf all and J. H.. Knight, " SCALE System Cross-section j

Validation with Shipping-cask Critical Experiments",

ME j.

Transactions, Vol. 33, p. 368, November 1979 4

i 8.

S.E.

Turner :and M.K.

Gurley,

" Evaluation of NITkWL-KENO i

Benchmark Calculations for High Density Spent Fuel Storage.

Racks",_

Nuclear Science and Encineerina, 80(2):230-237, February 1982.

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i c'ww ewre g, v' y,ges w wwewww ey w uw w, tm M y* wur

>gn m y-'=,

• * +cw* S ev et e a

ero c e

> w a.,ew w_

w reew ww -er m+-

. e -v -wwe e vr trvwe k - % =w

+w-re-e-.

emm w" 'w v y - yg =

a q

9.

J.C. Manaranche, et. al., " Dissolution and Storage Experiment with 4.75% U-235 Enriched UO, Rods", Nuclear Techno1gn, Vol.

50, pp 148, September 1980 10.

A.M.

Hathout, et.

al.,

" Validation of Three Cross-section Libraries Used with the SCALE System for Criticality Analy-sis", Oak Ridge National Laboratory, NUREG/CR-1917, 1981.

11.

S.R.

Baerman, et.

al.,

" Critical Separation between Sub-critical Clusters of 4.29 Wt. 4 *U Enriched UO Rods in Water with Fixod Neutron Poisons", Bate 11e Pacific Nor,thwest Labora-tories, NUREG/CR/OO73, May 1978 (with August 1979 errata).

A 4

i A-6

.... ~ -..

. _. ~.. _ _ _. _ _ _. _ _. _ _.

1 1

Tcblo 1 RESULTS OF 27-GROUP (SCALE) NITAWL-KENO-Sa Calc LATIONS

(

OF B&W CRITICAL EXPERIMENTS k

j Experiment Calculated a

Number k,,,

}

)

I 0.9922 1 0.0006 II 0.9917 1 0.0005 III 0.9931 1 0.0005 IX 0.9915 1 0.0006 X

0.9903 1 0.0006 XI 0.9919 1 0.0005 4

XII 0.9915 1 0.0006 i

XIII

f.. " 9 4 5 O.0006 XIV 0.9902 1 0.0006 XV 0.9836 1 0.0006 XVI 0.9863 i 0.0006 XVII 0.9875 0.0006 XVIII 0.9880 1 0.C006 XIX 0.9882 1 0.0005 i

XX 0.9885 0.0006 XXI 0.9890 1 0.0006 Hean 0.9899 1 0.00070)

Dias (95%/95%)

0.0101 1 0.0018 di Standard Deviation of the Mean, calculated from the k,,, values.

l I

i A-7 l

l l

l l

l 4

o Table 2 a

i RESULTS OF 27-GROUP (SCALE) HITAWL-KENO-Sa CAIEULATIONS L

OF FRENCH and B)NL CRITICAL EXPERIMENTS i

I French Experiments i

separation Critical Calculated Distance, em Height, cm k,,,

O 23.8 1.0231 1 0.0036 2.5 24.48 1.0252 i O.0043 5.0 31.47 1.0073 i O.0013 10.0 64.34 0.9944 i O.0014 BlNL Experiments Calculated Case Expt. No.

k,,,

No Absorber 004/032 0.9964 i O.0034 SS Plates (1.05 B) 009 0.9988 i O.0038 SS Plates (1.62 B) 011 1.0032 i O.0033 i

SS Plates (1.62 B) 012 0.9986 i O.0036 SS Plates 013 0.9980 1 0.0038 SS Plates 014 0.9936 i O.0036 Zr Plates 030 1.0044 i O.0035 Mean O. 99' s i O.0037 A-8

~

.. _ _... _ _.. ~.. _. _.. _. _.,, _ _. - _ _ _ _.. _

.....~ -..-.._

j Table 3 RESULTS OF CASMO-3 AND HITAWL-KENO-5a BENCHMARK (INTERCOMPARISON) CALCUIATIONS i

Enrichment (D k,

Wt. 4 U-235 NITAWL-KENO-Sa )

CASMO-3 l8k!

tt 2.5 0.8408 i O.0016 0.8379 0.0029 1

3.0 0.8331 i O.0016 0.8776 0.0055 j

3.5 0.5097 i O.0016 0.9090 0.0007 4.0 0.9334 i O.0016 0.9346 0.0012 a

4.5 0.9569 i O.0018 0.9559 0.0010 1

5.0 0.9766 i O.0018 0.9741 0.0025 Mean 0.0023 W

i Infinite array of assemblies typical of high-density spent fuel storage racks.

(Il k, from NITAWL-KENO-Sa corrected for bias.

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UNITED STATES OF AMERICA i

NUCLEAR REGULATORY COMMISSION in the hatter of NORTHEAST NUCLEAR EllERGY COMPANY Docket No.(s) 50-336-OLA (Millstone Nuclear Power Station, Unit No. 2)

CERTIFICATE OF SERVICE 1 hereby certify that copies of the foregoing NOTE CCMN RE NU RESPONSE 8/7..

have been served upon the following persons by U.S. mail, first class, except as otherwise noted and in accordance with the requirements of 10 CFR Sec. 2.712.

Office of Commission Appellate Administrative Judge Adjudication Ivan W. Smith, Chairman U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Washington, DC 20555' U.S. Nuclear Regulatory Commission Washington, DC 20555 Administrative Judge Administrative Judge Charles N. Kelber Jerry R. Kline 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 John T. Hull, Esq.

Richard M. Kacich-Office of the General Counsel Director, Nucle;r Licensing U.S. Nuclear Regulatory Commission Northeast Utilities Washington, DC 20555 P. O. Box 270

!!artford, - CT 06101 Patricia R. Nowicki Nicholas S, Reynolds, Esq.

Associate Director John A. MacEvoy, Esq.

EARTHVISION, Inc.

Winston & Strawn 42 Highland Drive 1400 L Street, N.W.

South Windsor, CT 06074

. ashington, DC 20005 W

i

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

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Docket No.(s)50-336-OLA l

NOTE CCMN RE NU RESPONSE 8/7..

3 3

I

)

Mary Ellen Marucci Michael J. Pray, AIA 104 Brownell Street 87 Blinman Street New Haven, CT 06511 New London CT 06320 DatedatRockvillgg2M' 27 day of August AM E

i Me of the Secretary of the commission I

l' r

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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION in the Mattar of NORTHEAST NUCLEAR ENERGY COMPANY Docket No.(s) 50-336-OLA (Millstone Nuclear Pow a Station, Unit No. 2)

CERTIFICATE OF SERVICE I hereby certify that copies of the foregoing RE-SERVED--NOTE CCMN TO NV..*

have been served upon the following persons by U.S. mail, first class, except as otherwise noted and in accordance with the requirements of 10 CFR Sec. 2.712.

Office of Commission Appellate Administrative Judge Adjudication Ivan W. Smith, Chairman U.S. Nuclear Regulatory Commission Atomic Safety and-Licensing Board Washington, DC 20555.

U.S. Nuclear Regulatory Commission Washington, DC 20555 Administrative Judge Administrative Judge Charles N. Kelber Jerry R. Kline Atomic Safety and Licensing Board Atomic Safety and Licensing Peard-U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555-John T. Hull, Esq.

Richard M. Kacich Office of the General Counsel Director, Nuclear Licensing U.S. Nuclear Regulatory Commission Northeast Utilities Washington, DC 20555 P. O. Box 270 Hartford, CT 06101 Patricia R. Nowicki

• Mitzi Sc. Bowman Associate Director Coordinator EARTHVISION, INC.

DON'T WASTE CONNECTICUT 42 Highland Drive 97 Longhill Terrace South Windsor, CT 06074 New Haven, CT 06515 t

4 Docket No.(s)50-336-OLA RE-SERVED--NOTE CCMN TO NV..*

Nicholas S. Reynolds. Esq.

John A. MacEvoy, Esq.

Mary Ellen Marucci Winston & Strawn 104 Brownell Street 1400 L Street, N.W.

New Haven, CT 06511 Washington, DC 20005 Michael J. Pray, AIA

• Frank X. Lo Sacco 87 Blinman Street 4 Glover Place, Box 1125 New London, CT 06320 Middletown, CT 06457

• Joseph M. Sullivan

• Rosemary Griffiths 17 Laurel Street 39 South Street Waterford, CT 06385 Niantic, CT 06357 Dated at Rockville, Md. this 28 day of August 1992 Office of the Secretary of the CommissT6n