Text

Redacted- License Renewal Application of the University of Wisconsin Nuclear Reactor; License R-74, Docket 20-156
	ML100740573
Person / Time
Site:	University of Wisconsin
Issue date:	10/17/2008
From:	Agasie R J; Univ of Wisconsin - Madison
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
	COWDREY, CHRIS, NRR/DPR/PRTA 415-2758
References
	TAC ME1585
	Download: ML100740573 (5)
	v • d • e

Nuclear Reactor Laboratory UWNR University of Wisconsin-Madison University Avenue, Room ME, Madison, Wl Tel:

FAX: email: reactor@engr.wisc.edu, http://reactor.engr.wisc.edu October 17, 2008 RSC 981 Director, Nuclear Reactor Regulations ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

License Renewal of the University of Wisconsin Nuclear Reactor; License R-74, Docket 50-156

Dear Sir:

The attached information is submitted in support of our application for license renewal dated May 9, 2000, as supplemented on May 6, 2004 and September 2, 2004. As a result of the extensive changes, a complete replacement copy of the Safety Analysis Report for Renewal of License R-74 for the University of Wisconsin Nuclear Reactor is enclosed.The changes described in the attachment are indicated by a vertical line in the margin of the enclosed revision 2 to the Safety Analysis Report for Renewal of License R-74 for the University of Wisconsin Nuclear Reactor, dated September 2008.In accordance with 10 CFR 50.30(b), I declare under penalty of perjury that the foregoing is true and correct.If you have any further questions, please contact me at (608) 262-3392.Robert J. ggas/e Reactor Director Attachment Enclosure cc: Daniel Hughes, NRC Project ManagerR-K UWNR Nuclear Reactor Laboratory University of Wisconsin-Madison 1513 University Avenue, Room 1215 ME, Madison, WI 53706-1687, Tel: (608) 262-3392, FAX: (608) 262-8590 email: reactor@engr.wisc.edu, http://reactor.engr.wisc.edu October 17, 2008 Director, Nuclear Reactor ATTN: Document Control Desk u.s. Nuclear Regulatory Commission Washington, DC 20555 RSC 981

Subject:

License Renewal of the University of Wisconsin Nuclear Reactor; License R-74, Docket 50-156

Dear Sir:

The attached information is submitted in support of our application for license renewal dated May 9, 2000, as supplemented on May 6, 2004 and September 2, 2004. As a result. of the extensive:

changes, a complete replacement copy of the Safety Analysis Report for Renewal of License R-74 for the University of Wisconsin Nuclear Reactor is enclosed.

The changes described in the attachment are indicated by a vertical line the margin of the enclosed revision 2 to the Safety Analysis Report for Renewal of License R-74 for the University of Wisconsin Nuclear Reactor, dated September 2008. In accordance with 10 CFR 50.30(b), I declare under penalty of perjury that the foregoing is true and correct. If you have any further questions, please,contact me at (608) 262-3392. Robert Reactor Director Attachment Enclosure cc: Daniel Hughes, NRC Project Manager SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference Chapter 1.2 1-1. Fixed typo.1.3 1-2 Updated excess reactivity (4.9 to 4.3) and shutdown margin (4 to 4.2) to reflect current core measurements.

1.3 1-2 Changed "TRIGA-FLIP High" to "TRIGA-FLIP," fixing typo.

1.3 1-3 Fixed typos.1.3 1-6 Expanded "STD fuel" to "TRIGA Standard fuel" for clarification.

1.4 1-6 Eliminated mention of sub-critical assembly and hotcell, since the hotcell is nowcontained within the restricted area and the subcritical assembly has been removed.

1.4 1-6 Corrected facilities wording since now all HVAC is non-shared use except offices.1.6 1-7 Removed sentence about iodine and strontium-90 inventory limits; iodine limits arecovered in Technical Specification 14.3.8.2, and limits on all fission products are covered in section 11.1.1.2, page 11-3, which states that all activity must be below 10 CFR 20 limits for effluent when assuming a months dilution with the ventilation system.1.8 1-8 Corrected initial criticality date from January to March 1961. It is not clear if anything significant occurred on January but it was not the initial criticality date as reported in chapter 4 of the SAR and as printed on the dedication plague displayed in the control room.

1.8- 1-8 Clarified that the initial replacement of 9 Standard fuel bundles with FLIP was for a mixed core.1.8 1-8 Fixed typo by adding "was initially critical in March 1974".1.8 1-8 Updated current MW Hours to over 20,000 (20,030 as of 9/16/2008) and added units of MWd.1.8 1-8 Updated most recent amendment number to 16 dated August 30, 2006.1.8 1-8,9 Updated facility modification history.SAR Revision 2 changes Page 1 of 14 Section Rev2 Page Change Reference Chapter 1 1.2 1-1 Fixed typo. 1.3 1-2 excess reactivity (4.9 to 4.3) and shutdown margin (4 to 4.2) to reflect current core measurements.

1.3 1-2 Changed "TRIGA-FLIP High" to "TRIGA-FLIP," fixing typo. 1.3 1-3 Fixed typos. 1.3 1-6 Expanded "STD fuel" to "TRIGA Standard fuel" for clarification.

J.4 1-6 Eliminated mention of sub-critical assembly and hotcell, since the hotcell is now -contained within the restricted area and the sub critical assembly has been removed. 1.4 1-6 Corrected facilities wording since now all HV AC is non-shared use except offices. 1.6 1-7 Removed sentence about iodirie and strontium-90 inventory limits; iodine limits are covered in Technical Specification 14.3.8.2, and limits on all fission products are covered in section 11.1.1.2, page 11-3, which states that all activity must be below 10 CFR 20 limits for effluent when assuming a months dilution with the ventilation

-system. 1.8 1-8 Corrected initial criticality date from January 5 tn to March 26 tn 1961. It is not clear if anything significant occurred on January 5 th , but it was not the initial criticality date as reported in chapter 4 of the SAR and as printed on the dedication plague displayed in the control room. 1.8-1-8 Clarified that the initial replacement of 9 Standard fuel bundles with FLIP was for a mixed core. 1.8 1-8 Fixed typo by adding "was initially critical in March 1974". 1.8 1-8 Updated current MW Hours to over 20,000 (20,030 as of 9/16/2008) and added units ofMWd. --1.8 1-8 Updated most recent amendment number to 16 dated August 30,2006. 1.8 1-8,9 Updated facility modification history. ---

SAR Revision 2 changes Page 2 of 14 Section Rev2 Page Change Reference Chapter 2 2.1.1.1 2-1 Updated Madison population with 2000 census data (changed from 191,000 to 208,000).2.1.1.1 2-1 Added clarification that 130m distance to nearest residence is 80m from building wall to avoid confusion with LEU report analysis for ground release.2.1.1.2 2-1 Changed reference from Figure 2-4 to 2-5, typo.2.1.1.2 2-1 Updated reactor lab room number from 130 to 1215, and clarified the site boundary 2.1.1.2 2-1 Added reference to ME floor maps for and floors. Also added reference to new building cross section figures.2.1.2 2-2 Updated to 2000 census data. New reference has replaced original reference

1. Response to NRC comment #4 2.1.2 2-2 Updated ME building occupancy estimate.

The initial estimate was 1182 assuming all lecture halls, computer labs, and offices were fully occupied, but this does not account for laboratories (such as engine labs), or hallways/lobbies (such as with Engineering Expo). Therefore the initial estimate was rounded up to 1500. The previous value was 300.2.1.2 2-2 Updated current Camp Randal capacity from 76,129 to 80,321. Source: UW Badger Athletics website (http://www.uwbadgers.com/facilities/camprandall/index_38.html).

Figure 2-3 2-3 Updated Madison city map using Microsoft Virtual Earth. Distance circles had been requested by NRC.Figure 2-4 2-4 Updated campus map using www.map.wisc.edu.

Figure 2-5 2-5 Updated engineering campus map using www.map.wisc.edu.

Figure 2-6 2-6 Updated ME basement map.Figure 2-7 2-7 Updated ME first floor map.Figure 2-8 2-8 Updated ME second floor map.

Figure 2-9 2-9 Updated ME third floor map.

Figure 2-10 2-10 Added new ME fourth floor map.SAR Revision 2 changes Page 2 of 14 Section Rev2 Page Change Reference Chapter 2* 2.1.1.1 2-1 Updated Madison population with 2000 census data (changed from 191,000 to 208,000).

2.1.1.1 2-1 Added clarification that 130m distance to nearest residence is 80m from building wall to avoid confusion with LEU report analysis for ground release. 2.1.1.2 2-1 Changed reference from Figure 2-4 to 2-5, typo. 2.1.1.2 2-1 Updated reactor lab room number from 130 to 1215, and clarified the site boundary definition.

2.1.1.2 2-1 Added reference to ME floor maps for 4 tn and 5 tn floors. Also added reference to new building cross section figures. 2.1.2 2-2 Updated to 2000 census data. Newreference has replaced original reference

1. Response to NRC comment #4 (I) 2.1.2 2-2 Updated ME building occupancy estimate.

The initial estimate was 1182 assuming all lecture halls, computer labs, and offices were fully occupied, but this does not account for laboratories (such as engine labs), or hallways/lobbies (such as with Engineering Expo ). Therefore the initial estimate was rounded up to 1500. The previous value was 300. 2.1.2 2-2 . Updated current Camp Randal capacity from 76,129 to 80,321. Source: UW Badger Athletics website (http://www.uwbadgers.comlfacilities/camp randall/index 38.html).

Figure 2-3 2-3 Updated Madison city map using Microsoft Virtual Earth. Distance circleshad been requested by NRC. Figure 2-4 2-4 Updated campus map using www.map.wisc.edu.

Figure 2-5 2-5 Updated engineering campus map using www.map.wisc.edu. Figure 2-6 2-6 Updated ME basement map. Figure 2-7 2-7 Updated ME first floor map. Figure 2-8 2-8 Updated ME second floor map. Figure 2-9 2-9 Updated ME third floor map. Figure 2-10 2-10 Added new ME fourth floor map.

SAR Revision 2 changes Page 3 Section Rev2 Page Change Reference Figure 2-11 2-11 Added new ME fifth floor map.Figure 2-12 2-12 Added new ME cross-section map looking north.Figure 2-13 2-13 Added new ME cross-section map looking east.Table 2-1 2-14 Updated with 2000 census data. Corresponding reference 1 has also been updated. Response to NRC comment #4 2.2.1 2-14 Updated National Guard annual flight missions from 1999 figure of approximately 3000 to 2006 figure of approximately 4000 (exact number is 8130 events).2.2.2 2-15 Updated 1999 traffic statistics with 2006 numbers for total events and commercial/general/military percentages.

Also corrected typo with runway headings (3100 should have been 3200).Figure 2-12 2-22 Updated Madison Well figure and made it full-page to make it easier to see.2.6 2-26 Updated reference

1. Response to NRC comment #4()2.6 2-26 Replaced references 3 and 4 (national guard and airport telephone interviews) with new reference 3 (email interview with airport).

All later references wererenumbered and the corresponding footnote labeling in the body of the chapter was also renumbered, but these renumbering changes were not redlined.Chapter 3.1 3-1 Added history of ME construction, new ventilation system and cooling system, design criteria.3.2 3-1 Updated 40 year history to 50.Chapter 4 4.1.1 4-1 Updated excess reactivity (4.2 to 4.3) and shutdown margin (4 to 4.2) to reflect current core measurements.

4.1.2 4-3 Fixed typos.4.1.2 Updated excess reactivity (4 to 4.3) and reactivity in control blades (6.9 to 7.1) to reflect current core measurements.

SAR Revision 2 changes . Page 3 of 14 Section Rev2 Page Change Reference Figure 2-11 2-11 Added new ME fifth floor map. Figure 2-12 2:..12 Added new ME cross-section map looking north. Figure 2-13 2-13 Added new ME cross-section map looking east. Table 2-1 2-14 Updated with 2000 census data. Corresponding reference 1 has also been updated. Response to NRC comment #4 (1) 2.2.1 2-14 Updated National Guard annual flight missions from 1999 figure of approximately 3000to 2006 figure of approximately 4000 (exact number is 8130 events). 2.2.2 2-15 Updated 1999 traffic statistics with 2006 numbers for total events and commercial/general/military percentages.

Also corrected typo with runway headings (310° should have been 320°). Figure 2-12 2-22 Updated Madison Well figure and made it full-page to make it easier to see. 2.6 2-26 Updated reference

1. Response to NRC comment #4(1) 2.6 2-26 Replaced references 3 and 4 (national guard and airport telephone interviews) with new reference 3 (email interview with airport).

All later references were renumbered and the corresponding footnote labeling in the body of the chapter was also renumbered, but these renumbering changes were not redlined.

-Chapter 3 3.1 3-1 Added history of ME construction, new ventilation system and cooling system, design criteria.

3.2 3-1 Updated40 year history to 50.

4.1.1 4-1 Updated excess reactivjty (4.2 to 4.3) and shutdown margin (4 to 4.2) to reflect current core measurements.

4.1.2 4-3 Fixed typos. 4.l.2 4:'5 Updated excess reactivity (4 to 4.3) and reactivity in control blades (6.9 to 7.1) to reflect current core measurements.

SAR Revision 2 changes Page 4 of 14 Section Rev2 Page Change Reference 4.1.3.# 4-5,6 Added outline numbering.

4.2 4-7 Added reference to 10 CFR 73.2 for definition of formula quantity HEU.4.2.1 4-8 Removed reference to aluminum irradiation thimble in 3-element fuel bundle assemblies, as they are no longer used nor is there any plan to use them in the future.4.2.1 4-8 Corrected Figure 4-2 reference, fixing typo. Response to NRC comment #7 4.2.2.# 4-13,14,15 Added outline numbering.

4.2.2.1 4-13 Corrected Figure 4-2 reference, fixing typo. Response to NRC comment #7 4.2.2.2 4-14 Clarified wording of safety blade section.4.2.5 4-20 Fixed typo.4.2.5 4-23 Added reference to Table 4-1 and 4-2 to explain codes in Figure 4-15.4.3 4-26 Updated 40 year history to 50.4.4 4-26 Included reference to section 13.1.3.2 for calculation of dose rates and integrated dose for floor occupant.

Stated that with evacuation, public dose would be less than 100 mrem (about 13 mrem).4.4.# 4-27 Added outline numbering.

4.5.1 4-27 Sections 4.5.1 and 4.5.2 were already combined, but outline numbering was revised to indicate section 4.5.1 rather than 4.5.1/2 for clarity.4.5.1.# 4-28,29,32, Added outline numbering.

34,37,40, 44,47,48 4.5.1.2.1 4-32 Fixed typo (reference to Figure 4-15)4.5.1.3.1 4-37 Fixed typo.4.5.1.3.2 4-43 Fixed typo.4.5.1.3.3 4-44,46,47 Fixed typos.4.5.2 4-51 Updated numbering from 4.5.3 to 4.5.2 (due to combination of previous sections) SAR Revision 2 changes Page 4 of 14 Section Rev2 Reference 4.1.3.# 4-5,6 Added outline numbering.

4.2 4-7 Added reference to 10 CFR 73.2 for-definition of formula quantity HEU. 4.2.1 4-8 Removed reference to aluminum irradiation thimble in 3-element fuel bundle assemblies, as they are no longer used nor is there any plan to use them in the future. 4.2.1 4-8 Corrected Figure 4-2 reference, fixing typo. Response to NRC , comment #7 (\) , 4.2.2.# 4-13,14,15 Added outline numbering.

4.2.2.1 4-13 Corrected Figure 4-2 reference, fixing typo. Response to NRC comment #7 (I) 4.2.2.2 4-14 Clarified wording of safety blade section. 4.2.5 4-20 Fixed typo. 4.2.5 4-23 Added reference to Table 4-1 and 4-2 to explain codes in Figure 4-15. 4.3 4-26 Updated 40 year history to 50. 4.4 4-26 Included reference to section 13.1.3.2 for calculation of dose rates and integrated dose for 3 rd floor occupant.

Stated that with evacuation, public dose would be less (' than 100 mrem (about 13 mrem). 4.4.# 4-27 Added outline numbering.

t 4.5.1 4-27 Sections 4.5.1 and 4.5.2 were already combined, but outline numbering was revised to indicate section 4.5.1 rather than 4.5.112 for clarity. 4.5.1.# 4-28,29,32, Added outline numbering.

34,37,40, , 44,47,48 4.5.1.2.1 4-32 Fixed typo (reference to Figure 4-15) 4.5.1.3.1 4-37 Fixed typo. 4.5.1.3.2 4-43 Fixed typo. 4.5.1.3.3 4-44,46,47 Fixed typos. 4.5.2 4-51 Updated numbering from 4.5.3 to 4.5.2 (due to combination of previous sections)

SAR Revision 2 changesPae5ol Page of 145-3,4 Split sections 5.5 and (water cleanup and makeup sections) for clarity. Also added description of waste system to section Previous Figure has also been updated and split into Figures 5-3, 5-4, and to separate pooi, cleanup, makeup, and waste systems.5-3,4 Updated flow rate and valve numbers for new demineralizer gpm, valve service out, valve 22).Figure Added new pool water systems schematic.

Figure Added new water cleanup system schematic.

Figure 5-4 Added new water waste system schematic.

Figure Added new water makeup system schematic.Updated reference to figures to account for added floor plan.Changed word roof to ceiling to clarify that the confinement ceiling is no longer thebuilding roof.

6.2.1 Eliminated mention of the console air conditioner and updated description to reflect new ventilation system.

Updated lab description to reflect construction changes, elimination of windows, change of doors, etc. Security details of doors and window are not specified due to Changed name of basement area *from "Nuclear Engineering Laboratory" to"Reactor Laboratory auxiliary support space." Deleted paragraph about future plans for room on north of lab.Corrected typo in outline numbering; should have been Figure Updated lab basement floor plan.

Figure 6-4 Added new lab first floor plan.Figure Updated diagram of lab facing south.

Figure 6-4 Updated diagram of lab facing north.________

Figure Updated diagram of lab facing east.

SAR Revision 2 changes Page 5 of 14 Section Rev2 Page Change Reference Chapter 5 5.5 5-3,4 Split sections 5.5 and 5.6 (water cleanup an4 makeup sections) for clarity. Also added description of waste system to section 5.5. Previous Figure 5-2 has also been updated and split into Figures 5-2, 5-3, 5-4, and 5-5 to separate pool, cleanup, makeup, and waste systems. 5.5 5-3,4 Updated flow rate and valve numbers for new demineralizer (18 gpm, valve 10102 service out, valve 22). Figure 5-2 5-5 Added new pool water systems schematic.

Figure 5-3 5-6 Added new water cleanup system schematic.

Figure 5-4 5-7 Added new water waste system schematic.

Figure 5-5 5-8 Added new water makeup system schematic.

' . Chapter 6. 6.2.1 6-1 Updated reference to figures to account for added 1 51 floor plan. 6.2.1 6-1 Changed word roof to ceiling to clarify that the confinement ceiling is no longer the building roof. 6.2.1 6-1 Eliminated mention ofthe console air conditioner and updated description to reflect new ventilation system. 6.2.1 6-1,2 Updated lab description to reflect construction changes, elimination of windows, change of doors, etc. Security details of doors and window are not specified due to SOL Changed name of basement area .from "Nuclear Engineering Laboratory" to "Reactor Laboratory auxiliary support space." Deleted paragraph about future plans for room on north of lab. 6.2.2 6-'2 Corrected typo in outline numbering; 6.2.3 should have been 6.2.2. Figure 6-1 6-3 Updated lab basement floor plan. Figure 6-2 6-4 Added new lab first floor plan. Figure 6-3 6-5 Updated diagram of lab facing south. Figure 6-4 6-6 Updated diagram of lab facing north. Figure 6-5 6-7 Updated diagram of lab facing east.'

SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference Figure 6-6 6-8 Updated diagram of lab facing west.Chapter 7 7.1 7-1 Updated to include digital fuel temp, computer pulse, all digital recorder.

Figure 7-1 7-2 Updated figure to reflect console upgrade. RSC RSC 803 (3)RSC 887 7.2.3.# 7-3,4 Added outline numbering.

7.2.3 7-3 Corrected references to bistables which should be relays.7.2.3.3 7-3 Clarified pulse output on console computer to reflect new pulse channel.7.2.3.5 7-4 Clarified console recorder.7.2.4 7-4 Updated history to 50 years.7.2.5 7-5 Removed reference to LogN not in operate scram which no longer exists. RSC 887 7-5,6,8,9 Added outline numbering.

7.3.3 7-5 Corrected typo, "PULSE" position to "SQUARE WAVE" position.7.3.4 7-6 Clarified pulse output on console computer to reflecf new pulse channel.7.4 7-10 Removed reference to LogN not in operate scram which was removed. RSC 887 (41 7.4 7-10 Clarified that pool level alarm is on high or low.Figure 7-2 7-11 Updated per console upgrade. RSC 887 (4)Figure 7-3 7-12 Updated per console upgrade. RSC 887 (4, 7.6.# 7-13,14,15 Added outline numbering.

7.6.1 7-13 Added note that area radiation level high alarms at UWPD and initiates evacuation.

7.6.1 7-13 Added Evacuation Alarm in Local annunciator.

RSC 893 (5)7.6.1 7-13 Updated SAM/CAM trouble annunciator.

RSC 896 7.6.1 7-13 Added Loss of Off-Site Power annunciator.

RSC 856 (7)7.6.1 7-14 Removed pn blower annunciator.

RSC 857 7.6.1 7-14 Updated to reflect UWPD name change.7.6.3 7-14 Updated pneumatic system panel description.

RSC 857 7 7.6.4 7-15 Updated ventilation system panel description, including new BP&TC EF-17. RSC 879 SAR Revision 2 changes Page 6 of 14 Section Rev2 Page Change Reference Figure 6-6 6-8 Updated diagram of lab facing west. Chapter 7 7.1 7-1 Updated to include digital fuel temp, computer pulse, all digital recorder.

Figure 7-1 7-2 Updated figure to reflect console upgrade. RSC 773 (L) RSC 803 (3) RSC'887 (4) 7.2.3.# 7-3,4 Added outline numbering.

7.2.3 7-3 Corrected references to bistables which should be relays. 7.2.3.3 7-3 Clarified pulse output on console computer to reflect new pulse channel. 7.2.3.5 7-4 Clarified console recorder.

7.2.4 7-4 Updated history to 50 years. 7.2.5 7-5 Removed reference to LogN not in operate scram which no longer exists. RSC 887 (4) 7.3.# 7-5,6,8,9 Added outline numbering.

7.3.3 7-5 Corrected typo, "PULSE" position to "SQUARE WAVE" position.

7.3.4 7-6 Clarified pulse output on console computer to reflecfnew pulse channel. 7.4 7-10 Removed reference to LogN not in operate scram which was removed. RSC 887 (4) 7.4 . 7-10 Clarified that pool level alarm is on high or low. Figure 7-2 7-11 Updated per console upgrade. RSC 887 (4) Figure 7-3 7-12 Updated per console upgrade. RSC 887 (4) 7.6.# 7-13,14,15 Added outline numbering.

7.6.1 7-13 Added note that area radiation level high alarms at UWPD and initiates evacuation.

7.6.1 7-13 Added Evacuation Alarm in Local annunciator.

RSC 893 (:l) 7.6.1 7-13 Updated SAM/CAM trouble annunciator.

RSC 896 (b) 7.6.1 7-13 Added Loss of Off-Site Power annunciator.

RSC 856 (I) 7.6.1 7-14 Removed pn blower annunciator.

RSC 857 (IS) 7.6.1 7-14 Updated to reflect UWPD name change. 7.6.3 7-14 Updated pneumatic system panel description.

RSC 857 (IS) 7.6.4 7-15 Updated ventilation system panel description, including new BP&TC EF-17. RSC 879 ('1)

SAR Revision 2 changes Page of 14Section Rev2 Page Change Reference 7.6.5 7-15 Added reference to chapter 5 for details of cooling system.7.6.6 7-15 Added whale system panel description for consistency.

7-15,16 Added outline numbering.

7.7.2 7-16 Updated stack air monitor description.

RSC 896 Chapter 8 All This entire chapter has been rewritten to reflect changes due to ME building construction, going into greater detail than previously. Also added electrical drawings.8.2 8-3 Added paragraphs describing UPS per facility mod, but also expanded the last RSC 895 0 sentence to reiterate that the UPS is not required "for maintaining the facility in safe shutdown, even for extended periods of time." Typo in RSC document referring to 6000 kVA has been corrected to 6000 VA.Chapter 9.1 9-1,2,3,4 Rewrote ventilation system description.

This is a complete rewrite which is based RSC 879 (9)on the RSC description.9-1,2,3,4 Added outline numbering.

Figure 9-1 9-5 Updated for new vent system. RSC 879 (9)Figure 9-2 9-6 Added new figure 9-2 for cross-section ventilation layout.Figure 9-3 9-7 Added new figure 9-3 for cross-section ventilation layout.Figure 9-4 9-8 Revised fuel rack diagram to reflect actual locations, and added North legend.9.2.2 9-9 Deleted potentially sensitive information on un-irradiated fuel storage.9.2.3 9-10 Corrected typo.

9.2.3 9-10 Added new dummy element with reduction AND increase in diameter.

RSC 869 RSC 919 9.3 9-13 Updated fire systems. RSC 894 (')9.3 9-13 Updated P&S to UWPD notation.9.4 9-13 Changed wording to auxiliary support space for consistency.

9.4 9-13 Removed description of 2intercom system which was uninstalled.

SAR Revision 2 changes Page 7 of 14 Section Rev2 Page Change Reference 7.6.5 7-15 Added reference to chapter 5 for details of cooling system. 7.6.6 7-15 Added whale system panel description for consistency.

7.7.# 7-15,16 Added outline numbering.

7.7.2 7-16 Updated stack air monitor description.

RSC 896 to) Chapter 8 All This entire chapter has been rewritten to reflect changes due to ME building construction, going into greater detail than previously.

Also added electrical drawings.

8.2 8-3 Added paragraphs describing UPS per facility mod, but also expanded the last RSC 895 (IU) sentence to reiterate that the UPS is not required "for maintaining-the facility in safe shutdown, even for extended periods of time." Typo in RSC document referring to 6000 k V A has been corrected to 6000 V A. Chapter 9 9.1 9-1,2,3,4 Rewrote ventilation system description.

This is a complete rewrite which is based RSC 879 on the RSC description.

9.1.# 9-1,2,3,4 Added outline numbering.

Figure 9-1 9-5 Updated for new vent system. RSC 879 Figure 9-2 9-6 Added new figure 9-2 for cross-section ventilation layout. Figure 9-3 9-7 Added new figure 9-3 for cross-section ventilation layout. Figure 9-4 9-8 Revised fuel rack diagram to reflect actual locations, and added North legend. 9.2.2 9-9 Deleted potentially sensitive information on un-irradiated fuel storage. 9.2.3 9-10 Corrected typo. 9.2.3 9-10 Added new dummy element with reduction AND increase in diameter.

RSC 869 (II) RSC 919 (12) 9.3 9-13 Updated fire RSC 894 t1j) 9.3 9-13 Updated P&S to UWPD notation.

9.4 9-13 Changed wording to auxiliary support space for consistency.

9.4 9-13 Removed description of 2 na intercom system which was uninstalled.

SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference 9.5 9-13 Removed old rooms, updated lab room number to 1215, basement area roomnumber to B 1215, and added B 9-14 Updated Agreement State license number.Chapter 10.2.# 10-1,3,6,8 Added outline numbering.

10.2.2 10-3 Revised "beams" to "beams of radiation" for clarification.

Figure 10-2 10-3 Updated Figure 10-2 to include core box and actually label TC and beam ports.10.2.3 10-6,7 Updated for new pneumatic system and modified to reflect the basement move. RSC 857 10.2.3 10-6 Clarified wording of activity limits. RSC 879 Figure 10-4 10-7 Updated for new pneumatic system. RSC 857 (10.2.4 10-8 Deleted description of old smaller hydraulic irradiation facility.10.3 10-11 Changed "if released with 30 days of dilution" to "when averaged over 30 days of dilution".

10.3 10-11 Removed outdated reference to 130 vs 131 activity limits (they are now the same). RSC 879 10.3 10-11 Changed 1 SRO to 2 SROs to reflect new commitment to RSC. RSC Charter 10.3 10-11 Fixed gender language.Chapter 11 11-1 Updated Agreement State license number.11 11-1 Updated reference to radiation safety regulations to remove url (since this has changed and may change again).11.1.1.1 1 1-1,2 Added outline numbering.

11.1.1.1.2 11-2 Updated Ar-41 release calculations for new vent system. RSC 879 11.1.1.1.2 11-3 Deleted redundant sentence that volatile or powder would be of importance.11.1.1.1.2 11-3,4 Updated dilution assumptions for new vent system. RSC 879 11.1.1.1.2 11-3 Updated descriptions for single fume hood.11.1.1.1.2 11-4 Clarified wording on RSC approval of non-routine samples.

RSC 879 11.1.2/3 11-5 Added note that these 2 sections are combined.SAR Revision 2 changes Page 8 of 14 Section Rev2 Page Change Reference 9.5 9-13 Removed old rooms, updated lab room number to 1215, basement area room . number to B1215, and added BI135. 9-14 Updated Agreement State license number. Chapter 10 10.2.# 10-1,3,6,8 Added outline numbering.

10.2.2 10-3 Revised "beams" to "beams of radiation" for clarification.

Figure 10-2 10-3 Updated Figure 10-2 to include core box and actually label TC and beam ports. 10.2.3 10-6,7 Updated for new pneumatic system and modified to reflect the basement move. RSC 857 (K) 10.2.3 10-6 Clarified wording of activity limits. RSC 879 ('.I) Figure 10-4 10-7 Updated for new pneumatic system. RSC 857 (K) 10.2.4 10-8 Deleted description of old smaller hydraulic irradiation facility.

10.3 10-11 Changed "if released with 30 days of dilution" to "when averaged over 30 days of dilution" . 10.3 10-11 Removed outdated reference to 130 vs 131 activity limits (they are now the same). RSC 879 10.3 10-11 Changed 1 SROto 2 SROs to reflect new commitment to RSC. RSC Charter (14) 10.3 10-11 Fixed gender language.

Chapter 11 11 11-1 Updated Agreement State license number. 11 11-1 Updated reference to radiation safety regulations to remove uri (since this has changed and may change again). 11.1.1.1.#

11-1,2 Added outline numbering.

11.1.1.1.2 11-2 Updated Ar-41 release calculations for new vent system. RSC 879 ('.I) 11.1.1.1.2 11-3 Deleted redundant sentence that volatile or powder would be of importance.

11.1.1.1.2 11-3,4 Updated dilution assumptions for new vent system. RSC 879 ('.I) 11.1.1.1.2 11-3 Updated descriptions for single fume hood. 11.1.1.1.2 11-4 Clarified wording on RSC approval of non-routine samples. RSC .879 ('.I) 11.1.2/3 11-5 Added note that these 2 sections are combined.

SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference 11.1.5 11-5 Changed students to experimenters to reflect that we will no longer TLD badge all RSC 922 (students.11.1.5 11-5,6 Paragraph about high radiation levels in experiments was revised to reference accesscontrol and postings in accordance with 10 CFR 20.1601.11.1.5 11-6 Tour/visitor dose paragraph was revised to reflect current use of non-radiation worker classification (based on University Radiation Safety Regulations).

11.2.2 11-7 Added radioactive sink to list of liquid wastes to holdup tank.

11.2.2 11-7 Included parenthetical reference to short description of waste holdup tank in section 5.5.11.2.3 11-7 The filter size for liquid waste discharge was supposed to be 0.5 micron (0.4 micron was a typo). The UWNR has always used 0.5 micron filters, as approved by the Reactor Safety Committee.

10 CFR 20.2003 says that waste must be readily soluble. NUREG-1556 Vol. 7 provides clarification of "readily soluble" by referencing Information Notice 94-07, but there is no guidance on what size filters to use to remove any potential dissolved solids.

UW Radiation Safety was consulted and stated that 0.5 microns is the common filter size used to remove any possible dissolved solids in order to comply with 10 CFR 20.2003.11.3 11-7 Removed website location from reference 1, since this has changed and may change again in the future.Chapter 12 12.1.2.# 12-1,2,3 Added outline numbering.

12.1.2.2 12-1 Updated department name that Radiation Safety is part of (now University Department of Environment, Health and Safety).12.1.2.3 12-1 Added responsibilities of Department Chair. ANS 15.4 provides very little detail for any of the levels. Level 1 just says "Individual responsible for the reactor facility's licenses or charter." In our case the chair is also responsible for appointing the reactor director.12.1.2.6 12-3 Fixed gender language.SAR Revision 2 changes Page 9 of 14 Section* Rev2 Page Change Reference 11.1.5 11-5 Changed students to experimenters to reflect that we will no longer TLD badge all RSC 922 (D) students.

11.1.5 11-5,6 Paragraph about high radiation levels in experiments was revised to reference access control and po stings in accordance with 10 CFR 20.1601. 11.1.5 I 11-6 Tour/visitor dose paragraph was revised to reflect current use of non-radiation worker classification (based on University Radiation Safety Regulations).

11.2.2 11-7 Added radioactive sink to list of liquid wastes to holdup tank. 11.2.2 11-7 Included ,parenthetical reference to short description of waste holdup tank in section 5.5. 11.2.3 11-7 The filter size for liquid waste discharge was supposed to be 0.5 micron (0.4 micron was a typo). The UWNR has always used 0.5 micron filters, as approved by the Reactor Safety Committee.

10 CFR 20.2003 says that waste must be readily soluble. NUREG-1556 Vol. 7provides clarification of "readily soluble" by referencing Information Notice 94-07, but there is no guidance on what size filters to use to remove any potential dissolved solids. UW Radiation Safety was consulted and stated that 0.5 microns is the common filter size used to remove any possible -dissolved solids in order to comply with 10 CFR 20.2003. 11.3 11-7 Removed website location from reference 1, since this has changed and may change . again in the future. Chapter 12 12.1.2.# 12-1,2,3 Added outline numbering.

12.1.2.2 12-1 Updated department name that Radiation Safety is part of (now University Department of Environment, Health and Safety). 12.1.2.3 12-1 Added responsibilities of Department Chair. ANS 15.4 provides very little detail for any of the levels. Levell just says "Individual responsible for the reactor facility's licenses or charter." In our case the chair is also responsible for appointing the reactor director.

12.1.2.6 12-3 Fixed gender language.

SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference 12.1.3 12-4 Corrected outline numbering typo.12.1.4 12-4,5 Revised description to reflect current Operator Candidacy Program. Also spelled out OJT as On the Job Training for clarification, and removed course url in case offuture change.

12.1.5 12-5 Updated department name that Radiation Safety is part of (now University Department of Environment, Health and Safety).12.2.1 12-7 Updated Radiation Safety name.12.2.3 12-7 Fixed gender language.12.2.4 12-8 Updated Radiation Safety name.12.3 12-9 Changed 1 SRO to 2 SROs required for temporary procedure changes, and fixedgender language.

12.4 12-10 Fixed gender language.Chapter 13.1 13-. Sections 13.1 and 13.2 were previously combined, but the outline numbering has 1,2,3,6,8-now been revised to eliminate 13.1/2 notation.

Also added note to clarify 12,14-16 combination of sections.13.1.1 .# 13-2,3,6 Added outline numbering.

13.1.1.4 13-3 Added clarification on assumed evacuation time (5 minutes to exit confinement, another 5 minutes to exit building).

Also added reference to Appendix A for clarification, and added sub-section numbering.

13.1.1.4 13-3 Revised self-contained breathing apparatus to powered air purifying respirator.

Table 13.1 13-4 Updated H and J values for new vent system. RSC 879 13.1.1.5 13-6,7 Updated for new vent system and made minor clarifications.

RSC 879 13.1.2.# 13-9,10 Added outline numbering.

13.1.3.# 13- Added outline numbering.

11,12,14 SAR Revision 2 changes Page 10 of 14 Section Rev2 Page Reference 12.1.3 12-4 Corrected outline numbering typo. 12.1.4 12-4,5 Revised description to reflect current Operator Candidacy Program. Also spelled out OJT as On the Job Training for clarification, and removed course urI in case of future change. 12.1.5 12-5 Updated department name that Radiation Safety is part of (now University Department of Environment, Health and Safety). 12.2.1 12-7 Updated Radiation Safety name. 12.2.3 12-7 Fixed gender language.

12.2.4 12-8 Updated Radiation Safety name. 12.3 12-9 Changed 1 SRO to 2 SROs required for temporary procedure changes, and fixed gender language.

12.4 12-10 Fixed gender language.

Chapter 13 13.1 13-. Sections 13.1 and 13.2 were previously combined, but the outline numbering has 1,2,3,6,8-now been revised to eliminate 13.112 notation.

Also added note to clarify 12,14-16 combination of sections.

13.1.1.# 13-2,3,6 Added outline numbering.

13.1.1.4 13-3 Added clarification on assumed evacuation time (5 minutes to exit confinement, another 5 minutes to exit building).

Also added reference to Appendix A for clarification, and added sub-section numbering.

13.1.1.4 13-3 Revised self-contained breathing apparatus to powered air purifying respirator.

Table 13.1 13-4 Updated H and J values for new vent system. RSC 879 l'l) 13.1.1.5 13-6,7 Updated for new vent system and made minor clarifications.

RSC 879 l'l) 13.1.2.# 13-9,10 Added outline numbering.

13.1.3.# 13-Added outline numbering. . 11,12,14 SAR Revision 2 changes Page of 14Section Rev2 Page Change Reference13.1.3.1 13-11 Updated pool drain time calculations.

There is no record of how the previous calculation was performed, but current calculation uses reasonable methods and assumptions with references.

13.1.3.2/3 13-12-14 Updated unshielded core dose calculations.

The calculated floor classroom dose-rates were significant, so an additional analysis was performed to model the integrated dose received to ensure it would be less than 1 00mrem during the evacuation. Integrated dose was calculated to be about 13 mrem. This section was also divided into. 13.1.3.2 for confinement doses, and 13.1.3.3 for unrestricted area doses.13.1.6 13-15 Fixed typo; missing section reference to 13.1.1 13.1.8 1345 Changed "incredible" to "not credible" for clarification.

13.1.9 13-16,17 Updated reference to Figure 2412 (was Figure 2.10) and deleted reference to drain thimble used for cooling tower blowdown (no longer exists). Also corrected typo referring to case 2 when it should be case 3.13.2/3 13-17 Updated section number from 13.3 to 13.2 andfrom 13.4 to 13.3 because of combination of sections 13.1/2 into 13.1.13.3 13-18 Added new reference 9 for the pool drain equation used in section 13.1.3.1 Chapter 14 All All outline numbering was revised to include a "TS" Prefix to avoid confusion with earlier chapters of the SAR.1.1 14-1 Updated summary to reflect elimination of Rev. 0 redlining.

1.3.2 14-5 Added new definition for non-secured experiment (just opposite of securedexperiment which was already defined).

2.2 14-10,11 Clarified between first and second LSSS in basis.3.1.2 13,14 Corrected control rod to control element.

3.2.5 14-22 Clarified wording for interlocks preventing application of air to fire transient rod.3.4 14-26 Renamed section heading to "Confinement." Response to NRC comment #11 SAR Revision 2 changes Page 11 of 14 Section Rev2 Page Change Reference 13.1.3.1 13-11 Updated pool drain time calculations.

There is no record of how the previous calculation was performed, but current calculation uses reasonable methods and assumptions with references.

13.1.3.2/3 13-12-14 Updated unshielded core dose calculations.

The calculated 3 ro floor classroom dose-rates were significant, so an additional analysis was performed to model the integrated dose received to ensure it would be less than 100mrem during the evacuation.

Integrated dose was calculated to be about 13 mrem. This section was also divided into. 13.1.3.2 for confinement doses, and 13.1.3.3 for unrestricted area doses. 13.1.6 13-15 Fixed typo; missing section reference to 13.1.1 13.1.8 13-'15 Changed "incredible" to "not credible" for clarification.

13.1.9 13-16,17 Updated reference to Figure 2-12 (was Figure 2.10) and deleted reference to drain thimble used for cooling tower blowdown (no longer exists). Also corrected typo referring to case 2 when it should be case 3. 13.2/3 13-17 Updated section number from 13.3 to 13.2 and from 13.4 to 13.3 because of combination of sections 13 .112 into 13.1. 13.3 13-18 Added new reference 9 for the pool drain equation used in section 13.1.3.1 Chapter 14 All All outline numbering was revised to include a "TS" Prefix to avoid confusion with earlier chapters of the SAR. 1.1 14-1 Updated summary to reflect elimination of Rev. 0 redlining.

1.3.2 14-5 Added new definition for non-secured experiment Gust opposite of secured experiment which was already defined). , 2.2 14-10,11 Clarified between first and second LSSS in basis. 3.1.2 14-13,14 Corrected control rod to control element. 3.2.5 14-22 Clarified wording for interlocks preventing application of air to fire transient rod. 3.4 14-26 Renamed section heading to "Confinement." Response to NRC comment #11 (I)

SAR Revision 2 changes Page 12 of 14 Section Rev2 Page Change Reference 3.4 14-27 Updated stack height from 17m to 26.5m, and corresponding reduction fraction from 10 to 2.6.3.7.2 14-29 Updated release concentration and dilution fraction for new vent system. RSC 879 3.8.1 14-30 Clarified limits on experiment reactivity are for. a single experiment.

3.8.2 14-31 Revised outline numbering for consistency.

3.8.2 14-31 Updated dilution fraction for new vent system. RSC 879 3.8.2 14-32 Changed formatting of basis list for clarification only.3.8.3 14-32 Clarified that the experimental capsule would be removed and inspected if failure occurs (fuel may be inspected if warranted, but would not be removed).4.2 14-36 Clarified wording.

Response to NRC comment # 14 {4.2 14-37 Updated 40 year history to 50 years.4.4 14-38 Renamed section heading to "Confinement." Response to NRC comment #11 )4.5 14-39 Revised the basis for quarterly checks on ventilation system.5.1 14-41 Changed stack height from 17m to 26.5m. RSC 879 5.2 14-41 Deleted outlet pipe from 15 foot requirement (outlet pipe requirement is specified in the next sentence).

5.3 14-42 Revised outline numbering for consistency.

6.1.1 14-45 Added reference to new Figure 14-1 (just copy of Figure 12-1).6.1.2 14-45 Fixed gender language.

Figure 14-1 14-46 Added new figure for clarification (copy of Figure 12-1).6.1.3 14-47 Clarified wording in sub-section 1-b.6.2.1 14-48 Revised list of required qualifications for RSC.6.3 14-49 Clarified authority of state license and updated license number.6.4 14-50 Fixed gender language.

6.5(2) 14-51 Fixed gender language and clarified wording.

6.6.2 14-52 Fixed gender language.6.7.1 14-53,54.

Reordered sub-sections under item 1 and revised outline numbering for consistency.

SAR Revision 2 changes Page 12 of 14 Section Rev2 Page Change Reference 3.4 14-27 Updated stack height from 17m to 26.5m, and corresponding reduction fraction from 10 to 2.6. 3.7.2 14-29 Updated release concentration and dilution fraction for new vent system. RSC 879 lV) 3.8.1 14-30 Clarified limits on experiment reactivity are fora single experiment.

3.8.2 14-31 Revised outline numbering for consistency.

3.8.2 14-31 Updated dilution fraction for new vent system. RSC 879 lV) 3.8.2 14-32 Changed formatting of basis list for clarification only. 3.8.3 14-32 Clarified that the experimental capsule would be removed and inspected if failure occurs (fuel may be inspected if warranted, but would not be removed).

4.2 14-36 Clarified wording. Response to NRC comment # 14 (I) 4.2 14-37 Updated 40 year history to 50 years. 4.4 14-38 Renamed section heading to "Confinement." Response to NRC comment #11 (\) 4.5 39 Revised the basis for quarterly checks on ventilation system. 5.1 14-41 Changed stack height from 17m to 26.5m. RSC 879 lV) 5.2 14-41 Deleted outlet pipe from 15 foot requirement (outlet pipe requirement is specified in the next sentence).

5.3 14-42 Revised outline numbering for consistency.

6.1.1 Added reference to new Figure 14-1 (just copy of Figure 12-1). 6.1.2 14-45 Fixed gender language.

Figure 14-1 14-46 Added new figure for clarification (copy of Figure 12-1). 6.1.3 14-47 Clarified wording in sub-section I-b. 6.2.1 14-48 Revised list of required qualifications for RSC. 6.3 14-49 Clarified authority of state license and updated license number. 6.4 14-50 Fixed gender language.

6.5(2) 14-51 Fixed gender language and clarified wording. 6.6.2 14-52 Fixed. gender language.

6.7.1 14-53,54 Reordered sub-sections under item 1 and revised outline numbering for consistency.

SAR Revision 2 changes Page of 14 Section Rev2 Page Change Reference Appendix Revised headings and order of entire Appendix to make it clear which sections of the SAR the calculations are supporting.Updated stack height from 17.1 to 26.5m. RSC 879 (9)Updated equation 2 to reflect new stack height. RSC 879 A-2 Updated equation 5 to reflect new building cross-section (12,200ft 2) RSC 879 (9)A-2 Updated equation 6 to reflect changes in equation 5. RSC 879 (9)A-2 Added number for equation 7 (previously not numbered).

A-2 Added number for equation 8 (previously not numbered).

A-2 Added number for equation 9 (previously not numbered).

A-3 Added number for equation .10 (previously not numbered).

A-3 Added number for equation 11 (previously not numbered).

A-3 Renumbered equation 9 to equation 12, and updated equation to reflect flow-rate RSC 879 (9)changing from 1000 scfm to 2700 scfm (also clarified the unit conversion).

A-3 Renumbered equation 10 to equation 13, and updated equation.to reflect changes in RSC 879 (9)equations 2 and 12.A-4 Renumbered equation 11 to equation 14, and updated equation to reflect changes in RSC 879 (9)equations 6 and 12.A-4 Updated factor from 10.17 to 2.6 because of new vent system. RSC 879 (9)A-4 Deleted and replaced sentence referring to building wake effects vs. reactor lab RSC 879 (9)wake effects being a 10 fold difference.

From now on all wake effects are calculated using the new ME building dimensions (12,200 ft2) since the reactor lab is entirely enclosed within the ME building.A-4 Clarified intermediate calculation of stack outlet concentration, and clarified that equation 15 (which uses equation 2) is assuming stack height.A-4 Renumbered equation 7 to equation 15, and updated equation to reflect changes in RSC 879 (9)equation 2.SAR Revision 2 changes Page 13 of 14 Section Rev2 ChanKe Reference Appendix A Revised headings and order of entire Appendix to make it clear which sections of the SAR the calculations are supporting.

A-I . Updated stack height from 17.1 to 26.5m. RSC . A-I Updated equation 2 to reflect new stack height. RSC A-2 Updated equation 5 to reflect new building cross-section (12,200ftL)

RSC A-2 Updated equation 6 to reflect changes in equation 5. RSC A-2 Added number for equation 7 (previously not numbered).

A-2 Added number for equation 8 (previously not numbered).

A-2 Added number for equation 9 (previously not numbered).

A-3 Added number for equation J 0 (previously not numbered). . A-3 Added number for equation 11 (previously not numbered) . A-3 Renumbered equation 9 to equation 12, and updated equation to reflect flow-rate RSC changing from 1000 scfm to 2700 scfm (also clarified the unit conversion).

A-3 Renumbered equation 10 to equation 13, and updated equation.to reflect changes in RSC equations 2 and 12. A-A Renumbered equation 11 to equation 14, and updated equation to reflect changes in RSC equations 6 and 12. A-4 Updated factor from 10.17 to 2.6 because of new vent system. RSC r-A-4 Deleted and replaced sentence referring to building wake effects vs. reactor lab RSC wake effects being a 10 fold difference.

From now on all wake effects are calculated using the new ME building dimensions (12,200 ft2) since the reactor lab is entirely enclosed within the ME building.

A-4 Clarified intermediate calculation of stack outlet concentration, and clarified that equation 15 (which uses equation 2) is assuming stack height. A-4 Renumbered equation 7 to equation 15, and updated equation to reflect changes in RSC equation 2.

SAR Revision 2 changes Page 14 of 14 Section Rev2 Page Change Reference A-4 Deleted calculation of discharge concentration using only reactor lab building cross- RSC 879 section (in support of section 11.1.1.1), now only calculates using entire ME building cross-section which is more appropriate.

A-4 Clarified that equation 16 is for zero stack height with building wake effect. RSC 879 A-4 Renumbered equation 9 to equation 16, and updated to reflect changes in equation 6. RSC 879 A-4 Updated final sentence because now only 2 calculations are done, not 3 (because the RSC 879 reactor-lab-only wake effect calculation was removed). Also updated reference to equation number 15 being more realistic, but added that the more conservative equation 16 was used in text.References (1) RSC 812, "NRC Request for Additional Information", May 6, 2004.(2) RSC 773, "Reactor Pulse Channel Digital Recorder Replacement", March 3, 2003.(3) RSC 803, "Replace Console Recorder with Honeywell Multitrend Paperless Recorder", March 12, 2004.(4) RSC 887, "Install New Neutron Measuring Channels; LCR, LogN, and Picoammeters", May 1, 2006.RSC 893, "Installation of New Building Reactor Evacuation Alarm System", August 4, 2006.(6) RSC 896, "Installation of New Reactor Stack Air and Continuous Air Monitors", August 4, 2006.(7) RSC 856, "Loss of Alternating Current SCRAM", October 2005.(8) RSC 857, "Pneumatic Tube Replacement", October 31, 2005.(9) RSC 879, "Modification to Reactor Ventilation System", March 21, 2006.(10) RSC 895, "Installation of Uninterruptible Power Supply", August 4, 2006.(11) RSC 869, "Report on 2005-2006 Annual Maintenance Activities", February 14, 2006.(12) RSC 919, "Report on 2006-2007 Annual Maintenance Activities", February 1, 2007.(13) RSC 894, "Installation of New Fire Detection and Suppression System", August 4, 2006.(14) RSC Charter, as revised on December 6, 2006. See also RSCM-72, "Minutes of the UWNR Safety Committee Meeting, December 6, 2006".(15) RSC 922, "Dosimetry for Students in NE 234, 427 and 428", March 14, 2007.SARRevision 2 changes Section Rev2 Page Change A-4 Deleted calculation of discharge concentration using only reactor lab building cross-section (in support of section 11.1.1.1), now only calculates using entire ME building cross-section which is more appropriate.

A-4 . Clarified that equation 16 is for zero stack height with building wake effect. A-4 Renumbered equation 9 to equation 16, and updated to reflect changes in equation 6. A-4 Updated final sentence because now only 2 calculations are done, not 3 (because the reactor-lab-only wake effect calculation was removed).

Also updated reference to equation number 15 being more realistic, but added that the more conservative equation 16 was used in text. . References (1) RSC 812, "NRC Request for Additional Information", May 6, 2004. (2) RSC 773, "Reactor Pulse Channel Digital Recorder Replacement", March 3,2003. (3) RSC 803, "Replace Console Recorder with Honeywell Multitrend Paperless Recorder", March 12,2004. (4) RSC 887, "Install New Neutron Measuring Channels; LCR, LogN, and Picoammeters", May 1,2006. (5) RSC 893, "Installation of New Building Reactor Evacuation Alarm System", August 4,2006. (6) RSC 896, "Installation of New Reactor Stack Air and Continuous Air Monitors", August 4,2006. (7) RSC 856, "Loss of Alternating Current SCRAM", October 31, 2005. (8) RSC 857, "Pneumatic Tube Replacement", OCtober 31,2005. (9) RSC 879, "Modification to Reactor Ventilation System", March 21,2006. (10) RSC 895, "Installation of Un interruptible Power Supply", August 4,2006. (11) RSC 869, "Report on 2005-2006 Annual Maintenance Activities", February 14,2006. (12) RSC 919, "Report on 2006-2007 Annual Maintenance Activities", February 1,2007. (13) RSG 894, "Installation of New Fire Detection and Suppression System", August 4, 2006. Page 14 of 14 Reference RSC 879 (lJ) . RSC 879 (lJ) RSC 879 (lJ) RSC 879 (lJ) (14) RSC Charter, as revised on December 6, 2006. See also RSCM-72; "Minutes of the UWNR Safety Committee Meeting, December 6, 2006". (15) RSC 922, "Dosimetry for Students in NE 234, 427 and 428", March 14,2007. r SAFETY REPORT FOR RENEWAL OF LICENSE R-74 FOR THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR Rev 2, September 2008* *

SAFETY ANALYSIS REPORT FOR RENEWAL OF LICENSE R-74 FOR THE UNIVERSITY OF WISCONSIN NUCLEAR REACTOR Rev 2, September 2008 REVISION HISTORY Rev 2, September This revision primarily consists of changes related to the remodeling of the Mechanical Engineering Building.

New floor plans and room layouts were described.

New equipment installed included the ventilation system, demineralizer, and the pneumatic tube system. The console instrumentation was significantly upgraded.

Because this revision impacted so much of the document, the opportunity was taken to make many minor corrections and clarifications.

As a result, this revision was issued as a complete replacement to Rev 1, since over 95% of pages were updated. The index was removed and will no longer be maintained.

All changes from Rev 1 are indicated with a vertical line in the margin for each line changed.Rev 1, August 2004 The reasons for this revision were a new reactor pool water cooling system, and a new computer-based pulse channel. Chapters 5 and 7 were completely revised, and the Table of Contents and Index were completely revised accordingly.

All changes from Rev 0 were indicated with a vertical line in the margin for each line changed. Change indications in Rev 0 (Chapter 14) have been removed.Rev April 2000 This was a major revision issued as part of the application for license renewal in 2000.

Sections were reorganized to conform with the format guidelines in NUREG-1537, but no substantial changes were made to the content of the SAR with the exception of the Technical Specifications (Chapter 14). These changes were indicated with a vertical line in the margin for each line changed.* REVISION HISTORY Rev 2, September 2008 This revision primarily consists of changes related to the remodeling of the Mechanical Engineering Building.

New floor plans and room layouts were described.

New equipment installed included the ventilation system, demineralizer, and the pneumatic tube system. The console instrumentation was significantly upgraded.

Because this revision impacted so much of the document, the opportunity was taken to make many minor corrections and clarifications.

As a result, this revision was issued as a complete replacement to Rev 1, since over 95% of pages were updated. The index was removed and will no longer be maintained.

All changes from Rev 1 are indicated with a vertical line in the margin for each line changed. Rev 1, August 2004 The reasons for this revision were a new reactor pool water cooling system, and a new based pulse channel. Chapters 5 and 7 were completely revised, and the Table of Contents and Index were completely revised accordingly.

All changes from Rev 0 were indicated with a vertical line in the margin for each line changed. Change indications in Rev 0 (Chapter 14) have been removed. Rev 0, April 2000

This was a major revision issued as part of the application for license renewal in 2000. Sections were reorganized to conform with the format guidelines in NUREG-153 7, but no substantial changes were made to the content of the SAR with the exception of the Technical Specifications (Chapter 14). These changes were indicated with a vertical line in the margin for each line changed .

1.3 1-2 2-1 2-1 2. Site Location and 2-1 2.

Specification and 2-12.1.1.2 Boundary and Zone Area Maps 2-1 2.1.2 Population 2-2 2-14 2.2.1 Locations and Routes 2-14 2.2.2 Air Traffic Analysis of Potential Accidents at Facilities General and Local Site Meteorology Temperature Precipitation 2-20 Regional Geology Site 2-24 2-24Maximum Earthquake Potential Vibratory Ground Surface Faulting Liquefaction 3 3.1 3.3 UWNR Safety Analysis Report Rev. 2 TC1Sp.20 Sept. 1 THE FACILITY ...................................................

1-1 1.1 Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 1-1 1.2 Summary and Conclusions on Principal Safety Considerations

....... 1-1

1.3 General Description of the Facility ...............................

1-2 1.4 Shared Facilities and Equipment

................................

1-6 1.5 Comparison With Similar Facilities

..............................

1-6 1.6 Summary of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 1-6 1.7 Compliance With the Nuclear Waste Policy Act of 1982 .............

1-7 1.8 Facility Modifications and History ...............................

1-8 2 SITE CHARACTERISTICS

.........................................

2-1 2.1 Geography and Demography

....................................

2-1 2.1.1 Site Location and Description

................................

2-1 2.1.1.1 Specification and Location .............................

2-1 2.1.1.2 Boundary and Zone Area Maps .........................

2-1 2.1.2 Population Distribution

.....................................

2-2 2.2 Nearby Industrial, Transportation, and Military Facilities

..........

2-14 2.2.1 Locations and Routes ......................................

2-14 2.2.2 Air Traffic ..............................................

2-15 2.2.3 Analysis of Potential Accidents at Facilities

....................

2-15 2.3 Meteorology

.................................................

2-15 2.3.1 General and Local Climate .................................

2-15

2.3.2 Site Meteorology

.........................................

2-16 2.3.2.1 Temperature

.......................................

2-16 2.3.2.2 Precipitation

.......................................

2-16 2.3.2.3 Winds ............................................

2-19 2.4 Hydrology

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2-20 2.5 Geology, Seismology, and Geotechnical Engineering

...............

2-23 2.5.1 Regional Geology .........................................

2-23 2.5.2 Site Geology .............................................

2-24 2.5.3 Seismicity

...............................................

2-24 2.5.4 Maximum Earthquake Potential

.............................

2-25 2.5.5 Vibratory Ground Motion ..................................

2-25 2.5.6 Surface Faulting ..........................................

2-25 2.5.7 Liquefaction Potential

.....................................

2-25 2.6 References

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2-26 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS

..........

3-1 3.1 Design Criteria ...............................................

3-1 3.2 Meteorological Damage ........................................

3-1 3.3 Water Damage ................................................

3-2 3.4 Seismic Damage ...............................................

3-2 3.5 Systems and Components

.......................................

3-2

UWNR Safety Analysis Report Rev. 2 TOC-l Sept. 2008 R eferences 3-2REACTOR .........................................4 -1 4.1 Summary Description 4.1.1 Introduction

...............................

4.1.2 Summary of Reactor Data

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4.1.3 Experimental Facilities

......................

4.1.3.1 Hydraulic Irradiation Facility (Whale) 4.1.3.2 Thermal Column .....................

4.1.3.3 Beam Ports

..........................4.1.3.4 Pneumatic Tube ......................

4.2 R eactor Core 4.2.1 Reactor Fuel ...............................

4.2.2 Control Elements ...........................

4.2.2.1 Control Blade Shrouds and Guide Tubes...4.2.2.2 Safety Blades ........................

4.2.2.3 Regulating Blade

.....................4.2.2.4 Transient Control Rod .................

4.2.3 Neutron Moderator and Reflector

..............

4.2.4 Neutron Startup Source

......................

4.2.5 Core Support Structure

......................

4.3 R eactor Pool 4.4 Biological Shield 4.4.1 Pool Surface Radiation Levels -N16 Activity .....4.4.2 Heating Effects in Shield and Thermal Column 4.5 Nuclear Design ................................

4-1 4-1 4-3 4-5 4-5 4-6 4-6 4-6 4-6........ 4-8 4.5.1 Normal Operating Conditions and Reactor Core Physics Parameters 4.5.1.1 Core Arrangements

................................

4.5.1.2 Standard TRIGA fuel cores ..........................

4.5.1.2.1 Reactivity Effects In Standard Fuel Cores ...4.5.1.3 Cores containing FLIP fuel ..........................

4.5.1.3.1 First mixed core- 9 FLIP bundles and 16 standard bundles ....................

4.5.1.3.2 Second mixed core -15 FLIP fuel bundles and 10 standard bundles ................

4.5.1.3.3 All-FLIP core 4.5.1.4 Isothermal Temperature Coefficient

...................

4.5.1.5 Pulse Param eters ..................................

4.5.2 Operating Lim its ........................................

4.6 Thermal-Hydraulic Design 4.7 R eferences

.................................................

4-13 4-13* 4-14 4-14 4-15 4-16 4-17 4-20 4-24 4-26 4-27 4-27 4-27 4-27 4-28 4-29 4-32 4-34 4-37 4-40 4-44 4-47 4-48 4-51 4-52 4-52UWNR Safety Analysis Report Rev. 2 Sept. 3.6 References

...................................................

3-2

4 REACTOR DESCRIPTION

.........................................

4-1 4.1 Summary Description

..........................................

4-1 4.1.1 Introduction

..............................................

4-1 4.1.2 Summary of Reactor Data ...................................

4-3 4.1.3 Experimental Facilities

.....................................

4-5 4.1.3.1 Hydraulic Irradiation Facility (Whale) ....................

4-5 4.1.3.2 Thermal Column ....................................

4-6 4.1.3.3 Beam Ports .........................................

4-6 4.1.3.4 Pneumatic Tube .....................................

4-6 4.2 Reactor Core .................................................

4-6 4.2.1 Reactor Fuel ..............................................

4-8 4.2.2 Control Elements .........................................

4-13 I 4.2.2.1 Control Blade Shrouds and Guide Tubes .................

4-13 I 4.2.2.2 Safety Blades ......................................

4-14 _I 4.2.2.3 Regulating Blade ...................................

4-14 I 4.2.2.4 Transient Control Rod ...............................

4-15 4.2.3 Neutron Moderator and Reflector

............................

4-16 4.2.4 Neutron Startup Source ....................................

4-17 4.2.5 Core Support Structure

....................................

4-20 4.3 Reactor Pool .................................................

4-24

4.4 Biological Shield .............................................

4-26 4.4.1 Pool Surface Radiation Levels -N 16 Activity ...................

4-27 4.4.2 Heating Effects in Shield and Thermal Column .................

4-27 4.5 Nuclear Design ..............................................

4-27 4.5.1 Normal Operating Conditions and Reactor Core Physics Parameters.

4-27 4.5.1.1 Core Arrangements

.................................

4-28 4.5.1.2 Standard TRIGA fuel cores ...........................

4-29 4.5.1.2.1 Reactivity Effects In Standard Fuel Cores .... 4-32 4.5.1.3 Cores containing FLIP fuel ...........................

4-34 4.5.1.3.1 First mixed core-9 FLIP bundles and 16 standard bundles .....................

4-37 4.5.1.3.2 Second mixed core -15 FLIP fuel bundles and 10 standard bundles .................

4-40 4.5.1.3.3 All-FLIP core .........................

4-44 4.5.1.4 Isothermal Temperature Coefficient

....................

4-47 4.5.1.5 Pulse Parameters

...................................

4-48 4.5.2 Operating Limits .........................................

4-51 4.6 Thermal-Hydraulic Design ....................................

4-52 4.7 References

..................................................

4-52 UWNR Safety Analysis Report Rev. 2 TOC-2 Sept. 2008

REACTOR 5-1Summary 5-1 5-4 5-4 Emergency System Design Design-Basis System Description Start-up Log Period Pulse Power 7.2.3.4 Safety 7-4 Temperature 7-47.2.4 System Performance 7-4 7-4 Mode Switch Manual Operation Square Wave Operation 7.3.4 Pulsing Operation Control Element Safety Blade Regulating Blade Control Transient Rod Control Automatic Level Control UWNR Safety Analysis Report Rev. 2 Sept.

5 REACTOR COOLANT SYSTEMS ...................................

5-1 5.1 Summary Description

.........................................

5-1 5.2 Primary Coolant System .......................................

5-1 5.3 Intermediate Coolant System ....................................

5-1 5.4 Campus Chilled Water System ..................................

5-3 5.5 Primary Coolant Cleanup System ........................ " ........ 5-3 5.6 Primary Coolant Makeup Water System ..........................

5-4 5.7 Nitrogen-16 Control System .....................................

5-4 5.8 Auxiliary Systems Using Primary Coolant .........................

5-4 5.9 References

...................................................

5-8 6 ENGINEERED SAFETY FEATURES .................................

6-1 6.1 Summary Description

..........................................

6-1 6.2 Detailed Descriptions

..........................................

6-1 6.2.1 Confinement

..............................................

6-1 6.2.2 Containment

..............................................

6-2 *6.2.3 Emergency Core Cooling System .............................

6-2 6.3 References

...................................................

6-2 7 INSTRUMENTATION AND CONTROL SySTEMS ....................

7-1 7.1 Summary Description

..........................................

7-1

7.2 Design of Instrumentation and Control Systems ....................

7-1 7.2.1 Design Criteria ............................................

7-1 7.2.2 Design-Basis Requirements

..................................

7-3 7.2.3 System Description

........................................

7-3 7.2.3.1 Start-up Channel ....................................

7-3 7.2.3.2 Log N -Period Channel ...............................

7-3 7.2.3.3 Pulse Power Channel .................................

7-3 7.2.3.4 Safety Channels .....................................

7-4 7.2.3.5 Temperature Measurements

............................

7-4 7.2.4 System Performance Analysis ................................

7-4 7.2.5 Conclusion

...............................................

7-4 7.3 Reactor Control System ........................................

7-5 7.3.1 Mode Switch .............................................

7-5 7.3.2 Manual Operation

.........................................

7-5 7.3.3 Square Wave Operation

.....................................

7-5 7.3.4 Pulsing Operation

..........................................

7-6 7.3.5 Control Element Operation

..................................

7-8 7.3.6 Safety Blade Control .......................................

7-8 7.3.7 Regulating Blade Control ....................................

7-8 7.3.8 Transient Rod Control ......................................

7-9 7.3.9 Automatic Level Control System ..............................

7-9 7.4 Reactor Protection System (Scram Circuits)

......................

7-10

UWNR Safety Analysis Report Rev. 2 TOC-3 Sept. 2008

Alarm and Indicator Indicator Lights 7-14 Pneumatic Tube System 7-14 7.6.4 Ventilation System 7-15 Cooling System Whale System Panel Area Radiation Stack Air Monitor Phase Electrical Power Phase Backed Electrical Power 240 4 wire Electrical Power, Transformer wire Electrical Power, Transformer Lighting Electrical Power Backed Lighting Electrical Reactor Laboratory Uninterruptible Power Supply Reactor Laboratory Support Space Uninterruptible Power Supply ...8-4 Mechanical Engineering Building Emergency Filters 9.1.4 Exhaust Variable Air Volume Boxes Emergency Venting Mode Beam Port and Thermal Column Ventilation System Pneumatic System Fume Hood 9-4 Fuel Fuel Fuel Bundle Maintenance and Measurements Systems Possession Use UWNR Safety Analysis Report Rev. 2 TC4Sp.20 TOC-4 Sept. 7.5 Engineered Safety Features Actuation Systems ....................

7-13 7.6 Control Console and Display Instruments

........................

7-13

7.6.1 Alarm and Indicator System

.................................

7-13 7.6.2 Indicator Lights ..........................................

7-14 7.6.3 Pneumatic Tube System Panel ...............................

7-14 7.6.4 Ventilation System Panel ...................................

7-15 7.6.5 Cooling System Panel .....................................

7-15 7.6.6 Whale System Panel .......................................

7-15 7.7 Radiation Monitoring Systems .................................

7-15 7.7.1 Area Radiation Monitors ...................................

7-15 7.7.2 Stack Air Monitor ........................................

7-16 7.8 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

7-17 8 ELECTRICAL POWER SYSTEMS ...................................

8-1 8.1 Normal Electrical Power Systems ................................

8-1 8.1.1277/480 VAC 3 Phase Electrical Power ........................

8-1 8.1.2277/480 VAC 3 Phase Backed Up Electrical Power ...............

8-1 8.1.3 2404 wire V AC Electrical Power, Transformer T1 ...............

8-2 8.1.420811203 wire VAC Electrical Power, Transformer T2 ...........

8-2 8.1.5 277 V AC Lighting Electrical Power ...........................

8-3 8.1.6 277 V AC Backed Up Lighting Power .................

8-3 8.2 Emergency Electrical Power Systems .............................

8-3 8.2.1 Reactor Laboratory Uninterruptible Power Supply .................

8-3 8.2.2 Reactor Laboratory Support Space Uninterruptible Power Supply .... 8-4

8.2.3 Mechanical Engineering Building Emergency Power .............

8-4 9 AUXILIARY SYSTEMS ............................................

9-1 9.1 Heating; Ventilation, and Air Conditioning Systems ................

9-1 9.1.1 Air Handling Unit ...........

.' .............................

9-1 9.1.2 Exhaust Fans .............................................

9-1 9.1.3 Filters ..................................................

9-2 9.1.4 Exhaust Variable Air Volume Boxes ..........................

9-2 9.1.5 Emergency Venting Mode ..................................

9-3 9.1.6 Beam Port and Thermal Column Ventilation System ..............

9-3 9.1. 7 Pneumatic System Fume Hood Exhaust ........................

9-4 9.2 Handling and Storage of Reactor Fuel ............................

9-8 9.2.1 Fuel Handling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

9-8 9.2.2 Fuel Storage ..............................................

9-9 9.2.3 Fuel Bundle Maintenance and Measurements

....................

9-9 9.3 Fire Protection Systems and Programs ...........................

9-13 9.4 Communication Systems ......................................

9-13 9.5 Possession and Use of Byproduct, Source, and Special Nuclear Material ......................................

9-13 UWNR Safety Analysis Report Rev. 2 TOC-4 Sept. 2008

9-1410.2.1 Thermal Column 10.2.2 Beam Ports Pneumatic 10.2.4 Grid Box Irradiation Facilities 10-12 Radiation Sources Airborne Radiation Releases from abnormal reactor operations..

11.1.1.1.2 Releases from normal reactor operations

...11-1 11-1 11-1 11-1 11-1 11-21.2 Liquid Radioactive Sources Solid Radioactive Sources 11.1.2 Radiation Protection Program ALARA 11.1.4 Radiation Monitoring and Surveying 11.1.5 Radiation Exposure Control and Dosimetry Contamination Control Environmental 11.2.1 Radioactive Waste Management Program 11.2.2 Radioactive Waste Control

.... .11-4 11-4 11-5 11-5

Release of Radioactive Waste 12-1 12-1 12.1.1 Structure 12-1 12.1.2 12-1 12.1.2.1 University Radiation Safety Committee............12-1 12.1.2.2 University Radiation Safety (Part of University Department of Environment, Health and Safety)

Chair, Engineering Physics Department 12-1 UWNR Safety Analysis Report Rev. 2 TC5Sp.20 Sept. * *

9.6 Cover Gas Control in closed Primary Coolant Systems .............

9-14 9.7 Other Auxiliary Systems ......................................

9-14 9.8 References

..................................................

9-14 10 EXPERIMENTAL FACILITIES AND UTILIZATION

..................

10-1 11 10.1 Summary Description

.........................................

10-1 10.2 Experimental Facilities

........................................

10-1 10.2.1 Thermal Column ...................................

10-1 10.2.2 Beam Ports ........................................

10-3 10.2.3 Pneumatic Tube ....................................

10-6 10.2.4 Grid Box Irradiation Facilities . . . . . . . . . . . . . . . . . . . . . . .

.. 10-8 10.3 Experiment Review ..........................................

10-11 10.4 References

.................................................

10-12 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT

..........................................

11-1 11.1 Radiation Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 11-1 11.1.1 Radiation Sources ..................................

11-1 11.1.1.1 Airborne Radiation Sources . . . . . . . . . . . . . . . . . . . . . . . .

.. 11-1 11.1.1.1.1 Releases from abnormal reactor operations . .. 11-1 11.1.1.1.2 Releases from normal reactor operations . . . .. 11-2 11.1.1.2 Liquid Radioactive Sources . . . . . . . . . . . . . . . . . . . . . . . .

.. 11-4 11.1.1.3 Solid Radioactive Sources . . . . . . . . . . . . . . . . . . . . . . . . .

.. 11-4 11.1.2 Radiation Protection Program .........................

l1-S 11.1.3 ALARA Program ................................. " 11-S 11.1.4 Radiation Monitoring and Surveying

................

' .... l1-S 11.1.S Radiation Exposure Control and Dosimetry

..............

11-S 11.1.6 Contamination Control ..............................

11-6 11.1.7 Environmental Monitoring. . . . . . . . . . . . . . . . . . . . . . . . . .

.. 11-6 11.2 Radioactive Waste Management

................................

11-6 11.2.1 Radioactive Waste Management Program ................

11-6 11.2.2 Radioactive Waste Control ...........................

11-7 11.2.3 Release of Radioactive Waste .........................

11-7 11.3 References 11-7 12 CONDUCT OF OPERATIONS

......................................

12-1 12.1 Organization

................................................

12-1 12.1.1 Structure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-1 12.1.2 Responsibility. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-1 12.1.2.1 University Radiation Safety Committee

...........

12-1 12.1.2.2 University Radiation Safety (Part of University Department of Environment, Health and Safety) ...............

12-1 12.1.2.3 Chair, Engineering Physics Department

...........

12-1 UWNR Safety Analysis Report Rev. 2 TOC-S Sept. 2008 12.1.2.4 12.1.2.5 12.1.2.6 12.1.2.7 12.1.2.8 Reactor Director ..........

Reactor Safety Committee Reactor Supervisor

........Senior Operators (alternate SuReactor operators

.........12.1.3 Staffi ng .......................

12.1.4 Selection and Training of Personnel 12.1.5 Radiation Safety ................

12.2 Review and Audit Activities 12.2.1 Composition and Qualifications

...12.2.2 Charter and Rules ...............

12.2.3 Review Function ...............

12.2.4 Audit Function

.................

12.3 Procedures 12.4 Required Actions 12.5 R eports .................................

R ecords Emergency Planning Security Planning Quality Assurance 12.10 Operator Training and Requalification 12.11 Startup Plan 12.12 Environmental Reports R eferences .........ors)

.........)........ .o °....°.. °,.. ....°..° .° ., , ., .

..

..12-2 12-2 12-2 12-3 12-3 12-4 12-4 12-5 12-7 12-7 12-7 12-7 12-8 12-8 12-9[2-10[2-13 12-14[2-15 12-15 12-16 12-16 12-16 12-16 Accident Analysis Initiating Events and Determination of C onsequences 13.1.1 Maximum Hypothetical Accident ......................

13-1 13.1.1.1 Fission Product Inventory in Fuel Element .........

13-2 13.1.1.2 Fission Product Release Fraction .................

13-2 13.1.1.3 Activity in Pool W ater .........................

13-2 13.1.1.4 Fission Product Release to Air within the Reactor Laboratory

..................................

13-3 13.1.1.5 Release of Fission Products to Unrestricted Areas 13-6 13.1.2 Insertion of Excess Reactivity

.........................

13-8 13.1.2.1 Fuel Temperatures from Operation at the Scram Point 13-9 13.1.2.2 Temperature after Pulse

.......................

13-10 13.1.3 Loss of Coolant ...................................

13-11 13.1.3.1 Possible Means of Water Loss ..................

13-11 13.1.3.2 Radiation Levels in Confinement Due to Unshielded C ore ......................................

13-12 UWNR Safety Analysis Report Rev. 2 Sept. 12.1.2.4 Reactor Director ..............................

12-2

12.1.2.5 Reactor Safety Committee . . . . . . . . . . . . . . . . . . . .

.. 12-2 12.1.2.6 Reactor Supervisor

............................

12-2 12.1.2.7 Senior Operators (alternate Supervisors)

...........

12-3 12.1.2.8 Reactor operators

.............................

12-3 12.1.3 Staffing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-4 12.1.4 Selection and Training of Personnel

....................

12-4 12.1.5 Radiation Safety ....................................

12-5 12.2 Review and Audit Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-7 12.2.1 Composition and Qualifications

.......................

12-7 12.2.2 Charter and Rules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-7 12.2.3 Review Function ...................................

12-7 12.2.4 Audit Function .....................................

12-8 12.3 Procedures

..................................................

12-8 12.4 Required Actions .............................................

12-9 12.5 Reports ....................................................

12-10 12.6 Records ....................................................

12-13 12.7 Emergency Planning . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-14 12.8 Security Planning

........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-15 12.9 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 12-15 12.10 Operator Training and Requalification

.........................

12-16 12.11 Startup Plan ................................................

12-16 12.12 Reports ......................................

12-16

12.13 References

.......................................

.........

12-16 13 ACCIDENT ANALYSIS ...........................................

13-1 13.1 Accident Analysis Initiating Events and Determination of Consequences. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 13-1 13 .1.1 Maximum Hypothetical Accident ;.....................

13-1 13.1.1.1 Fission Product Inventory in Fuel Element .........

13-2 13.1.1.2 Fission Product Release Fraction .................

13-2 13.1.1.3 Activity in Pool Water .........................

13-2 13.1.1.4 Fission Product Release to Air within the Reactor Laboratory

..................................

13-3 13 .1.1.5 Release of Fission Products to Unrestricted Areas ... 13-6 13.1.2 Insertion of Excess Reactivity

.........................

13-8 13.1.2.1 Fuel Temperatures from Operation at the Scram Point 13-9 13.1.2.2 TemperatureafierPulse

.......................

13-10 13.1.3 Loss of Coolant ...................................

13-11 13 .1.3.1 Possible Means of Water Loss . . . . . . . . . . . . . . . .

.. 13-11 13.1.3.2 Radiation Levels in Confinement Due to Unshielded Core ......................................

13-12 UWNR Safety Analysis Report Rev. 2 TOC-6 Sept. 2008

13.1.3.4 Radiation Levels in Unrestricted Areas Due to Unshielded Core Fuel Temperature After Loss of Pool Water ...13.1.4 Loss of Coolant Flow Mishandling or Malfunction of Fuel Experiment Malfunction Loss of Normal Electrical Power External Events Mishandling or Malfunction of Equipment

..

13-14 13-14 13-14 Operating Conditions Reactor Experiments and Irradiation Reactor Components 1.3.4 Reactor Instrumentation: Excess Reactivity Shutdown Margin Pulse Limits 3.1.4 Core 3.1.5 Reactivity Fuel Operable Control Rods Reactivity Insertion Rates (Scram Other Pulsed Operation 14-1 14-1 14-1 14-1 14-1 14-2 14-414-7 14-9 14-9 14-10 14-13 14-13 14-13 14-13 14-14 14-15 14-16 14-17 14-18 14-18 14-19 14-19 14-20 14-20 14-22 14-22 14-23 14-233.2.4 Reactor Safety System Table 3.2.4 Reactor Safety System Channels Table Backup Shutdown Mechanisms

...

Bypassing Channels

...........

UWNR Safety Analysis Report Rev. 2 TC7Sp.20 Sept. 13.1.3.3 Radiation Levels in Unrestricted Areas Due to

Unshielded Core .............................

13-12 13.1.3.4 Fuel Temperature After Loss of Pool Water ....... 13-14 13.1.4 Loss of Coolant Flow. . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 13-14 13.1.5 Mishandling or Malfunction of Fuel ...................

13-14 13.1.6 Experiment Malfunction

............................

13-15 13.1.7 Loss of Normal Electrical Power ......................

13-15 13.1.8 External Events ...................................

13-15 13.1.9 Mishandling or Malfunction of Equipment

..............

13-16 13.2 Summary and Conclusions

....................................

13-17 13.3 References

.................................................

13-17 14 TECHNICAL SPECIFICATIONS

...................................

14-1 TS 1 INTRODUCTION

......................................

..... 14-1 I TS 1.1 Scope ............................................

14-1 I TS 1.2 Format ...........................................

14-1 I TS 1.3 Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-1 I TS 1.3.1 Reactor Operating Conditions

...................

14-2 I TS 1.3.2 Reactor Experiments and Irradiation

..............

14-4 I TS 1.3.3 Reactor Components

..........................

14-5 I TS 1.3.4 Reactor Instrumentation:

......................

14-7 I TS 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS . 14-9 I

TS 2.1 Safety Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-9 I TS 2.2 Limiting Safety System Settings . . . . . . . . . . . . . . . . . . .

.. 14-10 I TS 3 LIMITING CONDITIONS FOR OPERATION

..................

14-13 I TS 3.1 Reactor Core Parameters

..........................

14-13 I TS 3.1.1 Excess Reactivity . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-13 I TS 3.1.2 Shutdown Margin. . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-13 I TS 3.1.3 Pulse Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-14 I TS 3.1.4 Core Configurations

..........................

14-15 I TS 3.1.5 Reactivity Coefficients

........................

14-16 I TS 3.1.6 Fuel Parameters

.............................

14-17 I TS 3.2 Reactor Control and Safety Systems .................

14-18 I TS 3.2.1 Operable Control Rods .......................

14-18 I TS 3.2.2 Reactivity Insertion Rates (Scram time) ..........

14-19 I TS 3.2.3 Other Pulsed Operation Limitations

.............

14-19 I TS 3.2.4 Reactor Safety System

........................

14-20 I Table 3.2.4 Reactor Safety System Channels ............

14-20 I TS 3.2.5 Interlocks

..................................

14-22 I Table 3.2.5 Interlocks

...............................

14-22 I TS 3.2.6 Backup Shutdown Mechanisms

.................

14-23 I TS 3.2.7 Bypassing Channels ..........................

14-23 I

UWNR Safety Analysis Report Rev. 2 TOC-7 Sept. 2008 TS 3.2.8 Control Systems and Instrumentation Required for Operation

...........................

Table 3.2.8 Instrumentation and Controls Required for O peration ....................................

TS 3.3 Reactor Pool Water Systems ....................

TS 3.4 Confinement

.................................

TS 3.5 Ventilation Systems ...........................

TS 3.6 Emergency Power ............................

TS 3.7 Radiation Monitoring Systems and Effluents

......TS 3.7.1 Monitoring Systems ......................

Table 3.7.1 Radiation Monitoring Systems ..........TS 3.7.2 Effluent (Argon-41)

Discharge Limit ........TS 3.8 Experim ents .................................

TS 3.8.1 Reactivity Limits ........................

TS 3.8.2 M aterials ..............................

TS 3.8.3 Experiment Failure and Malfunctions

........TS 3.9 Facilitv SDecific LCOs ............................. 14-24 14-24 14-25 14-26 14-27 14-28 14-28 14-28 14-28 14-29 14-30 14-30 14-31 14-32 14-32 TS 4 SURVEILLANCE REQUIREMENTS

...................

.......14-334.1 Reactor Core Parameters

..........................

TS 4.2 Reactor Control and Safety Systems .................

TS 4.3 Coolant System s ..................................

TS 4.4 Confinem ent .....................................

TS 4.5 Ventilation Systems ...............................

TS 4.6 Emergency Electrical Power Systems ................

TS 4.7 Radiation Monitoring Systems and Effluents

..........

TS 4.7.1 Radiation Monitoring Systems Applicability

.......TS 4.7.2 Effluents

...................................

TS 4.8 Experim ents .....................................

TS 4.9 Facility-Specific Surveillance

.......................

DESIGN FEATURES ........................................

TS 5.1 Site and Facility Description

........................

TS 5.2 Reactor Coolant System ...........................

TS 5.3 Reactor Core and Fuel .............................

TS 5.4 Reactor Core .....................................

TS 5.5 Control Elements .................................

TS 5.6 Fissionable Material Storage .......................

ADMINISTRATIVE CONTROLS .............................

TS 6.1 O rganization

.....................................

TS 6.1.1 Structure

...................................

TS 6.1.2 Responsibility

..............................6.1.3 Staffi ng ....................................

TS 6.1.4 Selection and Training of Personnel

.............

TS 6.2 Review and Audit .................................

14-34 14-36 14-38 14-38 14-38 14-39 14-39 14-39 14-40 14-40 14-40 14-41 14-41 14-41 14-42 14-43 14-43 14-43 14-45 14-45 14-45 14-45 14-47 14-47 14-47 UWNR Safety Analysis Report Rev. 2 Sept. TS 3.2.8 Control Systems and Instrumentation Required for Operation

...............................

14-24

Table 3.2.8 Instrumentation and Controls Required for Operation

........................................

14-24 TS 3.3 Reactor Pool Water Systems ........................

14-25 TS 3.4 Confinement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-26 TS 3.5 Ventilation Systems ...............................

14-27 TS 3.6 Emergency Power ................................

14-28 TS 3.7 Radiation Monitoring Systems and Effluents . . . . . . . .

.. 14-28 TS 3.7.1 Monitoring Systems ..........................

14-28 Table 3.7.1 Radiation Monitoring Systems ..............

14-28 TS 3.7.2 Effluent (Argon-4l)

Discharge Limit ............

14-29 TS 3.8 Experiments

.....................................

14-30 TS 3.8.1 Reactivity Limits ............................

14-30 TS 3.8.2 Materials

..................................

14-31 TS 3.8.3 Experiment Failure and Malfunctions

............

14-32 TS 3.9 Facility Specific LCOs .............................

14-32 TS 4 SURVEILLANCE REQUIREMENTS

..........................

14-33 TS 4.1 Reactor Core Parameters

..........................

14-34 TS 4.2 Reactor Control and Safety Systems .................

14-36 TS 4.3 Coolant Systems ..................................

14-38 TS 4.4 Confinement

.....................................

14-38 TS 4.5 Ventilation Systems ...............................

14-38

TS 4.6 Emergency Electrical Power Systems ................

14-39 TS 4.7 Radiation Monitoring Systems and Effluents

..........

14-39 TS 4.7.1 Radiation Monitoring Systems Applicability

....... 14-39 TS 4.7.2 Effluents

...................................

14-40 TS 4.8 Experiments

.....................................

14-40 TS 4.9 Facility-Specific Surveillance

.......................

14-40 TS 5 DESIGN FEATURES ........................................

14-41 TS 5.1 Site and Facility Description . . . . . . . . . . . . . . . . . . . . . .

.. 14-41 TS 5.2 Reactor Coolant System ...........................

14-41 TS 5.3 Reactor Core and Fuel .............................

14-42 TS 5.4 Reactor Core .....................................

14-43 TS 5.5 Control Elements .................................

14-43 TS 5.6 Fissionable Material Storage .......................

14-43 TS 6 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-45 TS 6.1 Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-45 TS 6.1.1 Structure

...................................

14-45 TS 6.1.2 Responsibility. . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-45 TS 6.1.3 Staffing ....................................

14-47 TS 6.1.4 Selection and Training of Personnel

.............

14-47 TS 6.2 Review and Audit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-47 UWNR Safety Analysis Report Rev. 2 TOC-8 Sept. 2008

Composition and Qualifications.................14-48 Charter and 14-48 Review Function 14-486.2.4 Audit 14-49Radiation 14-496.4 Experiment Review and Approval 1Required 14-52 Action to be Taken in Case of Safety Limit Violation 14-52 Action to be Taken in the Event of an Occurrence of the Type Identified in and 6.7.2(1)c 14-52Reports 14-53 Operating Reports 14-53 Special Reports 14-55Records 14-56Records to be Retained for a Period of at least Five Years or for the Life of the Component Involved if Less than Five Years................14-56 Records to be Retained for at Least One Cycle 14-56 Records to be Retained for the Lifetime of the Reactor 14-56 14-57 15-1 Financial Ability to Construct a Non-Power Reactor Financial Ability to Operate a Non-Power Reactor Financial Ability to Decommission the Facility 15.4 OTHER Prior Use of Reactor Medical Use of Non-Power Appendix Calculation Methods for Atmospheric Release of Radioactivity Models Used for Calculations in Sections 11.1.1.1.2 and B. Sample Calculations Supporting Section 11.1.1.1.2 Calculations Supporting Section 13.1.1.4 Whole Body Exposure Calculations Supporting Section 13.1.1.4 (2), Dose to the Lungs Calculations Supporting Section and Table A-4 Appendix B Supporting Documents B-I Task Order No. 2 Under Master Task Agreement No. UWNR Safety Analysis Report Rev. 2 TC9Sp.20 Sept. * *

15 TS 6.2.1 Composition and Qualifications

................

14-48 TS 6.2.2 Charter and Rules ............................

14-48 TS 6.2.3 Review Function ............................

14-48 TS 6.2.4 Audit Function . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-49 TS 6.3 Radiation Safety ..... , . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-49 TS 6.4 Procedures

......................................

14-50 TS 6.5 Experiment Review and Approval . . . . . . . . . . . . . . . . .

.. 14-51 TS 6.6 Required Actions .................................

14-52 TS 6.6.1 Action to be Taken in Case of Safety Limit Violation 14-52 TS 6.6.2 Action to be Taken in the Event of an Occurrence of the Type Identified in 6.7.2(1)b., and 6.7.2(1)c

........ 14-52 TS 6.7 Reports .........................................

14-53 TS 6.7.1 Operating Reports ...........................

14-53 TS 6.7.2 Special Reports .............................

14-55 TS 6.8 Records .........................................

14-56 TS 6.8.1 Records to be Retained for a Period of at least Five Years or for the Life of the Component Involved if Less than Five Years .................................

14-56 TS 6.8.2 Records to be Retained for at Least One Cycle ..... 14-56 TS 6.8.3 Records to be Retained for the Lifetime of the Reactor Facility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.. 14-56 TS 7 References

.................................................

14-57 FINANCIAL QUALIFICATIONS

...................................

15-1 15.1 Financial Ability to Construct a Non-Power Reactor ...............

15-1 15.2 Financial Ability to Operate a Non-Power Reactor .................

15-1 15.3 Financial Ability to Decommission the Facility ....................

15-2 15.4 References

..................................................

15-2 16 OTHER LICENSE CONSIDERATIONS

.............................

16-1 16.1 Prior Use of Reactor Components

...............................

16-1 16.2 Medical Use of Non-Power Reactors .............................

16-1 Appendix A Calculation Methods for Atmospheric Release of Radioactivity

...... A-I A. Models Used for Calculations in Sections 11.1.1.1.2 and 13.1.1.5 ..... A-I B. Sample Calculations Supporting Section 11.1.1.1.2 . . . . . . . . . . . . . . .

.. A-2 C. Calculations Supporting Section 13.1.1.4 (1), Whole Body Exposure .. A-2 D. Calculations Supporting Section 13.1.1.4 (2), Dose to the Lungs . . . . .. A-3 E. Calculations Supporting Section 13.1.1.5 and Table 13.1 . . . . . . . . . .

.. A-3 References

.......................................................

A-4 Appendix B Supporting Documents

.........................................

B-1 Task Order No.2 Under Master Task Agreement No. C96-175937

..........

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This page is intentionally left blank.

UWNR Safety Analysis Report Rev. 2 Toe-IO Sept. 2008

THE FACILITYIntroduction This document is prepared as part of an application for renewal of License R-74.The University of Wisconsin has operated a teaching and research reactor, licensed as R-74 under Docket 50-156 since 1961. The reactor supports teaching as a facility of the Engineering Physics Department, other departments of the university, and other educational institutions.

Research use of the reactor supports the department, other University of Wisconsin departments, numerous other educational institutions, and some non-educational groups.The reactor is located on the campus of the university in a building located at 1513 UniversityAvenue in Madison, Wisconsin.

It currently operates at a 1000 kW steady-state power with pulsing capability to 1000 MW.The original Hazards Summary Report has been amended a number of times over the operating history, and the present version of the Safety Analysis Report has been kept up to date by issuance of replacement pages at each annual report submission.

As a request for license renewal progressed, however, it was determined that the Safety Analysis Report should be replaced with this completely new version structured in accordance with the guidance in NUREG 1537,"Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors" dated February 1996.1.2 Summary and Conclusions on Principal Safety Considerations Analysis of possible accident scenarios is included in Chapter 13. As a TRIGA-type reactor, the primary safety features stem from the use of a fuel with a strong negative prompt temperature coefficient of reactivity which limits excursions from reactivity insertions, thus preventing fuel damage from credible reactivity accidents.

Ejection of the transient rod from the core when the core is operating at the power level scram point will result in no fuel damage. Since experiments are limited to the same reactivity worth as the transient rod, experiment failure cannot result inmore severe transients.

In addition, the operating power level of 1000 kW results in a decay heat potential in the fuel small enough that loss of reactor coolant does not result in fuel damage or release of fission products.In extreme accident conditions in which operation is taking place with already damaged fuel, releases to the public are shown to be nominal except for a combination of loss of pool water and loss of the ventilation system concurrent with the fuel damage. Even with these extremely conservative assumptions, analysis of this accident shows the event will result in exposure to the public that would be classified as an Alert. More realistic assumptions used in the accident UWNR Safety Analysis Report Rev. 2Sept. * *

1 THE FACILITY 1.1 Introduction This document is prepared as part of an application for renewal of License R -74. The University of Wisconsin has operated a teaching and research reactor, licensed as R-74 under Docket 50-156 since 1961. The reactor supports teaching as a facility of the Engineering Physics Department, other departments of the university, and other educational institutions.

Research use of the reactor supports the department, other University of Wisconsin departments, numerous other educational institutions, and some non-educational groups. The reactor is located on the campus of the university in a building located at 1513 University Avenue in Madison, Wisconsin.

It currently operates at a 1000 kW steady-state power with pulsing capability to 1000 MW. The original Hazards Summary Report has been amended a number of times over the operating history, and the present version of the Safety Analysis Report has been kept up to date by issuance of replacement pages at each annual report submission.

As a request for license renewal progressed, however, it was determined that the Safety Analysis Report should be replaced with this completely new version structured in accordance with the guidance in NUREG 1537, "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors" dated February 1996 . 1.2 Summary and Conclusions on Principal Safety Considerations Analysis of possible accident scenarios is included in Chapter 13. As a TRIGA-type reactor, the primary safety features stem from the use of a fuel with a strong negative prompt temperature coefficient of reactivity which limits excursions from reactivity insertions, thus preventing fuel damage from credible reactivity accidents.

Ejection of the transient rod from the core when the core is operating at the power level scram point will result in no fuel damage. Since experiments are limited to the same reactivity worth as the transient rod, experiment failure cannot result in more severe transients.

In addition, the operating power level of 1000 kW results in a decay heat potential in the fuel small enough that loss of reactor coolant does not result in fuel damage or release of fission products.

In extreme accident conditions in which operation is taking place with already damaged fuel, releases to the public are shown to be nominal except for a combination of loss of pool water and loss of the ventilation system concurrent with the fuel damage. Even with these extremely conservative assumptions, analysis of this accident shows the event will result in exposure to the public that would be classified as an Alert. More realistic assumptions used in the accident UWNR Safety Analysis Report Rev. 2 1-1 Sept. 2008 analysis indicates that the maximum hypothetical accident would result in emissions and radiation exposure within those allowed by 10 CFR Part 20.General Description of the Facility The University of Wisconsin Nuclear Reactor is located in the Mechanical Engineering Building on the Madison campus of the University within the city of Madison, WI. The building also contains classrooms, laboratories, shops, and staff offices for the departments of MechanicalEngineering, Industrial Engineering, and Engineering Physics departments.

Figure is a pictorial view of the reactor. The reactor is a heterogeneous pool-type nuclear reactor currently fueled with TRIGA-FLIP fuel modified to adapt to 4-element bundle assemblies.

The coolant is light water which circulates through the core by natural convection.The core is reflected by water and graphite.

Maximum steady-state power level is 1000 KW.A 7 by 9 grid, surrounded by a core box, positions fuel, reflector, and control elements.

Three shim-safety blades, a transient control rod, and a regulating blade control core reactivity.

The transient control rod is guided by a tube replacing a fuel element in a central fuel bundle, while the control blades move vertically in two shrouds extending the length of the core. The grid box and control element drive mechanisms are supported by a suspension frame from the reactor bridge.Cold, clean core excess reactivity is about 4.3 % Ak/k. Control elements having a scram capability provide a shut-down margin of about 4.2 % Ak/k.Proposed technical specifications for the facility are included in this report as Chapter 14.

SUMMARY

OF REACTOR DATA Responsible Organization The University of Wisconsin Location Madison, Wisconsin Purpose Teaching and Research Fuel Type TRIGA and TRIGA-FLIP Hydride in 4 element clusters Number of elements in standard 1000 KW Core U vv N aiLULY Ana ys s Repoit ev. Sept. /_VV6 analysis indicates that the maximum hypothetical accident would result in emissions and radiation exposure within those allowed by 10 CFR Part 20. 1.3 General Description of the Facility The University of Wisconsin Nuclear Reactor is located in the Mechanical Engineering Building on the Madison campus of the University within the city of Madison, WI. The building also contains classrooms, laboratories, shops, and staff offices for the departments of Mechanical Engineering, Industrial Engineering, and Engineering Physics departments.

Figure 1-1 is a pictorial view of the reactor. The reactor is a heterogeneous pool-type nuclear reactor currently fueled with TRIGA-FLIP fuel modified to adapt to 4-element bundle assemblies.

The coolant is light water which circulates through the core by natural convection.

The core is reflected by water and graphite.

Maximum steady-state power level is 1000 KW. A 7 by 9 grid, surrounded by a core box, positions fuel, reflector, and control elements.

Three shim-safety blad.es, a transient control rod, and a regulating blade control core reactivity.

The transient control rod is guided by a tube replacing a fuel element in a central fuel bundle, while the control blades move vertically in two shrouds extending the length of the core. The grid box and control element drive mechanisms are supported by a suspension frame from the reactor bridge. I Cold, clean core excess reactivity is about 4.3 % Llklk. Control elements having a scram I* capability provide a shut-down margin of about 4.2 % Llklk. Proposed technical specifications for the facility are included in this report as Chapter 14.

SUMMARY

OF REACTOR DATA Responsible Organization Location Purpose Fuel Type Number of elements in standard 1000 KW Core UWNR Safety Analysis Report Rev. 2 The University of Wisconsin Madison, Wisconsin Teaching and Research TRIGA and TRIGA-FLIP Hydride in 4 element clusters 1-2 Sept. 2008 * *

Control Safety elements Regulating-servo element Transient controlThree vertical blades One vertical blade One rod Experimental Facilities Thermal columnBeam Ports Pneumatic Tube Hydraulic in-pool irradiation facilities One, 40-inch square graphite, 4) th 2 x nv Four, 6-inch diameterx 1010 nv at shield side of shutter;about 8 x at core end of port One, 2-inch (sample size 1 1/4 inch diameter by 5-1/2 inches long), 4) th 4 x 1012 nv.Presently three, 2.5 inch (sample size up to 1 7/8 diameter by 4 inches long)

4) th = 8 x 1012 -2.4 x nv, depending upon location Thermal neutron fluxes for isotope production include the above, plus large irradiation spaces outside the core with thermal neutron fluxes of around 1.3 x 1013 nv.UWNR Safety Analysis Report Rev. 2Sept. * *

Control Safety elements Regulating-servo element Transient control Three vertical blades One vertical blade One rod Experimental Facilities Thermal column Beam Ports Pneumatic Tube One, 40-inch square graphite, th = 2 x 10 8 nv Four, 6-inch diameter th = 1 -3 x 10 10 nv at shield side of shutter; about 8 x 1011 at core end of port One, 2-inch (sample size 1 114 inch diameter by 5-112 inches long), th = 4 x 10 12 nv. Hydraulic in-pool irradiation facilities Presently three, 2.5 inch (sample size up to 1 7/8 diameter by 4 inches long) th = 8 x 10 12 -2.4 X 10\3 nv, depending upon location Thermal neutron fluxes for isotope production include the above, plus large irradiation spaces outside the core with thermal neutron fluxes of around 1.3 x 10\3 nv . UWNR Safety Analysis Report Rev. 2 1-3 Sept. 2008 Figure Open Pool Reactor UWNR Safety Analysis Report Rev. 0 1-4April 2000

Figure 1-1 UWNR Open Pool Reactor UWNR Safety Analysis Report Rev. 0 1-4 April 2000
Dimensions Pool Standard 1000 KW core Grid box Control blades Fuel Element Diameter Length Fueled length Nuclear Characteristics 1 MW Steady state:Maximum thermal neutron flux Maximum fast neutron flux 1000 MW Pulse Maximum thermal neutron flux Maximum fast neutron flux Prompt temperature coefficient of reactivity Void coefficient of reactivity 8 x 12 x 27-1/2 ft. deep 15 x 17 x 15 inches high 9 x 7 array of 3-inch modules 10-1/2 inches wide 3.2 x nv 3.0 x nv 3.2 x 1016 nvx 1016 nv-1.26 x 10-4 AK/°C-.2 x 10-4 AK/% void UWNR Safety Analysis Report Rev. 0April 2000
Dimensions Pool 8 x 12 x 27-112 ft. deep Standard 1000 KW core 15 x 17 x 15 inches high Grid box 9 x 7 array of 3-inch modules Control blades10-112 inches wide Fuel Element Diameter -Length -Fueled length -Nuclear Characteristics
1 MW Steady state: Maximum thermal neutron flux 3.2 x 10 13 nv Maximum fast neutron flux 3.0 x 1013 nv 1000 MW Pulse Maximum thermal neutron flux 3.2 x 10 16 nv Maximum fast neutron flux 3.0 x 10 16 nv Prompt temperature coefficient of reactivity
-1.26 x 10-4 LiKlc Void coefficient of reactivity
-.2 x 10-4 LiKl% void
UWNR Safety Analysis Report Rev. a 1-5 April 2000 Prompt neutron lifetime 42 p.sec TRIGA Standard fuel, 24 pisec FLIP Effective delayed neutron fraction 0.007 1.4 Shared Facilities and Equipment The Reactor Laboratory and supporting laboratories are an integral part of the Mechanical Engineering Building, and thus share walls, water supplies, sewage, and main electricaldistribution with the remainder of the building.
Heating Ventilation
& Air Conditioning (HVAC) systems are dedicated to non-shared use except for those HVAC systems in officespaces. The restricted area contains only reactor-related activities.
1.5 Comparison With Similar Facilities The best indication of reactor characteristics is the performance of the facility itself, which has been in routine operation with the present operational core since August, 1979.The reactor at Washington State University is very similar to UWNR, having also originally been a General Electric Open Pool Reactor which was converted to TRIGA fuel, and eventually partially converted to FLIP fuel. The reactor at Texas A & M University is also a converted core, though the original reactor was not built by GE. The pool size and experimental facility configuration differs on the three reactors, but basic reactor behavior and accident analysis are quite similar. In addition, the nuclear characteristics of UWNR are quite similar to the TRIGA Mark III prototype and other FLIP fueled reactors.
In chapter 4 of this report the similaritybetween the UWNR and the prototype is detailed.Summary of Operations Present plans and previous usage involve use of the reactor in performance of the following experiments:
1. Reactor Start-up and Operation;
2. Radiation Survey of the Reactor and Surroundings;
3. Control and Regulating Blade Calibration;
4. Measurement of Reactor Power and Calibration of Reactor Instruments;
5. Measurement of Shutdown Power Level;W iN tý_ 3MCLY I'MMYSIS tWPOR MeV- Sept. 2_008 Prompt neutron lifetime Effective delayed neutron fraction 1.4 Shared Facilities and Equipment 42 Ilsec TRIGA Standard fuel, 24 Ilsec FLIP 0.007 The Reactor Laboratory and supporting laboratories are an integral part of the Mechanical Engineering Building, and thus share walls, water supplies, sewage, and main electrical distribution with the remainder of the building.
Heating Ventilation
& Air Conditioning (HV AC) systems are dedicated to non-shared use except for those HV AC systems in office spaces. The restricted area contains only activities.
1.5 Comparison With Similar Facilities The best indication of reactor characteristics is the performance of the facility itself, which has been in routine operation with the present operational core since August, 1979.
The reactor at Washington State University is very similar to UWNR, having also originally been a General Electric Open Pool Reactor which was converted to TRIGA fuel, and eventually partially converted to FLIP fuel. The reactor at Texas A & M University is also a converted core,
though the original reactor was not built by GE. The pool size and experimental facility configuration differs on the three reactors, but basic reactor behavior and accident analysis are quite similar. In addition, the nuclear characteristics ofUWNR are quite similar to the TRIGA Mark III prototype and other FLIP fueled reactors.
In chapter 4 of this report the similarity between the UWNR and the prototype is detailed.
1.6 Summary of Operations Present plans and previous usage involve use of the reactor in performance of the following experiments:
1. Reactor Start-up and Operation;
2. Radiation Survey of the Reactor and Surroundings;
3. Control and Regulating Blade Calibration;
4. Measurement of Reactor Power and Calibration of Reactor Instruments;
5. Measurement of Shutdown Power Level; UWNR Safety Analysis Report Rev. 2 1-6 Sept. 2008 *
6. Measurement of Reactor Period, 7. Measurement of Temperature Coefficient of Reactivity;
8. Measurement of Void Coefficient of Reactivity;
9. Experiments Involving the Danger Coefficient Method;10. Experiments to Measure the Disadvantage Factor;Studies of Reactor Buckling and Delta K/K.12. Critical Mass Experiments;
13. Measurement of Thermal Neutron Cross Sections;
14. Delayed Neutron Emission;15. Activation Analysis;16. Experiments Utilizing Pile Oscillator Techniques;
17. Flux Distributions in Reactor and Effect of Absorbers on Flux Patterns;18. Shielding Experiments;
19. Experiments on the Production of Radioisotopes;
20. Neutron Diffractometer Measurements.
21. Neutron Radiography The above represents the experiments planned at present, but it is anticipated that further experi-ments (both for training and research) will be added.Compliance With the Nuclear Waste Policy Act of In accordance with a letter from the U. S. Department of Energy (R. L. Morgan) to the U. S.Nuclear Regulatory Commission (H. Denton) dated May 3, 1983, it has been determined that all universities operating non-power reactors have entered into a contract with DOE that providesthat DOE retain title to the fuel and DOE is obligated to take the spent fuel and/or high level waste for storage or reprocessing.
Because the University of Wisconsin has entered into such a contract with DOE, the applicable requirements of the Nuclear Waste Policy Act of 1982 have UWNR Safety Analysis Report Rev. 2Sept. 2008* *
6 . Measurement of Reactor Period, 7. Measurement of Temperature Coefficient of Reactivity;
8. Measurement of Void Coefficient of Reactivity;
9. Experiments Involving the Danger Coefficient Method; 10. Experiments to Measure the Disadvantage Factor; 11. Studies of Reactor Buckling and Delta KlK. 12. Critical Mass Experiments;
13. Measurement of Thermal Neutron Cross Sections;
14. Delayed Neutron Emission;
15. Activation Analysis;
16. Experiments Utilizing Pile Oscillator Techniques; 17 . Flux Distributions in Reactor and Effect of Absorbers on Flux Patterns;
18. Shielding Experiments;
19. Experiments on the Production of Radioisotopes;
20. Neutron Diffractometer Measurements.
21. Neutron Radiography The above represents the experiments planned at present, but it is anticipated that further ments (both for training and research) will be added. 1.7 Compliance With the Nuclear Waste Policy Act of 1982 In accordance with a letter from the U. S. Department of Energy (R. L. Morgan) to the U. S. Nuclear Regulatory Commission (R. Denton) dated May 3, 1983, it has been determined that all universities operating non-power reactors have entered into a contract with DOE that provides that DOE retain title to the fuel and DOE is obligated to take the spent fuel and/or high level waste for storage or reprocessing.
Because the University of Wisconsin has entered into such a contract with DOE, the applicable requirements of the Nuclear Waste Policy Act of 1982 have UWNR Safety Analysis Report Rev. 2 1-7 Sept. 2008 been satisfied the University of Wisconsin Nuclear Reactor Facility. copy of the current Task Order under Master Task Agreement (which is the successor to contract 76ERO is included in Appendix B.
Construction permit CPRR-55, authorizing construction of the University of Wisconsin NuclearReactor (UWNR), was issued on June License R-74 was issued November The expiration date of the license was set at 40 years after the issuance of the construction permit (June 2000). The University of Wisconsin Nuclear Reactor achieved initial criticality on March as a KW teaching and research reactor. After a license amendment dated October 1964, the power level was increased to KW on December 1964, using the original flat-plate aluminum clad fuel. Operations with the original core ended October after critical hours and megawatt hours of core exposure.cooling system was installed and the reactor was converted to a KW, TRIGA reactor with pulsing capability in Construction permnit CPRR-97 authorizing the changes was issued on June Amendment No. to the operating license for the conversion was issued on November Initial criticality with the TRIGA core occurred on November 14, After over megawatt hours of operation with the TRIGA core a partial refueling was necessary.
FLIP fuel was available to afford significantly improved core lifetimes, so a new Safety Analysis Report was submitted in April, describing facility characteristics and safeguards using standard fuel, FLIP fuel, and mixtures of the two fuel types in defined a mixed core with Standard fuel bundles replaced with FLIP (Fuel Life Improvement Program)fuel was initially critical in March 1974. Additional fuel replacements in January and August resulted in the present operating core, consisting entirely of FLIP fuel. The total fuel exposure since converting to TRIGA fuel is over 20,000 megawatt hours, or megawattnumber of other license amendments were issued during the term of the license, involving inclusion of security, training, and emergency plans. Several other amendments changed the amount and types of Special Nuclear Material used in connection with the license.
None of these changes had any effect on the operating characteristics of the reactor, and are therefore not detailed here. The most recent amendment was No. dated August new Safety Analysis Report was submitted in April 2000 as part of a license extension period and elimination of the electronic scram capability of the safety amplifier.
These changes were approved our Reactor Safety Committee based on a CFR analysis.
Neither of these items had been required Technical Specifications, since they did not provide protection for a pulsing TRIGA reactor. Sections and 7.4 describe the instrumentation as it has been changed.UWNR Safety Analysis Report Rev. 2 18Sp.20Sept. been satisfied by the University of Wisconsin Nuclear Reactor Facility.
A copy of the current Task Order under Master Task Agreement (which is the successor to contract DE-AC07-76ERO 1560) is included in Appendix B. 1.8 Facility Modifications and History Construction permit CPRR-55, authorizing construction of the University of Wisconsin Nuclear Reactor (UWNR), was issued on June 7 th , 1960. License R-74 was issued November 23,d, 1960. The expiration date of the license was set at 40 years after the issuance of the construction permit (June 6 th , 2000). The University of Wisconsin Nuclear Reactor achieved initial criticality on March 26 t\ 1961 as a 10 KW teaching and research reactor. After a license amendment dated October 22 nd , 1964, the power level was increased to 250 KW on December 7 t\ 1964, using the original flat-plate aluminum clad fuel. Operations with the original core ended October 13 th , 1967, after 2268.5 critical hours and 105.65 megawatt hours of core exposure.
A cooling system was installed and the reactor was converted to a 1000 KW, TRlGA reactor with pulsing capability in 1967. Construction permit CPRR-97 authorizing the changes was issued on June 7, 1967. Amendment No.8 to the operating license for the conversion was issued on November 13, 1967. Initial criticality with the TRlGA core occurred on November 14, 1967. After over 3,000 megawatt hours of operation with the TRlGA core a partial refueling was necessary.
FLIP fuel was available to afford significantly improved core lifetimes, so a new Safety Analysis Report was submitted in April, 1973 describing facility characteristics and
safeguards using standard fuel, FLIP fuel, and mixtures of the two fuel types in defined
compositions.
License amendment No. 10 was issued in response.
The initial partial refueling to a mixed core with 9 Standard fuel bundles replaced with FLIP (Fuel Life Improvement Program) fuel was initially critical in March 1974. Additional fuel replacements in January 1978 and August 1979 resulted in the present operating core, consisting entirely of FLIP fuel. The total fuel exposure since converting to TRlGA fuel is over 20,000 megawatt hours, or 833 megawatt days. A number of other license amendments were issued during the term of the license, involving inclusion of security, training, and emergency plans. Several other amendments changed the amount and types of Special Nuclear Material used in connection with the license. None of these changes had any effect on the operating characteristics of the reactor, and are therefore not detailed here. The most recent amendment was No. 16, dated August 30,2006. A new Safety Analysis Report was submitted in April 2000 as part of a license extension application.
Two changes were included in the new SAR: elimination of the reactor trip on short period and elimination of the electronic scram capability of the safety amplifier.
These changes were approved by our Reactor Safety Committee based on a 10 CFR 50.59 analysis.
Neither of these items had been required by Technical Specifications, since they did not provide protection for a pulsing TRlGA reactor. Sections 7.2.3 and 7.4 describe the instrumentation as it has been changed. UWNR Safety Analysis Report Rev. 2 1-8 Sept. 2008
A new cooling system was installed in 2003. Revision 1 of the SAR was issued in August 2004 mainly to reflect the new cooling system, which is described in Chapter 5.From 2004 through 2007 the Mechanical Engineering building surrounding the reactor laboratoryunder went construction. The center wing was completely demolished and rebuilt, and the east, west, and north wings, as well as the reactor auxiliary support spaces, were substantially renovated.
Some reactor systems were also renovated, but the reactor itself remained unchanged.
Revision 2 of the SAR was issued in September 2008 to address the relevant changes. The new building floor plans are shown in Chapter 2. A new ventilation system was installed, described primarily in Chapter 9 (the calculations in Chapter 11, Chapter 13, and Appendix A were also affected).
Several new reactor control console components were installed as part of a console instrumentation upgrade, and Chapter 7 was updated to reflect these changes.
A new pneumatic tube sample transfer system (which mates with the existing irradiation facility) was installed and is described in Chapter 10. These changes were approved by our Reactor Safety Committeebased on a 10 CFR 50.59 analysis.UWNR Safety Analysis Report Rev.
2Sept. * *
A new cooling system was installed in 2003. Revision 1 of the SAR was issued in August 2004 mainly to reflect the new cooling system, which is described in Chapter 5. From 2004 through 2007 the Mechanical Engineering building surrounding the reactor laboratory under went construction.
The center wing was completely demolished and rebuilt, and the east, west, and north wings, as well as the reactor auxiliary support spaces, were substantially renovated.
Some reactor systems were also renovated, but the reactor itself remained unchanged.
Revision 2 of the SAR was issued in September 2008 to address the relevant changes. The new building floor plans are shown in Chapter 2. A new ventilation system was installed, described primarily in Chapter 9 (the calculations in Chapter 11, Chapter 13, and Appendix A were also affected).
Several new reactor control console components were installed as part of a console instrumentation upgrade, and Chapter 7 was updated to reflect these changes. A new pneumatic tube sample transfer system (which mates with the existing irradiation facility) was installed and is described in Chapter 10. These changes were approved by our Reactor Safety Committee based on a 10 CFR 50.59 analysis . UWNR Safety Analysis Report Rev. 2 1-9 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2Sept.
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UWNR Safety Analysis Report Rev. 2 1-10 Sept. 2008
2 CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location and Description 2.1.1.1 Specification and Location The University of Wisconsin Nuclear Reactor Laboratory is located within the Mechanical Engineering Building at 1513 University Avenue on the University of Wisconsin-Madison (UW)campus. The reactor is located at On the United States Geological Survey (USGS) Madison West, WIS 15' Quadrangle topographical map, the Universal Transverse Mercator Coordinates are; .The UW campus is surrounded by the city of Madison in Dane county, Wisconsin.
Figure 2-1shows the location of Dane county within Wisconsin.
Madison, a city of approximately 208,000 residents (2000 Census statistics), is built around two lakes in the center of Dane county, Figure 2-2. Lake Mendota (15 square miles) lies northwest of Lake Monona (5 square miles) and the two lakes are only 2/3 of a mile apart at one point. The UW campus is set on this narrow neck of land between the two lakes, known as the isthmus, and on the southern shore of Lake Mendota.The Mechanical Engineering Building is near the southwestern border of the UW campus, where the nearest non UW owned property is 425 ft (130 m) from the reactor site (approximately 80m from the Mechanical Engineering Building west wall). The reactor is 2300 ft (700 m) south of the shore of Lake Mendota.2.1.1.2 Boundary and Zone Area Maps A map of the City of Madison detailing the general topography and the surrounding urban and rural zones up to a distance of 8 km is reproduced in Figure 2-3. The UW campus, which islocated on the southern shore of Lake Mendota, is shown in Figure 2-4. The Mechanical Engineering Building is located on the engineering campus which is the southwest comer of the UW campus as shown in Figure 2-5.The operations boundary is'defined as the Reactor Laboratory, Room 1215, of the MechanicalEngineering Building.
The site boundary is defined as the center and east wings of the Mechanical Engineering Building, but not including the north or west wings, plus the portion ofEngineering Drive (formally Johnson Drive) south. of the designated areas of the building.Figure Figure 2-7, Figure 2-8, Figure 2-9, Figure 2-10, and Figure 2-1,1 depict the floor plans of the Mechanical Engineering Building's basement, first floor, second floor, third floor, fourth floor, and fifth floor respectively.
Figure 2-12 and Figure depict cross sections of the building through the core centerline looking north and east, respectively.
The emergency preparedness zone is entirely within the operations boundary, as defined above.UWNR Safety Analysis Report Rev. 2 2-1 Sept. * *
2 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Site Location and Description 2.1.1.1 Specification and Location The University of Wisconsin Nuclear Reactor Laboratory is located within the Mechanical Engineering Building at 1513 University Avenue on the University of Wisconsin-Madison (UW) campus. The reactor is located at On the United States Geological Survey (USGS) Madison West, WIS 15' the Universal Transverse Mercator Coordinates are; The UW campus is surrounded by the city of Madison in Dane county, Wisconsin.
Figure 2-1 shows the location of Dane county within Wisconsin.
Madison, a city of approximately 208,000 residents (2000 Census statistics), is built around two lakes in the center of Dane county, Figure 2-2. Lake Mendota (15 square miles) lies northwest of Lake Monona (5 square miles) and the two lakes are only 2/3 of a mile apart at one point. The UW campus is set on this narrow neck of land between the two lakes, known as the isthmus, and on the southern shore of Lake Mendota. The Mechanical Engineering Building is near the southwestern border of the UW campus, where the nearest non UW owned property is 425 ft (130 m) from the reactor site (approximately 80m from the Mechanical Engineering Building west wall). The reactor is 2300 ft (700 m) south of the shore of Lake Mendota. 2.1.1.2 Boundary and Zone Area Maps A map of the City of Madison detailing the general topography and the surrounding urban and rural zones up to a distance of 8 km is reproduced in Figure 2-3. The UW campus, which is located on the southern shore of Lake Mendota, is shown in Figure 2-4. The Mechanical Engineering Building is located on the engineering campus which is the southwest corner of the UW campus as shown in Figure 2-5. The operations boundary is'defined as the Reactor Laboratory, Room 1215, of the Mechanical Engineering Building.
The site boundary is defined as the center and east wings of the Mechanical Engineering Building, but not including the north or west wings, plus the portion of Engineering Drive (formally Johnson Drive) south of the designated areas of the building.
Figure 2-6, Figure 2-7, Figure 2-8, Figure 2-9, Figure 2-10, and Figure 2-11 depict the floor plans of the Mechanical Engineering Building's basement, first floor, second floor, third floor, fourth floor, and fifth floor respectively.
Figure 2-12 and Figure 2-13 depict cross sections of the building through the core centerline looking north and east, respectively. The emergency preparedness zone is entirely within the operations boundary, as defined above . UWNR Safety Analysis Report Rev. 2 2-1 Sept. 2008 Figure 2-1 Wisconsin Map 3 &0 Figure 2-2 Dane County, Wisconsin 2.1.2 Population DistributionPopulation distributions estimated by the uniform density model and the 2000 Censusare shown in Table 2-1. The area around the campus is mature residential, and the central business district is only a short distance away. Population is therefore quite stable in the immediate surrounding areas of the UW campus. The 8 kilometer radius includes much more sparsely-settled regions to the north of Lake Mendota, and population in this region is likely to increase markedly in the future. The nearest permanent residence is approximately 150 m west of the reactor site.Transient population around the reactor include students present in classrooms and offices in the Mechanical Engineering Building during the months of September through May as well as spectators attending sporting events at Camp Randall Stadium which is approximately 250 m due south of the reactor. The maximum number of students present in the east and center wings of the Mechanical Engineering Building at any one time is estimated at 1500. The maximum number of spectators contained by Camp Randall Stadium is 80,321.UWNR Safety. Analysis Report Rev. 2 2-2 Sept. Figure 2-1 Wisconsin Map Figure 2-2 2.1.2 Population Distribution Dane County, Wisconsin Population distributions estimated by the uniform density model and the 2000 Census 1 are shown in Table 2-1. The area around the campus is mature residential, and the central business district is only a short distance away. Population is therefore quite stable in the immediate surrounding areas of the UW campus. The 8 kilometer radius includes much more sparsely-settled regions to the north of Lake Mendota, and population in this region is likely to increase markedly in the future. The nearest permanent residence is approximately 150 m west of the reactor site. Transient popUlation around the reactor include students present in classrooms and offices in the Mechanical Engineering Building during the months of September through Mayas well as spectators attending sporting events at Camp Randall Stadium which is approximately 250 m due south of the reactor. The maximum t:J.umber of students present in the east and center wings of the Mechanical Engineering Building at anyone time is estimated at 1500. The maximum number of spectators contained by Camp Randall Stadium is 80,321. . UWNR Safety Analysis Report Rev. 2 2-2 Sept. 2008 * *
Figure City of Madison, Wisconsin UWNR Safety Analysis Report Rev. 2 2-3 Sept. *
F ig u re 2-3 City o f Madison, W i sco n s i n
UWNR Safety Analysis R eport R ev. 2 2-3 Sept. 2008
~ j7 University of Wisconsin Campus Figure 2-4 UWNR Safety Analysis Report Rev. 2 2-4 Sept. * * " , Figure 2-4 .. ....,. , '-. " , University of Wisconsin Campus UWNR Safety A n a l ysis Report Rev. 2 2-4 . .. , . .. , T -':' .. 1-;' -'0 , , ,_to * , -*
Sept. 2008 * *
University Ave. 774,7-Fw I ii jot i nq St~ j4h it4ýhIHti F JL-40Regent St.Figure 2-5 UW Engineering Campus UWNR Safety Analysis Report Rev. 2Sept. * *
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Figure 2-6 Mechanical Engineering-Basement UWNR Safety Analysis Report Rev. 2 2-6 Sept. 2008
Figure 2-7 Mechanical Engineering
-First Floor UWNR Safety Analysis Report Rev. 2 2-7 Sept. 2008* *
Figure 2-7 Mechanical Engineering
-First Floor UWNR Safety Analysis Report Rev. 2 2-7 Sept. 2008 Figure Mechanical Engineering
-Second Floor UWNR Safety Analysis Report Rev. 2Sept. *
Figure 2-8 Mechanical Engineering
-Second Floor UWNR Safety Analysis Report Rev. 2 2-8 Sept. 2008
Figure Mechanical Engineering
-Third Floor UWNR Safety Analysis Report Rev. 2 2-9 Sept. 2008*
Figure 2-9 Mechanical Engineering
-Third Floor
UWNR Safety Analysis Report Rev. 2 2-9 Sept. 2008 Figure 2-10 Mechanical Engineering
-Fourth Floor UWNR Safety Analysis Report Rev. 2 2-10 Sept. I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I t I I t I I t I I t Figure 2-10 Mechanical Engineering
-Fourth Floor UWNR Safety Analysis Report Rev. 2 2-10 * *
Sept. 2008 Figure 2-11 Mechanical Engineering
-Fifth Floor UWNR Safety Analysis Report Rev. 2 2-11 Sept. 2008* *
Figure 2-11 Mechanical Engineering
-Fifth Floor UWNR Safety Analysis Report Rev. 2 2-11 Sept. 2008 I I I I I I I I I I Mechanical Engineering Cross Section through Core Centerline, looking North Figure 2-12 UWNR Safety Analysis Report Rev. 2 2-12 Sept. Figure 2-12 Mechanical Engineering Cross Section through Core Centerline, looking North UWNR Safety Analysis Report Rev. 2 2-12 Sept. 2008 * *
Figure 2-13 Mechanical Engineering Cross Section through Core Centerline, looking East UWNR Safety Analysis Report Rev. 2 2-13 Sept. 2008*
Figure 2-13 Mechanical Engineering Cross Section through Core Centerline, looking East
UWNR Safety Analysis Report Rev. 2 2-13 Sept. 2008 TABLE 2-1 Population Distribution Distance from Facility(kilometers)
Estimated 2000 Population' 1 12,595 2 37,814 4 71,780 6 118,919 8 192,575 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2.1 Locations and Routes The UW campus is surrounded, mainly, by a residential district to the south and west, while to the east is primarily a commercial business district with city and state government office buildings.
No industrial facilities are in the vicinity of the reactor.A railroad spur of the Wisconsin
& Southern Railroad Company runs through campus and is 100m to the Reactor Laboratory at its closest approach.
A rail car holding yard which is a part of this spur is approximately 300 m northwest.
The primary commodity transported over this spur or resident at the rail car holding yard is coal 2.The reactor is located approximately 4.5 km north of a bypass highway, known as the Beltline, for US highways 12, 14, 18 and 151. Interstates 90 and 94 are located approximately 10 km tothe east and northeast of the reactor site.There are no military facilities in the Madison area with the exception of the Wisconsin Air National Guard which is located on a military ramp of the Dane County Regional Airport. While the Wisconsin Air National Guard flies approximately 4000 missions annually, the flight pattern of these missions typically are north of Lake Mendota and the City of Madison. At no time do any of these mission flights carry live ammunition 3.More information about the Dane County Regional Airport facility is reported in section 2.2.2 Air Traffic.UWNR Safety Analysis Report Rev. 2 2-14 Sept. TABLE 2-1 Population Distribution Distance from Facility(kilometers)
Estimated 2000 Population 1 1 12,595 2 37,814 4 71,780 6 118,919 8 192,575 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2.1 Locations and Routes The UW campus is surrounded, mainly, by a residential district to the south and west, while to the east is primarily a commercial business district with city and state government office buildings.
No industrial facilities are in the vicinity of the reactor. A railroad spur of the Wisconsin
& Southern Railroad Company runs through campus and is 100m to the Reactor Laboratory at its closest approach.
A rail car holding yard which is a part of this spur is approximately 300 m northwest.
The primary commodity transported over this spur or resident at the rail car holding yard is coaf. The reactor is located approximately 4.5 km north of a bypass highway, known as the Beltline, for US highways 12, 14, 18 and 151. Interstates 90 and 94 are located approximately 10 km to the east and northeast of the reactor site. There are no military facilities in the Madison area with the exception of the Wisconsin Air National Guard which is located on a military ramp of the Dane County Regional Airport. While the Wisconsin Air National Guard flies approximately 4000 missions annually, the flight pattern of these missions typically are north of Lake Mendota and the City of Madison. At no time do any of these mission flights carry live ammunition 3* More information about the Dane County Regional Airport facility is reported in section 2.2.2 Air Traffic. UWNR Safety Analysis Report Rev. 2 2-14 Sept. 2008 * *
2.2.2 Air Traffic The Dane County Regional Airport is approximately 8 km to the north east of the reactor site.This is the only commercial airport near the reactor. While there are three smaller air fields within 16 km from the reactor in the communities surrounding Madison, these air fields are for general aviation only.The Dane County Regional Airport has 3 runways with the following outbound headings; 360°(north)/1800, 320'/140', and 2100/300.
None of these headings have trajectories that take commercial traffic directly over the reactor immediately before. arrival or after departure.
The airport is serviced by several commercial express carriers and is utilized by the Wisconsin Air National Guard as well as general aviation aircraft 3.Of the 115,613 events (arrival and departure are counted as two separate events) in 2006, 59% were classified as general aviation, 34%commercial carriers/taxi, and 7% military.
Due to the infrequent arrival of commercial traffic, the air traffic control tower does not place inbound traffic in holding patterns around the city ofMadison4.2.2.3 Analysis of Potential Accidents at Facilities There are no industrial, transportation or military facilities within the vicinity of the reactor site that have the potential for accidents with consequences significant to impact the ReactorLaboratory. While a railroad spur passes within 100 m of the reactor facility, this spur transports non hazardous cargo and other major ground transportation routes are located at great distances from the Reactor Laboratory.
Due to the frequency and flight paths of commercial and military air traffic the probability of occurrence of an accident is considered extremely low.
2.3 Meteorology 2.3.1 General and Local Climate Madison has the typical continental climate of interior North America with a large annual temperature range and with frequent short period temperature changes. The range of extreme temperatures is from about 110 to -40 'F. Winter temperatures (December
-February) average near 20 'F and the summer (June -August) average temperature is in the upper 60s. Daily temperatures average below 32 'F about 120 days and above 40 'F for about 210 days of the year.Madison lies in the path of the frequent cyclones and anticyclones which move eastward over this area during fall, winter and spring.
In summer, the cyclones have diminished intensity and tend to pass farther north. The most frequent air masses are of polar origin. Occasional outbreaks of arctic air affect this area during the winter months. Although northward movingtropical air masses contribute considerable cloudiness and precipitation, the true Gulf air mass UWNR Safety Analysis Report Rev. 2Sept. * *
2.2.2 Air Traffic The Dane County Regional Airport is approximately 8 km to the north east of the reactor site. This is the only commercial airport near the reactor. While there are three smaller air fields within 16 km from the reactor in the communities surrounding Madison, these air fields are for general aviation only. The Dane County Regional Airport has 3 runways with the following outbound headings; 360 0 (north)1180°, 320°1140°, and 210°/30°.
None of these headings have trajectories that take commercial traffic directly over the reactor immediately before arrival or after departure.
The airport is serviced by several commercial express carriers and is utilized by the Wisconsin Air National Guard as well as general aviation aircraft 3* Of the 115,613 events (arrival and departure are counted as two separate events) in 2006,59% were classified as general aviation, 34% commercial carriers/taxi, and 7% military.
Due to the infrequent arrival of commercial traffic, the air traffic control tower does not place inbound traffic in holding patterns around the city of Madison 4* 2.2.3 Analysis of Potential Accidents at Facilities There are no industrial, transportation or military facilities within the vicinity of the reactor site that have the potential for accidents with consequences significant to impact the Reactor Laboratory.
While a railroad spur passes within 100 m of the reactor facility, this spur transports non hazardous cargo and other major ground transportation routes are located at great distances from the Reactor Laboratory.
Due to the frequency and flight paths of commercial and military air traffic the probability of occurrence of an accident is considered extremely low. 2.3 Meteorology 2.3.1 General and Local Climate Madison has the typical continental climate of interior North America with a large annual temperature range and with frequent short period temperature changes. The range of extreme temperatures is from about 110 to -40 OF. Winter temperatures (December
-February) average near 20 OF and the summer (June -August) average temperature is in the upper 60s. Daily temperatures average below 32 OF about 120 days and above 40 OF for about 210 days of the year. Madison lies in the path of the frequent cyclones and anticyclones which move eastward over this area during fall, winter and spring. In summer, the cyclones have diminished intensity and tend to pass farther north. The most frequent air masses are of polar origin. Occasional outbreaks of arctic air affect this area during the winter months. Although northward moving tropical air masses contribute considerable cloudiness and precipitation, the true Gulf air mass UWNR Safety Analysis Report Rev. 2 2-15 Sept. 2008 does not reach this area in winter, and only occasionally at other seasons. Summers are pleasant, with only occasional periods of extreme heat or high humidity.There are no dry and wet seasons, but about 60 percent of the annual precipitation falls in the five months of May through September. Cold season precipitation is lighter, but lasts longer. During July, August, and September rainfall is mostly from thunderstorms and tends to be erratic and variable.
Average occurrence of thunderstorms is just under 7 days per month during this period.
Tornadoes are infrequent.
Dane County has about one tornado in every three to five years.The ground is covered with 1 inch or more of snow about 60 percent of the time from about December 10 to near February 25 in an average winter. The soil is usually frozen from the first of December through most of March with an average frost penetration of 25 to 30 inches.", 5 2.3.2 Site Meteorology The summary of meteorological conditions for Madison is based on the records obtained from the International Station Meteorological Climate Summary6 jointly produced by the National Oceanic and Atmospheric Administration (NOAA), the United States Air Force (USAF) and United States Navy. The data specifically compiled for Madison was obtained from the National Weather Service and unless specifically noted, is for the period of record from 1948 to 1995.The Reactor Laboratory does not have a continuing onsite meteorological data measurements program. All future meteorological data will be obtained from the National Weather Service station in Madison.2.3.2.1 Temperature The monthly average and daily average extreme temperatures for the Madison area are shown in Table 2-2. The record extreme temperatures in the Madison area, as reported by the NationalWeather Service, have ranged from a low of-37 'F in January 1951 to a high of 107 'F in July 1936.2.3.2.2 Precipitation The Madison area normally receives an annual average of 31.6 inches of precipitation.
The monthly average precipitation data is reported in Table 2-3. The record maximum level of precipitation to fall in Madison in one year, is reported by the National Weather Service to be 52.9 inches in 1881. The greatest 24-hour rain fall total for Madison was 5.25 inches on July 15-16, 1950. The 48-hour, 100-yr. return period rainfall for south central Wisconsin is estimated to be 7.82 inches 7.The annual average snow fall for Dane County is 43.7 inches.
The monthly average snow fall data is reported in Table 2-4. The record maximum snow to fall during the winter season is UWNR Safety Analysis Report Rev. 2Sept. does not reach this area in winter, and only occasionally at other seasons. Summers are pleasant, with only occasional periods of extreme heat or high humidity.
There are no dry and wet seasons, but about 60 percent of the annual precipitation falls in the five months of May through September.
Cold season precipitation is lighter, but lasts longer. During July, August, and September rainfall is mostly from thunderstorms and tends to be erratic and variable.
Average occurrence of thunderstorms is just under 7 days per month during this period. Tornadoes are infrequent.
Dane County has about one tornado in every three to five years. The ground is covered with 1 inch or more of snow about 60 percent of the time from about December 10 to near February 25 in an average winter. The soil is usually frozen from the first of December through most of March with an average frost penetration of25 to 30 inches."s . 2.3.2 Site Meteorology The summary of meteorological conditions for Madison is based on the records obtained from the International Station Meteorological Climate Summary6 jointly produced by the National Oceanic and Atmospheric Administration (NOAA), the United States Air Force (USAF) and United States Navy. The data specifically compiled for Madison was obtained from the National Weather Service and unless specifically noted, is for the period of record from 1948 to 1995. The Reactor Laboratory does not have a continuing onsite meteorological data measurements program. All future meteorological data will be obtained from the National Weather Service station in Madison. 2.3.2.1 Temperature The monthly average and daily average extreme temperatures for the Madison area are shown in Table 2-2. The record extreme temperatures in the Madison area, as reported by the National Weather Service, have ranged from a low of -37 OF in January 1951 to a high of 107 OF in July 1936. 2.3 .2.2 Precipitation The Madison area normally receives an annual average of 31.6 inches of precipitation.
The monthly average precipitation data is reported in Table 2-3. The record maximum level of precipitation to fall in Madison in one year, is reported by the National Weather Service to be 52.9 inches in 1881. The greatest 24-hour rain fall total for Madison was 5.25 inches on July 15-16, 1950. The 48-hour, 100-yr. return period rainfall for south central Wisconsin is estimated to be 7.82 inches 7* The annual average snow fall for Dane County is 43.7 inches. The monthly average snow fall data is reported in Table 2-4. The record maximum snow to fall during the winter season is UWNR Safety Analysis Report Rev. 2 2-16 Sept. 2008 * *
reported by the National Weather Service as 76.1 inches during the winter of 1978-79. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months state record for heaviest snow occurred December 27-28, 1904 in Neillsville, Wisconsin in which 26 inches of snow fell.TABLE 2-2 Average Temperatures for the Madison AreaAverage Monthly Average Daily Average Daily Temperature Maximum Temperature Minimum Temperature (OF) (OF) (OF)January 16.8 25.7 8.0 February 21.3 30.6 12.0 March 32.3 41.9 22.7 April 46.1 57.6 34.7 May 57.5 70.0 44.9 June 67.0 79.4 54.5 July 71.4 83.3 59.5 August 69.2 81.0 57.4 September 60.4 72.3 48.6 October 49.5 60.9 38.2 November 35.3 44.0 26.5 December 22.6 30.7 14.4 Year 45.8 56.5 35.1 UWNR Safety Analysis Report Rev. 2Sept. * *
reported by the National Weather Service as 76.1 inches during the winter of 1978-79. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months state record for heaviest snow occurred December 27-28, 1904 in Neillsville, Wisconsin in which 26 inches of snow fell. TABLE 2-2 Average Temperatures for the Madison Area Average Monthly Average Daily Average Daily Temperature Maximum Temperature Minimum Temperature COF) COF) COF) January 16.8 25.7 8.0 February 21.3 30.6 12.0 March 32.3 41.9 22.7 April 46.1 57.6 34.7 May 57.5 70.0 44.9 June 67.0 79.4 54.5 July 71.4 83.3 59.5 August 69.2 81.0 57.4 September 60.4 72.3 48.6 October 49.5 60.9 38.2 November 35.3 44.0 26.5 December 22.6 30.7 14.4 Year 45.8 56.5 35.1 UWNR Safety Analysis Report Rev. 2 2-17 Sept. 2008 TABLE 2-3 Monthly Precipitation Data for the Madison AreaAverage Monthly Total Average Monthly Precipitation Maximum Precipitation (inches) (inches)January 1.14 2.45 February 1.14 2.77 March 2.17 5.04 April 3.02 7.11 May 3.14 6.26 June 3.83 9.95 July 4.05 10.93 August 3.90 9.49 September 3.12 9.22 October 2.29 5.63 November 2.15 5.13 December 1.66 4.09 UWNR Safety Analysis Report Rev. 2Sept. TABLE 2-3 Monthly Precipitation Data for the Madison Area
Average Monthly Total Average Monthly Precipitation Maximum Precipitation (inches) (inches) January 1.14 2.45 February 1.14 2.77 March 2.17 5.04 April 3.02 7.11 May 3.14 6.26 June 3.83 9.95 July 4.05 10.93 August 3.90 9.49 September 3.12 9.22 October 2.29 5.63
November 2.15 5.l3 December 1.66 4.09 UWNR Safety Analysis Report Rev. 2 2-18
Sept. 2008 TABLE 2-4Monthly Snowfall Data for the Madison AreaMonthly Average Monthly Maximum Year Total Snowfall Snowfall (inches) (inches)January 10.3 31.8 1929 February 7.7 37.0 1994 March 8.5 28.4 1923 April 2.3 17.4 1973 May 0.1 1935 June 0 July 0 August 0 September Trace October 0.2 5.0 1917 November 3.8 18.3 1985 December 10.9 32.8 1987 2.3.2.3 Winds The average annual wind speed in the Madison area is 9.8 mph. The prevailing winds during the months of November through March are from the west-northwest direction, the remaining months of April though October the prevailing winds are from the south'. Table 2-5 reports the frequency of surface wind direction versus wind speed. The record gust in the Madison area occurred in June 1975 when wind gusts were reported at 83 mph from the west.UWNR Safety Analysis Report Rev. 2Sept. * *
TABLE 2-4 Monthly Snowfall Data for the Madison Area Monthly Average Monthly Maximum Year Total Snowfall Snowfall (inches) (inches) January 10.3 31.8 1929 February 7.7 37.0 1994 March 8.5 28.4 1923 April 2.3 17.4 1973 May 0.1 5.0 1935 June 0 July 0 August 0 September Trace October 0.2 5.0 1917 November 3.8 18.3 1985 December 10.9 32.8 1987 2.3.2.3 Winds The average annual wind speed in the Madison area is 9.8 mph. The prevailing winds during the months of November through March are from the west-northwest direction, the remaining months of April though October the prevailing winds are from the souths. Table 2-5 reports the frequency of surface wind direction versus wind speed. The record gust in the Madison area occurred in June 1975 when wind gusts were reported at 83 mph from the west. UWNR Safety Analysis Report Rev. 2 2-19 Sept. 2008 TABLE 2-5 Frequency of Surface Wind Direction versus Wind Speed Speed (knots)Direction 1 4-6 7-10 11-16 17-21 22-27 28-33 34-40 41-47 >=48 Percent Wind Speed (knots)N 0.5 1.4 1.6 1.3 0.2 * *
0 0 4.7 8.2 NNE 0.3 0.7 1.1 0.8 0.2 *
0 0 0 3.3 9.30.6 1.1 1.4 1.1 0.2 *
0 0 0 4.6 9 ENE 0.6 1.7 1.7 1.2 0.2 *
0 0 0 8.2 E 0.51 1.1 1.4 0.9 0.1 * *
0 0 3.8 7.8 ESE 0.31 0.9 1.2 0.7 0.1
0 0 0 0 3.2 8.5 SE 0.4 1.1 1.5 0.7 0.1
0 0
0 3.8 8.5SSE 0.4 1.4 1.7 1
0 0 0 0 4.9 8.40.6 2.8 4.1 2.9 0.5 0.1 * *
0 10.4 8.8 SSW 0.3 1.1 2.4 2.3 0.4 0.1 *
0 0 6.8 10.1 SW 0.3 1.1 2.4 2 0.4 0.1 * *
0 6.5 10.3 WSW 0.3 1.1 2 1.5 0.3 0.1 * *
0 5.5 10 W. 0.4 1.6 2.5 1.9 0.5 0.1 *
0 0 6.7 9.5 WNW 0.4 1.9 2.9 2.6 0.6 0.1 *
0 0 8.6 9.7 NW 0.6 1.6 2.4 2.5 0.6 0.1 * *
0 8.3 10.1 NNW 0.5 1.6 1.6 1.6 0.3 0.1 *
0 0 5.8 9 VAR 0 0 0 0 0 0 0 0 0 0 0 0 CLM 0 0 0 0 0 0 0 0 0 7.6 0ALL 6.9 22.2 32.1 25.2 4.7 1 0.1 *
0 100 8.5= PERCENT <.05 2.4 Hydrology Madison is located just northeast of the "Drifiless Area" of southwestern Wisconsin.
The glacially shaped topography of Madison and the surrounding area in central Dane County is irregular, ranging from flat or gently rolling to hilly. The most prominent geomorphic features were glacially formed. Among these are Lakes Mendota, Monona, Waubesa, Kegonsa and Wingra. Glacial drift covers the entire area except for local areas of bedrock outcrop 9.In the vicinity of the reactor, a glacial deposit exists which contains clay and large boulders.Although this deposit may be as much as 100 feet thick, it is probably less than twenty. Thereason for the variation in thickness is that the bed-rock sandstone which underlies the deposit is very uneven. The bed-rock consists of Cambrian sandstones which are 700 to 800 feet thick and which are permeable to water.UWNR Safety Analysis Report Rev. 2 2-20 Sept. TABLE 2.:5 Frequency of Surface Wind Direction versus Wind Speed Speed (knots) Direction 1 -3 4-6 7-10 11-16 17-21 22-27 28-33 34-40 41-47 >=48 Percent Wind Speed (knots) N 0.5 1.4 1.6 1.3 0.2 * *
0 0 4.7 8.2 NNE 0.3 0.7 1.1 0.8 0.2 *
0 0 0 3.3 9.3 NE 0.6 1.1 1.4 1.1 0.2 *
0 0 0 4.6 9 ENE 0.6 1.7 1.7 1.2 0.2 *
0 0 0 5.5 8.2 E 0.5 1.1 1.4 0.9 0.1 * *
0 0 3.8 7.8 ESE 0.3 0.9 1.2 0.7 0.1
0 0 0 0 3.2 8.5 SE 0.4 1.1 1.5 0.7 0.1
0 0
0 3.8 8.5 SSE 0.4 104 1.7 1 0.1
0 0 0 0 4.9 804 S 0.6 2.8 4.1 2.9 0.5 0.1 * *
0 lOA 8.8 SSW 0.3 1.1 2.4 2.3 0.4 0.1 *
0 0 6.8 ' 10.1 SW 0.3 1.1 2.4 2 0.4 0.1 * *
0 6.5 10.3 WSW 0.3 1.1 2 1.5 0.3 0.1 * *
0 5.5 10 W. 0.4 1.6 2.5 1:9 0.5 0.1 *
0 0 6.7 9.5 WNW 0.4 1.9 2.9 2.6 0.6 0.1 *
0 0 8.6 9.7 NW 0.6 1.6 2.4 2.5 0.6 0.1 * *
0 8.3 10.1 NNW 0.5 1.6 1.6 1.6 0.3 0.1 *
0 0 5.8 9 VAR 0 0 0 0 0 0 0 0 0 0 0 0 CLM 0 0 0 0 0 0 0 0 0 0 7.6 0 ALL 6.9 22.2 32.1 25.2 4.7 1 0.1 *
0 100 8.5 * = PERCENT < .05 2.4 Hydrology Madison is located just northeast of the "Driftless Area" of southwest em Wisconsin.
The glacially shaped topography of Madison and the surrounding area in central Dane County is irregular, ranging from flat or gently rolling to hilly. The most prominent geomorphic features were glacially formed. Among these are Lakes Mendota, Monona, Waubesa, Kegonsa and Wingra. Glacial drift covers the entire area except for local areas of bedrock outcrop9.
In the vicinity of the reactor, a glacial deposit exists which contains clay and large boulders.
Although this deposit may be as much as 100 feet thick, it is probably less than twenty. The reason for the variation in thickness is that the bed-rock sandstone which underlies the deposit is very uneven. The bed-rock consists of Cambrian sandstones which are 700 to 800 feet thick and which are permeable to water. UWNR Safety Analysis Report Rev. 2 2-20 Sept. 2008 * *
The Yahara River is the main drainage feature in the Madison area. The Yahara River gradient is essentially flat where it flows through the lakes in Madison. Water surface elevations are controlled mostly by dams at the outlets of lakes Mendota and Waubesa. The ground water flowfrom the reactor site is generally toward the Lake Mendota -Yahara River -Lake Monona system. Thus, the general flow is toward the east and south. Historical data obtained from the USGS Wisconsin Geological and Natural History Survey 1" indicated the annual average water table in the vicinity of the reactor is approximately 60 feet below the surface.Madison obtains its drinking water supply from several deep well aquifers drilled, typically several hundred feet, into the Cambrian sandstone described above. The location of these wells is shown on Figure 2-12, and they supply the University as well as the city. All of these wells are cased from ground level into the sandstone so as to keep out ground water from the glacial deposit. The closest well to the reactor is Madison City Well 27 located 2,000 feet southeast.
Due to the large drainage capacity of the Yahara river and the outlet dams of lakes Mendota andWausbesa flooding is not a serious problem in most of Madison. The 100 year flood is estimated to increase the surface level of Lake Mendota approximately 3 feet causing the lake to over flow its existing banks by about 30 feet 1 1.UWNR Safety Analysis Report Rev. 2 2-21 Sept. * *
The Yahara River is the main drainage feature in the Madison area. The Yahara River gradient is essentially flat where it flows through the lakes in Madison. Water surface elevations are controlled mostly by dams at the outlets oflakes Mendota and Waubesa. The ground water flow from the reactor site is generally toward the Lake Mendota -Yahara River -Lake Monona system. Thus, the general flow is toward the east and south. Historical data obtained from the USGS Wisconsin Geological and Natural History SurveylO indicated the annual average water table in the vicinity of the reactor is approximately 60 feet below the surface. Madison obtains its drinking water supply from several deep well aquifers drilled, typically several hundred feet, into the Cambrian sandstone described above. The location of these wells is shown on Figure 2-12, and they supply the University as well as the city. All of these wells are cased from ground level into the sandstone so as to keep out ground water from the glacial deposit. The closest well to the reactor is Madison City Well 27 located 2,000 feet southeast.
Due to the large drainage capacity of the Yahara river and the outlet dams of lakes Mendota and Wausbesa flooding is not a serious problem in most of Madison. The 100 year flood is estimated to increase the surface level of Lake Mendota approximately 3 feet causing the lake to over flow its existing banks by about 30 feet!! . UWNR Safety Analysis Report Rev. 2 2-21 Sept. 2008 Figure 2-12 City of Madison Well Water Supply UWNR Safety Analysis Report Rev. 2 2-22Sept. 2008

Figure 2-12 City of Madison Well Water Supply UWNR Safety Analysis Report Rev. 2 2-22 , Sept. 2008
Geology, Seismology, and Geotechnical Engineering 2.5.1 Regional Geology The Midwestern United States does not lie on, or anywhere near, a tectonic plate boundary.The region is in the middle of the North American Plate, hundreds of miles from both the eastern and western edges. The Midwest, however, has a series of faults around the Mississippi Valley, Figure 2-13, the most active of which is the New Madrid Fault System. Thesefaults were formed by the tearing open of the ancient continental crust almost 5 million years ago. This region is known as the New Madrid Seismic Zone which includes the states of Missouri, Arkansas, Louisiana, Illinois, Indiana and parts of Kentucky and Tennessee12.Wisconsin is not associated with this region. The New Madrid Fault System, located near New Madrid, Missouri is greater than 500 miles from Madison, Wisconsin and poses little or no seismic hazard.Figure 2-13 New Madrid Seismic Zone UWNR Safety Analysis Report Rev. 2Sept. * *
2.5 Geology, Seismology, and Geotechnical Engineering 2.5.1 Regional Geology The Midwestern United States does not lie on, or anywhere near, a tectonic plate boundary.
The region is in the middle of the North American Plate, hundreds of miles from both the eastern and western edges. The Midwest, however, has a series of faults around the Mississippi Valley, Figure 2-13, the most active of which is the New Madrid Fault System. These faults were formed by the tearing open of the ancient continental crust almost 5 million years ago. This region is known as the New Madrid Seismic Zone which includes the states of Missouri, Arkansas, Louisiana, Illinois, Indiana and parts of Kentucky and Tennessee 12* Wisconsin is not associated with this region. The New Madrid Fault System, located near New Madrid, Missouri is greater than 500 miles from Madison, Wisconsin and poses little or no seismic hazard. MISSOURI ,/' ILLINOIS MtSS1SSIPPI
* * (NO (A NA ! TENNESSEE
* * * * * * : ALABAMA * * * * * * * *
Figure 2-13 New Madrid Seismic Zone UWNR Safety Analysis Report Rev. 2 2-23 Sept. 2008 2.5.2 Site Geology As discussed in section 2.4 Hydrology, the local geology of the reactor site includes a layer of glacial deposits which rests on the Cambrian sandstone bedrock.
The layer of the glacial deposit is variable, however, soil bores at the reactor site indicate this layer is approximately 16 feet deep, Table 2-613'14,11.
The Cambrian sandstone layer below the glacial drift is approximately 700 to 800 feet thick. Below the sandstone is impermeable basement rock. There are nogeological structures of consequence in the vicinity of the reactor site.TABLE 2-6 Depth of Glacial Deposit to Cambrian SandstoneDistance from Reactor Site Depth of Glacial Deposit (Feet) (Feet)0 1616.54 200 1413.25 300 9.5 350 17.5 400 15 450 9.5 2.5.3 Seismicity As discussed in section 2.5.1, Regional Geology, Wisconsin is located in a geologically stable region of the United States. The closest active fault is greater than 500 miles in the New Madrid Seismic Zone. While no seismic events have occurred at the reactor site, there are records of earthquakes occurring within 200 km of the reactor site. A review of the USGS earthquake database" 6 for all earthquakes of modified Mercalli intensity greater then IV or magnitude greater than 3.0 which is within 200 km of the reactor site resulted in the data reported in Table 2-7.UWNR Safety Analysis Report Rev. 2 2-24 Sept. 2.5.2 Site Geology As discussed in section 2.4 Hydrology, the local geology of the reactor site includes a layer of glacial deposits which rests on the Cambrian sandstone bedrock. The layer of the glacial deposit is variable, however, soil bores at the reactor site indicate this layer is approximately 16 feet deep, Table 2_6 13 ,14,15. The Cambrian sandstone layer below the glacial drift is approximately 700 to 800 feet thick. Below the sandstone is impermeable basement rock. There are no geological structures of consequence in the vicinity of the reactor site. TABLE 2-6 Depth of Glacial Deposit to Cambrian Sandstone Distance from Reactor Site Depth of Glacial Deposit (Feet) (Feet) 0 16 150 16.5 175 4 200 14 250 13.25 300 9.5 350 17.5 400 15 450 9.5 2.5.3 Seismicity As discussed in section 2.5.1, Regional Geology, Wisconsin is located in a geologically stable region of the United States. The closest active fault is greater than 500 miles in the New Madrid Seismic Zone. While no seismic events have occurred at the reactor site, there are records of earthquakes occurring within 200 km of the reactor site. A review of the USGS earthquake database 16 for all earthquakes of modified Mercalli intensity greater then IV or magnitude greater than 3.0 which is within 200 km of the reactor site resulted in the data reported in Table 2-7. UWNR Safety Analysis Report Rev. 2 2-24 Sept. 2008 * *
TABLE 2-7 History of Seismic Events Within 200 km of Madison, Wisconsin.
Date Magnitude Intensity Distance (Richter) (Modified Mercalli) (km)August 20, 1804 4.4 VI 178 May 27, 1881 4.6 VI 198 May 26, 1909 5.1 VII 195 January 2, 1912 4.5 VI 190 November 12, 1934 4.0 VI 196 September 15, 1972 4.5 VI 158 September 9, 1985 3.0 V 178 2.5.4 Maximum Earthquake Potential The determination of earthquake potential and frequency is based on data from previous events.Because there are few historic moderate to large earthquakes in the vicinity of the reactor, analysis is difficult.
The maximum earthquake potential may be inferred from data supplied by the USGS for the 50 year peak ground acceleration estimate" 7.The estimated 50 year peak ground acceleration due to a seismic event in the vicinity at the reactor is less than 0.01 g.2.5.5 Vibratory Ground MotionDue to insufficient data from previous seismic events the vibratory ground motion can only beinferred from the peak ground acceleration data of section 2.5.4 of less than 0.01 g.2.5.6 Surface FaultingBased on the distance to any known active faults and the stable site geology, surface faulting is not considered to be a credible event in the vicinity of the reactor site.2.5.7 Liquefaction PotentialBased on the distance to any known active faults and the stable site geology, the liquefaction potential is considered to be insignificant.
UWNR Safety Analysis Report Rev. 2Sept. * *
TABLE 2-7 History of Seismic Events Within 200 km of Madison, Wisconsin.
Date Magnitude Intensity Distance (Richter) (Modified Mercalli) (km) August 20, 1804 4.4 VI 178 May 27, 1881 4.6 VI 198 May 26,1909 5.1 VII 195 January 2, 1912 4.5 VI 190 November 12, 1934 4.0 VI 196 September 15, 1972 4.5 VI 158 September 9, 1985 3.0 V 178 2.5.4 Maximum Earthquake Potential The determination of earthquake potential and frequency is based on data from previous events . Because there are few historic moderate to large earthquakes in the vicinity of the reactor, analysis is difficult.
The maximum earthquake potential may be inferred from data supplied by the USGS for the 50 year peak ground acceleration esti mate 17. The estimated 50 year peak ground acceleration due to a seismic event in the vicinity at the reactor is less than 0.01 g. 2.5.5 Vibratory Ground Motion Due to insufficient data from previous seismic events the vibratory ground motion can only be inferred from the peak ground acceleration data of section 2.5.4 of less than 0.01 g. 2.5.6 Surface Faulting Based on the distance to any known active faults and the stable site geology, surface faulting is not considered to be a credible event in the vicinity of the reactor site. 2.5.7 Liquefaction Potential Based on the distance to any known active faults and the stable site geology, the liquefaction potential is considered to be insignificant.
UWNR Safety Analysis Report Rev. 2 2-25 Sept. 2008 References
1. Prepared by the Applied Population Laboratory, UW-Madison/Extension. Source: U.S.
Census Bureau; 2000 Census of Population and Housing, summary File 1 (SF1);generated by Applied Population Laboratory using Census 2000 Summary File 1 (SF1)Wisconsin CD-ROM.2. Meighan, Ben. Union and Southern Railroad Company. Telephone Interview.
January 24, 2000.3. Lenss, Marty. Director of Operations/Public Safety, Dane County Regional Airport.Email Interview (lenss@msnairport.com).
March 27, 2007.4. Donovan, Greg. Telephone Interview.
February 5, 2000.International Station Meteorological Climate Summary. Version 4.0. US ISMCS StationClimatic Narrative. National Climatic Data Center, September 1996.6. International Station Meteorological Climate Summary. Version 4.0. US ISMCS StationDatabase. National Climatic Data Center, September 1996.7. Huff, F.A., and Angel, J.R.. Rainfall Frequency Atlas of the Midwest. Midwestern Climate Center (MCC) Research Report 92-03. 1992.8. Knox, P.N.
Wind Atlas of Wisconsin.
Wisconsin Geological and Natural History Survey.Bulletin 94. 1996.9. Flood Insurance Study. Federal Emergency Management Agency. Community Number 550083. March 5, 1996.10. Groundwater Observation Network. Wisconsin Geology and Natural History Survey.September 13, 1999. http://wi.water.usgs.gov/gw/
11. National Flood Insurance Program. Flood Insurance Rate Map Community Panel Number 550083 0010 D. March 8, 1974.12. Atkinson, William. The Next New Madrid Earthquake.
Southern Illinois University Press. Carbondale IL,1989.13. Engineering Centers Facility Soil Boring Log. Woodward-Clyde International-America.
Waukesha, WI. July 9, 1998.UWNR Safety Analysis Report Rev. 2Sept. 2.6 References
1. Prepared by the Applied Population Laboratory, UW-MadisonlExtension.
Source: U.S. Census Bureau; 2000 Census of Population and Housing, summary File 1 (SF1); generated by Applied Population Laboratory using Census 2000 Summary File 1 (SF1) Wisconsin CD-ROM. 2. Meighan, Ben. Union and Southern Railroad Company. Telephone Interview.
January 24, 2000. 3. Lenss, Marty. Director ofOperations/Public Safety, Dane County Regional Airport. Email Interview(lenss@msnairp011.com).
March 27, 2007. 4. Donovan, Greg. Telephone Interview.
February 5, 2000. 5. International Station Meteorological Climate Summary. Version 4.0. US ISMCS Station Climatic Narrative.
National Climatic Data Center, September 1996. 6. International Station Meteorological Climate Summary. Version 4.0. US ISMCS Station Database.
National Climatic Data Center, September 1996. 7. Huff, F.A., and Angel, J.R .. Rainfall Frequency Atlas of the Midwest. Midwestern Climate Center (MCC) Research Report 92-03. 1992. 8. Knox, P.N. Wind Atlas of Wisconsin.
Wisconsin Geological and Natural History Survey. Bulletin 94. 1996. 9. Flood Insurance Study. Federal Emergency Management Agency. Community Number 550083. March 5, 1996. 10. Groundwater Observation Network. Wisconsin Geology and Natural History Survey. September 13, 1999. http://wi.water.usgs.gov/gw/
11. National Flood Insurance Program. Flood Insurance Rate Map Community Panel Number 5500830010 D. March 8, 1974. 12. Atkinson, William. The Next New Madrid Earthquake.
Southern Illinois University Press. Carbondale IL, 1989. 13. Engineering Centers Facility Soil Boring Log. Woodward-Clyde International-America.
Waukesha, WI. July 9, 1998. UWNR Safety Analysis Report Rev. 2 2-26 Sept. 2008 * * *
14. Preliminary Soils Exploration of Lot 17 Parking Structure.
Soils & Engineering Services Inc. Madison, WI. August 4, 1992.15. Subsurface Exploration of Proposed Lot 17 Parking Structure.
Terracon Consultants Inc.Naperville, IL. August 18, 1997.16. National Earthquake Information Center. United States Geological Survey.January 18, 2000. http://wwwneic.cr.usgs.gov/neis/epic/epic.html
17. National Seismic Hazards Mapping Project. United States Geological Survey.November 19, 1999. http://geohazards.cr.usgs.gov/eg UWNR Safety Analysis Report Rev. 2Sept. * *
14 . Preliminary Soils Exploration of Lot 17 Parking Structure.
Soils & Engineering Services Inc. Madison, WI. August 4, 1992. 15. Subsurface Exploration of Proposed Lot 17 Parking Structure.
Terracon Consultants Inc. Naperville, IL. August 18, 1997. 16. National Earthquake Information Center. United States Geological Survey. January 18, 2000. http://wwwneic.cr.
usgs.gov/neis/epic/epic.html
17. National Seismic Hazards Mapping Project. United States Geological Survey. November 19,1999. http://geohazards.cr.usgs.gov/eq UWNR Safety Analysis Report Rev. 2 2-27 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2Sept.
This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 2 2-28 Sept. 2008
OF Design CriteriaWhen the University of Wisconsin Nuclear Reactor was to be upgraded by increasing authorized power to 1000 kW the principal design criterion was to assure the facility could withstand loss of pool water and any other credible accident with no hazard to the public, without reliance on engineered safety features.
This criterion was met by selecting stainless-steel-clad TRIGA fuel due to the well-documented characteristics of this fuel type'. Details of the physical mechanisms and characteristics that cause TRIGA and FLIP fuel to exhibit the prompt negative temperature coefficient responsible for the fuel characteristics are given in the reference and a number ofother documents and are not included here. Detailed analysis for this facility (Chapter 13 of this report) agree with the conclusions in the reference. The design criteria that result in this negligible safety risk are the result of the fuel composition and cladding, not of specific features provided in the equipment and building that surrounds the reactor.
The Mechanical Engineering Building which houses the reactor laboratory was completed in 1930. Extensive remodeling of the north, east, and west wings, and construction of a new centerwing was completed in 2007. Extensive remodeling of the room that became the Reactor Laboratory took place in 1960 when the reactor was installed.
The original reactor installation used fuel and components manufactured by General Electric, and the specifications to which structures were built were those stated by General Electric. Specific design criteria were not stated. The original ventilation system was designed in 1962 and installed in 1963. The design specifications stated only the desired flow rates and stack height. The current ventilation system was installed in 2006. The design specifications included the desired flow rates and stack height, and are detailed in section 9.1. The original cooling system was designed in 1966 and installed in 1967. The only design specification was the heat removal rate required.
The current cooling system was installed in 2002. The design specifications included the heat removal rate required, and are detailed in chapter 5. During the 1966 upgrade, the diffuser system was alsoinstalled. This system was designed and fabricated by General Atomic to their design specifications.
Conversion to TRIGA fuel took place in 1969, and the auxiliaries for pulsing operation (transient rod and drive) were designed and built by General Atomic to their specifications.
All building modifications and equipment additions were in conformance with the building codes in existence at that time.Meteorological Damage There are no design criteria for the protection of facility structures from meteorologicalconditions except that all facility structures were constructed to applicable building codes in existence at the time. The Reactor Laboratory has endured approximately 50 years of Wisconsin weather with no meteorological damage. Furthermore, no facility structures are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis).
UWNR Safety Analysis Report Rev. 2Sept. * *
3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Design Criteria When the University of Wisconsin Nuclear Reactor was to be upgraded by increasing authorized power to 1000 kW the principal design criterion was to assure the facility could withstand loss of pool water and any other credible accident with no hazard to the public, without reliance on engineered safety features.
This criterion was met by selecting stainless-steel-clad TRlGA fuel due to the well-documented characteristics of this fuel type l. Details ofthe physical mechanisms and characteristics that cause TRlGA and FLIP fuel to exhibit the prompt negative temperature coefficient responsible for the fuel characteristics are given in the reference and a number of other documents and are not included here. Detailed analysis for this facility (Chapter 13 ofthis report) agree with the conclusions in the reference.
The design criteria that result in this negligible safety risk are the result of the fuel composition and cladding, not of specific features provided in the equipment and building that surrounds the reactor. The Mechanical Engineering Building which houses the reactor laboratory was completed in 1930. Extensive remodeling of the north, east, and west wings, and construction ofa new center wing was completed in 2007. Extensive remodeling of the room that became the Reactor Laboratory took place in 1960 when the reactor was installed.
The original reactor installation used fuel and components manufactured by General Electric, and the specifications to which structures were built were those stated by General Electric.
Specific design criteria were not stated. The original ventilation system was designed in 1962 and installed in 1963. The design specifications stated only the desired flow rates and stack height. The current ventilation system was installed in 2006. The design specifications included the desired flow rates and stack height, and are detailed in section 9.1. The original cooling system was designed in 1966 and installed in 1967. The only design specification was the heat removal rate required.
The current cooling system was installed in 2002. The design specifications included the heat removal rate required, and are detailed in chapter 5. During the 1966 upgrade, the N I6 diffuser system was also installed.
This system was designed and fabricated by General Atomic to their design specifications.
Conversion to TRlGA fuel took place in 1969, and the auxiliaries for pUlsing operation (transient rod and drive) were designed and built by General Atomic to their specifications.
All building modifications and equipment additions were in conformance with the building codes in existence at that time. 3.2 Meteorological Damage There are no design criteria for the protection of facility structures from meteorological conditions except that all facility structures were constructed to applicable building codes in existence at the time. The Reactor Laboratory has endured approximately 50 years of Wisconsin weather with no meteorological damage. Furthermore, no facility structures are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis) . UWNR Safety Analysis Report Rev. 2 3-1 Sept. 2008 Water Damage There are no design criteria for the protection of facility structures, systems and components from water damage. The possibility of flooding due to the lake system of Dane County is considered insignificant due to the distance of the Reactor Laboratory to the Lake Mendota flood plain as described in Chapter 2, section 2.4, Hydrology.
Furthermore, no facility structures, systems andcomponents susceptible to water damage are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis).
3.4 Seismic Damage There are no design criteria for the protection of facility structures, systems and components from seismic damage except that all facility structures were constructed to applicable building codes in existence at the time. The probability of a seismic event in the vicinity of the reactor site is considered insignificant due to the stable regional geology (see Chapter 2, section 2.5, Geology, Seismology and Geotechnical Engineering).
Furthermore, no facility structures, systems andcomponents, including the reactor pool, susceptible to seismic damage are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis).
3.5 Systems and Components At the time of original construction of the Reactor Laboratory, design bases were not provided by General Electric for facility systems and components.
With the upgrade to TRIGA and TRIGA-FLIP fuel, accident analyses, including NUREG/CR 2387 and Chapter 13, show that by the design of TRIGA fuel, reliance upon other systems, structures and components are not necessary to ensure safety of the general public. Therefore, with the exception of the fuel, no other facility structure, system or component is assumed to be operable in this SAR for the mitigation of any accident.Nevertheless, experience gained on facility systems and components over many years of operation have shown these systems to be highly reliable. Descriptions of system design and operation of these systems are discussed in the succeeding chapters of this SAR.3.6 References 1.. NUREG/CR2387, Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, Hawley and Kathren, Pacific Northwest Laboratory, April 1982 UWNR Safety Analysis Report Rev. 0April 2000 3.3 Water Damage There are no design criteria for the protection of facility structures, systems and components from water damage. The possibility of flooding due to the lake system of Dane County is considered insignificant due to the distance of the Reactor Laboratory to the Lake Mendota flood plain as described in Chapter 2, section 2.4, Hydrology.
Furthermore, no facility structures, systems and components susceptible to water damage are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis).
3.4 Seismic Damage There are no design criteria for the protection of facility structures, systems and components from seismic damage except that all facility structures were constructed to applicable building codes in existence at the time. The probability of a seismic event in the vicinity ofthe reactor site is considered insignificant due to the stable regional geology (see Chapter 2, section 2.5, Geology, Seismology and Geotechnical Engineering).
Furthermore, no facility structures, systems and components, including the reactor pool, susceptible to seismic damage are assumed to be operable in this SAR for the mitigation of any accident (see Chapter 13, Accident Analysis).
3.5 Systems and Components
At the time of original construction ofthe Reactor Laboratory, design bases were not provided by
General Electric for facility systems and components.
With the upgrade to TRIGA and TRIGA-FLIP fuel, accident analyses, including NUREG/CR 2387 and Chapter 13, show that by the design ofTRIGA fuel, reliance upon other systems, structures and components are not necessary to ensure safety of the general public. Therefore, with the exception of the fuel, no other facility structure, system or component is assumed to be operable in this SAR for the mitigation of any accident.
Nevertheless, experience gained on facility systems and components over many years of operation have shown these systems to be highly reliable.
Descriptions of system design and operation of these systems are discussed in the succeeding chapters of this SAR. 3.6 References
1. NUREG/CR2387, Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors, Hawley and Kathren, Pacific Northwest Laboratory, April 1982 UWNR Safety Analysis Report Rev. 0 3-2 April 2000
4 REACTOR 4.1 Summary Description 4.1.1 Introduction The reactor was constructed and installed by the Atomic Power Equipment Department of the General Electric Company. The present modification employs a core composed of TRIGA-FLIP fuel supplied by the General Atomic Company.Initial criticality was achieved on March 26 th 1961. The original maximum steady state power level was 10 kW. Power was increased to 250 kW on December 7 th 1964 and again increased to the present maximum steady state power level of 1,000 kW on November 1 4 th 1967. Operation with FLIP fuel began in March 1974 with a mixed core containing 9 FLIP bundles. In January 1978 an additional 6 FLIP bundles were added. In August 1979 the conversion to FLIP fuel core was completed.
Figure 4-1 is a pictorial cutaway view of the reactor. The reactor is a heterogeneous pool-type, fueled with TRIGA or TRIGA-FLIP fuel which is cooled by natural convection.
The fuel is currently all 70% enriched in Uranium U23 5 , although 20% enriched fuel can also be used. Light water acts as both coolant and moderator as well as being a biological shield. The core is reflected on two sides by graphite and on two sides by water, the water-reflected areas being utilized as irradiation facility locations.
The top and bottom reflector region is partially graphite and partially water.A 7-by-9 grid, surrounded by a core box, positions fuel, reflectors, control elements, and irradiation facilities.
Core reactivity is changed and controlled by three shim safety blades, aregulating blade, and a transient control rod. All control elements move vertically in shrouds positioned in the core box or inside a fixed tube as is the case of the transient rod. A suspension frame supports the grid box and control element drive mechanisms.
The suspension frame, in turn, is supported by the reactor bridge.
Cold, clean core excess reactivity in the present operational core is about 4.3 % Ak/k. Control elements (control blades and the transient rod) provide a shutdown margin of about 4.2 % Ak/k.UWNR Safety Analysis Report Rev. 2 4-1 Sept. * *
4 REACTOR DESCRIPTION 4.1 Summary Description 4.1.1 Introduction The reactor was constructed and installed by the Atomic Power Equipment Department of the General Electric Company. The present modification employs a core composed ofTRIGA-FLIP fuel supplied by the General Atomic Company. Initial criticality was achieved on March 26 th 1961. The original maximum steady state power level was 10 kW. Power was increased to 250 kW on December 7th 1964 and again increased to the present maximum steady state power level of 1,000 kW on November 14th 1967. Operation with FLIP fuel began in March 1974 with a mixed core containing 9 FLIP bundles. In January 1978 an additional 6 FLIP bundles were added. In August 1979 the conversion to FLIP fuel core was completed.
Figure 4-1 is a pictorial cutaway view of the reactor. The reactor is a heterogeneous pool-type, fueled with TRIGA or TRIGA-FLIP fuel which is cooled by natural convection.
The fuel is currently all 70% enriched in Uranium U m , although 20% enriched fuel can also be used. Light water acts as both coolant and moderator as well as being a biological shield. The core is reflected on two sides by graphite and on two sides by water, the water-reflected areas being utilized as irradiation facility locations.
The top and bottom reflector region is partially graphite and partially water. A 7-by-9 grid, surrounded by a core box, positions fuel, reflectors, control elements, and irradiation facilities.
Core reactivity is changed and controlled by three shim safety blades, a regulating blade, and a transient control rod. All control elements move vertically in shrouds positioned in the core box or inside a fixed tube as is the case of the transient rod. A suspension frame supports the grid box and control element drive mechanisms.
The suspension frame, in tum, is supported by the reactor bridge. Cold, clean core excess reactivity in the present operational core is about 4.3 % ilklk. Control elements (control blades and the transient rod) provide a shutdown margin of about 4.2 % ilklk . UWNR Safety Analysis Report Rev. 2 4-1 Sept. 2008 Figure 4-1 University of Wisconsin Nuclear Reactor (UWNR)UWNR Safety Analysis Report Rev. 0 4-2 April 2000*
Figure 4-1 University of Wisconsin Nuclear Reactor (UWNR) UWNR Safety Analysis Report Rev. 0 4-2 April 2000
4.1.2 Summary of Reactor Data Responsible Organization Location Purpose Fuel Type Number of elements in standard 1000 kW core The University of Wisconsin Madison, Wisconsin Teaching and ResearchTRIGA Hydride in 4 element clusters Control Safety elements Regulating-servo element Transient control Experimental Facilities Three vertical blades One vertical blade One rod Thermal Column One, 40-inch square graphite, dO th 2 x 10 8 nvBeam Ports Four, 6-inch diameterth = 1 -3 x 1010 nv at shield side of shutter;about 8 x 1011 nv at core end of port Pneumatic tube One, 2-inch (sample size 1-1/4 inch diameter by 5-1/2 inches long), d4 th = 5 x 1012 nvThermal neutron fluxes for isotope production include the above, plus large irradiation spaces outside the core with thermal neutron fluxes of around 1.3 x nv.Reactor Materials Fuel -moderator material U-Zr H 1.6 UWNR Safety Analysis Report Rev. 2 4-3 Sept. * *
4.1.2 Summary of Reactor Data Responsible Organization Location Purpose Fuel Type Number of elements in standard 1000 k W core Control Safety elements Regulating-servo element Transient control Experimental Facilities Thermal Column Beam Ports Pneumatic tube The University of Wisconsin Madison, Wisconsin Teaching and Research TRIGA Hydride in 4 element clusters
Three vertical blades One vertical blade One rod One, 40-inch square graphite, th = 2 x 10 8 nv Four, 6-inch diameter th = 1 -3 x 10 10 nv at shield side of shutter; about 8 x lOll nv at core end of port One, 2-inch (sample size 1-1/4 inch diameter by 5-1/2 inches long), th = 5 x 10 12 nv Thermal neutron fluxes for isotope production include the above, plus large irradiation spaces outside the core with thermal neutron fluxes of around 1.3 x 1013 nv. Reactor Materials Fuel -moderator material U-Zr H L6 UWNR Safety Analysis Report Rev. 2 4-3 Sept. 2008 U235 enrichment 70%U235 content/element (average)
Burnable poison 1.5wt.% 1'Cladding 20 mil sta Reflector Water an Coolant Light wat Control Boral & s transient Structural material Aluminur Shield Concrete Dimensions Pool 8 x 12 x 2 Standard 1000 kW Core 15 x 17 x Grid box 9 x 7 arra Fuel element Diameter Length Active Length Nuclear characteristics 1 MW Steady State:
Maximum thermal neutron flux Maximum fast neutron flux qatural Erbium inless steelgraphite er tainless steel; borated graphite for rod n and water 27-1/2 ft. deep 15 inches high y of 3 inch modules 3.2 x nvx 1013 nv UWNR Safety Analysis Report Rev. 0 4-4 April 2000 U 235 enrichment U 235 content/element (average)
Burnable poison Cladding Reflector Coolant Control Structural material Shield Dimensions Pool Standard 1000 kW Core Grid box Fuel element Diameter Length Active Length Nuclear characteristics 1 MW Steady State: 70% -1.5wt.% Natural Erbium 20 mil stainless steel Water and graphite Light water Boral & stainless steel; borated graphite for transient rod Aluminum Concrete and water 8 x 12 x 27-112 ft. deep 15 x 17 x 15 inches high 9 x 7 array of 3 inch modules --Maximum thermal neutron flux 3.2 X 10 13 nv Maximum fast neutron flux 3.0 X 1013 nv UWNR Safety Analysis Report Rev. 0 4-4 April 2000 . * *
1000 MW Pulse Maximum thermal neutron flux Maximum fast neutron flux Core Loading (Standard 1000 kW core)Operating excess reactivity Reactivity in control blades Average prompt temperature coefficient of reactivityVoid coefficient of reactivity Prompt neutron lifetime Effective delayed neutron fraction 3.2 x 1016 nv 3.0 x 1016 nv-4.3% AKeff-7.1% AKeff-1.26 x 10-4 AK/°C-.2 x 10-4 void 2.4 x second 0.007 4.1.3 Experimental Facilities Facilities are provided to permit use of radiations from the reactor in experimental work without endangering personnel. These facilities include three hydraulic irradiation facilities
("whales"), four beam ports, one thermal column, and a pneumatic transfer system ("rabbit").
4.1.3.1Hydraulic Irradiation Facility (Whale)Aluminum pipes of 2-7/16" internal diameter extend from approximately 18" below the pool surface to grid box positions on the periphery of the core. These pipes draw sample bottles made of polyethylene down and position them approximately at the center line of the fuel.
Two sample containers can be loaded in each tube. The addition of a second sample bottle, however, causes the natural rotation of the first bottle to stop. Thermal Neutron Fluxes in these positions are approximately 1013 nv.Irradiations are also conducted in the other reflector regions surrounding the core.
Irradiation baskets may be inserted in any vacant grid position, and irradiations can also be conducted outside the grid box in specially fabricated enclosures.
UWNR Safety Analysis Report Rev.
2 4-5 Sept. * *
1000 MW Pulse Maximum thermal neutron flux Maximum fast neutron flux Core Loading (Standard 1000 k W core) Operating excess reactivity Reactivity in control blades Average prompt temperature coefficient of reactivity Void coefficient of reactivity Prompt neutron lifetime Effective delayed neutron fraction 4.1.3 Experimental Facilities 3.2 X 10 16 nv ** 11
-1.26 x 1 0-4
-.2 x 10-4 SK/% void 2.4 x 10-5 second 0.007 Facilities are provided to permit use of radiations from the reactor in experimental work without endangering personnel.
These facilities include three hydraulic irradiation facilities
("whales"), four beam ports, one thermal column, and a pneumatic transfer system ("rabbit").
4.1.3.1 Hydraulic Irradiation Facility (Whale) Aluminum pipes of2-7116" internal diameter extend from approximately 18" below the pool surface to grid box positions on the periphery of the core. These pipes draw sample bottles made of polyethylene down and position them approximately at the center line of the fuel. Two sample containers can be loaded in each tube. The addition of a second sample bottle, however, causes the natural rotation of the first bottle to stop. Thermal Neutron Fluxes in these positions are approximately 10 13 nv. Irradiations are also conducted in the other reflector regions surrounding the core. Irradiation baskets may be inserted in any vacant grid position, and irradiations can also be conducted outside the grid box in specially fabricated enclosures . UWNR Safety Analysis Report Rev. 2 4-5 Sept. 2008
The thermal column is a graphite-filled horizontal penetration through the biological shieldwhich provides neutrons in the thermal energy range (about eV) for irradiation experiments.
The column, which is about feet long, is filled with about feet of graphite. small experimental air chamber (40" x 40" x 24") between the face of the graphite and the thermal column door has conduits for service connections (air, water, electricity) to the biological shield face. The compensated ion chambers for the safety and logN instrumentation channels are located in the thermal column.Personnel in the building are protected against gamma radiation from the column a denseconcrete door which closes the column at the biological shield. The door moves on tracks set into the concrete floor perpendicular to the shield face.4.1.3.3 Beam Ports Four 6-inch beam ports penetrate the shield and provide fluxes of both fast and thermal neutrons for experimental use. The ports are air filled tubes, welded shut at the core ends and provided with water-tight covers on the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel.shutter assembly, made of lead encased in aluminum, is opened for irradiations a lifting device. When closed, the shutter shields against gamma rays from the shut-down core, allowing experiments to be loaded and unloaded without excessive radiation exposure to personnel.Shielding plugs are installed in the outer end of each port. The plugs, made of dense concrete in aluminum casings, have spiral conduits for passage of instrument leads.pneumatic tube system conveys samples from a basement room to an irradiation position beside the core. The "rabbits" used in the system will convey samples up to 1-1/4 inches diameter and inches long. The system operates as a closed loop with carbon dioxide cover gas limiting generation of Ar 4' activity.The reactivity effect from a sample in the pneumatic tube is restricted to less than 0.2% Testsrun with water and cadmium samples indicate that sample reactivity effects will normally be less than Static reactivity measurements will be run for samples of fissionable material or particularly strong absorbers such as some of the rare earths. The reactor may be operated with either standard (20% enriched) enriched) fuel as described in section 4.2. In addition, mixed cores containing fuel of both types may be loaded..UXVNR Safety Analysis Report Rev. 2 46Sp.20 4-6 Sept. 4.1.3.2 Thennal Column The thennal column is a graphite-filled horizontal penetration through the biological shield which provides neutrons in the thennal energy range (about 0.025 eV) for irradiation experiments.
The column, which is about 8 feet long, is filled with about 6 feet of graphite.
A small experimental air chamber (40" x 40" X 24") between the face of the graphite and the thennal column door has conduits for service connections (air, water, electricity) to the biological shield face. The compensated ion chambers for the safety and 10gN instrumentation channels are located in the thennal column. Personnel in the building are protected against gamma radiation from the column by a dense concrete door which closes the column at the biological shield. The door moves on tracks set into the concrete floor perpendicular to the shield face. 4.1.3.3 Beam Ports Four 6-inch beam ports penetrate the shield and provide fluxes of both fast and thennal neutrons for experimental use. The ports are air filled tubes, welded shut at the core ends and provided with water-tight covers on the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel. A shutter assembly, made of lead encased in aluminum, is opened for irradiations by a lifting
device. When closed, the shutter shields against gamma rays from the shut-down core, allowing
experiments to be loaded and unloaded without excessive radiation exposure to personnel.
Shielding plugs are installed in the outer end of each port. The plugs, made of dense concrete in aluminum casings, have spiral conduits for passage of instrument leads. 4.1.3.4 Pneumatic Tube A pneumatic tube system conveys samples from a basement room to an irradiation position beside the core. The "rabbits" used in the system will convey samples up to 1-114 inches diameter and 5-112 inches long. The system operates as a closed loop with carbon dioxide cover gas limiting generation of Ar 41 activity.
The reactivity effect from a sample in the pneumatic tube is restricted to less than 0.2% p. Tests run with water and cadmium samples indicate that sample reactivity effects will nonnally be less than 0.01 % p. Static reactivity measurements will be run for samples of fissionable material or particularly strong absorbers such as some of the rare earths.
4.2 Reactor Core The reactor may be operated with either standard (20% enriched) or FLIP (70% enriched) fuel as described in section 4.2.1. In addition, mixed cores containing fuel of both types may be loaded . UWNR Safety Analysis Report Rev. 2 4-6 Sept. 2008
Since funding is not presently available for replacing the FLIP fuel with LEU with burnable poison, the core that will usually be operated is composed only of FLIP fuel in order to maintain the radiation levels of this fuel above the point at which it is self-protecting.
The timing of funding for LEU fuel is unknown at present. If funding for converting the entire core is received it may be necessary to revert to cores of mixed standard and FLIP cores before loading a new core in order to assure that less than a formula quantity of (as defined in 10 CFR 73.2)becomes non-self-protecting.
If funding for LEU replacement is received a few fuel bundles at a time, reversion to a mixed standard/FLIP core may not be required, but further analysis of cores of mixtures of FLIP and the new LEU fuel would be required.
Such cores are not considered in this SAR.The use of the reactor as a training and research tool requires flexibility of core arrangement.
These arrangements are subject, however, to the following criteria:
a. A mixed core must contain at least 9 FLIP bundles.b. Any FLIP fuel must be located in a central contiguous region.c. The core must be a close packed array except for single fuel element (not fuel bundle) positions or grid positions on the core periphery.
d. Calculations indicate that operation will be within safety limits on power generation per element and fuel temperature.
UWNR Safety Analysis Report Rev. 2 4-7-Sept. * *
Since funding is not presently available for replacing the FLIP fuel with LEU with burnable poison, the core that will usually be operated is composed only of FLIP fuel in order to maintain the radiation levels of this fuel above the point at which it is self-protecting.
The timing of funding for LEU fuel is unknown at present. If funding for converting the entire core is received it may be necessary to revert to cores of mixed standard and FLIP cores before loading a new core in order to assure that less than a formula quantity ofHEU (as defined in 10 CFR 73.2) becomes non-self-protecting.
If funding for LEU replacement is received a few fuel bundles at a time, reversion to a mixed standard/FLIP core may not be required, but further analysis of cores of mixtures of FLIP and the new LEU fuel would be required.
Such cores are not considered in this SAR. The use of the reactor as a training and research tool requires flexibility of core arrangement.
These arrangements are subject, however, to the following criteria:
a. A mixed core must contain at least 9 FLIP bundles. b. Any FLIP fuel must be located in a central contiguous region. c. The core must be a close packed array except for single fuel element (not fuel bundle) positions or grid positions on the core periphery.
d. Calculations indicate that operation will be within safety limits on power generation per element and fuel temperature.
UWNR Safety Analysis Report Rev. 2 4-7 Sept. 2008 4.2.1 Reactor Fuel The fuel is of the TRIGA four-element bundle type developed to provide a simple means ofconverting reactors using flat-plate fuel to TRIGA reactors.
A variant bundle, called athree-element bundle, has only three fuel elements installed; the fourth space is used for an aluminum control rod guide tube or an instrumented fuel element. Figure 4-2 shows a four-element bundle, a three-element bundle containing a control rod guide tube, and a three-element bundle containing an instrumented fuel element (the conduit for the thermocouple leads is shown cut short)..Figure 4-2 Fuel Bundles Sept. 2008 UWNR Safety Analysis Report Rev. 2 4-8 4.2.1 Reactor Fuel The fuel is of the TRIGA four-element bundle type developed to provide a simple means of converting reactors using flat-plate fuel to TRIGA reactors.
A variant bundle, called a three-element bundle, has only three fuel elements installed; the fourth space is used for an aluminum control rod guide tube or an instrumented fuel element. Figure 4-2 shows a four-element bundle, a three-element bundle containing a control rod guide tube, and a three-element bundle containing an instrumented fuel element (the conduit for the thermocouple leads is shown cut short) .. UWNR Safety Analysis Report Rev. 2 Figure 4-2 4-8 *
Fuel Bundles
Sept. 2008 The four-element bundle (Figure 4-3) consists of bottom adapter, top adapter, and four TRIGAelements. The bottom adapter of the bundle fits the existing grid plate as did the original flat-plate fuel elements.
The end fittings on individual TRIGA elements are threaded into the bottom adapter until a flange on the element seats firmly against the adapter, providing rigid cantilever-type support. The top adapter serves both as a handling fitting and as a spacer for the upper ends of the fuel elements.
A sliding fit between this adapter and the fuel element end fittings allows for differential expansion of the elements. This top fitting can be removed with remote handling tools to disassemble the bundles for replacement of individual fuel elements or for shipping spentelements for reprocessing.
The individual fuel elements (Figure 4-4) are quite similar to the TRIGA elements used on TRIGA reactors using the standard TRIGA grid plates. The differences are (1) reduction of diameter from inches to maintain the proper metal-to-water ratio in this core; (2) the bottom end fixture is threaded; (3) flats on the stainless steel bottom end fixture provide wrench surfaces for disassembly without stressing the cladding; and (4) the top end fixture is modified to allow the top end fitting to be locked in place.The TRIGA elements used at UWNR are of two types, standard and FLIP. Both have outside dimensions, clad thickness, and construction as shown in Figure 4-4. The two types differ asshown in the following table:
Design Parameter Standard Fuel FLIP Fuel Fuel moderator material U-Zr H 1.7 U-Zr U235 enrichment 20% 70%U235 content/element (average)Burnable Poison None Natural erbium Erbium content 1.5 wt The FLIP fuel was designed to extend the lifetime of TRIGA fuel, and was used in step-wise additions of fuel to the University of Wisconsin Nuclear Reactor. The reactor is currently operating with a core consisting entirely of FLIP fuel. Fuel bundles contain only one type of fuel, and the top adapters for FLIP fuel bundles are marked (by notches machined into the top of thetop adapter) to facilitate identification during underwater fuel handling.UWNR Safety Analysis Report Rev. 0 4-9 April 2000* *
The four-element bundle (Figure 4-3) consists of bottom adapter, top adapter, and four TRIGA elements.
The bottom adapter of the bundle fits the existing grid plate as did the original plate fuel elements.
The end fittings on individual TRIGA elements are threaded into the bottom adapter until a flange on the element seats firmly against the adapter, providing rigid type support. The top adapter serves both as a handling fitting and as a spacer for the upper ends of the fuel elements.
A sliding fit between this adapter and the fuel element end fittings allows for differential expansion of the elements.
This top fitting can be removed with remote handling tools to disassemble the bundles for replacement of individual fuel elements or for shipping spent elements for reprocessing.
The individual fuel elements (Figure 4-4) are quite similar to the TRIGA elements used on TRIGA TRIGA grid plates. The differences are (1) reduction of diameter from _ inches to maintain the proper metal-to-water ratio in this core; (2) the bottom end fixture is threaded; (3) flats on the stainless steel bottom end fixture provide wrench surfaces for disassembly without stressing the cladding; and (4) the top end fixture is modified to allow the top end fitting to be locked in place. The TRIGA elements used at UWNR are of two types, standard and FLIP. Both have outside dimensions, clad thickness, and construction as shown in Figure 4-4. The two types differ as shown in the following table: Design Parameter Standard Fuel FLIP Fuel Fuel moderator material U-Zr HI.7 U-Zr H1.6 U 23S enrichment 20% 70% U 23S content/element * * (average)
Burnable Poison None Natural erbium Erbium content 1.5 wt % The FLIP fuel was designed to extend the lifetime of TRIGA fuel, and was used in step-wise additions of fuel to the University of Wisconsin Nuclear Reactor. The reactor is currently operating with a core consisting entirely of FLIP fuel. Fuel bundles contain only one type of fuel, and the top adapters for FLIP fuel bundles are marked (by notches machined into the top of the top adapter) to facilitate identification during underwater fuel handling.
UWNR Safety Analysis Report Rev. 0 4-9 April 2000 Figure 4-3 Four-element Bundle Assembly UWNR Safety Analysis Report Rev. 0 4-10 April 2000*
Figure 4-3 Four-element Bundle Assembly UWNR Safety Analysis Report Rev. 0 4-10 April 2000
Figure 4-4 Fuel Element Construction UWNR Safety Analysis Report Rev. 0 4-11 April 2000* *
Figure 4-4 Fuel Element Construction UWNR Safety Analysis Report Rev. 0 4-11 April 2000 Figure 4-5 shows an instrumented element. This element is fitted with three thermocouples.
At least one such element (inserted into the vacant position of a three-element bundle) is included in every core. The sensing tips in the thermocouples are located at the vertical centerline of the fuel section and one inch above and below the centerline.
The thermocouple leads pass through a seal in a stainless steel tube which provides a water-tight conduit carrying the lead-out wires above the surface of the pool water.Figure 4-5 Instrumented Fuel Element UWNR Safety Analysis Report Rev. 0 4-12 April 2000 Figure 4-5 shows an instrumented element. This element is fitted with three thennocouples.
At
least one such element (inserted into the vacant position of a three-element bundle) is included in every core. The sensing tips in the thennocouples are located at the vertical centerline of the fuel section and one inch above and below the centerline.
The thennocouple leads pass through a seal in a stainless steel tube which provides a water-tight conduit carrying the lead-out wires above the surface of the pool water.
Figure 4-5 Instrumented Fuel Element UWNR Safety Analysis Report Rev. 0 4-12 April 2000
4.2.2 Control ElementsBoth blade and rod shaped control elements are used.4.2.2.1 Control Blade Shrouds and Guide Tubes Each blade type control element, both safety and regulating, is guided throughout its travel by a shroud as shown in Figure 4-6. The shroud consists of two thinaluminum plates 38 inches high, separated by aluminum spacers to provide a 1/8-inch water annulus. The shrouds can be removed, if necessary, by use of a grapple hook. Small flow holes at the bottom of the shroud minimize the effect of viscous damping on the scram time.Rod shaped control elements are guided by a guide tube as shown in one of the three-element bundles of Figure 4-2. Holes drilled in the sides of the guide tube allow for water displacement when the control rod is fired out during pulsing operation or dropped in response to a scram condition.
27.4"- I IE-Figure 4-6 Control Blade Shroud UWNR Safety Analysis Report Rev. 2 4-13 Sept. * *
4.2.2 Control Elements Both blade and rod shaped control elements are used. 4.2.2.1 Control Blade Shrouds and Guide Tubes Each blade type control element, both safety and regulating, is guided throughout its travel by a shroud as shown in Figure 4-6. The shroud consists of two thin aluminum plates 38 inches high, separated by aluminum spacers to provide a 1I8-inch water annulus. The shrouds can be removed, if necessary, by use of a grapple hook. Small flow holes at the bottom of the shroud minimize the effect of viscous damping on the scram time. Rod shaped control elements are guided by a guide tube as shown in one of the element bundles of Figure 4-2. Holes drilled in the sides of the guide tube allow for water displacement when the control rod is fired out during pulsing operation or dropped in response to a scram condition.
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Figure 4-6 Control Blade Shroud UWNR Safety Analysis Report Rev. 2 4-13 Sept. 2008 4.2.2.2 Safety BladesReactor control for startup and shutdown is accomplished by three blade-type control elements, Figure 4-7, with a total shutdown worth between 6.9 and 11 per cent A KCff. The poison section is boral sheet (boron carbide and aluminum sandwiched between aluminum side plates). Each safety blade is 40.5 inches long. When a blade is full in the bottom of the blade overlaps the bottom of the active fuel by 1.5 inches.BORAL SHEET ALUMINUM CLADDING" Figure 4-7 Shim/Safety Blade4,2.2.3 Regulating Blade The regulating blade, Figure 4-8 (shown upside down for ease in reading the dimensions), is a stainless-steel sheet with curls on the vertical edges, about 11 inches wide and 40 inches long, supported and guided in the same manner as the safety blades. It is used to compensate for small changes of reactivity during normal reactor operation and may be actuated by a servo-control channel.UWNR Safety Analysis Report Rev. 2 4-14 Sept. 4.2.2.2 Safety Blades Reactor control for startup and shutdown is accomplished by three blade-type control elements, Figure 4-7, with a total shutdown worth between 6.9 and 11 per cent D. K eff* The poison section is boral sheet (boron carbide and aluminum sandwiched between aluminum side plates). Each safety blade is 40.5 inches long. When a blade is full in the bottom of the blade overlaps the bottom of the active fuel by 1.5 inches. YP" BORAL SHEET 1/8" ALUMINUM CLADDING ---+--4.2.2.3 TOP OF ACTIVE FUEL ELEIvJENTS

+----

e ,REACTOR CORE T'C'TTOi; OF FU:::L ELil,lENTS Ie, 'j" HOi'" Figure 4-7 Shim/Safety Blade , Regulating Blade The regulating blade, Figure 4-8 (shown upside down for ease in reading the dimensions), is a stainless-steel sheet with curls on the vertical edges, about 11 inches wide and 40 inches long, supported and guided in the same manner as the safety blades. It is used to compensate for small changes of reactivity during normal reactor operation and may be actuated by a servo-control channel. UWNR Safety Analysis Report Rev. 2 4-14 Sept. 2008 * *

Regulating Blade 4.2.2.4 Transient Control Rod The transient control rod is boron carbide or borated graphite contained in a inch diameter stainless steel or aluminum tube The poison section is approximately inches long. This rod is guided laterallythe aluminum guide tube in a special three-element fuel bundle. hold-~down tube extends from this guide tube to the top of the reactor structure and holds the three-element bundle in place during transient rod movement.
~Transient Control Rod UWNR Safety Analysis Report Rev. 2 4-15 Sept. * * * -, *l i I -I I @J q I I I q I I I II I I I II \ll II "'--II I! " I II I I I II I I I I I I II I I I II I I _.-----r----.----..
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.. _--"-Figure 4-8 Regulating Blade 4.2.2.4 Transient Control Rod The transient control rod is boron carbide or borated graphite contained in a 1.25 inch diameter stainless steel or aluminum tube (Figure 4-9). The poison section is approximately 15 inches long. This rod is guided laterally by the aluminum guide tube in a special three-element fuel bundle. A down tube extends from this guide tube to the top of the reactor structure and holds the three-element bundle in place during transient rod movement.
UWNR Safety Analysis Report Rev. 2 4-15 Figure 4-9 Transient Control Rod Sept. 2008 4.2.3 Neutron Moderator and Reflector Pool water serves as moderator for the core and as reflector above, below, and on those sides of the core not provided with graphite reflectors or special reflector elements designed to condition the quality of a beam being extracted from the core. (Individual fuel elements contain an internal 3.5 inch long graphite end reflector above and below the fueled portion, so the core top and bottom are actually reflected by a mixture of graphite and water.)The reflectors are standard General Electric Company reflectors furnishedat initial startup of UWNR. The nominal 3-inch square reflector elements are made of AGOT grade graphite clad with aluminum (Figure 4-Reflector element lifting handles are diagonal to facilitate identification when viewing the core and storage racks.Special reflectors are the same size as the graphite reflectors; but may consist of solid aluminum, hollow aluminum, or combinations of graphite sections with the center portion replaced with solid aluminum, voids, or gamma absorbers such as lead or bismuth.Such special reflectors are used for irradiation facilities or to adjust the mix of thermal, epithermal, and fast neutrons transmitted to experimental facilities.
-GRAPHITE 1/16" ALuMINUM CLADDING TOP VIEW IT 3" NON Figure 4-10 Graphite Reflector ElementUWNR Safety Analysis Report Rev. 0 4-16 April 2000 4.2.3 Neutron Moderator and Reflector Pool water serves as moderator for the core and as reflector above, below, and on those sides of the core not provided with graphite reflectors or special reflector elements designed to condition the quality of a beam being extracted from the core. (Individual fuel elements contain an internal 3.5 inch long graphite end reflector above and below the fueled portion, so the core top and bottom are actually reflected by a mixture of graphite and water.) The reflectors are standard General Electric Company reflectors furnished at initial startup ofUWNR. The nominal 3-inch square reflector elements are made of AGOT grade graphite clad with aluminum (Figure 4-10). Reflector element lifting handles are diagonal to facilitate identification when viewing the core and storage racks. Special reflectors are the same size as the graphite reflectors; but may consist of solid aluminum, hollow aluminum, or combinations of graphite sections with the center portion replaced with solid aluminum, voids, or gamma absorbers such as lead or bismuth. Such special reflectors are used for irradiation facilities or to adjust the mix of thermal, epithermal, and fast neutrons transmitted to experimental facilities.
UWNR Safety Analysis Report Rev. 0 SIDE IrV:li GRAPHITE TOP VIEW ---
I---3" NOM --I Figure 4-10 Graphite Reflector Element 4-16 April 2000 * *
4.2.4 Neutron Startup Source The neutron source is a radium-beryllium source irradiated to give an output greater thanneutrons/second.
It is encapsulated in a 0.515 inch diameter by 3.10 inch long stainless steel welded cylindrical capsule, which in turn contains two 1.25 inch long welded stainless steel capsules.The source fits into a source holder (Figure 4-11).The source holder, in turn, fits into an irradiation basket (Figure 4-12) occupying one grid module adjacent to the active core. The source is usually left in for full power operation, and will, with the normal operating cycle, maintain its output of about neutrons/second.
If the source is not left in during full power operation, neutron emission rate will decrease over a long time period to approximately 106 neutrons/second.
UWNR Safety Analysis Report Rev. 2 4-17 Sept.
4.2.4 Neutron Startup Source *
The neutron source is a _ radium-beryllium source irradiated to give an output greater than 10 7 neutrons/second.
It is encapsulated in a 0.515 inch diameter by 3.10 inch long stainless steel welded cylindrical capsule, which in tum contains two 1.25 inch long welded stainless steel capsules.
The source fits into a source holder (Figure 4-11). The source holder, in tum, fits into an irradiation basket (Figure 4-12) occupying one grid module adjacent to the active core. The source is usually left in for full power operation, and will, with the normal operating cycle, maintain its output of about 10 7 neutrons/second.
If the source is not left in during full power operation, neutron emission rate will decrease over a long time period to approximately 10 6 neutrons/second.
UWNR Safety Analysis Report Rev. 2 4-17 Sept. 2008
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r f : * .. Figure 4-12 Irradiation Basket UWNR Safety Analysis Report Rev. 2 t .!\ .:: ,,0-. -.. i.e 4-19 T-0 !II ,I " '"" Ii Ii II 'I ,I II II II " :1 I' " ': Ii II II II q II 'I Ii I, II II Ii I' II 'ii I, I II Ii Ii II II II I Ii II ,I I' I I II /\ If v i' II I II P r:J:" -P, .\:--1$* 0 I '!' ' I , , , , I """""* ./ l. r/'-4r ' .. , .... \ II. '" Sept. 2008 4.2.5 Core Support Structure The core is suspended from an all-aluminum frame, Figure 4-13, which extends from the grid box to a height of about one foot above the pool surface. One of the hollow comer posts of the suspension frame serves as a guide for the gamma chamber used in pulsing operation.
The other three comer posts may also be used to position detectors in positions above the core.The reactor bridge (mounted over the pool) supports the core suspension frame. The all-steel, prefabricated bridge was bolted together in the field and aligned with shims.A locating plate, made of 0.5-inch steel, spans the upper end of the suspension frame. It is bolted to the bridge and aligns the four control blade drive mechanisms and the transient rod drive withthe core. The five mechanisms work through individual clearance holes, each mechanism being secured to the locating plate.
The plate and mechanisms are not removable as a unit to prevent accidental withdrawal of the control elements.
The fission counter drive is mounted on a portion of the hand railing support structure.
Four 4-inch square 6061 aluminum suspension tubes (0.25 inch wall thickness) extend from the bridge to the grid box, and support the grid box by bolted connections.
The aluminum grid box, Figure 4-14, encloses and supports the 6-inch thick cast aluminum grid plate which is machined to locate and support the control element shrouds and bottom end fittings of fuel bundles,reflectors and in-core experimental facilities such as hydraulic irradiation positions and irradiation baskets. Figure 4-15 shows the grid position designation system, location of experimental facilities and radiation detectors relative to the grid box, and letter and number codes used in later descriptions to identify fuel and reflector reactivity worths.An aluminum coolant header (not shown in the figures) mates with the bottom of the grid box and forms a transition to the coolant piping originally provided (but not used) for future use with a forced convection cooling system. An opening in the side of the header, 24 inches wide by 12 inches high, allows cooling water flow for natural convection.
A diffuser pump and jet above the core deflects the cooling water streaming from the core to reduce N16 activity above the core.UWNR Safety Analysis Report Rev. 2 4-20 Sept. 4.2.5 Core Support Structure The core is suspended from an all-aluminum frame, Figure 4-13, which extends from the grid box to a height of about one foot above the pool surface. One of the hollow comer posts of the suspension frame serves as a guide for the gamma chamber used in pulsing operation.
The other three comer posts may also be used to position detectors in positions above the core. The reactor bridge (mounted over the pool) supports the core suspension frame. The all-steel, prefabricated bridge was bolted together in the field and aligned with shims. A locating plate, made ofO.5-inch steel, spans the upper end of the suspension frame. It is bolted to the bridge and aligns the four control blade drive mechanisms and the transient rod drive with the core. The five mechanisms work through individual clearance holes, each mechanism being secured to the locating plate. The plate and mechanisms are not removable as a unit to prevent accidental withdrawal of the control elements.
The fission counter drive is mounted on a portion of the hand railing support structure.
Four 4-inch square 6061 aluminum suspension tubes (0.25 inch wall thickness) extend from the bridge to the grid box, and support the grid box by bolted connections.
The aluminum grid box, Figure 4-14, encloses and supports the 6-inch thick cast aluminum grid plate which is machined to locate and support the control element shrouds and bottom end fittings of fuel bundles, reflectors and in-core experimental facilities such as hydraulic irradiation positions and * , irradiation baskets. Figure 4-15 shows the grid position designation system, location of
experimental facilities and radiation detectors relative to the grid box, and letter and number codes used in later descriptions to identify fuel and reflector reactivity worths. An aluminum coolant header (not shown in the figures) mates with the bottom of the grid box and forms a transition to the coolant piping originally provided (but not used) for future use with a forced convection cooling system. An opening in the side of the header, 24 inches wide by 12 inches high, allows cooling water flow for natural convection.
A diffuser pump and jet above the core deflects the cooling water streaming from the core to reduce N I6 activity above the core. UWNR Safety Analysis Report Rev. 2 4-20 Sept. 2008
Figure 4-13 Core Suspension UWNR Safety Analysis Report Rev. 2 4-21 Sept. 2008*
Figure 4-13 Core Suspension
UWNR Safety Analysis Report Rev. 2 4-21 Sept. 2008 Figure 4-14 Grid Box and Grid Plate UWNR Safety Analysis Report Rev. 2 4-22 Sept. *
Figure 4-14 Grid Box and Grid Plate UWNR Safety Analysis Report Rev. 2 4-22 Sept. 2008
Figure 4-15 Grid Arrangement With Fuel and Reflector Codes See Table 4-1 and 4-2 for an explanation of coding in the figure above.UWNR Safety Analysis Report Rev. 2 4-23 Sept. *
Figure 4-15 Grid Arrangement With Fuel and Reflector Codes See Table 4-1 and 4-2 for an explanation of coding in the figure above.
UWNR Safety Analysis Report Rev. 2 4-23 Sept. 2008 The aluminum-lined concrete pool, is feet wide, 12 feet long, and 1/2 feet deep. The reinforced concrete pool walls also serve as the biological shield (further details on the pooi walls are found in the next section).
The pool is penetrated experimental ports.Piping systems connecting with the pool are discussed in detail in Chapter In summary, piping connections are built to preclude accidental26 loss of pooi water failure of components located outside of the pool. When possible, pipes enter and leave the pool above the water surface (primary cooling system, diffuser system, and hydraulic irradiation tube water system) and are-equipped with passive siphon breakers that prevent loss of more than a few inches of water even in the event of a pipe break or system mis-operation.
Two inch aluminum pipes intended for use in a forced-convection cooling system were in the concrete at initial construction. (See Figure especially the note concerning anti-siphonloop). One of these pipes penetrates the pooi wall about 14 feet below the pool curb, but is closed-with flanges on both the inside and outside ends, preventing loss of pool water unless both flanges fail. The other inch pipe was looped inside -the concrete and equipped with a siphon breaker that extends from the top of the loop to above the pool curb. One end of the loop is flanged closed outside the shield, while the other end vertically penetrates the bottom of the pool within the coolant header (below the grid box). the outer flange of this pipe were to fail, this pipe could___drain the pool to 14 feet below the pool curb. At Fiue41RacoPolndSed UWNR Safety Analysis Report Rev. 2 424Sp.08 4-24 Sept. 4.3 Reactor Pool The aluminum-lined concrete pool, Figure 4-16, is 8 feet wide, 12 feet long, and 27 112 feet deep. The reinforced concrete pool walls also serve as the biological shield (further details on the pool walls are found in the next section). The pool is penetrated by experimental ports. Piping systems connecting with the pool are discussed in detail in Chapter 5. In summary, all piping connections are built to preclude accidental loss of pool water by failure of components located outside of the pool. When possible, pipes enter and leave the pool above the water surface (primary cooling system, diffuser system, and hydraulic irradiation tube water system) and are equipped with passive siphon breakers that prevent loss of more than a few inches of water even in the event of a pipe break or system operation.
Two 8 inch aluminum pipes intended for use in a forced-convection cooling system were imbedded in the concrete at initial construction.
Figure 5-2, especially the note concerning anti-siphon loop). One of these pipes penetrates the pool wall about 14 feet below the pool curb, but is closed with flanges on both the inside and outside ends, preventing loss of pool water unless both flanges fail. The other 8 inch pipe was looped inside the concrete and equipped with a siphon breaker that extends from the top of the loop to above the pool curb. One end of the loop is flanged closed outside the shield, while the other end vertically penetrates the bottom of the pool within the coolant header (below the grid box). If the outer flange of this pipe were to fail, this pipe could drain the to 14 feet below the curb. At UWNR Safety Analysis Report Rev. 2 4-24 Sept. 2008 * *
The pool makeup and purification system supply pipe penetrates the pool 38 inches below the pool surface, thus limiting water loss to that level reduction upon failure. The discharge pipe from this system is discussed in the paragraph immediately above.The 4 beam ports penetrate the pool wall at mid-core level. The beam ports are operated with a watertight flange on the outside end which will prevent leakage should the in-pool portion of the beam port fail. The in-pool portion of the tubes are aluminum pipes with welded end closure andbolted flanged connection to the beam port shutter assembly.
The four beam ports have acommon drain system, but the discharge valve for the drain system is maintained closed during operation.
The beam ports also have vent connections which connect to the Beam Port and Thermal Column Ventilation system (see Chapter 9, Section 9.1). The vent connections are equipped with valves (normally kept open for ventilation flow), but which may be closed if a beam port leak occurs. Finally, each beam port vent has a check valve which allows air flow only out of the vent, thus preventing pressure differences between the beam ports from causingcirculation between beam ports.
The thermal column case also penetrates the pool wall. It is of welded aluminum construction and has no valves or flanges which could be opened to drain the pool. The pool water is kept within the following limits: Temperature (at Core cooling water inlet) <130 0 F Resistivity
>2 x 10' Ohm cm Radioactivity
<10 CFR Part 20 Appendix B Table 3 values for radioisotopes with >24 hour half-life The reactor core is cooled by natural convection of pool water through the core.
The 130'F temperature limit is imposed by demineralizer resin tolerance and by humidity control considerations.
UWNR Safety Analysis Report Rev. 2 4-25 Sept. * *
The pool makeup and purification system supply pipe penetrates the pool 38 inches below the pool surface, thus limiting water loss to that level reduction upon failure. The discharge pipe from this system is discussed in the paragraph immediately above. The 4 beam ports penetrate the pool wall at mid-core level. The beam ports are operated with a watertight flange on the outside end which will prevent leakage should the in-p'ool portion of the beam port fail. The in-pool portion of the tubes are aluminum pipes with welded end closure and bolted flanged connection to the beam port shutter assembly.
The four beam ports have a common drain system, but the discharge valve for the drain system is maintained closed during operation.
The beam ports also have vent connections which connect to the Beam Port and Thermal Column Ventilation system (see Chapter 9, Section 9.1). The vent connections are equipped with valves (normally kept open for ventilation flow), but which maybe closed if a beam port leak occurs. Finally, each beam port vent has a check valve which allows air flow only out of the vent, thus preventing pressure differences between the beam ports from causing circulation between beam ports . The thermal column case also penetrates the pool wall. It is of welded aluminum construction and has no valves or which could be to drain the The pool water is kept within the following limits: Temperature (at Core cooling water inlet) Resistivity Radioactivity
<130°F >2 x 10 5 Ohm -cm <10 CFR Part 20 Appendix B Table 3 values for radioisotopes with >24 hour half-life The reactor core is cooled by natural convection of pool water through the core. The 130°F temperature limit is imposed by demineralizer resin tolerance and by humidity control considerations . UWNR Safety Analysis Report Rev. 2 4-25 Sept. 2008 The resistivity limit is set to reduce corrosion effects, extending the expected lifetime of the fuel elements and controlling water radioactivity. Routine checks of resistivity are made to determine the necessity of regenerating the demineralizer.
The radioactivity of the pool water is continuously monitored by an area monitor station located near the demineralizer.
Should the pool water reach the activity limit above, the reading on this area monitor will increase during periods when the reactor is not operating.
In addition, water samples are routinely analyzed for activity by other methods which give a more exact identification of quantity and type of activity present.No problem has been experienced in maintaining pool water radioactivity below the indicated limits in nearly 50 years of operation.
4.4 Biological Shield The reactor is shielded by concrete and water (See Figure 4-16). At normal pool level the core is covered by 20 feet of water. The shield at core level consists of about 3 feet of water plus 8 feet (9 feet on thermal column side) of ordinary concrete.
Denser concrete is used in the thermal column door and beam port plugs.
Calculations and measurements of radiation levels for 1000 kW operation are (excepting N16 activity) discussed below: Surface of shield, excepting beam port and thermal column openings -less than 1.5 mrem/hr.Pool surface (leakage radiation) (No N16) less than 15 mrem/hr."Hot spots" -measurements have shown that higher radiation levels exist around the beam ports and thermal column. Measurements of the maximum radiation levels at these "hot spots" at 1000 kW are about 10 mrem/hr around the beam ports and 40 mrem/hr at the hottest spot around the thermal column door. The dose one foot away from the hot spots is about 5 mrem/hr.Since the third and fourth-floor classrooms and offices, and fifth-floor mechanical room, are above the level of the pool curb, an analysis of the effect of complete water loss on persons inthese areas was performed and results included in the updated Emergency Plan. The computer code MCNP5 was used to model the dose rate, as described in section 13.1.3.2.
Since the biological shield does not offer any shielding to the central wing third floor classrooms, the dose in these classrooms would be greater than any location on the fourth or fifth floors. The building evacuation alarm would evacuate people from these areas before the pool was completely drained, such that the total integrated dose received by evacuating members of the public would be about 13 mrem.UWNR Safety Analysis Report Rev. 2 4-26 Sept. The resistivity limit is set to reduce corrosion effects, extending the expected lifetime of the fuel
elements and controlling water radioactivity.
Routine checks of resistivity are made to determine the necessity of regenerating the demineralizer.
The radioactivity of the pool water is continuously monitored by an area monitor station located near the demineralizer.
Should the pool water reach the activity limit above, the reading on this area monitor will increase during periods when the reactor is not operating.
In addition, water samples are routinely analyzed for activity by other methods which give a more exact identification of quantity and type of activity present. No problem has been experienced in maintaining pool water radioactivity below the indicated I* limits in nearly 50 years of operation.
4.4 Biological Shield The reactor is shielded by concrete and water (See Figure 4-16). At normal pool level the core is covered by 20 feet of water. The shield at core level consists of about 3 feet of water plus 8 feet (9 feet on thermal column side) of ordinary concrete.
Denser concrete is used in the thermal column door and beam port plugs. Calculations and measurements of radiation levels for 1000 kW operation are (excepting N l6 activity) discussed below: Surface of shield, excepting beam port and thermal column openings -less than 1.5 mremlhr. Pool surface (leakage radiation) (No N 16) less than 15 mremlhr. "Hot spots" -measurements have shown that higher radiation levels exist around the beam ports and thermal column. Measurements of the maximum radiation levels at these "hot spots" at 1000 k Ware about 10 mremlhr around the beam ports and 40 mremlhr at the hottest spot around the thermal column door. The dose one foot away from the hot spots is about 5 mremlhr. Since the third and fourth-floor classrooms and offices, and fifth-floor mechanical room, are above the level of the pool curb, an analysis ofthe effect of complete water loss on persons in these areas was performed and results included in the updated Emergency Plan. The computer code MCNP5 was used to model the dose rate, as described in section 13.1.3.2.
Since the biological shield does not offer any shielding to the central wing third floor classrooms, the dose in these classrooms would be greater than any location on the fourth or fifth floors. The building evacuation alarm would evacuate people from these areas before the pool was completely drained, such that the total integrated dose received by evacuating members of the public would be about 13 mrem. UWNR Safety Analysis Report Rev. 2 4-26 Sept. 2008 *
4.4.1 Pool Surface Radiation Levels -Activity The radiation level due to activity at the pool surface directly above the core when operating at 1000 kW would range from 80 to 220 mrem/hr if no N-16 control system were in operation (variability is due to changes in surface flow patterns).
The N-16 diffuser system is normally in operation, however, reducing the dose rates at the pool surface to 2-4 mremlhour at the pool surface. These radiation levels are low enough that no hazard will exist to personnel outside theReactor Laboratory or in normally occupied levels within the Reactor Laboratory.
Radiation levels on the walkway surrounding the pool are around 20 mrem/hr while the reactor is operating at 1000 kW without the diffuser operating and <0.5 mrem/hour with the diffuser operating.
All of the Reactor Laboratory outside of the console area is posted as a radiation area and a radioactive materials area. A cable and switch arrangement is positioned on the north stairway to the pool surface so that an alarm will be sounded should entry to that area be made while the reactor is operating, thus assuring that personnel will not enter the area without knowledge of the reactor operator.The south stairway, leading from the console area to the pool surface does not have a cable and switch arrangement as does the north stairway.
Access to these stairs is gained only through the console area and is well monitored.
No difficulty has been experienced in maintaining radiation doses to individuals well below those doses permitted in 10 CFR 20.4.4.2 Heating Effects in Shield and Thermal Column Heating effects caused by absorption of gamma radiation and fast neutrons are within allowablelimits. For all calculations, it was assumed that the pool water was at the 130' temperature limit, and the reactor was operated continuously at 1.5 MW.The heating in the concrete shield is approximately 20% of the maximum suggested by Rockwell'.
Analysis of the heating rate in the lead shield for the thermal column indicates that the maximum temperature of the lead will be less than 217°F.
Calculation of the graphite temperature in the thermal column indicates a maximum of 244°F.4.5 Nuclear Design 4.5.1 Normal Operating Conditions and Reactor Core Physics Paremeters Note: NUREG-1537 specifies separate sections for "Normal Operating Conditions" and "Reactor Core Physics Parameters." These two sections are combined to enable concise inclusion of the measured core parameters of the several cores which have been operated under the license.
UWNR Safety Analysis Report Rev. 2 4-27 Sept. 2008* *
4.4.1 Pool Surface Radiation Levels -N I6 Activity The radiation level due to N I6 activity at the pool surface directly above the core when operating at 1000 kW would range from 80 to 220 mremlhr ifno N-16 control system were in operation (variability is due to changes in surface flow patterns).
The N -16 diffuser system is normally in operation, however, reducing the dose rates at the pool surface to 2-4 mremlhour at the pool surface. These radiation levels are low enough that no hazard will exist to personnel outside the Reactor Laboratory or in normally occupied levels within the Reactor Laboratory.
Radiation levels on the walkway surrounding the pool are around 20 mremlhr while the reactor is operating at 1000 kW without the diffuser operating and <0.5 mremlhour with the diffuser operating.
All of the Reactor Laboratory outside of the console area is posted as a radiation area and a radioactive materials area. A cable and switch arrangement is positioned on the north stairway to the pool surface so that an alarm will be sounded should entry to that area be made while the reactor is operating, thus assuring that personnel will not enter the area without knowledge of the reactor operator.
The south stairway, leading from the console area to the pool surface does not have a cable and switch arrangement as does the north stairway.
Access to these stairs is gained only through the console area and is well monitored.
No difficulty has been experienced in maintaining radiation doses to individuals well below those doses permitted in 10 CFR 20. 4.4.2 Heating Effects in Shield and Thermal Column Heating effects caused by absorption of gamma radiation and fast neutrons are within allowable limits. For all calculations, it was assumed that the pool water was at the 130° temperature limit, and the reactor was operated continuously at 1.5 MW. The heating in the concrete shield is approximately 20% of the maximum suggested by Rockwell I . Analysis of the heating rate in the lead shield for the thermal column indicates that the maximum temperature of the lead will be less than 217°F. Calculation of the graphite temperature in the thermal column indicates a maximum of 244°F. 4.5 Nuclear Design 4.5.1 Normal Operating Conditions and Reactor Core Physics Paremeters Note: NUREG-1537 specifies separate sections for "Normal Operating Conditions" and "Reactor Core Physics Parameters." These two sections are combined to enable concise inclusion of the measured core parameters of the several cores which have been operated under the license. UWNR Safety Analysis Report Rev. 2 4-27 Sept. 2008 4.5.1.1 Core ArrangementsThe use of the reactor as a training and research tool requires flexibility of core arrangement.
Permitted arrangements are subject, however, to the following criteria: a. A mixed core must contain at least 9 FLIP fuel bundles (clusters)
b. FLIP fuel must be located in a central contiguous region c. The core must be a close packed array except for single fuel element positions or fuel bundle positions on the core periphery d. Calculations indicate that operation of a specific core will be within technical specification limits on power generation per element and fuel temperature.When the Safety Analysis Report for converting UWNR from flat-plate to TRIGA fuel was written, expected performance was based on computations and on the behavior of a "prototype" TRIGA Mark reactor, the Torrey Pines Reactor at General Atomics. The prototype reactor used individual TRIGA fuel elements in a right circular cylindrical array typical of TRIGA reactors; UWNR uses four-element bundles in a rectangular arrangement in the grid box provided for the original flat-plate fuel. The uranium loading in the prototype was 8 wt%uranium, while UWNR has a uranium loading of 8.5 wt%. Both the prototype and initial UWNRTRIGA cores had stainless steel clad, and both used 20% enriched uranium. The heat transfer characteristics were quite similar, although the diameter of the clad for UWNR was slightly smaller to fit to the grid box array spacing. UWNR also differed by having shrouds dividing the core box into three regions. These shrouds guide control blades, but also introduce water gaps within the core lattice.The prototype was operated for many years at steady-state power levels up to 1500 kW and thousands of pulses up to 6000 MW. In this report, although the prototype performance characteristics are indicated and sometimes compared to UWNR, most of the information is based on the measured performance of the cores which are currently operable under the presentlicense and technical specifications.
The current core is an all-FLIP (stainless steel clad) configuration consisting of 23 FLIP fuel bundles. Cores with 9 and 15 FLIP fuel bundles also have been operated for significant times, and have been thoroughly tested for conformance to technical specifications and the predictions and descriptions in this and the previous Safety Analysis Report. It is planned that the all-FLIP core will continue to be the operating core until funding (and analysis) for refueling with LEU is completed, at which time this SAR will be amended. It may become necessary to revert to the 15- or 9-bundle FLIP cores before refueling, however, in order to maintain less than a formula quantity of HEU fuel at non-self protecting levels during the return of the HEU to DOE.UWNR Safety Analysis Report Rev. 2 4-28 Sept. 4.5.1.1 Core Arrangements The use of the reactor as a training and research tool requires flexibility of core arrangement.
Permitted arrangements are subject, however, to the following criteria:
a. A mixed core must contain at least 9 FLIP fuel bundles (clusters)
b. FLIP fuel must be located in a central contiguous region c. The core must be a close packed array except for single fuel element positions or fuel bundle positions on the core periphery
d. Calculations indicate that operation of a specific core will be within technical specification limits on power generation per element and fuel temperature.
When the Safety Analysis Report for converting UWNR from flat-plate to TRIGA fuel was written, expected performance was based on computations and on the behavior of a "prototype" TRIGA Mark III reactor, the Torrey Pines Reactor at General Atomics. The prototype reactor used individual TRIGA fuel elements in a right circular cylindrical array typical of TRIGA reactors; UWNR uses four-element bundles in a rectangular arrangement in the grid box provided for the original flat-plate fuel. The uranium loading in the prototype was 8 wt% uranium, while UWNR has a uranium loading of 8.5 wt%. Both the prototype and initial UWNR TRIGA cores had stainless steel clad, and both used 20% enriched uranium. The heat transfer characteristics were quite similar, although the diameter of the clad for UWNR was slightly smaller to fit to the grid box array spacing. UWNR also differed by having shrouds dividing the core box into three regions. These shrouds guide control blades, but also introduce water gaps within the core lattice. The prototype was operated for many years at steady-state power levels up to 1500 kW and thousands of pulses up to 6000 MW. In this report, although the prototype performance characteristics are indicated and sometimes compared to UWNR, most of the information is based on the measured performance of the cores which are currently operable under the present license and technical specifications.
The current core is an all-FLIP (stainless steel clad) configuration consisting of23 FLIP fuel bundles. Cores with 9 and 15 FLIP fuel bundles also have been operated for significant times, and have been thoroughly tested for conformance to technical specifications and the predictions and descriptions in this and the previous Safety Analysis Report. It is planned that the all-FLIP core will continue to be the operating core until funding (and analysis) for refueling with LEU is completed, at which time this SAR will be amended. It may become necessary to revert to the 15-or 9-bundle FLIP cores before refueling, however, in order to maintain less than a formula quantity of HEU fuel at non-self protecting levels during the return of the HEU to DOE. UWNR Safety Analysis Report Rev. 2 4-28 Sept. 2008 * *
4.5.1.2 Standard TRIGA fuel cores 2This core, shown in Figure 4-17, and a succeeding core enlarged to 30 fuel bundles because of fuel burnup was operated from November 1967 until March 1974 when it was shut down.Measured core parameters for the initial 25 bundle version of the core are presented below.Several variations in the peripheral reflector configuration were included in this series of cores, including up to 18 reflector elements around the periphery, and some cores with special voided-center or Bismuth-center reflectors.
Core Designation A25-RIOExcess reactivity 4.82 % p Shutdown Margin 5.17 % p Transient Rod worth 2.10 % p FP reactivity defect 2.95 % p Peak pulse power 1930 MWPrompt neutron lifetime 42E-6 sec Fig UWNR Safety Analysis Report Rev. 2 4-29 Sept. * *
4.5.1.2 Standard TRIG A fuel cores 2 This core, shown in Figure 4-17, and a succeeding core enlarged to 30 fuel bundles because of fuel bumup was operated from November 1967 until March 1974 when it was shut down. Measured core parameters for the initial 25 bundle version of the core are presented below. Several variations in the peripheral reflector configuration were included in this series of cores, including up to 18 reflector elements around the periphery, and some cores with special center or Bismuth-center reflectors.
Core Designation A25-RI0 Excess reactivity 4.82 % P Shutdown Margin 5.17 % P Transient Rod worth 2.10 % P FP reactivity defect 2.95 % P Peak pulse power 1930 MW Prompt neutron lifetime 42E-6 sec Figure 4-17 Standard TRIGA Fuel Core UWNR Safety Analysis Report Rev. 2 4-29 Sept. 2008 Themeasured fuel temperatures (Figure 4-18) in the UWNR standard TRIGA core almost matched those in the prototype, although this could have differed significantly, depending upon instrumented element placement in the two cores. The UWNR instrumented element was located near the core center in grid position D4NW as indicated in Figure 4-17.I 1.S1.
~-F ~!-7-tIýREACTOR POWER, KILOWATTS Figure 4-18 Fuel Temperature-Standard Fuel UWNR Safety Analysis Report Rev. 2 4-30 Sept. The measured fuel temperatures (Figure 4-18) in the UWNR standard TRlGA core almost
matched those in the prototype, although this could have differed significantly, depending upon instrumented element placement in the two cores. The UWNR instrumented element was located near the core center in grid position D4NW as indicated in Figure 4-17. C,.) o POWER, KILOWATTS Figure 4-18 Fuel Temperature-Standard Fuel , UWNR Safety Analysis Report Rev. 2 4-30 Sept. 2008 *
When the reactivity loss from power operation for UWNR with the original TRIGA core is compared to the prototype (Figure 4-19) it is apparent that the power defect for UWNR is significantly larger than for the prototype.
In both cases, the core had been pulsed a significant number of times before the temperature measurements were made, so the difference is not from clad stretching in pulsing. The large loss was considered to be a function of the different core geometry and reflection making the leakage change more drastically with temperature in UWNR.O.0 f f am 1.0 4=t~4pi~ .... .. ... .~+~+LU 0oo 200 4oo Reactor Power, Kilowatts160c Figure 4-19 Power Defect vs. Power- Standard TRIGA Core When pulsing behavior of the two cores was compared, other differences were expected and found. The pulsing behavior differed from the typical TRIGA core primarily due to the water gaps in the control blade shrouds and the graphite reflector, both of which increased neutron lifetime, resulting in longer periods for the same pulsed reactivity insertion, and thus broader pulses. Later graphs, Figure 4-35 through Figure 4-37, compare the pulsing behavior of this and all of the other pulsing cores with that of the prototype TRIGA reactor. Note that the pulsed reactivity addition limit for the standard TRIGA fuel was 2.1 % p , instead of the 1.4 % p for cores containing FLIP fuel.
Pulsing behavior differences will be discussed later in this report.UWNR Safety Analysis Report Rev. 2 4-31 Sept. * *
When the reactivity loss from power operation for UWNR with the original TRIGA core is compared to the prototype (Figure 4-19) it is apparent that the power defect for UWNR is significantly larger than for the prototype.
In both cases, the core had been pulsed a significant number of times before the temperature measurements were made, so the difference is not from clad stretching in pulsing. The large loss was considered to be a function of the different core geometry and reflection making the leakage change more drastically with temperature in UWNR. > ... u
II II: -+. '-== =. -,--.
'i?!F.;::::;: --.* : . "-+ . '. .. -' "--!" . '. Reactor Power, Kilowatts Figure 4-19 Power Defect vs. Power-Standard TRIGA Core :!:::=::r:
...;.0::';
,-,-:-F:-._. ---= When pulsing behavior of the two cores was compared, other differences were expected and . found. The pulsing behavior differed from the typical TRIGA core primarily due to the water gaps in the control blade shrouds and the graphite reflector, both of which increased neutron lifetime, resulting in longer periods for the same pulsed reactivity insertion, and thus broader pulses. Later graphs, Figure 4-35 through Figure 4-37, compare the pulsing behavior of this and all of the other pUlsing cores with that of the prototype TRIGA reactor. Note that the pulsed reactivity addition limit for the standard TRIG A fuel was 2.1 % p, instead ofthe 1.4 % P for cores containing FLIP fuel. Pulsing behavior differences will be discussed later in this report . UWNR Safety Analysis Report Rev. 2 4-31 Sept. 2008 4.5.1.2.1 Reactivity Effects In Standard Fuel Cores The reactivity effect of fuel bundles in different lattice positions have been estimated from measured and calculated values. The position codes in Table 4-1 are those shown in Figure 4-15. The reactivity value given is the worth of adding or removing a fuel bundle while the remainder of a -bundle core is already present. The worth of the bundles whenadded in an approach to critical would be radically different, since cores are loaded as compact cores during the loading sequence; that is, the fuel loading plan is planned to assure that the next fuel bundleloaded will have a smaller reactivity effect than the bundles previously loaded.The reactivity effects of reflector variations also have been measured and/or estimated and are indicated in Tables 4-2 and 4-3.First, the worth of both a graphite reflector element and a voided reflector element areindicated relative to a water reflector, with the position codes being those shown in Figure 4-15.Table 4-1 Fuel Bundle Worths in UWNR Cores-all in % Position Water Reflected Graphite Reflected A 2.60 4.0 B 1.95 3.46 C 1.22 2.18 C' 1.16 1.15 D 0.77 E 1.76 2.76 F 1.85 Table 4-2 Reflector Element Reactivity Worths Position Replace water Replace water with graphite with air % p1 1.228 -0.239 2 0.180 -0.198 3 0.058 -0.064 4 0.076 -0.083 5 0.115 -0.126 6 0.157 -0.172 7 0.029 UWNR Safety Analysis Report Rev. 2 4-32 Sept. 4.5.1.2.1 Reactivity Effects In Standard Fuel Cores The reactivity effect of fuel bundles in Table 4-1 different lattice positions have been estimated from measured and calculated values. The position codes in Table 4-1 are those shown in Figure 4-15. The reactivity value given is the worth of adding or removing a fuel bundle while the remainder of a.-bundle core is already present. The worth of the bundles when added in an approach to critical would be radically different, since cores are loaded as compact cores during the loading sequence; that is, the fuel loading plan is planned to assure that the next fuel bundle loaded will have a smaller reactivity effect than the bundles previously loaded. The reactivity effects of reflector variations also have been measured and/or estimated and are indicated in Tables 4-2 and 4-3. First, the worth of both a graphite reflector element and a voided reflector element are indicated relative to a water reflector, with the position codes being those shown in Figure 4-15 . UWNR Safety Analysis Report Rev. 2 Position A B C C' D E F Table 4-2 Position 1 2 3 4 5 6 7 4-32 Fuel Bundle Worths in UWNR all in % p Water Reflected Graphite Reflected 2.60 4.0 1.95 3.46 1.22 2.18 1.16 1.15 0.77 -1.76 2.76 1.85 1.55 Reflector Element Reactivity Worths Replace water Replace water with graphite with air % p %p 1.228 -0.239 0.180 -0.198 0.058 -0.064 0.076 -0.083 0.115 -0.126 0.157 -0.172 0.029 -------Sept. 2008 * *
Next, the effect of other changes that affect reflection are indicated.
Most of these were measured in a core of standard TRIGA fuel.Table 4-3Reactivity Effect of Reflector Region Changes Condition Result-%p Flooding all 4 beam ports +0.0005Flooding pneumatic tube +0.002 Pneumatic tube sampleswater filled
-0.0003 Cadmium filled Dropping fuel bundle on top of +0.5 core Adding fuel bundle on side of core +0.77 UWNR Safety Analysis Report Rev. 2 4-33 Sept. * *
Next, the effect of other changes that affect reflection are indicated.
Most ofthese were measured in a core of standard TRIGA fuel. Table 4-3 Reactivity Effect of Reflector Region Changes Condition Result-%p Flooding all 4 beam ports +0.0005 Flooding pneumatic tube +0.002 Pneumatic tube samples water filled -0.0003 Cadmium filled Dropping fuel bundle on top of +0.5 core Adding fuel bundle on side of core +0.77 UWNR Safety Analysis Report Rev. 2 4-33 Sept. 2008 4.5.1.3 Cores containing FLIP fuelThe longer operating lifetime for FLIP fuel was the major reason for selecting this fuel type for refueling the University of Wisconsin Nuclear Reactor. The higher enrichment of FLIP fuel coupled with erbium poisoning provides the longer operating lifetime, but it also causes changes in operating characteristics relative to standard fuel. The prototype FLIP core was also the Torrey Pines TRIGA Mark III fueled with FLIP fuel.
The most marked changes from use of FLIP fuel are a reduction of prompt neutron cycle time to about 1 OE-6 seconds at beginning of core life (20E-6 at end of core life) and a temperature coefficient that is strongly temperaturedependent. (Figure 3-16, page 3 of reference) 3.These data are for the prototype reactor; values in UWNR were expected to and do differ because of the water gap in the control blade shrouds and the graphite reflector, making the neutron lifetime considerably longer in UWNR FLIP cores than it was in the standard TRIGA core.In addition, the harder spectrum in a FLIP core leads to power peaking in regions near water gaps. This leads, in a compact core, to a peaking factor within a FLIP element of 1.43. If a large water-filled flux trap is located adjacent to an element, the peaking factor in the element can increase to 2.65 peak/average within the cell.Thermal and hydraulic parameters of FLIP fuel remain the same as standard fuel.FLIP fuel elements are not mixed with standard elements in the same fuel bundle at Wisconsin.
Thus, the smallest increment of FLIP fuel addition possible will be three FLIP elements (in a bundle containing the transient rod guide tube).
Placing such a bundle in the center of a 5 x 5 array of standard TRIGA fuel leads to the highest value of power peaking possible, with resultant, power generation of 31.2 kW in each element. Although no operation with this core isanticipated or desired, other TRIGA reactors have operated with power generation rates at least as high as 32kW per element.Addition of less than five FLIP fuel bundles (24 FLIP elements) was not considered useful for a full power operating core, since it would not provide sufficient additional reactivity to compensate for burnup in the standard elements.Calculations were performed for cores with 1, 2, 5, 9, 15 and 25 FLIP -bundles in central contiguous regions of the core. All calculations were for a 5 x 5 array of fuel bundles with the transient rod guide tube in the fuel bundle at grid-position D5. Calculations were performed with a two-dimensional diffusion theory code (Exterminator 2). Standard seven group cross sections obtained from Gulf-General Atomic were used in the calculations. The accuracy of the calculations was checked by analysis of cores with known values of Keff and power density.Results on calculation of mixed cores and FLIP cores were found to be consistent with similar calculations performed elsewhere 4.Subsequent computations using a 3-dimensional diffusion theory code (DIF-3D) in support of use of LEU fuel agreed well with the results for Exterminator-2 for the all-FLIP core.UWNR Safety Analysis Report Rev.
2 4-34 Sept. 4.5.1.3 Cores containing FLIP fuel The longer operating lifetime for FLIP fuel was the major reason for selecting this fuel type for refueling the University of Wisconsin Nuclear Reactor. The higher enrichment of FLIP fuel coupled with erbium poisoning provides the longer operating lifetime, but it also causes changes in operating characteristics relative to standard fuel. The prototype FLIP core was also the Torrey Pines TRIGA Mark III fueled with FLIP fuel. The most marked changes from use of FLIP fuel are a reduction of prompt neutron cycle time to about 10E-6 seconds at beginning of core life (20E-6 at end of core life) and a temperature coefficient that is strongly temperature dependent. (Figure 3-16, page 3 ofreference)3.
These data are for the prototype reactor; values in UWNR were expected to and do differ because of the water gap in the control blade shrouds and the graphite reflector, making the neutron lifetime considerably longer in UWNR FLIP cores than it was in the standard TRIGA core. In addition, the harder spectrum in a FLIP core leads to power peaking in regions near water gaps. This leads, in a compact core, to a peaking factor within a FLIP element of 1.43. If a large water-filled flux trap is located adjacent to an element, the peaking factor in the element can increase to 2.65 peak/average within the cell. Thermal and hydraulic parameters of FLIP fuel remain the same as standard fuel. FLIP fuel elements are not mixed with standard elements in the same fuel bundle at Wisconsin.
Thus, the smallest increment of FLIP fuel addition possible will be three FLIP elements (in a
bundle containing the transient rod guide tube). Placing such a bundle in the center of a 5 x 5 array of standard TRIGA fuel leads to the highest value of power peaking possible, with resultant power generation of 31.2 kW in each element. Although no operation with this core is anticipated or desired, other TRlGA reactors have operated with power generation rates at least as high as 32kW per element. Addition ofless than five FLIP fuel bundles (24 FLIP elements) was not considered useful for a full power operating core, since it would not provide sufficient additional reactivity to compensate for burnup in the standard elements.
Calculations were performed for cores with 1,2,5,9, 15 and 25 FLIP -bundles in central contiguous regions of the core. All calculations were for a 5 x 5 array of fuel bundles with the transient rod guide tube in the fuel bundle at grid-position D5. Calculations were performed with a two-dimensional diffusion theory code (Exterminator 2). Standard seven group cross sections obtained from Gulf-General Atomic were used in the calculations.
The accuracy of the calculations was checked by analysis of cores with known values ofKeff and power density. Results on calculation of mixed cores and FLIP cores were found to be consistent with similar calculations performed elsewhere 4* Subsequent computations using a 3-dimensional diffusion theory code (DIF-3D) in support of use of LEU fuel agreed well with the results for Exterminator-2 for the all-FLIP core. UWNR Safety Analysis Report Rev. 2 4-34 Sept. 2008
FLIP fueled cores can experience significant power peaking which must be considered in permissible fuel arrangements and setting of limiting safety system settings.
The power produced in individual fuel elements was predicted from the computations done for safety analyses.
The following table shows both the power density in individual fuel elements and the worth of a fuel bundle loaded in a particular location (if applicable to the condition) for several different core arrangements analyzed.
See Figure 4-15 for position descriptions.
Table 4-4 Maximum Power Density and Reactivity Worth of Fuel Bundles- Cores Containing FLIP FUEL Core arrangement (keyed to fuel Power in Maximum Reactivity Effect of bundle locations in Figure 4-15) Element- kW Removing Fuel Bundle in NOTE: is a 3-element bundle Indicated Position-%p with transient rod guide tubeFLIP +20 Standard fuel bundles 21.4 (FLIP in Positions A & B)Replace FLIP in E5 with H 2 0 28.3 2.83 9 FLIP +16 Standard fuel bundles (FLIP in Positions A, B, and E) 18.1 ----Replace FLIP in with H20 20.0 0.93 Replace FLIP in or with H 2 0 25.9 1.69 Replace FLIP in D4 or D6 with H20 23.2 0.98 Replace FLIP in E4, C4, C6, or E6 with 22.3 1.49 H 2 0 15 FLIP +10 Standard fuel bundles (FLIP in Positions A, B, C, E, &F) 17.2 ----Replace FLIP in with H 2 0 19.0 0.87 Replace FLIP in or with H 2 0 24.6 1.65 Full FLIP- 25 FLIP fuel bundles (FLIP in all positions except D) 15.5 Replace FLIP in with H 2 0 17.2 0.79 Replace FLIP in C6 or D6 with H 2 0 20.0 0.51 Replace FLIP in or with H 2 0 22.1 1.42 Reference to the table above (Table 4-4) shows that power generation in any individual element is well below 23 KW in all compact FLIP fuel arrangements.
Further, the presence of a 3-inch square water gap in the FLIP fuel region will result in power generation rates below 23 KW/element in most of the cores.
The initial operational mixed core contained nine FLIP fuel bundles (35 elements), and the calculations indicate that flux traps could not be permitted for full power operation in this arrangement for locations D4, and D6 if the maximum kW/element is to be kept below UWNR Safety Analysis Report Rev. 2 4-35 Sept. * *
FLIP fueled cores can experience significant power peaking which must be considered in permissible fuel arrangements and setting of limiting safety system settings.
The power produced in individual fuel elements was predicted from the computations done for safety analyses.
The following table shows both the power density in individual fuel elements and the worth of a fuel bundle loaded in a particular location (if applicable to the condition) for several different core arrangements analyzed.
See Figure 4-15 for position descriptions.
Table 4-4 Maximum Power Density and Reactivity Worth of Fuel Bundles-Cores Containing FLIP FUEL Core arrangement (keyed to fuel Power in Maximum Reactivity Effect of bundle locations in Figure 4-15) Element-kW Removing Fuel Bundle in NOTE: D5 is a 3-element bundle Indicated Position-%p with transient rod guide tube 5 FLIP +20 Standard fuel bundles 21.4 ----(FLIP in Positions A & B) Replace FLIP in E5 with H 2 O 28.3 2.83 9 FLIP + 16 Standard fuel bundles (FLIP in Positions A, B, and E) 18.1 ----Replace FLIP in D5 with H 2 O 20.0 0.93 Replace FLIP in C5 or E5 with H 2 O 25.9 1.69 Replace FLIP in D4 or D6 with H 2 O 23.2 0.98 Replace FLIP in E4, C4, C6, or E6 with 22.3 1.49 H 2 O 15 FLIP + 1 0 Standard fuel bundles (FLIP in Positions A, B, C, E, &F) 17.2 ----Replace FLIP in 05 with H 2 O 19.0 0.87 Replace FLIP in C5 or E5 with HP 24.6 1.65 Full FLIP-25 FLIP fuel bundles (FLIP in all positions except 0) 15.5 ----Replace FLIP in D5 with H 2 O 17.2 0.79 . Replace FLIP in C6 or D6 with HP 20.0 0.51 Replace FLIP in C5 or E5 with H 2 O 22.1 1.42 Reference to the table above (Table 4-4) shows that power generation in any individual element is well below 23 KW in all compact FLIP fuel arrangements.
Further, the presence of a 3-inch square water gap in the FLIP fuel region will result in power generation rates below 23 KW/element in most of the cores. The initial operational mixed core contained nine FLIP fuel bundles (35 elements), and the calculations indicate that flux traps could not be permitted for full power operation in this arrangement for locations C5, E5, D4, and D6 if the maximum kW/element is to be kept below UWNR Safety Analysis Report Rev. 2 4-35 Sept. 2008 kW The combination of fuel bundle proximity to control blade shrouds and the transient rod guide tube causes the greatest power peaking in any of these cores.It is also apparent from comparing Tables 4-1 and 4-4 that the reactivity worth of an individual FLIP bundle is lower than that of a standard fuel bundle, even in mixed cores.The Technical Specifications under which the facility has operated since conversion to FLIP fuel required at least nine FLIP fuel bundles elements) in a central contiguous location with no water gaps larger than a single element except on the core periphery.
As a resul t, the maximum power density in any fuel element at kW was limited to kW for any of the cores considered.
This is approximately higher than the maximum in an all standard fuel core.Three different FLIP-containing cores have been operated at Characteristics of each core, measured during the startup and acceptance testing of each core, are shown in the following sections.
During core test programs, one quadrant of each core containing FLIP fuel was mapped for temperature moving an instrumented fuel element into each unique core position.Interpretation of the fuel measurements was complicated instrumented element failures so that measurements were made with different instrumented elements in different cores as explained below. The standard fuel temperature measurements were made using two different instrumented elements, since the original standard instrumented element failed before the bundle FLIP core was tested. There was, however, only one standard instrumented element in the core at a time, so no comparisons between the indication of the elements in the same core position are available.
Two instrumented FLIP elements were available, and both were used in the temperature mapping. The individual FLIP instrumented elements were both placed in at least one common position to enable comparison between the indication of the different elements in the same core position.
However, because one instrumented element had all thermocouples fail, three different instrumented FLIP elements have been used in the tested cores, and widely varying temperatures (as much as C) were measured when these different elements were placed in the same core position.
This makes interpretation of the predicted and measured fuel temperatures more difficult, but the conclusion reached during the test programs was that the results were reasonably consistent with the predicted values, considering that the non-instrumented fuel assemblies probably have as large a range of heat transfer characteristics as the instrumented elements do. Data tables from these core test programs include fuel temperatures as predicted and measured.
Safety Analysis Report Rev. 2 436Sp.20 4-36 Sept. 23 kW. The combination of fuel bundle proximity to control blade shrouds and the transient rod
guide tube causes the greatest power peaking in any of these cores. It is also apparent from comparing Tables 4-1 and 4-4 that the reactivity worth of an individual FLIP bundle is lower than that of a standard fuel bundle, even in mixed cores. The Technical Specifications under which the facility has operated since conversion to FLIP fuel required at least nine (9) FLIP fuel bundles (35 elements) in a central contiguous location with no water gaps larger than a single element except on the core periphery.
As a result, the maximum power density in any fuel element at 1,000 kW was limited to 18.1 kW for any of the cores considered.
This is approximately 11 % higher than the maximum in an all standard fuel core.
Three different FLIP-containing cores have been operated at UWNR. Characteristics of each core, measured during the startup and acceptance testing of each core, are shown in the following sections.
During core test programs, one quadrant of each core containing FLIP fuel was mapped for temperature by moving an instrumented fuel element into each unique core position.
Interpretation of the fuel measurements was complicated by instrumented element failures so that measurements were made with different instrumented elements in different cores as explained below. The standard fuel temperature measurements were made using two different instrumented elements, since the original standard instrumented element failed before the 15-bundle FLIP core was tested. There was, however, only one standard instrumented element in the core at a time, so no comparisons between the indication of the elements in the same core position are available.
Two instrumented FLIP elements were available, and both were used in the temperature mapping. The individual FLIP instrumented elements were both placed in at least one common position to enable comparison between the indication of the different elements in the same core position.
However, because one instrumented element had all thermocouples fail, three different instrumented FLIP elements have been used in the tested cores, and widely varying temperatures (as much as 130° C) were measured when these different elements were placed in the same core position.
This makes interpretation of the predicted and measured fuel temperatures more difficult, but the conclusion reached during the test programs was that the results were reasonably consistent with the predicted values, considering that the instrumented fuel assemblies probably have as large a range of heat transfer characteristics as the instrumented elements do. Data tables from these core test programs include fuel temperatures as predicted and measured.
UWNR Safety Analysis Report Rev. 2 4-36 Sept. 2008 *
4.5.1.3.1 First mixed core- 9 FLIP bundles and 16 standard bundles 5 This initial mixed core, Table 4-4, Figure 4-20 and Figure 4-21, was operated from March 1974 through December 1977.Some parameters of this core were: Core Designation Excess reactivityShutdown margin Transient Rod worth FP reactivity defect Peak pulse powerPrompt neutron lifetime F25-R10 4.05 % 3.60 % 1.37 % p 1.92 % p 805 MW 29.7E-6 sec Figure 4-20 9-Bundle FLIP Core UWNR Safety Analysis Report Rev. 2 4-37 Sept. 2008* *
4.5.1.3.1 First mixed core-9 FLIP bundles and 16 standard bundles 5 This initial mixed core, Table 4-4, Figure 4-20 and Figure 4-21, was operated from March 1974 through December 1977. Some parameters of this core were: Core Designation Excess reactivity Shutdown margin Transient Rod worth FP reactivity defect Peak pulse power Prompt neutron lifetime F25-RlO 4.05 % P 3.60 % P 1.37 % P 1.92 % P 805MW 29.7E-6 sec Figure 4-20 9-Bundle FLIP Core UWNR Safety Analysis Report Rev. 2 4-37 Sept. 2008 Core Position 1) 5 NW SW5 NW K 4 N'NW 1) 4 ;NE NW SW SE KW in H1eqrent16.1 18.1 17.415.417.0 16.1 17.4 16.26.7 Pred Itted.Temp., 372 360 372 363360 361, 342 342361 350 350 363 350 272 239 Measured temperatures Bdle 41 FLIP Bdle 42 FLIP Bdle 22 STD Can't Cahnt-Can't 402 395 402 397 382 383 384397 399 400 398 Measure Measure Measure 365365:361 343 346 348 343 35 9 353 363 342 F SE P 4- NE SWF 3 NE, NW SW SL NW SW SE D 3 SW SE.8.061.35.8 6.8.4.996 6.6 6.9: 8.3 257 266 238 223 230 242 210 ,204237 242 250 290 238 260 280 233 216 200 222175 240 269 275 246 280 260-242 261 Figure 4-21 Power/element and Temperature
-9 FLIP Bundle Core UWNR Safety Analysis Report Rev. 2 4-38 Sept. KW Pred Ictt*d (0 )
Temperatures'., e
Core. 'Position F.lement TemE' °c .8d1e III ,FLIP Bdle 42}o'LIP Bdle 22 S'l'D \) 5 NE 18.1 372 Can't Measure NW 16.1 360 Can't Measure SW IB.l 372 Cari"t Measure 5 NK 17.4 363 402 }65 NW 16.1 35.0 395 350 SW 16.,9 360 402 365 *SK rhO 361 397 361 K4 Ni'; 15.4 342 382 343 NW 15.4 342 383 346 SW 16.1 350 384 348 SI': 17.0 361 380 341 l) 4 NK 16.1 350 397 359 NW 16.1 350 399 353 SW 17.4 363 400 363 SE, 16.2 350 398 342 F 5 NE A,o', 9.3 272 290
sit 6.7 239 238 F 4, NE 8.0 257 , 260 NW 266 280 SW 6.3 238 233 SE 5 * .7 223 216 F 3 5.B 230 200 NW 6.8 222 SW 4.9 210 185 SI;; 4,'6 204 175 EJ NE 7.4 250 240 NW .6.6 237 269 SW 6.,9 242 275 SE 7 .. 4, 250 246 D 3 SW 6.9 .242 280 SF. ,8.3 261 260 Figure 4-21 Power/element and Temperature
-9 FLIP Bundle Core
UWNR Safety Analysis Report Rev. 2 4-38 Sept. 2008 Fuel temperatures in the 9-Bundle FLIP core are shown in Figure 4-22. Instrumented elements were located in grid position F5NE for bundle 22 (standard fuel), grid position C4SW for bundle 41, and grid position E5NE for bundle 42. The instrumented element in bundle 42 was located immediately adjacent to the transient rod guide tube. Although the power density was much higher in the FLIP fuel in this partial FLIP core, the fuel temperatures were reasonable.

-== 3 : -6 00. .:7 :0 oo0: Figure 4-22 Fuel Temperatures vs. Power- 9 Bundle FLIPPower Defect vs. Power -9 Bundle FLIP The power defect was strongly affected by the FLIP fuel. See Figure 4-23. Since the temperature coefficient in FLIP fuel becomes more negative as temperature rises, the total power defect during normal full power operation is smaller, while accident response remains essentially the same as in standard TRIGA fuel.i I5I I h I TM , I.K ---..... --- -.. ..

.!-- -
....... ---- :-:---
-
70 ','.5 Figure 4-23 Power Defect vs. Power -9 Bundle FLIP Analysis Report Rev. 2 4-39 UWNR Safety Sept. * *
Fuel temperatures in the 9-Bundle FLIP core are shown in Figure 4-22. Instrumented elements were located in grid position F5NE for bundle 22 (standard fuel), grid position C4SW for bundle 41, and grid position E5NE for bundle 42. The instrumented element in bundle 42 was located immediately adjacent to the transient rod guide tube. Although the power density was much higher in the FLIP fuel in this partial FLIP core, the fuel temperatures were reasonable. ---, .. =-==::.:-:..:..:: .... -..=.:. -:.-===-=.
---_=:.-=--i "
"_":JOOO --------'\" . -'1 -_ .. _----------
Figure 4-22 Fuel Temperatures vs. Power-9 Bundle FLIP Power Defect vs. Power -9 Bundle FLIP The power defect was strongly affected by the FLIP fuel. See Figure 4-23. Since the temperature coefficient in FLIP fuel becomes more negative as temperature rises, the total power defect during normal full power operation is smaller, while accident response remains essentially the same as in standard TRIGA fuel. UWNR Safety Analysis Report Rev. 2 4-39 . Sept. 2008 4.5.1.3.2 Second mixed core -15 FLIP fuel bundles and 10 standard bundles 6 This core, shown in Figure 4-24 and Figure 4-25, was operated from January 1978 until June 1979, although some initial high power operation with a core intermediate between this and the all FLIP core was done to make the initial shipment self-protecting before the final batch of fuel was received.Some parameters for this core were:
Core Designation G25-R1O Excess reactivity 3.87 % p Shutdown Margin 3.90 % p Transient Rod worth 1.38 % p FP reactivity defect 1.75 % p Peak pulse power 930 MWPrompt neutron lifetime 24E-6 sec Figure 4-24 15 Bundle FLIP core UWNR Safety Analysis Report Rev. 2 4-40 Sept. 4.5.1.3.2 Second mixed core-15 FLIP fuel bundles and 10 standard bundles 6 This core, shown in Figure 4-24 and Figure 4-25, was operated from January 1978 until June 1979, although some initial high power operation with a core intermediate between this and the all FLIP core was done to make the initial shipment self-protecting before the final batch of fuel was received.
Some parameters for this core were: Core Designation Excess reactivity Shutdown Margin Transient Rod worth FP reactivity defect Peak pulse power Prompt neutron lifetime G25-RlO 3.87 % P 3.90 % P 1.38 % P 1.75 % P 930MW 24E-6 sec Figure 4-24 15 Bundle FLIP core UWNR Safety Analysis Report Rev. 2 4-40 Sept. 2008 * *
Core Position DS NE NW SW ES NE NW SW SE E4 NE SW SE D4 NE NW SW SE E3 NE NW SW SE D3 NE NW SW SENE NW SW SEF4 NE NW SW SE KW in Element 17.2 16.0 17.2 16.5 15.3 15.8 15.8 11.9 14.6 15.0 13.3 12.516.4 12.6 9.7 9.9 11.2 10.6 11.0 10.4 10.4 11.0 8.8 8.8 6.4 6.4 7.5 8.4 6.1 5.5 Predicted Temp.(26 or 41)*C 363 349 363 355 342347 304 334 338 318 309 340 354 310 277 280 294 288 293 285 285 293 267 267 237 237 252 262 231 222 Measured Bdle 41 Center TC CAN'CANN CAN'269 237 290 247 294 Temperatures Bdle 42 Center TC T MEAS T MEAS T MEAS 435 361 361 340 355 345 322Bdle 26 Center TC URE URE URE 400 359 415 301 250 271246 263 220 213 203 220 177 163 F3 NE 5.4 220 NW .6.3 236 SW 4.8 208 SE 4.5 200 Figure 4-25 Power/element and Temperature
-15 FLIP Bundle Core UWNR Safety Analysis Report Rev. 2 4-41 Sept.
Measured Temperatures (OC) KW in Predicted Temp. Bd1e 41 Bd1e 42 Bd1e 26 Core Position Element or 41)OC Center TC Center TC Center TC OS NE 17.2 363 CAN' T MEA SUR E NW 16.0 349 CAN ' T MEASURE SW 17.2 363 CAN 'T MEASURE E5 NE 16.5 355 435 NW 15.3 342 SW 15.8 347 361 SE 15.8 347 361 E4 NE 11.9 304 340 NW 14.6 334 355 SW 15.0 338 345 SE 13.3 318 322 04 NE 12.5 309 NW 15.2 340 SW 16.4 354 269 400 SE 12.6 310 359 E3 NE 9.7 277 237 415 NW 9.9 280 290
SW n.2 294 247 SE 10.6 288 294 03 NE 11.0 293 NW 10.4 285 SW 10.4 285 301 SE 11.0 293 250 FS NE 8.8 267 271 NW 8.8 267 SW 6.4 237 SE 6.4 237 237 F4 NE 7.5 252 246 NW 8.4 262 263 SW 6.1 231 220 SE 5.5 222 213 F3 NE 5.4 220 203 NW 6.3 236 220 SW 4.8 208 177 SE 200 163 Figure 4-25 Power/element and Temperature
-15 FLIP Bundle Core
UWNR Safety Analysis Report Rev. 2 4-41 Sept. 2008 With the -bundle core, FLIP fuel characteristics become more pronounced.
The standard fuel instrumented element in standard fuel bundle 25 was located in grid position F5NE, and the measured temperatures are shown in Figure 4-26. I
-°C I.~~z+XhI2I~ZI774 17 Figure 4-26 Standard Fuel Temperature vs. Power -15 Bundle FLIP Figure 4-27 shows the temperatures for the instrumented element in fuel bundle F41, located in grid position E3NE. The unusual behavior of the bottom thermocouple (41-B in the figure) wasdue to the beginnings of failure of the thermocouple due to internal shorting... ........ .... ..---- ------- L------1]I Figure 4-27 Bundle 41 Fuel Temperature vs. Power -15 Bundle FLIP UWNR Safety Analysis Report Rev. 2 4-42 Sept. With the .bundle core, FLIP fuel characteristics become more pronounced.
The standard fuel instrumented element in standard fuel bundle 25 was located in grid position F5NE, and the
measured temperatures are shown in Figure 4-26. =F--i ! Figure 4-26 Standard Fuel Temperature vs. Power -15 Bundle FLIP Figure 4-27 shows the temperatures for the instrumented element in fuel bundle F41, located in grid position E3NE. The unusual behavior of the bottom thermocouple.
(41-B in the figure) was due to the beginnings of failure of the thermocouple due to internal shorting.
Figure 4-27 Bundle 41 Fuel Temperature vs. Power -15 Bundle FLIP UWNR Safety Analysis Report Rev. 2 4-42 Sept. 2008 *
Figure 4-28 shows the temperature for the fuel bundle F42 instrumented element vs. reactor power.........400TEMPT-2 --------- -- Figure 4-28 Bundle 42 Fuel Temperature vs. Power -15 Bundle FLIP The power defect for this core is shown in Figure 4-29. Again, the effect of the FLIP fuel results in a still lower power defect with the higher amount of power generation in the FLIP portion of this core. Pulsing behavior also showed a further reduction in the prompt neutron lifetime and thus faster periods for the same reactivity input in a pulse.Figure 4-29 Power Defect vs. Power -15 Bundle FLIP UWNR Safety Analysis Report Rev. 2 4-43 Sept. * *
Figure shows the temperature for the fuel bundle F42 instrumented element vs. reactor power. 400 200 100;---'
-=:-n_ : -_. _
__ -: _" o---t-I -__':-.1-I ,
---:;:

:::::::-1-:-

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1-'------

:::::t===-I Figure Bundle 42 Fuel Temperature vs. Power -15 Bundle FLIP The power defect for this core is shown in Figure 4;.29. Again, the effect of the FLIP fuel results in a still lower power defect with the higher amount of power generation in the FLIP portion of this core. Pulsing behavior also showed a further reduction in the prompt neutron lifetime and thus faster periods for the same reactivity input in a pulse. Figure 4-29 Power Defect vs. Power -15 Bundle FLIP UWNR Safety Analysis Report Rev. 2 4-43 Sept. 2008 4.5.1.3.3 All-FLIP core 7 The reactor has been operated with cores containing only FLIP fuel since July 1979. Two variants of this core have been commonly used, differing only in the number of graphite reflectors used and location of irradiation facilities. The arrangement shown in Figure 4-30 with the characteristics indicated below and in Figure 4-31 is the most-used variant.Some measured parameters for this core were: Core Designation 123-RIO Excess reactivity 4.23 % pShutdown Margin 3.80 % p Transient Rod worth 1.395% p FP reactivity defect 1.53 % p Peak pulse power 950 MW Prompt neutron lifetime 23E-6 sec UWNR Safety Analysis Report Rev. 2 4 -44 Sept. 4.5.1.3.3 All-FLIP core 7 The reactor has been operated with cores containing only FLIP fuel since July 1979. Two variants of this core have been commonly used, differing only in the number of graphite reflectors used and location of irradiation facilities.

The arrangement shown in Figure 4-30 with the characteristics indicated below and in Figure 4-31 is the most-used variant. Some measured parameters for this core were: Core Designation I23-R10 Excess reactivity 4.23 % P Shutdown Margin Transient Rod worth FP reactivity defect Peak pulse power Prompt neutron lifetime 3.80 % P 1.395% P 1.53 % P 950MW 23E-6 sec UWNR Safety Analysis Report Rev. 2 4-44 Sept. 2008 * *
Core Position 05 NE NW SW KW in Element 16.0 14.9 16.0 Predicted Temp.(#42)°c 490 470 490 TEMPERATURES AT FULL#41 Bot Ctr Top Bot POWER#42 Ctr, urere Can Can't't't M M Mas ea E5 NE NW SW SE E4 NE NW SW SESW SE E3 NE NW SW SE F5 NE SE F4 NE NW SW SE F3 NE NW SW SE 15.5 14.4 14.6 14.6 11.913.7 12.1 15.2 151.1 14.2 12.2 9.911.6 7.9 9.8 11.0 7.5 6.8 8.7 8.1 5.9 6.7 480 460 465 465 415 442 450 420 475 470 455 420 385 395 415 345 380 405 340 325 360 345 300 325 365 382 353 340 362 338 300 315 295334 350 328 330 350 326 305 325 305 390 405 380 492 432 407 426 380 343 277 259 375 337 300 275 290 270 275 285 265 241 224 450 400 350 341 301 268 410 442 340 305 318 362 220 242 362 395 300 270 282 320 245 220 325 352 270 240 241 280 280 198 Figure 4-31 Power/element and Temperature
-All FLIP Core UWNR Safety Analysis Report Rev. 2 4-45 Sept.
TEMPERATURES AT FULL POWER KW in Predicted Temp. #41 #42 Core Position
(#42)OC Bot ctr IQE. Bot err, -05 NE 16.0 490 Can I t M e a.s u r e NW '14.9 470 Can I t Mea sur e SW 16.0 490 Can I t Mea sur e E5 HE 15.5 480 365 382 353 NW 14.4 460 SW 14.6 465 SE 14.6 465 340 362 338 E4 NE 11.9 415 300 315 295 NW 13.6 442 334 350 328 SW 13.7 450 330 350 326 SE 12.1 420 305 325 305 04 SW 15.2 475 390 405 380 492 432 407 SE 15.1 470 426 380 343
E3 NE 14.2 455 277 259 375 337 300 NW 12.2 420 275 290 270 SW 9.9 385 275 285 265 SE 10.6 395 241 224 F5 NE 11.6 415 450 400 350 SE 7.9 345 341 301 268 F4 NE 9.8 380 410 362 325 NW 11.0 405 442 395 352 SW 7.5 340 340 300 270 SE 6.8 325 305 270 240 F3 NE 8.7 360 318 282 241 NW 8.1 345 362 320 280 SW 5.9 300 220 245 280 SE 6.7 325 242 220 198
Figure 4-31 Power/element and Temperature
-All FLIP Core UWNR Safety Analysis Report Rev. 2 4-45 Sept. 2008 Fuel temperatures in the all-FLIP core are shown in Figure 4-32 and Figure 4-33. The bottom thermocouple in fuel bundle 41 had developed a short from one side of the couple to ground, and thus reads well below the actual temperature in this graph. The instrumented element in bundle F41 was in grid position E3NE, while that for bundle F42 was in grid position D4SW next to the transient rod guide tube.Figure 4-32 Bundle 41 Fuel Temperature vs. Power -All FLIP Figure 4-33 Bundle 42 Fuel Temperature vs. Power -All FLIP UWNR Safety Analysis Report Rev. 2 4-46 Sept. Fuel temperatures in the all-FLIP core are shown in Figure 4-32 and Figure 4-33. The bottom
thermocouple in fuel bundle 41 had developed a short from one side of the couple to ground, and thus reads well below the actual temperature in this graph. The instrumented element in bundle F41 was in grid position E3NE, while that for bundle F42 was in grid position D4SW next to the transient rod guide tube. -----t Figure 4-32 Bundle 41 Fuel Temperature vs. Power -All FLIP
400 ., ! 300 0.. UJO , , -200 100 -r; , Figure Bundle 42 Fuel Temperature vs. Power -All FLIP UWNR Safety Analysis Report Rev. 2 4-46 Sept. 2008
Figure 4-34 shows the power defect versus power for this core. In this all-FLIP core the powerdefect for licensed full power decreased slightly from the 15-bundle FLIP core value. At the end of 1999, after more than 477 MWd operation, the power defect at licensed full power remains at 1.51 % AK/K. 0.Figure 4-34 Power Defect vs. Power -All FLIP 4.5.1.4 Isothermal Temperature Coefficient The coolant water temperature in the prototype was varied over wide ranges (200 to 60'C) to measure the resulting reactivity change. The measurements were made at power levels of less than 10 watts. The coefficient is slightly positive with a net gain in available reactivity of 0.077%over the range indicated. The average coefficient,0.00 9%/°C, is small enough that it is essentially negligible for normal operating conditions.
The effect of the water gap left in the shrouds when the control blades are withdrawn was expected to increase the temperature coefficient by about 20% in the UWNR, giving atemperature coefficient estimated at 0.0024%/°C.
This value was small enough to be considered negligible for normal operating conditions. Values measured during startup testing were 0 and0.0042, but with vary large uncertainty in the values because of other possible reactivity variations that might occur (bubbles, variation of rest position of control blades and the transient rod, and other extremely small variations).
Considering these other variations it is not possible to see any change in reactivity in the UWNR cores as the bulk water temperature changes for either the standard, mixed, or all-FLIP cores.UWNR Safety Analysis Report Rev. 2 4-47 Sept. * *
Figure 4-34 shows the power defect versus power for this core. In this all-FLIP core the power defect for licensed full power decreased slightly from the IS-bundle FLIP core value. At the end of 1999, after more than 477 MWd operation, the power defect at licensed full power remains at 1.51 % ilKlK. Figure 4-34 Power Defect vs. Power -All FLIP 4.5.1.4 Isothermal Temperature Coefficient The coolant water temperature in the prototype was varied over wide ranges (20* to 60*C) to measure the resulting reactivity change. The measurements were made at power levels of less than 10 watts. The coefficient is slightly positive with a net gain in available reactivity of 0.077% over the range indicated.
The average coefficient,0.0019%I"C, is small enough that it is essentially negligible for normal operating conditions.
The effect of the water gap left in the shrouds when the control blades are withdrawn was expected to increase the temperature coefficient by about 20% in the UWNR, giving a temperature coefficient estimated at 0.0024%I"C.
This value was small enough to be considered negligible for normal operating conditions.
Values measured during startup testing were 0 and 0.0042, but with vary large uncertainty in the values because of other possible reactivity variations that might occur (bubbles, variation of rest position of control blades and the transient rod, and other extremely small variations).
Considering these other variations it is not possible to see any change in reactivity in the UWNR cores as the bulk water temperature changes for either the standard, mixed, or all-FLIP cores. UWNR Safety Analysis Report Rev. 2 4-47 Sept. 2008 4.5.1.5 Pulse Parameters Measurements were made of the various parameters relating to pulsing operation of the prototype and of the UWNR cores. The most important of these are given below for step insertions of reactivity up to 2.1% AK/K for the standard-fueled cores and 1.4% AK/K for the cores that contain FLIP fuel. The data for the prototype TRIGA Mark III core, the UWNR standard TRIGA core (both water and graphite reflected), and the mixed cores are indicated in the figures referenced below.Period and Pulse Width During pulsing operation the o reactor is placed in a super- j prompt-critical condition in which 1.4 the asymptotic period is related to : H4 the prompt reactivity insertion
'!HT .divided into the prompt neutron cycle time. The pulse width is inversely related to the prompt reactivity insertion. Behavior of the different cores and the ... T prototype is indicated in Figure 4- Till 35 with points of inverse period and FWHM shown on the same -2 .2o graph. The plots show the results 2 : E.of plotting the reciprocal of the * : 'measured period versus the V prompt reactivity insertion.
Since N the period data were obtained 1 :r from an oscillographic recording of the reactor power versus time at ! " a portion of the pulse before fuel -41.temperature limiting effects have begun, the accuracy of the Lit measurements is not as good as -1 for other parameters, and some .8. .......2 ....o scatter in the data are expected.
Insertion, ýaXK As can be seen, the minimum period in standard fuel obtained Figure 4-35 InversenPeriod and Inverse Width at Half-I for reactivity insertions of 2.1% vs. Prompt Reactivity Insertion AK/K is about 2.6 msec while that of the FLIP core is 3.4 msec for a 1.4% AK/K insertion. As FLIP fuel replaces standard fuel in the mixed cores, the decrease in prompt neutron cycle time results in a different straight line plot for each core. Further, it becomes apparent on the all-FLIP UWNR Safety Analysis Report Rev. 2 4-48 Sept. 4.5.1.5 Pulse Parameters Measurements were made of the various parameters relating to pulsing operation of the prototype and of the UWNR cores. The most important of these are given below for step insertions of reactivity up to 2.1 %
for the standard-fueled cores and 1.4%
for the cores that contain FLIP fuel. The data for the prototype TRIGA Mark III core, the UWNR standard TRIGA core (both water and graphite reflected), and the mixed cores are indicated in the figures referenced below. Period and Pulse Width I During pulsing operation the reactqr is placed in a prompt-critical condition in which the asymptotic period is related to the prompt reactivity insertion divided into the prompt neutron cycle time. The pulse width is inversely related to the prompt reactivity insertion.
Behavior of the different cores and the :; .*1 I .. .., c: o u i L prototype is indicated in Figure 4-! 35 with points of inverse period and FWHM shown on the same graph. The plots show the results of plotting the reciprocal of the ... .. % . .. measured period versus the ; prompt reactivity insertion.
Since 5 the period data were obtained from an oscillographic recording of the reactor power versus time at 1 .., a portion of the pulse before fuel :. temperature limiting effects have II begun, the accuracy of the . .1 " " I; I Prompt Re8ctiyfty-
'nsertfon, measurements is not as good as for other parameters, and some scatter in the data are expected.
As can be seen, the minimum period in standard fuel obtained for reactivity insertions of 2.1 % Figure 4-35 Inverse.Period and Inverse Width at Half-Max vs. Prompt Reactivity Insertion is about 2.6 msec while that of the FLIP core is 3.4 msec for a 1.4%
insertion.
As FLIP fuel replaces standard fuel in the mixed cores, the decrease in prompt neutron cycle time results in a different straight line plot for each core. Further, it becomes apparent on the all-FLIP UWNR Safety Analysis Report Rev. 2 4-48 Sept. 2008 * *
core that the faster cycle time, even for the smaller prompt reactivity insertion allowed for the mixed cores, causes the data for larger reactivity insertions to depart from a straight line. This is because the transient rod has not completed its travel before the reactor reaches a substantial power level, thus resulting in a longer period as negative reactivity is inserted by the fuel temperature coefficient.
Pulse Width The width of the power pulse is [z 7.most conveniently described as the time interval between half-power points. Also shown in Figure 4-35 ". .,t .'.. 't are plots of the reciprocal of the : .: i -...measured width versus prompt V l , i reactivity insertion.
Each of the I;i-,:- , ..: r. .cores has a different straight line o :t ii.'1.I.L L .14, ~i h ,ý;h plot, again due to the increasingly 4-........t short prompt neutron cycle time as ,,.-. .! ,," the amount of FLIP fuel increases.
2 ,.... ~ ~ Ih i-f...' ...N li~iii , ...When the pulse widths (Full Width ,, : :..... , l.t;: .., '"...at Half Maximum power) of the t.-various cores are compared to their 2 01i prompt period as shown in Figure ..+-!..1,...4.
j ., tl4.: ,. -:. , :I : ...i m .+' ...+ ..m .',
4-36, all of the cores conform fairly -I 441.well to the same straight line I..,--r because the difference in prompt = * ,.j,,.1 .,II.. ...:..l 1 , , neutron cycle time is not a factor -r.. .. .
.in the relationship between these ... ,. ..... ..+"' ....; , two variables as it is in the reactivity-period and reactivity-i/ '; F ' -,, " <. ... ..............
1,14,+i+ f"i+'+: FWHM relationships... , .., r~02 FWHM vs Period UWNR Safety Analysis Report Rev. 2 4-49 Sept. * *
core that the faster cycle time, even for the smaller prompt reactivity insertion allowed for the mixed cores, causes the data for larger reactivity insertions to depart from a straight line. This is because the transient rod has not completed its travel before the reactor reaches a substantial power level, thus resulting in a longer period as negative reactivity is inserted by the fuel temperature coefficient.
Pulse Width The width of the power pulse is most conveniently described as the time interval between half-power points. Also shown in Figure 4-35 are plots of the reciprocal of the measured width versus prompt reactivity insertion.
Each of the cores has a different straight line plot, again due to the increasingly short prompt neutron cycle time as the amount of FLIP fuel increases.
2 When the pulse widths (Full Width at Half Maximum power) of the various cores are compared to their prompt period as shown in Figure 4-36, all of the cores conform fairly well to the same straight line because the difference in prompt neutron cycle time is not a factor in the relationship between these two variables as it is in the reactivity-period and FWHM relationships.
Period, Milliseconds Figure 4-36 FWHM vs Period UWNR Safety Analysis Report Rev. 2 4-49 Sept. 2008 Peak Power Peak power in a pulse is proportional to .: =., the square of the prompt reactivity li:i !;' ..R 01!insertion, while energy generated in a pulse ...' ...., is directly related to the prompt reactivity '.'.i 7 insertion.
Figure 4-37 shows the inter- .T , relationship between maximum transient .... .. i .I., power and pulse width. Peak power is plotted against the square of inverse ....FWHM in order to get a straight line plot. 1200,<....
The standard and FLIP fueled cores show 4 11;differences due to the shorter prompt .l" neutron cycle time. Figure 4-38 shows the relationship Half-Width (1)!
103between initial period and peak power and seems to be fit fairly well by a single Figure 4-37 Peak Power vs Inverse Half-width straight line. Note that the prompt Squared reactivity insertion is limited to 1.4 AK/K for the FLIP cores.
For a given core configuration, the peak 0' 1 power, integral power in the prompt burst, and width of the pulse are determined by the reactivity insertion made. It can be seen from the plots that the peak power is -controllable over a rather wide range since
! J'1 this parameter is very nearly proportional !to (AK/K- 0.7%)2. Pulse width and It! integral powers, on the other hand, are approximately linear functions of reactivity insertions above prompt critical
........so that their range is more limited..I .... .o
... .................. N.ewtor Figure 4-38 Peak Power vs. Reactor PeriodUWNR Safety Analysis Report Rev. 2 4-50 Sept. Peak Power Peak power in a pulse is proportional to the square of the prompt reactivity insertion, while energy generated in a pulse is directly related to the prompt reactivity insertion.
Figure 4-37 shows the relationship between maximum transient power and pulse width. Peak power is plotted against the square of inverse FWHM in order to get a straight line plot. The standard and FLIP fueled cores show differences due to the shorter prompt neutron cycle time. Figure 4-38 shows the relationship between initial period and peak power and seems to be fit fairly well by a single straight line. Note that the prompt reactivity insertion is limited to 1.4 % ilKlK for the FLIP cores. For a given core configuration, the peak power, integral power in the prompt burst, and width of the pulse are determined by the reactivity insertion made. It can be seen from the plots that the peak power is controllable over a rather wide range since this parameter is very nearly proportional to (ilKlK -0.7%)2. Pulse width and integral powers, on the other hand, are approximately linear functions of reactivity insertions above prompt critical so that their range is more limited. UWNR Safety Analysis Report Rev. 2 Figure 4-37 Peak Power vs Inverse Half-width Squared _.ector Perfod, Mllilleconds Figure 4-38 Peak Power vs. Reactor Period 4-50 Sept. 2008 * *
4.5.2 Operating Limits For previous operation of the reactor at 1 MW, the reactivity has been less than 4.9 % AK/K above clean cold critical.
The reactivity is allocated approximately as indicated below: Power Coefficient 1.75% AK/K Xenon Poisoning 1.75% LK/K Control & Flux Balancing 1.40% AK/KNo specific amount of negative reactivity available from control element action is specified, since the requirement on minimum shutdown margin assures safe shutdown from any operating condition.
The minimum shutdown margin with the most reactive control element and any un-scrammable control elements full out will not be less than 0.2% AK/K. Shutdown margin is verified by calibrated control element positions and by rod-drop measurements.
The limitations on cores containing FLIP fuel will maintain power density to levels capable of natural convection cooling during power operation up to 1.5 MW power. This limitation will also assure that power density in any fuel element will be below that at which loss of reactor coolant will result in fuel damage.
In addition, the maximum reactivity for an experiment is limited to 1.4 % AK/K All in-pool experiments will be constrained at least as well as the fuel bundles. In-core experiments are designed so they are constrained by the grid or grid box structure, although part of their support may be from other pool structure.
Should an experiment having the maximum reactivity worth allowed for all experiments (1.4%AK/K) fail, the resulting step change in reactivity worth would be less than that deliberately inserted during pulsing operation.
Should the beam ports and pneumatic tube flood while the reactor is operating at full power, a step reactivity addition of 0.07% AK/K would result.
This reactivity change is so small that it would not cause any disruption of normal operation.
If a gross departure from procedure were to be made and a fuel element bundle were added to the outside of the core while operating at full power, the maximum reactivity that would result would be about 0.7% AK//K. This is a reactivity smaller than that routinely inserted during pulsing operation.
Despite the built-in safeguards and inherent safety of the reactor and its fuel, great attention is paid to proper supervision of operation and adherence to procedures approved by competent authority.
It is the policy of the University of Wisconsin that standard operating procedures are carefully prepared and reviewed, strictly followed, and kept current. Likewise, competent UWNR Safety Analysis Report Rev. 2 4-51 Sept. * *
4.5.2 Operating Limits For previous operation of the reactor at 1 MW, the reactivity has been less than 4.9 % ilKlK above clean cold critical.
The reactivity is allocated approximately as indicated below: Power Coefficient Xenon Poisoning Control & Flux Balancing 1.75% t.KlK 1.75% t.KlK 1.40% t.KlK No specific amount of negative reactivity available from control element action is specified, since the requirement on minimum shutdown margin assures safe shutdown from any operating condition.
The minimum shutdown margin with the most reactive control element and any un-scrammable control elements full out will not be less than 0.2% ilKlK. Shutdown margin is verified by calibrated control element positions and by rod-drop measurements.
The limitations on cores containing FLIP fuel will maintain power density to levels capable of natural convection cooling during power operation up to 1.5 MW power. This limitation will also assure that power density in any fuel element will be below that at which loss of reactor coolant will result in fuel damage. In addition, the maximum reactivity for an experiment is limited to 1.4 % ilKlK All in-pool experiments will be constrained at least as well as the fuel bundles. In-core experiments are designed so they are constrained by the grid or grid box structure, although part of their support may be from other pool structure.
Should an experiment having the maximum reactivity worth allowed for all experiments (1.4% ilKlK) fail, the resulting step change in reactivity worth would be less than that deliberately inserted during pulsing operation.
Should the beam ports and pneumatic tube flood while the reactor is operating at full power, a step reactivity addition of 0.07% ilKlK would result. This reactivity change is so small that it would not cause any disruption of normal operation.
If a gross departure from procedure were to be made and a fuel element bundle were added to the outside of the core while operating at full power, the maximum reactivity that would result would be about 0.7% ilKlIK. This is a reactivity smaller than that routinely inserted during pulsing operation.
Despite the built-in safeguards and inherent safety of the reactor and its fuel, great attention is paid to proper supervision of operation and adherence to procedures approved by competent authority.
It is the policy of the University of Wisconsin that standard operating procedures are carefully prepared and reviewed, strictly followed, and kept current. Likewise, competent UWNR Safety Analysis Report Rev. 2 4-51 Sept. 2008 supervision assures that operation is kept within the limits set by licenses, technical specifications, existing procedures, and general good practice.4.6 Thermal-Hydraulic Design General Atomics has done extensive thermal-hydraulic computations over the years, including one study specifically directed at the use of four-element bundles of TRIGA fuel as a replacement fuel for reactors which were originally fueled with flat-plate fuel elements and operated at power levels up to 2,000 kW with natural convection cooling 8.The Puerto Rico Nuclear Center TRIGA-FLIP reactor used 4-element fuel bundles and operated at power levels much higher than the 1,000 kW steady-state power level of UWNR with natural convection cooling 9.The thermal and hydraulic design for operation of the Torrey Pines Thermionic Reactor, section 3.3 of the reference, describes the core parameters for a very similar core'°. The conclusions of these references were that four element bundles of TRIGA fuel could be used for power levels up to 2000 kW with natural convection cooling. The University of Wisconsin Nuclear Reactor has been operated with natural convection cooling at steady-state power levels up to 1,000 kW for many years with no cooling problems and no fuel damage. Many other TRIGA reactors have been operated at power levels up to 1.5 MW with natural convection cooling and no cooling problems.
4.7 References
1. Reactor Shielding Design Manual, Theodore Rockwell III, Editor, McGraw Hill Book Company, 1956, page 178.2. Memo No. 4, Report on Refueling the University of Wisconsin Nuclear Reactor, R. J.Cashwell, Nuclear Engineering Department, University of Wisconsin, March 1968.3. GA-9064, Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor, Section 3.2 and Figure 3-16, General Atomics, Jan. 5, 1970.4. Same as 2.Core Test Program UWNR Mixed TRIGA-FLIP Core (9 FLIP Bundles),R.
J. Cashwell, Nuclear Engineering Department, University of Wisconsin, July 1974.6. Core Test Program UWNR Mixed TRIGA-FLIP Core (15 FLIP Bundles),R.
J. Cashwell, Nuclear Engineering Department, University of Wisconsin, February 1978.7. Core Test Program All FLIP Core) ,R. J. Cashwell, Nuclear Engineering Department, University of Wisconsin, January 1980.8. Steady State Thermal Analysis for the Proposed Use of TRIGA Fuel Elements in MTR Reactors, GA-5708, General Atomics, 1965.UWNR Safety Analysis Report Rev. 2 4-52 Sept. supervision assures that operation is kept within the limits set by licenses, technical specifications, existing procedures, and general good practice.
4.6 Thermal-Hydraulic Design General Atomics has done extensive thermal-hydraulic computations over the years, including one study specifically directed at the use of four-element bundles ofTRIGA fuel as a replacement fuel for reactors which were originally fueled with flat-plate fuel elements and operated at power levels up to 2,000 kW with natural convection cooling 8* The Puerto Rico Nuclear Center TRIGA-FLIP reactor used 4-element fuel bundles and operated at power levels much higher than the 1,000 kW steady-state power level ofUWNR with natural convection cooling 9* The thermal and hydraulic design for operation of the Torrey Pines Thermionic Reactor, section 3.3 of the reference, describes the core parameters for a very similar core lO* The conclusions of these references were that four element bundles ofTRIGA fuel could be used for power levels up to 2000 kW with natural convection cooling. The University of Wisconsin Nuclear Reactor has been operated with natural convection cooling at steady-state power levels up to 1,000 kW for many years with no cooling problems and no fuel damage. Many other TRIGA reactors have been operated at power levels up to 1.5 MW with natural convection cooling and no cooling problems.
4.7 References
1. Reactor Shielding Design Manual, Theodore Rockwell III, Editor, McGraw Hill Book Company, 1956, page 178. 2. Memo No.4, Report on Refueling the University of Wisconsin Nuclear Reactor, R. J. Cashwell, Nuclear Engineering University of Wisconsin, March 1968. 3. GA-9064, Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor, Section 3.2 and Figure 3-16, General Atomics, Jan. 5, 1970. 4. Same as 2. 5. Core Test Program UWNR Mixed TRIGA-FLIP Core (9 FLIP Bundles),R.
J. Cashwell, Nuclear Engineering Department, University of Wisconsin, July 1974. 6. Core Test Program UWNR Mixed TRIGA-FLIP Core (15 FLIP Bundles),R.
J. Cashwell, Nuclear Engineering Department, University of Wisconsin, February 1978. 7. Core Test Program All FLIP Core) ,R. J. Cashwell, Nuclear Engineering Department, University of Wisconsin, January 1980. 8. Steady State Thermal Analysis for the Proposed Use ofTRIGA Fuel Elements in MTR Reactors, GA-5708, General Atomics, 1965. UWNR Safety Analysis Report Rev. 2 4-52 Sept. 2008 * * *
9. Safeguards Summary Report for the TRIGA-FLIP Reactor at the Puerto Rico NuclearCenter, PRNC 123, Puerto Rico Nuclear Center, November 11, 1969.10. Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor, GA-9064, Gulf General Atomics, January 5, 1970.UWNR Safety Analysis Report Rev. 2 4-53 Sept. * *
9 . Safeguards Summary Report for the TRlGA-FLIP Reactor at the Puerto Rico Nuclear Center, PRNC 123, Puerto Rico Nuclear Center, November 11, 1969. 10. Safety Analysis Report for the Torrey Pines TRlGA Mark III Reactor, GA-9064, Gulf General Atomics, January 5, 1970 . UWNR Safety Analysis Report Rev. 2 4-53 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2 4-54 Sept.
This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 2 4-54 Sept. 2008 REACTOR Summary Description The pool water is cooled by the system shown schematically in Figure The design basis of the reactor coolant system is to dissipate 1.0 MW with primary temperatures approximately 80 'F and prevent the inadvertent loss of pool water. The system, however, performs no safety function.The system consists of three loops; the closed-loop primary coolant system, the closed-loopintermediate coolant system and the closed-loop campus chilled water system. Heat from the primary coolant system is transferred to the intermediate coolant system through the primary heat exchanger. Heat from the intermediate coolant system is then rejected to the campus chilledwater system through the intermediate heat exchanger.
The system is designed to maintain apressure gradient towards the pool in order to prevent the inadvertent loss of pool water.5.2 Primary Coolant System The primary coolant system is composed of a pump, isolation valves and various devices used to extract flow rate, temperatures and pressures.
Stainless steel components and piping are used in the system in order to maintain primary water quality more easily.
The primary system continuously circulates pool water through the primary heat exchanger.
The intake and outletdiffusers include siphon breaker holes to preclude draining more than 1 foot of water even in the case of a pipe rupture.
This will maintain at least 19 feet of water above the active core.5.3 Intermediate Coolant System The intermediate coolant system consists of a pump, isolation valves and various devices used to extract temperatures and pressures.
Stainless steel components and piping are used in the system to maintain reactor grade quality water in the intermediate coolant system. Circulation in this system is maintained by the intermediate pump, discharging through the intermediate heat exchanger, where it will reject heat to the campus chilled water system. The cold water will then circulate through the primary heat exchanger to cool the primary water and return to the pump suction.The intermediate coolant system is equipped with a pressurized expansion tank and a make-up water system. The expansion tank will accommodate volumetric changes in the intermediate system process fluid and maintain the intermediate system pressure above the primary coolant system pressure under both static and operational conditions.
By maintaining the intermediate system pressure higher than the primary system, should a leak occur, it would result in intermediate water entering the primary system thereby protecting against the inadvertent loss of pool water. A pressure sensor provides indication at the control console and an interlock prevents starting the primary pump unless the intermediate loop pump is running.UWNR Safety Analysis Report Rev. 1August 2004* *
5 REACTOR COOLANT SYSTEMS 5.1 Summary Description The pool water is cooled by the system shown schematically in Figure 5-1. The design basis of the reactor coolant system is to dissipate 1.0 MW with primary temperatures approximately 80 of and prevent the inadvertent loss of pool water. The system, however, performs no safety function.
The system consists of three loops; the closed-loop primary coolant system, the closed-loop intermediate coolant system and the closed-loop campus chilled water system. Heat from the primary coolant system is transferred to the intermediate coolant system through the primary heat exchanger.
Heat from the intermediate coolant system is then rejected to the campus chilled water system through the intermediate heat exchanger.
The system is designed to maintain a pressure gradient towards the pool in order to prevent the inadvertent loss of pool water. 5.2 Primary Coolant System The primary coolant system is composed of a pump, isolation valves and various devices used to extract flow rate, temperatures and pressures.
Stainless steel components and piping are used in the system in order to maintain primary water quality more easily. The primary system continuously circulates pool water through the primary heat exchanger.
The intake and outlet diffusers include siphon breaker holes to preclude draining more than 1 foot of water even in the case of a pipe rupture. This will maintain at least 19 feet of water above the active core. 5.3 Intermediate Coolant System The intermediate coolant system consists of a pump, isolation valves and various devices used to extract temperatures and pressures.
Stainless steel components and piping are used in the system to maintain reactor grade quality water in the intermediate coolant system. Circulation in this system is maintained by the intermediate pump, discharging through the intermediate heat exchanger, where it will reject heat to the campus chilled water system. The cold water will then circulate through the primary heat exchanger to cool the primary water and return to the pump suction. The intermediate coolant system is equipped with a pressurized expansion tank and a make-up water system. The expansion tank will accommodate volumetric changes in the intermediate system process fluid and maintain the intermediate system pressure above the primary coolant system pressure under both static and operational conditions.
By maintaining the intermediate system pressure higher than the primary system, should a leak occur, it would result in intermediate water entering the primary system thereby protecting against the inadvertent loss of pool water. A pressure sensor provides indication at the control console and an interlock prevents starting the primary pump unless the intermediate loop pump is running . UWNR Safety Analysis Report Rev. I 5-1 August 2004 E~*(n)Figure Cooling System SchematicUWNR Safety Analysis Report Rev. 1August 2004* c:-o. co E E :::l & Q. Q; Ol c:-13 't: x & UJ ro Q)
J: 2 0. co E '6 :::l Q) Q. E 0. 2 0 0 c --' 2 Q; Ol co c '6 co ..c Q) U E x UJ 2 ro Q) J: "0 2 :c 0 Figure 5-1 Cooling System Schematic UWNR Safety Analysis Report Rev. 1 5-2 August 2004
Should leakage occur, it could be detected in three ways. First, the intermediate loop will be unable to maintain pressure and a low pressure annunciator will alarm at the reactor control console. Second, the pool level float switch will be actuated by high pool water level should as much as 150 gallons of intermediate water enter the pool system.
Finally, if the integrity of theintermediate water heat exchanger is also compromised, an influx of degraded qualityintermediate water will increase conductivity in the pool water.5.4 Campus Chilled Water System The campus chilled water system consists of carbon steel piping, pump, isolation valves andvarious devices used to extract temperatures and pressures.
Circulation in this system is maintained by the chilled water pump, taking a suction on the main campus chilled water system.The pump discharges through a filter and into the intermediate heat exchanger, where it cools the intermediate loop water and returns to the main campus chilled water system.The campus chilled water loop is maintained at a higher pressure than the intermediate system.This pressure gradient will insure that in the extremely unlikely event of leaks in both the primary coolant system and intermediate coolant system heat exchangers and loss of intermediate system pressure that inadvertent loss of pool water will be physically impossible.Primary Coolant Cleanup System Water connections through the biological shield are shown in Figure 5-2. The pool clean-up system is shown schematically in Figure Water is circulated from the pool surface, through the pump, through the demineralizer, and then into the pool under the core box and coolant header. The pump maintains about 18 gallons/minute flow through the demineralizer.
The demineralizer is a mixed-bed type with provisions for regeneration of resins or discharge of spent resin and loading with new resin. A water softener supplies softened water for regeneration of the demineralizer.
Flow from the demineralizer to the pool is through valve 10102, check valve 22 which prevents back flow, and valve 719 into the 8 inch pipe loop and into the bottom of the grid box. The 8 inch line is equipped with a siphon breaker at the top of the pool so that rupture of the line at the demineralizer outlet or of the 8 inch line outside the shield cannot drain the pool to a level that will uncover the core. A second 8 inch line is flanged off on both ends. The 8 inch lines were originally installed to allow a forced-convection cooling mode, but the lines are used only as indicated above.A two inch line whose rupture could have caused loss of pool water has been permanently plugged inside the concrete shield and is presently sealed off outside the shield. A pool drain line and valve have been eliminated.
There are no valves in the system that, if opened, can drain the pool.UWNR Safety Analysis Report Rev. 2Sept. * *
Should leakage occur, it could be detected in three ways. First, the intermediate loop will be unable to maintain pressure and a low pressure annunciator will alarm at the reactor control console. Second, the pool level float switch will be actuated by high pool water level should as much as 150 gallons of intermediate water enter the pool system. Finally, ifthe integrity ofthe intermediate water heat exchanger is also compromised, an influx of degraded quality intermediate water will increase conductivity in the pool water. 5.4 Campus Chilled Water System The campus chilled water system consists of carbon steel piping, pump, isolation valves and various devices used to extract temperatures and pressures.
Circulation in this system is maintained by the chilled water pump, taking a suction on the main campus chilled water system. The pump discharges through a filter and into the intermediate heat where it cools the intermediate loop water and returns to the main campus chilled water system. The campus chilled water loop is maintained at a higher pressure than the intermediate system. This pressure gradient will insure that in the extremely unlikely event of leaks in both the primary coolant system and intermediate coolant system heat exchangers and loss of intermediate system pressure that inadvertent loss of pool water will be physically impossible.
5.5 Primary Coolant Cleanup System Water connections through the biological shield are shown in Figure 5-2. The pool clean-up system is shown schematically in Figure 5-3. Water is circulated from the pool surface, through the pump, through the demineralizer, and then into the pool under the core box and coolant header. The pump maintains about 18 gallons/minute flow through the demineralizer.
The demineralizer is a mixed-bed type with provisions for regeneration of resins or discharge of spent resin and loading with new resin. A water softener supplies softened water for regeneration of the demineralizer.
Flow from the demineralizer to the pool is through valve 10102, check valve 22 which prevents back flow, and valve 719 into the 8 inch pipe loop and into the bottom ofthe grid box. The 8 inch line is equipped with a siphon breaker at the top ofthe pool so that rupture ofthe line at the demineralizer outlet or of the 8 inch line outside the shield cannot drain the pool to a level that will uncover the core. A second 8 inch line is flanged off on both ends. The 8 inch lines were originally installed to allow a forced-convection cooling mode, but the lines are used only as indicated above. A two inch line whose rupture could have caused loss of pool water has been permanently plugged inside the concrete shield and is presently sealed off outside the shield. A pool drain line and valve have been eliminated.
There are no valves in the system that, if opened, can drain the pool. UWNR Safety Analysis Report Rev. 2 5-3 Sept. 2008 Should valve number 5 (shown in Figure 5-3) be left open upon placing the system in its normal operating condition, as much as 400 gallons of pool water could be pumped to the holdup tank.No further loss of water would then occur, since check valve 22 will prevent reverse flow from the 8 inch pipe loop to the demineralizer and the siphon breaker at the top of the loop will prevent additional water loss.Wastes from demineralizer regeneration and waste poured down the reactor laboratory floor drain or radioactive sink are collected in the waste system holdup tank. The waste system consists of a 2000 gallon holdup tank, pump, and filter, and is shown schematically in Figure 5-4. The holdup tank is periodically sampled, analyzed, and then pumped out into the sanitary sewer through 0.5.micron filters to preclude any particulate activity from being discharged.
All operations involving the cleanup system are performed by written checklist-type procedures designed to prevent draining of the pool.5.6 Primary Coolant Makeup Water System The pool makeup water system is shown schematically in Figure 5-5. Normally, makeup water is supplied by the still. The still delivers water to a system of storage tanks from which it is pumped (by the pool recirculating pump) into the pool to maintain pool water level. Althoughdistilled water is normally used for makeup, alternate flow paths allow softened or city water to be fed through the demineralizer into the pool. In either case, impurities in make-up water are reduced to less than 1 ppm before going into the pool.All operations involving the makeup system are performed by written checklist-type procedures designed to prevent draining of the pool.5.7 Nitrogen-16 Control System Nitrogen-16 suppression is accomplished by ajet-type diffuser system. This system pumps about 80 gallons of water per minute from near the pool surface through a single nozzle having a 0.75inch wide, 6.5 inch long opening.
The nozzle is located 5.5 feet above and 0.5 feet east of the core, with the diffusing stream directed downward at a 45 degree angle toward the west end of the pool.The pump for the N-16 suppression system is located on the outside east face of the reactor's concrete pool shield structure, about 8 feet below the pool surface. The system is constructed with siphon breaker holes which preclude draining more than one foot of water from the pool in the event of a pipe rupture.Auxiliary Systems Using Primary Coolant There are none.UWNR Safety Analysis Report Rev. 2 5-4 Sept. Should valve number 5 (shown in Figure 5-3) be left open upon placing the system in its normal
operating condition, as much as 400 gallons of pool water could be pumped to the holdup tanle No further loss of water would then occur, since check valve 22 will prevent reverse flow from the 8 inch pipe loop to the demineralizer and the siphon breaker at the top of the loop will prevent additional water loss. Wastes from demineralizer regeneration and waste poured down the reactor laboratory floor drain or radioactive sink are collected in the waste system holdup tank. The waste system consists of a 2000 gallon holdup tank, pump, and filter, and is shown schematically in Figure 5-4. The holdup tank is periodically sampled, analyzed, and then pumped out into the sanitary sewer through OSmicron filters to preclude any particulate activity from being discharged.
All operations involving the cleanup system are performed by written checklist-type procedures designed to prevent draining of the pool.
5.6 Primary Coolant Makeup Water System The pool makeup water system is shown schematically in Figure 5-5. Normally, makeup water is supplied by the still. The still delivers water to a system of storage tanks from which it is pumped (by the pool recirculating pump) into the pool to maintain pool water level. Although distilled water is normally used for makeup, alternate flow paths allow softened or city water to be fed through the demineralizer into the pool. In either case, impurities in make-up water are
reduced to less than 1 ppm before going into the pool. All operations involving the makeup system are performed by wrItten checklist-type procedures designed to prevent draining of the pool.
5.7 Nitrogen-16 Control System Nitrogen-16 suppression is accomplished by a jet-type diffuser system. This system pumps about 80 gallons of water per minute from near the pool surface through a single nozzle having a 0.75 inch wide, 6.5 inch long opening. The nozzle is located 5.5 feet above and 0.5 feet east of the core, with the diffusing stream directed downward at a 45 degree angle toward the west end of the pool. The pump for the N -16 suppression system is located on the outside east face of the reactor's concrete pool shield structure, about 8 feet below the pool surface. The system is constructed with siphon breaker holes which preclude draining more than one foot of water from the pool in the event of a pipe rupture. 5.8 Auxiliary Systems Using Primary Coolant There are none. UWNR Safety Analysis Report Rev. 2 5-4 Sept. 2008
Pool Curb 891.11t Pool Overflow L Deminerolizer- SuctionLines Anti-siphon loop (For Future use)Bbnd with drain-ore eno Beam Port Floor8666f£tDemineraiizer-I Return V/ste Pool Floor 863.6Ft SI From WoteinSystem L Pit 70 Water CLeanup 706 System Figure 5-2 Pool Water Systems UWNR Safety Analysis Report Rev. 2Sept. 2008* *
Pool Cl)rb E 891.1 pt Port Floor E 866.6ft Figure 5-2 I I I I I I .-)+1 I I I I 8' Lines I Anti-Siphon loop I Future F: I Blinci flange with cirain-r---I one end Pool Water Systems UWNR Safety Analysis Report Rev. 2 5-5 Pool Overf'low I Der'linerolizer Suction I I I I I I I I To \,Jater CleQnl)p 706 SysteM Pool Floor E 863.6f't Sept. 2008 Figure 5-3 Water Cleanup System Schematic UWNR Safety Analysis Report Rev. 2 5-6 Sept. > OJ U:: n: §
5l----<i5-1----Y(-5>-!JD.
'-II-r+-------, Figure 5-3 Water Cleanup System Schematic UWNR Safety Analysis Report Rev. 2 5-6 *
Sept. 2008
5Water Waste System Schematic UWNR Safety Analysis Report Rev. 2Sept.
c C/J ...... '-<:: ;J> ::l r::.. '-<:: Vl ....... Vl ('1) '"d 0 ::+ N VI I -...J C/J .g ;+ N o o 00 .... = Ut I ('1) .., Vl ...... ('1) C/J '-<:: Vl ...... ('1) a C/J g. ('1) a (s. DE'r1lnero.Uzer RinsE' Outlet fron \'/0. tE'r CtE'o.nup SystE"rrt M pool To TC Vent SysteM fo.nk-o-Meter LevE"l Indicator A5 Vent \*/'J.s-::.e
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... .. .. .. B 113 5 B ..._ _ _ ...!,I ( Figure 5-5 Water Makeup System SchematicReferences There are none.UWNR Safety Analysis Report Rev. 2 5-8 Sept. 2008 I I I I I I I I' Ba.cl< .... pCltyl.lat
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.. C:teo.l"lup Syste" 01.11 Hold 1 ..... 1< ,., 'V, f-+-l r; .... l 4-,-+ 'V, '" [ eva 'V, SV< " ] [ 'V, nCOM'" Cv' " J CV' [ AC',.I f-+-c C,tyI.l4 t ... \0 LJ 1.I .. 5yst ... VD2 CI.IIO i Softent'r
[(p " " S:tlll (\.Ill 8-" I S "0>0 " I I I " I I I I I I I -I I I I I , "I I I ON .... I I IJ _____ __________
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J m r SH til t .... p 1 I ROOM Bl135B jCClo'>dl'nSQtP .. t ..... ________________
--.-1 ,,, Figure S-S Water Makeup System Schematic S.9 References There are none. UWNR Safety Analysis Report Rev. 2 5-8
Held T""I;.
D,sUI.d \Jilt ... 100nl< Sept. 2008
Engineered safety features are not required for this reactor due to low operating power and good fission product retention in the fuel. confinement with a controlled ventilation system is provided, however, to reduce the consequence of fission product release from fuel or experiment malfunctions to even lower levels Confinement The Reactor Laboratory is a 43 foot room of conventional construction within theMechanical Engineering Building ,with a ceiling height of approximately feet in most of the room. The portion of the ceiling above the console area is at a height of 22 feet. through show the outlines of the room and location of major reactor components.
The floor of the room is concrete laid on the ground. The walls are concrete and brick. The ceiling is a 1-1/2 inch steel deck with 2 inches of rigid insulation and a built-up surface.The console area is located in the southwest corner of the Reactor Laboratory.
It is separated on the north and east sides from the laboratory proper wire reinforced glass provided to reduce noise originating from the cooling system and other pumps and equipment.
Two doors, one on the east and one on the north of the console area open into the remainder of the Reactor Laboratory.
The non-shared use Reactor Laboratory ventilation system provides both supply and exhaust air to the console area. The console is therefore within the confinement system of theReactor Laboratory The Reactor Laboratory has no exterior windows, but the control room does have a single borrowed-light window on the south side looking into the visitor's center, room which has two large windows facing the parking lot and stadium. The control room window does not open.There are three single doors; one opening at ground level on the control room south wall into the visitor's center (which connects to the Mechanical Engineering lobby through a locked door), one opening at ground level on the east wall into a locked vestibule (which connects to a public hallway), and one opening at basement level on the west wall into the Reactor Laboratoryauxiliary support space. One double door opens at basement level on the east wall into the Reactor Laboratory auxiliary support space. doors have narrow viewing windows, are interior doors which do not open to public, un-restricted areas, and are fire doors and are not weather-stripped.
No special seals are provided for lines that penetrate the walls. doors are normally closed and locked for security and air flow control considerations.
UWNR Safety Analysis Report Rev. 2 61Sp.20Sept. * *
6 ENGINEERED SAFETY FEATURES 6.1 Summary Description Engineered safety features are not required for this reactor due to low operating power and good fission product retention in the fuel. A confinement with a controlled ventilation system is provided, however, to reduce the consequence of fission product release from fuel or experiment malfunctions to even lower levels 6.2 Detailed Descriptions 6.2.1 Confinement The Reactor Laboratory is a 43 by 70 foot room of conventional construction within the Mechanical Engineering Building ,with a ceiling height of approximately 36 feet in most of the room. The portion of the ceiling above the console area is at a height of22 feet. Figure 6-1 through Figure 6-6 show the outlines of the room and location of major reactor components.
The floor of the room is concrete laid on the ground. The walls are concrete and brick. The ceiling is a 1-1/2 inch steel deck with 2 inches of rigid insulation and a 4-ply, built-up surface. The console area is located in the southwest comer of the Reactor Laboratory.
It is separated on the north and east sides from the laboratory proper by wire reinforced glass provided to reduce noise originating from the cooling system and other pumps and equipment.
Two doors, one on the east and one on the north of the console area open into the remainder of the Reactor Laboratory.
The non-shared use Reactor Laboratory ventilation system provides both supply and exhaust air to the console area. The console is therefore within the confinement system of the Reactor Laboratory The Reactor Laboratory has no exterior windows, but the control room does have a single borrowed-light window on the south side looking into the visitor's center, room 1101, which has two large windows facing the parking lot and stadium. The control room window does not open. There are three single doors; one opening at ground level on the control room south wall into the visitor's center (which connects to the Mechanical Engineering lobby through a locked door), one opening at ground level on the east wall into a locked vestibule (which connects to a public hallway), and one opening at basement level on the west wall into the Reactor Laboratory auxiliary support space. One double door opens at basement level on the east wall into the Reactor Laboratory auxiliary support space. All doors have narrow viewing windows, are interior doors which do not open to public, un-restricted areas, and are fire doors and are not weather-stripped.
No special seals are provided for lines that penetrate the walls. All doors are normally closed and locked for security and air flow control considerations . UWNR Safety Analysis Report Rev. 2 6-1 Sept. 2008 The Reactor Laboratory auxiliary support space surround s the Reactor Laboratory on the west, north, and east basement level (See Figure This auxiliary support space contains small rooms having concrete walls on the order of 8 inches thick. The small rooms house thepneumatic tube equipment (B 1135C) and dispatch station (B 1135B), a sample preparation room (B 1135D), air activity monitor equipment (B 1135E), and both general storage (B 1215D) and radioactive storage areas (BI135A andB1215C).
A small hot cell is located on the east end (B1215C), an instrumentation shop is located in the north-west comer, while the west end of the room is a counting laboratory with HPGe, Nal, and proportional swipe counters. The east end is used for teaching nuclear engineering laboratory classes (NE 427 and 428).The ground-floor level of the Mechanical Engineering Building to the east of the Nuclear Reactor Laboratory houses an office area for the Reactor Laboratory.
The Reactor Laboratory is a restricted area. All doors are kept locked at all times except when authorized personnel are in the room. Keys are issued to a small number of authorized personnel.
6.2.2 Containment No containment is needed or provided.6.2.3 Emergency Core Cooling SystemNo emergency core cooling system is required due to the low operating power.References There are no references for this chapter.
UWNR Safety Analysis Report Rev. 2Sept. The Reactor Laboratory auxiliary support space surrounds the Reactor Laboratory on the west, north, and east basement level (See Figure 6-1). This auxiliary support space contains small rooms having concrete walls on the order of 8 inches thick. The small rooms house the pneumatic tube equipment (Bl135C) and dispatch station (Bl135B), a sample preparation room (Bl135D), air activity monitor equipment (Bl135E), and both general storage (B1215D) and radioactive storage areas (Bl135A and.B1215C).
A small hot cell is located on the east end (B 1215C), an instrumentation shop is located in the north-west corner, while the west end of the room is a counting laboratory with HPGe, NaI, and proportional swipe counters.
The east end is used for teaching nuclear engineering laboratory classes (NE 427 and 428). The ground-floor level ofthe Mechanical Engineering Building to the east of the Nuclear Reactor Laboratory houses an office area for the Reactor Laboratory.
The Reactor Laboratory is a restricted area. All doors are kept locked at all times except when authorized personnel are in the room. Keys are issued to a small number of authorized personnel.
6.2.2 Containment No containment is needed or provided.
6.2.3 Emergency Core Cooling System No emergency core cooling system is required due to the low operating power. 6.3 References There are no references for this chapter.
UWNR Safety Analysis Report Rev. 2 6-2 Sept. 2008 . * *
Figure Reactor Laboratory Basement Floor Plan UWNR Safety Analysis Report Rev. 2 6-3Sept. 2008
*
UWNR Safety Analysis Report Rev. 2 6-3 Sept. 2008 I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I I Figure 6-2 Reactor Laboratory First Floor Plan UWNR Safety Analysis Report Rev. 2 6-4 Sept. 2008*
Figure 6-2 Reactor Laboratory First Floor Plan UWNR Safety Analysis Report Rev. 2 6-4 Sept. 2008
Figure Reactor Confinement Cross Section Through Core Centerline, Facing South UWNR Safety Anslysis Report Rev. 2Sept. 2008*
Figure 6-3 Reactor Confinement Cross Section Through Core Centerline, Facing South
UWNR Safety Anslysis Report Rev. 2 6-5 Sept. 2008 ..
Figure 6-4 Reactor Confinement Cross Section Through Core Centerline, Facing North UWNR Safety Anslysis Report Rev. 2 6-6 Sept. 2008 I I. I I I I I I I I I I I I I I I I I Figure 6-4 Reactor Confinement Cross Section Through Core Centerline, Facing North I UWNR Safety Anslysis Report Rev. 2 6-6 Sept. 2008 * *
Figure Reactor Confinement Facing East UWNR Safety Anslysis Report Rev. 2Sept. * *
Figure 6-5 Reactor Confinement Cross Section Through Core Centerline, Facing East UWNR Safety Anslysis Report Rev. 2 6-7 Sept. 2008 T U WNK Safety Analysis Keport Key. 2Sept. *
melmellt Cross SectIOn Through Core Centerline, Facing West UWNR Safety Analysis Report Rev. 2 6-8 Sept. 2008
7 CONTROL Summary Description The reactor operates in three standard modes: Mode 1 Manual or automatic operation at power levels up to 1,000 KW.Mode 2 Square-wave operation (reactivity insertions to reach a desired steady state power level essentially instantaneously) at power levels between 100 and 1,000 KW.Mode 3 Pulsed operation produced by rapid transient rod withdrawal that results in a step insertion of reactivity up to the reactivity limit established in the Technical Specifications.
A selector switch is provided to select manual, automatic, square-wave, or pulsing modes of operation.
Operation is from a console displaying all pertinent reactor operation conditions.
Instrumentation is entirely analog, except for a digital chart recorder, digital fuel temperature indication, and a computer based pulse recorder used to display the power trace during pulsing operation.Design of Instrumentation and Control Systems Four ranges of radiation-based power instrumentation, with significant overlap, are provided to cover the operating range from source level to the maximum permitted pulse power.
Fuel temperature is also measured and used by the reactor protection system. In addition, otherprocess variables are measured, but not used in the reactor protection system. Figure 7-1 shows the instrumentation and control system for the UWNR.7.2.1 Design Criteria The instrumentation and control system provides the following functions:
Provides the operator with information on the status of the reactorProvides the means for insertion and withdrawal of control elements* Provides for automatic control of reactor power levelProvides the means for detecting over-power or fuel over-temperature and automatically scram the control elements to terminate the conditionProvides auxiliary trip functions based on possible loss of operability of the channels providing the overpower protectionProvides a record of operation and radioactivity discharged from the stack UWNR Safety Analysis Report Rev. 2Sept. * *
7 INSTRUMENTATION AND CONTROL SYSTEMS 7.1 Summary Description The reactor operates in three standard modes: Mode 1 Manual or automatic operation at power levels up to 1,000 KW. Mode 2 Square-wave operation (reactivity insertions to reach a desired steady state power level essentially instantaneously) at power levels between 100 and 1,000 KW. Mode 3 Pulsed operation produced by rapid transient rod withdrawal that results in a step insertion of reactivity up to the reactivity limit established in the Technical Specifications.
A selector switch is provided to select manual, automatic, square-wave, or pulsing modes of operation.
Operation is from a console displaying all pertinent reactor operation conditions.
Instrumentation is entirely analog, except for a digital chart recorder, digital fuel temperature indication, and a computer based pulse recorder used to display the power trace during pulsing operation.
7.2 Design of Instrumentation and Control Systems Four ranges of radiation-based power instrumentation, with significant overlap, are provided to cover the operating range from source level to the maximum permitted pulse power. Fuel temperature is also measured and used by the reactor protection system. In addition, other process variables are measured, but not used in the reactor protection system. Figure 7-1 shows the instrumentation and control system for the UWNR. 7.2.1 Design Criteria The instrumentation and control system provides the following functions:
Provides the operator with information on the status of the reactor
Provides the means for insertion and withdrawal of control elements
Provides for automatic control of reactor power level
Provides the means for detecting over-power or fuel over-temperature and automatically scram the control elements to terminate the condition
Provides auxiliary trip functions based on possible loss of operability of the channels providing the overpower protection
Provides a record of operation and radioactivity discharged from the stack UWNR Safety Analysis Report Rev. 2 7-1 Sept. 2008 I onOo[-I 0LLJ -L<L 2" 0Lio < 0<0 <0 wLiLw00 w0>1n S .0.0 0<<< COO~Q i Ceo0-0 0-0Cnc 00 Li a:0 00 Oh- Li:2L0~
~ ~ ~ ~ ~ F 7- IntuettonadCnrlBlc iga UWNR Safety Analysis Report Rev. 2 7-2 Sept. 2008 c ;;0 rn rt' >-::s e. '< en -. en "0 0 ::t. -< N -...,J I N rn (1) "0 r+ N o o 00 ... C ""l -...l I .... -::s en ...... ""I S (1) ::s ...... a o* ::s 5-n 0 8 -to -0 () u -. 3
STARTUP -+LOG N CHANNEL -I-SAFETY CHANNELS ,-CHANNEL +-SERVO CONTR0Y.RADIA TION MONITOR+-PULSE -------.l 4 FUEL TEMPERA CHANNEL CHANNELS CHANNEl' CHANNEL FISSION COUNTER LOG COUNT RATE AMPLIFIER HIGH VOLTAGE DISTRIBUTION AND PULSING CURRENT LIMITING COM PENS A TED ION CHAMBER LOG N & PERIOD AMPLIFIER MUL TI-PEN RECORDER THERMOCOUPLES COMPENSA TED ION CHAMBER COMPENSA TED ION CHAMBER I METERI I BLADEI-eJ , TRANSI # 2 ROD R G BLADE SAFETY MAGNETS POSITION INDICATION AND TRANSIENT ROD DRIVES
LAB AREA MON. GM SENSORS EXPERIMENTAL FACILITY MON. STACK AND CONT. AIR MONITORS GASEOUS ACTIVITY MONITOR GAMMA-RAY ION CHAMBER DIGITAL RECORDER FUEL '----------jl TEMP METER FUEL COUPLE
7.2.2 Design-Basis Requirements The primary design basis for TRIGA reactor safety is the safety limit on fuel temperature.
A trip on high fuel temperature is set to assure that the fuel temperature will not be exceeded.
Since fuel temperature measurement includes time lag due to thermocouple response time, a reactortrip based on reactor power level as measured by a neutron sensing system is provided.7.2.3 System Description 7.2.3.1 Start-up Channel As shown in Figure , the sensing element for this channel is a fission counter with a drive which can be positioned by the console operator.
The counter has a range from 2 nv to 106 nv.Since the counter is moveable, its effective range is thus from about 2 micro-watts to 2 MW.The pulses from the startup counter are amplified and converted to a logarithmic count-rate displayed on a meter and recorded.
The amplified pulses may also be sent to a scaler that is used for subcritical measurements.
The amplifier includes a normally open relay which allows control element withdrawal only if the count rate is greater than 2 counts/second.
Anothernormally closed relay provides protection to the fission counter by preventing insertion of thefission counter drive when the count rate is too high. The start-up channel, in the full in position, overlaps the low end of the safety channel range instruments.7.2.3.2 Log N -Period Channel This channel monitors the power level of the reactor over the range from 0.1 watt to full power.The Log N -period amplifier detects the signal from a compensated ionization chamber and amplifies the signal to provide a 7-decade logarithmic display proportional to power level. The amplifier also extracts period (startup rate) information.
The Log N signal is recorded and operates a normally open relay used in pulse and square wave modes to prevent firing the transient rod when above 1 kW. The period signal is recorded, displayed on a meter on theconsole, and fed to the automatic control channel when the mode switch is in AUTO mode.7.2.3.3 Pulse Power ChannelCurrent from a gamma ionization chamber (or an uncompensated neutron ionization chamber) is fed to a digital data acquisition channel. In "PULSE" mode, a signal concurrent with firing the transient rod causes data to be recorded and displayed on the console computer. Information on peak power and integrated power in the pulse is automatically computed and displayed.
Peak fuel temperature after the pulse is also recorded and displayed on the console computer.UWNR Safety Analysis Report Rev.
2Sept. * *
7.2.2 Design-Basis Requirements The primary design basis for TRIGA reactor safety is the safety limit on fuel temperature.
A trip on high fuel temperature is set to assure that the fuel temperature will not be exceeded.
Since fuel temperature measurement includes time lag due to thermocouple response time, a reactor trip based on reactor power level as measured by a neutron sensing system is provided.
7.2.3 System Description 7.2.3.1 Start-up Channel As shown in Figure 7-1, the sensing element for this channel is a fission counter with a drive which can be positioned by the console operator.
The counter has a range from 2 nv to 10 6 nv. Since the counter is moveable, its effective range is thus from about 2 micro-watts to 2 MW. The pulses from the startup counter are amplified and converted to a logarithmic count-rate displayed on a meter and recorded.
The amplified pulses may also be sent to a scaler that is used for subcritical measurements.
The amplifier includes a normally open relay which allows control element withdrawal only ifthe count rate is greater than 2 counts/second.
Another normally closed relay provides protection to the fission counter by preventing insertion of the fission counter drive when the count rate is too high. The start-up channel, in the full in position, overlaps the low end of the safety channel range instruments.
7.2.3.2 Log N -Period Channel This channel monitors the power level of the reactor over the range from 0.1 watt to full power. The Log N -period amplifier detects the signal from a compensated ionization chamber and amplifies the signal to provide a 7 -decade logarithmic display proportional to power level. The amplifier also extracts period (startup rate) information.
The Log N signal is recorded and operates a normally open relay used in pulse and square wave modes to prevent firing the transient rod when above 1 kW. The period signal is recorded, displayed on a meter on the console, and fed to the automatic control channel when the mode switch is in AUTO mode. 7.2.3.3 Pulse Power Channel Current from a gamma ionization chamber (or an uncompensated neutron ionization chamber) is fed to a digital data acquisition channel. In "PULSE" mode, a signal concurrent with firing the transient rod causes data to be recorded and displayed on the console computer.
Information on peak power and integrated power in the pulse is automatically computed and displayed.
Peak fuel temperature after the pulse is also recorded and displayed on the console computer.
UWNR Safety Analysis Report Rev. 2 7-3 Sept. 2008 I 7.2.3.4 Safety ChannelsTwo safety channels monitor reactor power level from about 0.1 watt to full power. The signal from each channel originates in a compensated ionization chamber. The chamber signal is fed into a solid state picoammeter.
The picoammeter includes normally open relay contacts which open on an overpower condition to cause a reactor scram. Should either or both safety channel scram signals be present, the reactor shuts down.
The power level scram trip point is set to 1.25 times the maximum operating level. The safety channels provide an additional interlock signal to pulse and square wave logic to prevent firing the transient rod unless the picoammeters are on the full-power range.7.2.3.5 Temperature Measurements Fuel element internal temperature is indicated at the console. It causes an alarm and scram at the limiting safety system setting.
The temperature of the bulk pool water is measured at the core inlet by a thermocouple.
This temperature is indicated on the console recorder and causes an alarm and a scram on high temperature.
Primary cooling, intermediate cooling, and campus chilled water systems inlet and outlet temperatures, and demineralizer inlet temperature are indicated on the console recorder.
An alarm on this recorder indicates excessive temperature at any of these points.7.2.4 System Performance Analysis The instrumentation and control systems have been in routine operation for the almost 50 years of operating history. All of the measuring instruments have been replaced over the years withinstruments incorporating advances in electronics but meeting or exceeding the original designcriteria. The switch to entirely solid-state electronics has resulted in marked improvement in stability and reliability.
Limiting safety system settings, limiting conditions for operation, surveillance requirements, and action statements concerning the control and instrumentation systems are detailed in the proposed Technical Specifications in Chapter 14.7.2.5 ConclusionOperation during the term of the license has shown the instrumentation and control system to be capable of performing all intended functions with excellent stability and reliability.
The system is expected to continue to perform the intended functions.
UWNR Safety Analysis Report Rev. 2 7-4 Sept. 7.2.3.4 Safety Channels Two safety channels monitor reactor power level from about 0.1 watt to full power. The signal from each channel originates in a compensated ionization chamber. The chamber signal is fed into a solid state picoammeter.
The picoammeter includes normally open relay contacts which open on an overpower condition to cause a reactor scram. Should either or both safety channel scram signals be present, the reactor shuts down. The power level scram trip point is set to 1.25 times the maximum operating level. The safety channels provide an additional interlock signal to pulse and square wave logic to prevent firing the transient rod unless the picoammeters are on the full-power range. 7.2.3.5 Temperature Measurements Fuel element internal temperature is indicated at the console. It causes an alarm and scram at the limiting safety system setting. The temperature of the bulk pool water is measured at the core inlet by a thermocouple.
This temperature is indicated on the console recorder and causes an alarm and a scram on high temperature.
Primary cooling, intermediate cooling, and campus chilled water systems inlet and outlet temperatures, and demineralizer inlet temperature are indicated on the console recorder.
An alarm on this recorder indicates excessive temperature at any of these points. 7.2.4 System Performance Analysis The instrumentation and control systems have been in routine operation for the almost 50 years of operating history. All of the measuring instruments have been replaced over the years with instruments incorporating advances in electronics but meeting or exceeding the original design criteria.
The switch to entirely solid-state electronics has resulted in marked improvement in .. stability and reliability.
Limiting safety system settings, limiting conditions for operation, surveillance requirements, and action statements concerning the control and instrumentation systems are detailed in the proposed Technical Specifications in Chapter 14. 7.2.5 Conclusion Operation during the term of the license has shown the instrumentation and control system to be capable of performing all intended functions with excellent stability and reliability.
The system* is expected to continue to perform the intended functions.
UWNR Safety Analysis Report Rev. 2 7-4 Sept. 2008 * *
The safety limits of fuel temperature and reactor power are adequately protected the safety channels and the fuel temperature channels, each of which will cause a reactor shutdown if the limiting safety system settings are exceeded. Additional components which cause trips on loss of conditions necessary for continued power operation (loss of high voltage to detectors, loss of water from the pool, pool water temperature above the temperature used in calculation of event consequences, loss of electrical power) provide assurance that the primary protection equipment will operate as planned.
Mode Switch The mode switch and associated logic circuits provide the following capabilities and operating restrictions for the different positions.
The switch is a cam operated switch which is rotatedthrough consecutive positions.
From rotating clockwise places the switch in mode; rotating counter-clockwise places the switch first in WAVE" mode, then in mode. Returning from to requires going through theWAVE" position. details the conditions and restrictions invoked in the different mode switch positions.
Manual Operation For manual operation the control elements are slowly withdrawn to obtain the desired power level. At this level the reactor may continue to be operated manually or it may be switched to automatic control. The automatic control channel maintains power level servo control of the regulating blade, transient rod, or #2 control blade. shows a block diagram of the control system.
Square Wave Operation This mode is provided for those applications which require that the power level be brought rapidly to some high level, held there for a period of time, and then reduced rapidly producing a square wave of power.In the square wave mode the reactor is brought to a level of less than watts in the manual mode. The mode switch is then changed to the square wave position. Changing of the mode switch to the WAVE" position removes an interlock that prevents application of airto the transient rod unless the transient rod is in the full "IN" position. preadjusted step reactivity change i then made to bring the reactor to preset power levels between and kW. The reactivity step change is made with the transient rod. The automatic control system inserts additional reactivity required to maintain the preset power level as the fuiel heats up. The operator must manually augment the reactivity inserted the servo because the transient roddoes not have sufficient worth to overcome the power defect at high power levels. The linear UWNR Safety Analysis Report Rev. 2 75Sp.20Sept. * *
The safety limits of fuel temperature and reactor power are adequately protected by the safety channels and the fuel temperature channels, each of which will cause a reactor shutdown if the limiting safety system settings are exceeded.
Additional components which cause trips on loss of conditions necessary for continued power operation (loss of high voltage to detectors, loss of water from the pool, pool water temperature above the temperature used in calculation of event consequences, loss of electrical power) provide assurance that the primary protection equipment will operate as planned. 7.3 Reactor Control System 7.3.1 Mode Switch The mode switch and associated logic circuits provide the following capabilities and operating restrictions for the different positions.
The switch is a cam operated switch which is rotated through consecutive positions.
From" MANUAL", rotating clockwise places the switch in "AUTO" mode; rotating counter-clockwise places the switch first in "SQUARE WAVE" mode, then in "PULSE" mode. Returning from "PULSE" to "MANUAL" requires going through the "SQUARE WAVE" position.
Table 7-1 details the conditions and restrictions invoked in the different mode switch positions.
7.3.2 Manual Operation For manual operation the control elements are slowly withdrawn to obtain the desired power level. At this level the reactor may continue to be operated manually or it may be switched to automatic control. The automatic control channel maintains power level by servo control ofthe regulating blade, transient rod, or #2 control blade. Figure 7-1 shows a block diagram of the control system. 7.3.3 Square Wave Operation This mode is provided for those applications which require that the power level be brought rapidly to some high level, held there for a period of time, and then reduced rapidly producing a square wave of power. In the square wave mode the reactor is brought to a level of less than 1000 watts in the manual mode. The mode switch is then changed to the square wave position.
Changing of the mode switch to the "SQUARE WAVE" position removes an interlock that prevents application of air to the transient rod unless the transient rod is in the full "IN" position.
A preadjusted step reactivity change is then made to bring the reactor to preset power levels between 100 and 1000 kW. The reactivity step change is made with the transient rod. The automatic control system inserts additional reactivity required to maintain the preset power level as the fuel heats up. The operator must manually augment the reactivity inserted by the servo because the transient rod does not have sufficient worth to overcome the power defect at high power levels. The linear UWNR Safety Analysis Report Rev. 2 7-5 Sept. 2008 power level scram is maintained at 1.25 P max. and an interlock prevents initiation of this mode if the range switch is not on the full power range setting.7.3.4 Pulsing Operation The reactor is brought to a power level of less than 1000 watts in steady state mode. The mode switch is then changed to pulsing mode. When the switch is in pulsing mode the normal neutron channels are disconnected and a high level pulsing chamber is connected to read out the peak power of the pulse on the console computer. Changing of the mode switch to the "PULSE" position removes an interlock that prevents application of air to the transient rod unless the transient rod is in the full "IN" position.
Fuel temperature is recorded during pulsing operation.
The pulse channels are also indicated on Figure N~ nali~l/5: ys ]3I s £pitRv. Sept. /-VVO power level scram is maintained at 1.25 P max. and an interlock prevents initiation of this mode
if the range switch is not on the full power range setting. 7.3.4 Pulsing Operation The reactor is brought to a power level of less than 1000 watts in steady state mode. The mode switch is then changed to pulsing mode. When the switch is in pulsing mode the normal neutron channels are disconnected and a high level pulsing chamber is connected to read out the peak power of the pulse on the console computer.
Changing of the mode switch to the "PULSE" position removes an interlock that prevents application of air to the transient rod unless the transient rod is in the full "IN" position.
Fuel temperature is recorded during pulsing operation.
The pulse channels are also indicated on Figure 7-1 . UWNR Safety Analysis Report Rev. 2 7-6 Sept. 2008 *
Table Mode Switch Functions Mode Switch Conditions/restrictions Position Manual MANUAL Transient rod may be fired only if scram is reset and transient rod drive is at the "IN" position.Automatic AUTO Same as Manual, except automatic control system can controlpower, subject to period limit Square- SQUARE Transient rod can be fired from other than full in position only if Wave WAVE scram is reset, both safety channels are on top range, and power level does not exceed 1 kW as indicated by the LogN channel.
The period channel in the LogN amplifier is defeated (restored after a short time delay when the switch is returned to "MANUAL" position).
Automatic control system can control power if actual power is within +5% of scheduled power.Pulse PULSE Period channel remains defeated.
Prohibits control blade withdrawal.
If the scram is reset, both safety channels are on top range, and power level does not exceed 1 kW as indicated' by the LogN channel;(a) Transient rod can be fired from other than full in position, (b) High voltage is removed from the fission counter, safety channel CICs , and the LogN CIC.(c) Signals from all CICs are directed to ground rather than to instrument input.(d) The transient rod drops automatically 15 seconds or less after it is fired (e) The pulse power level channel is sent a signal causing it to record the pulse power trace.When returning to the "SQUARE/WAVE" position, high voltage is restored to the detectors immediately and the signals to the neutron measuring instruments are restored after a short time delay (to prevent damage to instrument inputs from the transients resultingfrom high voltage restoration).
UWNR Safety Analysis Report Rev. 2Sept. Table 7-1 Mode Switch Functions
Mode Switch Conditions/restrictions Position Manual MANUAL Transient rod may be fired only if scram is reset and transient rod drive is at the "IN" position.
Automatic AUTO Same as Manual, except automatic control system can control power, subject to period limit Square-SQUARE Transient rod can be fired from other than full in position only if Wave WAVE scram is reset, both safety channels are on top range, and power level does not exceed 1 kW as indicated by the LogN channel. The period channel in the LogN amplifier is defeated (restored after a short time delay when the switch is returned to "MANUAL" position).
Automatic control system can control power if actual power is within +/-5% of scheduled power. Pulse PULSE Period channel remains defeated.
Prohibits control blade withdrawal.
If the scram is reset, both safety channels are on top range, and power level does not exceed 1 kW as indicated' by the LogN channel; (a) Transient rod can be fired from other than full in position, (b) High voltage is removed from the fission counter, safety channel CICs , and the LogN CIC . * (c) Signals from all CICs are directed to ground rather than to instrument input. (d) The transient rod drops automatically 15 seconds or less after it is fired (e) The pulse power level channel is sent a signal causing it to record the pulse power trace. When returning to the "SQUARE/WAVE" position, high voltage is restored to the detectors immediately and the signals to the neutron measuring instruments are restored after a short time delay (to prevent damage to instrument inputs from the transients resulting from high voltage restoration) .
UWNR Safety Analysis Report Rev. 2 7-7 Sept. 2008 7.3.5 Control Element Operation There are five control elements; three shim-safety blades, a transient control rod, and a regulating blade.. The shim-safety blades and the transient control rod have scram capability.
The following conditions must be met before any control element drive can be withdrawn (raised), either manually or the automatic control system: 1 No scram conditions present and scram relays reset;2. Count-rate on startup channel greater than 2 counts per second;Fission Counter not in motion;4. Console key switch set to position;There are no interlocks or permissives which restrict insertion (lowering) of control element drives. Insertion is accomplished placing the individual momentary-contact control switches in the position, or a maintained-contact switch which inserts all control element drives to the limit. The three shim/safety blade drives also automatically run to the limit when a SCRAM has occurred.
Safety Blade Control The three safety blades are manually controlled individual pistol-grip, switches with LOWER, OFF, and RAISE positions, with spring return to OFF. One safety blade may be selected to be controlled the automatic level control system.
The position of each safety blade is indicatedseparate digital read-outs, and the indicator lights on the console show when each drive is at its or limit and when the blade magnets are engaged with the armatures Position indication is accurate to +/-0O.02 inches.The safety blades will scram from any position when stationary or during withdrawal and insertion.
In the event of a scram the safety blade drives automatically run in to their limits.
Regulating Blade Control The regulating blade has identical position indication and and limit indication.
It is manually controlled a separate pistol-grip switch and may be controlled the automatic level control system.UWNR Safety Analysis Report Rev. 2 78Sp.20Sept. 7.3.5 Control Element Operation There are five control elements; three shim-safety blades, a transient control rod, and a regulating blade. The shim-safety blades and the transient control rod have scram capability.
The following conditions must be met before any control element drive can be withdrawn (raised), either manually or by the automatic control system: 1. No scram conditions present and scram relays reset; 2. Count-rate on startup channel greater than 2 counts per second; 3. Fission Counter not in motion; 4. Console key switch set to "ON" position; There are no interlocks or permissives which restrict insertion (lowering) of control element drives. Insertion is accomplished by placing the individual momentary-contact control switches in the "IN" position, or by a maintained-contact "RUNDOWN" switch which inserts all control element drives to the "IN" limit. The three shim/safety blade drives also automatically run to the "IN" limit when a SCRAM has occurred.
7.3.6 Safety Blade Control The three safety blades are manually controlled by individual pistol-grip switches with LOWER, OFF, and RAISE positions, with spring return to OFF. One safety blade may be selected to be controlled by the automatic level control system. The position of each safety blade is indicated by separate digital read-outs, and the indicator lights on the console show when each drive is at its "IN" or "OUT" limit and when the blade magnets are engaged with the armatures.
Position indication is accurate to +/-0.02 inches. The safety blades will scram from any position when stationary or during withdrawal and insertion.
In the event of a scram the safety blade drives automatically run in to their "IN" limits. 7.3.7 Regulating Blade Control The regulating blade has identical position indication and "IN" and "OUT" limit indication.
It is manually controlled by a separate pistol-grip switch and may be controlled by the automatic level control system. UWNR Safety Analysis Report Rev. 2 7-8 Sept. 2008 * *
7.3.8 Transient Rod Control Manual movement of the transient rod drive is controlled by the console-mounted, switch/light, push buttons which not only control movement, but also indicate in and out limits.
Position indication is accurate to 0.02 inches. The transient rod drive may be selected to be controlledby the automatic level control system.Air pressure is used to fire the transient rod to the selected position in pulse and square-wavemodes and to engage and hold the transient rod at the drive position in manual and automatic modes. Additional lighted push-button switches are installed to support these functions.
Illumination of the "READY/FIRE T ROD" indicator/switch indicates that the permissives for firing are met and the rod has not been fired.
Illumination of the "ENG'D/AIR" indicator/switch indicates movement of the transient rod from the full-in rest position as a result of air having been applied. (Applying air while the transient rod drive is in the full in position causes the piston in the drive to move upward by compressing the spring inside the shock absorber).
Pressing the "ENG'D/AIR" switch removes the air from the drive, causing the transient rod to drop to the full-in rest position.Since the transient rod control is capable of introducing step changes in reactivity up to the Technical Specification limit, logic circuits are provided to assure the transient control rod is fired only under the appropriate conditions.
7.3.9 Automatic Level Control System The servo amplifier (level controller) controls reactor power level in automatic and square-wavemodes. The servo amplifier output drives either the regulating blade, transient rod, or a safety blade as selected by a servo element selector switch on the console. Only one element at a time may be selected; selecting another replaces the previously selected element. The servo amplifier responds to a power level signal from one of the safety channel picoammeters and controls speed and direction of the servo element through a servo motor.In automatic mode, the automatic level control channel uses period information from the Log N-period channel to limit control element withdrawal to maintain a period longer than a pre-selected level. In square-wave mode a servo error circuit is employed.
This circuit allows servo operation only when the servo error is less than 5%. Servo error (the difference between scheduled power in % and the picoammeter indication in % of full scale) is indicated at the console in both "square-wave" and "automatic" modes.
Additional indicators on the reactor control console are provided to indicate "AUTO ON" and scheduled power.UWNR Safety Analysis Report Rev. 2Sept. * *
7.3.8 Transient Rod Control Manual movement of the transient rod drive is controlled by the console-mounted, switch/light, push buttons which not only control movement, but also indicate in and out limits. Position indication is accurate to 0.02 inches. The transient rod drive may be selected to be controlled by the automatic level control system. Air pressure is used to fire the transient rod to the selected position in pulse and square-wave modes and to engage and hold the transient rod at the drive position in manual and automatic modes. Additional lighted push-button switches are installed to support these functions.
Illumination of the "READY/FIRE T ROD" indicator/switch indicates that the permissives for firing are met and the rod has not been fired. Illumination of the "ENG'D/AIR" indicator/switch indicates movement of the transient rod from the full-in rest position as a result of air having been applied. (Applying air while the transient rod drive is in the full in position causes the piston in the drive to move upward by compressing the spring inside the shock absorber).
Pressing the "ENG'D/ AIR" switch removes the air from the drive, causing the transient rod to drop to the full-in rest position.
Since the transient rod control is capable of introducing step changes in reactivity up to the Technical Specification limit, logic circuits are provided to assure the transient control rod is fired only under the appropriate conditions.
7.3.9 Automatic Level Control System The servo amplifier (level controller) controls reactor power level in automatic and square-wave modes. The servo amplifier output drives either the regulating blade, transient rod, or a safety blade as selected by a servo element selector switch on the console. Only one element at a time may be selected; selecting another replaces the previously selected element. The servo amplifier responds to a power level signal from one of the safety channel picoammeters and controls speed and direction of the servo element through a servo motor. In automatic mode, the automatic level control channel uses period information from the Log N -period channel to limit control element withdrawal to maintain a period longer than a selected level. In square-wave mode a servo error circuit is employed.
This circuit allows servo operation only when the servo error is less than 5%. Servo error (the difference between scheduled power in % and the picoammeter indication in % of full scale) is indicated at the console in both "square-wave" and "automatic" modes. Additional indicators on the reactor control console are provided to indicate "AUTO ON" and scheduled power. UWNR Safety Analysis Report Rev. 2 7-9 Sept. 2008 7.4 Reactor Protection System Scram CircuitsThe scram circuit, which initiates shutdown by dropping the shim-safety blades and the transient rod, is shown in Figure 7-2. Scram is accomplished by de-energizing the scram relays under one of the following conditions:
1. Manual scram;2. Fuel temperature above LSSS;3. Power level greater than 1.25 P max;4. High voltage failure in control console;Loss of control power;6. Coolant temperature at core coolant entrance above 130'F;7. Pool water level high or low.In addition to these scram functions, the transient control rod logic includes a timer which causes the transient rod to drop to the full in position within 15 seconds after the transient rod has been fired in pulse mode only.The key-operated console MASTER switch, designated 4S2 on the figures, is a cam-operated switch with a large number of contacts which are selectively operated in the three switch positions. (OFF, ON, and TEST). Six different contact sets must be closed (and are closed only in the "ON" position of the switch) in order to reset the scram relays and apply power to the trip amplifier which supplies DC voltage to the shim-safety blade magnets.A normally-open contact of Relay 6K1A in the transient rod logic circuit opens when the relay is de-energized, resulting in the drop of the transient rod.Normally-open contacts of both 6K 1 A and 6K 1 B are in series with the alternating current power supply to the trip amplifier as shown in Figure 7-3. Magnet power is turned off if either or both of the scram relays de-energizes.
u WNR Safety Analysis Report Rev. 2Sept. 7.4 Reactor Protection System Scram Circuits The scram circuit, which initiates shutdown by dropping the shim-safety blades and the transient rod, is shown in Figure 7-2. Scram is accomplished by de-energizing the scram relays under one of the following conditions:
1. Manual scram; 2. Fuel temperature above LSSS; 3. Power level greater than 1.25 P max; 4. High voltage failure in control console; 5. Loss of control power; 6. Coolant temperature at core coolant entrance above 130°F; 7. Pool water level high or low.
In addition to these scram functions, the transient control rod logic includes a timer which causes
the transient rod to drop to the full in position within 15 seconds after the transient rod has been fired in pulse mode only. The key-operated console MASTER switch, designated 4S2 on the figures, is a cam-operated switch with a large number of contacts which are selectively operated in the three switch positions. (OFF, ON, and TEST). Six different contact sets must be closed (and are closed only in the "ON" position of the switch) in order to reset the scram relays and apply power to the trip amplifier which supplies DC voltage to the shim-safety blade magnets. A normally-open contact of Relay 6KIA in the transient rod logic circuit opens when the relay is de-energized, resulting in the drop of the transient rod. I Normally-open contacts of both 6KIA and 6KIB are in series with the alternating current power supply to the trip amplifier as shown in Figure 7-3. Magnet power is turned off if either or both of the scram relays de-energizes.
UWNR Safety Analysis Report Rev. 2 7-10 Sept. 2008
wLL LI z ai > V_ I Ln M fl 0- Figure x LL L I < LO I fMO, LI U-ram Logic Elementary Diagram UWNR Safety Analysis Report Rev. 2 7-11 Sept. 2008* c::::: :;:0 C/l (t> > ::I tn* '0 o :4-:< tv -....) I ....... ....... C/l (1) '0 r+ tv o o 00 = ., rD -....l I N C/l () l' o (JQ ...... () tn -(1) S (1) ! ti 1 452 1 I 452 2 ..... TB10-11 TB5-1.11.12 113 TB6-1 505
TB20-2 RY100A 41 POOL LEVEL TB6-2 8 n IX) o o 506 7.............
opens with obsenc:e of 6K2 -"0 VAC ocross p'InS 1 &.8 0'[ 6f PNL 2 circuit 5 'ceiling lights a> 135 g TB6-3 i: 12AR2 J5-H 12K2-2 C r OPENS AT HIGH N. FLUX TB20-3 * '. A .... 12AR2 J5-J bLQoIe 2 " . cooling f'c>n Moo .... TB6-4 5Cl 7 ..... B 1 --6Kl-B TB10-12 115 452 .... / SCRAM 651-1 16 RESET 611 SCR Al'vl RESET 132 ... --------Er.j)
TB 6 -5 2 B 6K1A TB6-8 _.<_. 6Cl 6K2 from TB2-3 1 6K1A TB2-4 3;to2 11 /' ORR1 INLET TEMP. t OPENS AT 130F 10 503 TB2-5 11 j 19M1 OPENS ON HIGH FUEL 12 TB2-6 .... 504 11
--r-OPENS AT HIGH N. FLUX '1' T1B22A: J5-J 139 ('4 15Z1 CONTACTS ...........
+ OPEN ON HIGH --'-6 VOL TAGE FAILURE TB2-8 140 ,2 6K1C 10 :1 '1' 452 1 / TB5-8.9 T MAJ'I. SCRA.M 115 '1 M.'1N SCRA.ivl it' 115 ,TB6-9 TB2-10
TEMP.
??:7 i Ld Li I-CYbLiT)0L7 6iL L0 NJ (N 0 Ln N Li Figure Magnet Power Supply UWNR Safety Analysis Report Rev. 2Sept. c U'J ...... '-< > ::s -< t;Il (j;" (l) "0 0 ::4-(l) -< tv -....l , ...... tv U'J .g ;+ tv o o 00 * .... IJQ = ., 'nl -...l (JQ ::s (l) ...... >-0 0 :E (l) ""1 U'J .g "0 -'-< (A)T610-2 218 452 220 6K1A I \ I 8 110 SCRC,M')
SH,l / en I co m m GO ('., 651-3 SCRA,M RESET o 0T81-16 6 00T81-3 T83-3 T93-4 181-4 T83-6 12Z5 TRIP ELEIv1, DIA,G, 2196145 T61-50 o 181-6 T83-8 T8 2 _ 5 0 I 22 7 --------, 452 221 6K18 223 I I I I 0 183-7 5 TB2-1 T82-2
o 0 (A)T812-8 m '" '" P3 " ,,::If: PLRT OF 7Z1 T8 2 -6 0 1 __ 2_2_8'--__ TB2-3 T82-4 (A)TB11-9 (A)TB9-9 50.0-25W GO n GO P4 P3 '\.-1' '\. "R' DC SAME AS FOR SAME AS FOR 721 7Z1
There are no engineered safety features actuation systems.
Alarm and Indicator System When an abnormal condition develops, an audible signal sounds and a lighted annunciator begins to flash rapidly.
The operator may press the acknowledge button to silence the audible signal, at which time the audible signal stops and the lighted annunciator goes to steady illumination.
When the condition is corrected, the lighted annunciator goes to slow flash, and the light is extinguished when the operator presses the Reset button.The following conditions will actuate the alarm system: 1 High area radiation level (also gives an alarm at Police Department andinitiates building evacuation);
2. High experimental facility radiation level;Radiation monitor failed low;4. Evacuation Alarm in Local;5. Stack Air particulate or gaseous activity above normal level;Air particulate or gaseous activity above normnal level;Trouble in stack or continuous air monitors;Neutron flux exceeding times the normal value;Reactor period less than a preset level;Count rate on startup counter approaching saturation level;Any scram;12. Safety blade disengaged from magnet;Failure of high voltage power supply;14. Loss of Off-Site Power;UWNR Safety Analysis Report Rev. 2 Sept. * *
7.5 Engineered Safety Features Actuation Systems There are no engineered safety features actuation systems. 7.6 Control Console and Display Instruments 7.6.1 Alarm and Indicator System When an abnormal condition develops, an audible signal sounds and a lighted annunciator begins to flash rapidly. The operator may press the acknowledge button to silence the audible signal, at which time the audible signal stops and the lighted annunciator goes to steady illumination.
When the condition is corrected, the lighted annunciator goes to slow flash, and the light is extinguished when the operator presses the Reset button. The following conditions will actuate the alarm system: 1. High area radiation level (also gives an alarm at UW Police Department and initiates building evacuation);
2. High experimental facility radiation level; 3 . Radiation monitor failed low; 4. Evacuation Alarm in Local; 5. Stack Air particulate or gaseous activity above normal level; 6. CAM Air particulate or gaseous activity above normal level; 7. Trouble in stack or continuous air monitors;
8. Neutron flux exceeding 1.15 times the normal value; 9. Reactor period less than a preset level; 10. Count rate on startup counter approaching saturation level; 11. Any scram; 12. Safety blade disengaged from magnet; 13. Failure of high voltage power supply; 14 . Loss of Off-Site Power; UWNR Safety Analysis Report Rev. 2 7-13 Sept. 2008
15. Fuel element temperature high;Core inlet temperature above preset level;Cooling system temperatures above preset level;Intermediate coolant system low pressure.Thermal Column door open;
20. Chain switch across stair actuated or entry to High Radiation Area;21. Hold tank full;22. Water level in pool two or more inches above or below normal (also gives an alarm at UW Police Department);
7.6.2 Indicator Lights To provide operating information for the reactor operator, the following indicator lights are provided: 1. Scram reset;2. Safety blade magnet engaged;3. Power on;4. Control elements in (distinct light for each);5. Control elements.out (distinct light for each);
6. Automatic control on;7.6.3 Pneumatic Tube System Panel The pneumatic system control panel provides indication lights for system on, system purge in progress, isolation valves open, and rabbit in reactor. Buttons allow the console operator to start and stop the system, and emergency return a rabbit sample. The pneumatic system is described in section 10.2.3.U W N R Satety Analysis Report Rev. 2/-14 Sept. 2UooS 15. Fuel element temperature high; 16. Core inlet temperature above preset level; 17. Cooling system temperatures above preset level; 18. Intermediate coolant system low pressure.
19. Thermal Column door open; 20. Chain switch across stair actuated or entry to High Radiation Area; 21. Hold tank full; 22. Water level in pool two or more inches above or below normal (also gives an alarm at UW Police Department);
7.6.2 Indicator Lights To provide operating information for the reactor operator, the following indicator lights are provided:
1. Scram reset; 2. Safety blade magnet engaged; 3. Power on; 4. Control elements in (distinct light for each); 5. Control elements out (distinct light for each); 6. Automatic control on; 7.6.3 Pneumatic Tube System Panel The pneumatic system control panel provides indication lights for system on, system purge in progress, isolation valves open, and rabbit in reactor. Buttons allow the console operator to start and stop the system, and emergency return a rabbit sample. The pneumatic system is described in section 10.2.3. UWNR Safety Analysis Report Rev. 2 7-14 Sept. 2008 * *
7.6.4 Ventilation System Panel The ventilation system control panel provides indication for both of the main exhaust fans (EF-7 and EF-8), the air handling unit (AHU-5), the fume hood exhaust fan (EF- 13), and the Beam Port& Thermal Column exhaust fan (EF-17). Each fan has separate lights indicating whether the fan is running and whether the fan has power available to run if needed. All fans except the fume hood exhaust fan (EF-13) include switches for operation.
Digital indications are provided for stack exhaust flow-rate (scfm), differential pressure (inches of water column) across the mainexhaust filter bank, and static duct pressure (inches of water column) in the Beam Port &Thermal Column exhaust duct before the exhaust fan EF- 17. The ventilation system is described in section 9.1.7.6.5 Cooling System Panel Operating controls and indicators at the control console for the cooling system include switches to operate primary, intermediate, chilled water, and diffuser pumps (the hydraulic irradiation facility pump starts when the diffuser pump starts) along with indicators which light when the pumps have discharge pressure.
In addition, a pressure switch on the intermediate coolant system provides indication that the system is pressurized.
The cooling system is described in chapter 5.7.6.6 Whale System Panel The hydraulic irradiation facility (whale system) includes indicator lights for flow direction, sample in, and buttons to reverse flow direction, as well as indication that the whale pump hasdischarge pressure.
The whale system is described in section 10.2.4.7.7 Radiation Monitoring Systems 7.7.1 Area Radiation Monitors The radiation monitors are arranged into three systems; the primary area monitors, experimentalfacility area monitors, and air activity monitors.The primary area monitors are located as follows: 1. Demineralizer area;2. On the reactor bridge about one foot above the water surface;3. Beside the thermal column door;4. In the control console area.
All Area Radiation monitor units have ranges from 0.1 to 10000 mr/hr.UWNR Safety Analysis Report Rev. 2Sept. * *
7.6.4 Ventilation System Panel The ventilation system control panel provides indication for both of the main exhaust fans (EF-7 and EF-8), the air handling unit (AHU-5), the fume hood exhaust fan (EF-13), and the Beam Port & Thermal Column exhaust fan (EF-17). Each fan has separate lights indicating whether the fan is running and whether the fan has power available to run if needed. All fans except the fume hood exhaust fan (EF-13) include switches for operation.
Digital indications are provided for stack exhaust flow-rate (scfm), differential pressure (inches of water column) across the main exhaust filter bank, and static duct pressure (inches of water column) in the Beam Port & Thermal Column exhaust duct before the exhaust fan EF-17. The ventilation system is described in section 9.1. 7.6.5 Cooling System Panel Operating controls and indicators at the control console for the cooling system include switches to operate primary, intermediate, chilled water, and diffuser pumps (the hydraulic irradiation facility pump starts when the diffuser pump starts) along with indicators which light when the pumps have discharge pressure.
In addition, a pressure switch on the intermediate coolant system provides indication that the system is pressurized.
The cooling system is described in chapter 5. 7.6.6 Whale System Panel The hydraulic irradiation facility (whale system) includes indicator lights for flow direction, sample in, and buttons to reverse flow direction, as well as indication that the whale pump has discharge pressure.
The whale system is described in section 10.2.4. 7.7 Radiation Monitoring Systems 7.7.1 Area Radiation Monitors The radiation monitors are arranged into three systems; the primary area monitors, experimental facility area monitors, and air activity monitors.
The primary area monitors are located as follows: I. Demineralizer area; 2. On the reactor bridge about one foot above the water surface; 3. Beside the thermal column door; 4. In the control console area. All Area Radiation monitor units have ranges from 0.1 to 10000 mr/hr . UWNR Safety Analysis Report Rev. 2 7-15 Sept. 2008 Unit 1 supplies information on radiation level from the demineralizer.
It is set to alarm at a radiation level just above that expected in a normal run.
Unit 2, located just above the pool water level, alarms at a radiation level just above that reached during full power operation.
Unit 3 is located beside the thermal column. It too is set to alarm just above normal operating level. This unit will give an alarm if the thermal column door is left open when the reactor is operated at any substantial power.
Unit 4 indicates the dose rate in the console area. The 4 units indicated above are connected to the Reactor Laboratory evacuation alarm. An alarm from one of these units will sound the evacuation alarm if it is not corrected by the operator within 30 seconds (See procedure UWNR 150).The Experimental Facility Area Radiation Monitor is an area radiation monitor system installed to preclude the possibility of unknowingly generating high radiation levels by operating the reactor at high power levels with the beam ports open, or by return of an intensely radioactivepneumatic tube sample. The sensors for this system are installed on the walls of the ReactorLaboratory in direct line with the beam ports and at the pneumatic tube send-receive station. Thesystem gives visual and audible alarms at the console if the radiation level exceeds a preset value.The pneumatic tube monitor also provides local alarm and indication.
The monitors are normally set to alarm at a radiation level equivalent to a dose rate of 50-100 mrem/hr at the beam port flange (10 mrem/hr at the detector location).
The pneumatic tube monitor also is normally set to 10 mrem/hr, but the pneumatic tube operating procedure states that it may be set to a higher level if calculated sample activity is expected to result in a higher reading.7.7.2 Stack Air Monitor The stack air monitor measures both particulate and gaseous activity of the air discharged from the stack. Particulate activity is collected on filter paper and counted with a thin end-window sealed gas proportional tube. Gaseous activity is also measured with a gas proportional tube.The system operates by detecting activity.
Both particulate and gaseous activity levels are recorded, and provide annunciation should preset levels be exceeded.
In addition, gaseous activity levels are integrated to provide a record of total gaseous activity discharged from the stack.The sensitivity of the particulate activity monitor allows detection of concentrations of about 1.OE-10 [iC/ml of a material with a single particle emitted per disintegration. The efficiency is higher if more than one particle is emitted per disintegration.
The sensitivity of the gaseous activity monitor is such that a concentration of about 1.OE-6 [LCi/ml of Ar41 at the stack discharge can be detected by the instrument. The efficiency varies with the number of particles emitted by the isotope being detected.
The primary activity expected to be present in the stack discharge is Ar41 activity.An identical instrument, provided as a backup for the Stack Air Monitor, is operated as a Continuous Air Monitor. It samples the atmosphere immediately above the surface of the reactor pool, although it can be made to sample other locations when desired. The backup air activity monitor can be connected to the stack monitor flow path, should the stack monitor fail.UWNR Safety Analysis Report Rev. 2Sept. Unit 1 supplies infonnation on radiation level from the demineralizer.
It is set to alann at a radiation level just above that expected in a nonnal run. Unit 2, located just above the pool water
level, alanns at radiation level just above that reached during full power operation.
Unit 3 is located beside the thennal column. It too is set to alann just above nonnal operating level. This unit will give an alann if the thennal column door is left open when the reactor is operated at any substantial power. Unit 4 indicates the dose rate in the console area. The 4 units indicated above are connected to the Reactor Laboratory evacuation alann. An alann from one of these units will sound the evacuation alann ifit is not corrected by the operator within 30 seconds (See procedure UWNR 150). The Experimental Facility Area Radiation Monitor is an area radiation monitor system installed to preclude the possibility of unknowingly generating high radiation levels by operating the reactor at high power levels with the beam ports open, or by return of an intensely radioactive pneumatic tube sample. The sensors for this system are installed on the walls of the Reactor Laboratory in direct line with the beam ports and at the pneumatic tube send-receive station. The system gives visual and audible alanns at the console if the radiation level exceeds a preset value. The pneumatic tube monitor also provides local alann and indication.
The monitors are nonnally set to alarm at a radiation level equivalent to a dose rate of 50-100 mrem/hr at the beam port flange (10 mremlhr at the detector location).
The pneumatic tube monitor also is normally set to 10 mrem/hr, but the pneumatic tube operating procedure states that it may be set to a higher level if calculated sample activity is expected to result in a higher reading. 7.7.2 Stack Air Monitor The stack air monitor measures both particulate and gaseous activity of the air discharged from the stack. Particulate activity is collected on filter paper and counted with a thin end-window sealed gas proportional tube. Gaseous activity is also measured with a gas proportional tube. The system operates by detecting p activity.
Both particulate and gaseous activity levels are recorded, and provide annunciation should preset levels be exceeded.
In addition, gaseous activity levels are integrated to provide a record of total gaseous activity discharged from the stack. The sensitivity of the particulate activity monitor allows detection of concentrations of about 1.0E-I0 !lC/ml of a material with a single p particle emitted per disintegration.
The efficiency is higher if more than one p particle is emitted per disintegration.
The sensitivity of the gaseous activity monitor is such that a concentration of about 1.0E-6 !lCi/ml of Ar 41 at the stack discharge can be detected by the instrument.
The efficiency varies with the number of p particles emitted by the isotope being detected.
The primary activity expected to be present in the stack discharge is Ar41 activity.
An identical instrument, provided as a backup for the Stack Air Monitor, is operated as a Continuous Air Monitor. It samples the atmosphere immediately above the surface of the reactor pool, although it can be made to sample other locations when desired. The backup air activity monitor can be connected to the stack monitor flow path, should the stack monitor fail. UWNR Safety Analysis Report Rev. 2 7-16 Sept. 2008 *
References There are no references for this chapter.UWNR Safety Analysis Report Rev. 2Sept. 7.8 References
There are no references for this chapter. *
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UWNR Safety Analysis Report Rev. 2 7-18
Sept. 2008
There are no electrical power supplies that are critical for maintaining the facility in safe shutdown, even for extended periods of time.The Reactor Laboratory's electrical power originates in room B of the MechanicalEngineering Building.
The entire system is shown in Figure Utility power to the building is supplied circuits 1421 and 1422. Below is a description of each electrical circuit used in the Reactor Laboratory.Phase Electrical Poweris provided to the Reactor Laboratory a transformer and through a ground fault interrupted circuit breaker. The Laboratory itself has its own ground fault interrupter and circuit breaker This circuit connects to the Reactor Laboratory's panel (4R1) which is located behind the Reactor Console just outside of the Reactor Control Room.This panel supplies 480 to the Reactor Laboratory's Diffuser and Whale Pumps.Although both of these pumps are on the same circuit, each has its own motor starter. TheDiffuser Pump starter is located immediately behind the Reactor Console in the Reactor Control Room and the Whale Pump starter is located to the right of the Diffuser Pump starter. The Diffuser Pump also has a local disconnect which is located next to the Diffuser Pump on the Reactor's Shield Step. The Diffuser and Whale Pumps can be turned on and off using a single shared toggle switch located on a graphic mimic panel in the Reactor Control Console.
The panel also provides power to the loops of the Reactor Laboratory's Cooling System. Each loop (Primary, Intermediate and Chilled) has its own local disconnect switch located just below the Reactor's Control Room and each can be turned on and off using individual toggle switches located on a graphic mimic panel in the Reactor Control Console.Finally, the panel also supplies power to the Reactor Laboratory's overhead crane and to the primary windings of the Reactor Laboratory's two power transformers, TI and T2. These transformers are described in detail below.Phase Backed Electrical Power The Reactor Laboratory's backed up electrical power is supplied the sametransformer and through the same circuit breaker as the Laboratory's phase normal electrical power. This bus has its own breaker and passes through an Automatic Transfer Switch This automatically UWNR Safety Analysis Report Rev. 2 81Sp.20Sept. * *
8 ELECTRICAL POWER SYSTEMS 8.1 Normal Electrical Power Systems There are no electrical power supplies that are critical for maintaining the facility in safe shutdown, even for extended periods of time. The Reactor Laboratory's electrical power originates in room B 11 0 1 of the Mechanical Engineering Building.
The entire system is shown in Figure 8-1. Utility power to the building is supplied by circuits 1421 and 1422. Below is a description of each electrical circuit used in the Reactor Laboratory.
8.1.1 277/480 VAC 3 Phase Electrical Power 277/480 VAC is provided to the Reactor Laboratory by a 3000 kVA transformer and through a ground fault interrupted 3000 A circuit breaker. The Laboratory itself has its own 225 A ground fault interrupter and circuit breaker (#8). This circuit connects to the Reactor Laboratory's 277/480 panel (4Rl) which is located behind the Reactor Console just outside of the Reactor Control Room. This panel supplies 480 VAC to the Reactor Laboratory's N-16 Diffuser and Whale Pumps. Although both of these pumps are on the same circuit, each has its own motor starter. The Diffuser Pump starter is located immediately behind the Reactor Console in the Reactor Control Room and the Whale Pump starter is located to the right of the Diffuser Pump starter. The Diffuser Pump also has a local disconnect which is located next to the Diffuser Pump on the Reactor's Shield Step. The Diffuser and Whale Pumps can be turned on and off using a single shared toggle switch located on a graphic mimic panel in the Reactor Control Console. The 277/480 panel also provides power to the 3 loops of the Reactor Laboratory's Cooling System. Each loop (Primary, Intermediate and Chilled) has its own local disconnect switch located just below the Reactor's Control Room and each can be turned on and off using individual toggle switches located on a graphic mimic panel in the Reactor Control Console. Finally, the 277/480 panel also supplies power to the Reactor Laboratory's overhead crane and to the primary windings of the Reactor Laboratory's two power transformers, Tl and T2. These transformers are described in detail below. 8.1.2 277/480 VAC 3 Phase Backed Up Electrical Power The Reactor Laboratory's 277/480 VAC backed up electrical power is supplied by the same 277/480 VAC 3000 kVA transformer and through the same 3000 A circuit breaker as the Laboratory's 277/480 VAC 3 phase normal electrical power. This bus has its own 1600 A breaker and passes through an Automatic Transfer Switch (ATS). This ATS automatically UWNR Safety Analysis Report Rev. 2 Sept. 2008 switches the bus's source from utility to the Mechanical Engineering emergency diesel generator.After the ATS, this circuit passes though another 1600 A breaker and a 100 A breaker to a circuit within the Mechanical Engineering Building Panel 4CRl. This panel is located in room 1133 Mechanical Engineering.
A circuit in 4CR1 powers the Reactor Laboratory's backup/makeup air compressor.
When the Reactor Laboratory's compressed air system, provided by the Mechanical Engineering Building, falls below 75 psi, this local air compressor activates.
The Reactor Laboratory's ventilation system is also powered by this bus. Bus EMCC6 is connected to the Reactor Laboratory's 277/480 VAC backed up electrical power through two 400 A fuses. Bus EMCC6 powers the Reactor Laboratory's two ventilation fans, EF-7 and EF-8, as well as the Reactor Laboratory's air handling unit, AHU-5.Finally, the Reactor Laboratory's Support Space basement emergency lights are also powered by the 277/480 VAC backed up electrical bus through panel 4CR2.8.1.3 240 4 wire VAC Electrical Power, Transformer TI 240 4 wire VAC is provided by transformer TI. This transformer's primary winding is connected to 480 VAC from panel 4RI and its secondary winding feeds panel R. Both panel R and T are located behind the Reactor Console just outside of the Reactor Control Room.Panel R supplies power to the Thermal Column Door motor which allows the Thermal Column to be opened and closed by pressing a pair of buttons located on the Thermal Column Door itself.In addition, Panel R supplies 240 VAC power to several NEMA "twist-lock" type outlets locatedon the Reactor Shield at each beam port.8.1.4 208/120 3 wire VAC Electrical Power, Transformer T2 208/120 VAC is provided by transformer T2. This transformer's primary winding is connected to 480 VAC from Panel 4RI and its secondary winding feeds Panel 2. Both Panel 2 and T2 are located behind the Reactor Console just outside of the Reactor Control Room.Panel 2 supplies power to a pressure makeup pump in the Intermediate Loop of the Reactor's Cooling System. It also supplies power to the Demineralizer Pump in the Reactor's Primary Coolant Makeup System through an outlet located on the west wall by the Demineralizer.
In addition, Panel 2 powers the circulating pump of the Waste Holdup System.Various 120 VAC outlets throughout the lab as well as a set of 120 VAC and 208 VAC outlets at each beam port are also powered by Panel 2.Finally, though the use of an uninterruptible power supply (UPS), the Reactor's Control Console is powered through Panel 2. See section 8.2 for details of this UPS. All power to the Control Console is 120 VAC.UWNR Safety Analysis Report Rev. 2 8-2 Sept. switches the bus's source from utility to the Mechanical Engineering emergency diesel generator.
After the A TS, this circuit passes though another 1600 A breaker and a 100 A breaker to a circuit within the Mechanical Engineering Building Panel4CRl.
This panel is located in room 1133 Mechanical Engineering.
A circuit in 4CR1 powers the Reactor Laboratory's backup/makeup air compressor.
When the Reactor Laboratory's compressed air system, provided by the Mechanical Engineering Building, falls below 75 psi, this local air compressor activates.
The Reactor Laboratory's ventilation system is also powered by this bus. Bus EMCC6 is connected to the Reactor Laboratory's 277/480 V AC backed up electrical power through two 400 A fuses. Bus EMCC6 powers the Reactor Laboratory's two ventilation fans, EF-7 and EF-8, as well as the Reactor Laboratory's air handling unit, AHU-5. Finally, the Reactor Laboratory's Support Space basement emergency lights are also powered by the 277/480 VAC backed up electrical bus through paneI4CR2.
8.1.3 2404 wire VAC Electrical Power, Transformer T1 2404 wire VAC is provided by transformer T1. This transformer's primary winding is connected to 480 VAC from panel4RI and its secondary winding feeds panel R. Both panel R and T1 are located behind the Reactor Console just outside of the Reactor Control Room. Panel R supplies power to the Thermal Column Door motor which allows the Thermal Column to be opened and closed by pressing a pair of buttons located on the Thermal Column Door itself.
In addition, Panel R supplies 240 V AC power to several NEMA "twist-lock" type outlets located on the Reactor Shield at each beam port. 8.1.4 208/120 3 wire V AC Electrical Power, Transformer T2 2081120 VAC is provided by transformer T2. This transformer's primary winding is connected to 480 VAC from Panel4RI and its secondary winding feeds Panel 2. Both Panel 2 and T2 are located behind the Reactor Console just outside of the Reactor Control Room. Panel 2 supplies power to a pressure makeup pump in the Intermediate Loop of the Reactor's Cooling System. It also supplies power to the Demineralizer Pump in the Reactor's Primary Coolant Makeup System through an outlet located on the west wall by the Demineralizer.
In addition, Panel 2 powers the circulating pump of the Waste Holdup System. Various 120 VAC outlets throughout the lab as well as a set of 120 VAC and 208 VAC outlets at each beam port are also powered by Panel 2. Finally, though the use of an uninterruptible power supply (UPS), the Reactor's Control Console is powered through Panel 2. See section 8.2 for details of this UPS. All power to the Control Console is 120 V AC. UWNR Safety Analysis Report Rev. 2 8-2 Sept. 2008
8.1.5 277 VAC Lighting Electrical Power Most Reactor Laboratory lighting is supplied by the same 277/480 VAC 3000 kVA transformer and through the same 3000 A circuit breaker as the Laboratory's 277/480 3 phase electrical power. The lighting circuit has its own 200 A breaker and ground fault interrupter circuit.
This circuit is connected through a cutoff switch (located in room B 1140 Mechanical Engineering) to Panel BI 135 (located in room BI 135 Mechanical Engineering).
Panel BI 135 has variousbreakered circuits which then power the various lights within the Reactor Laboratory and the Reactor Laboratory's support space.8.1.6 277 VAC Backed Up Lighting Electrical Power A few selected lights within the Reactor Laboratory and the support space are powered by a separate 277 VAC circuit.
These lights are always on and are connected through an AutomaticTransfer Switch (ATS) which transfers the circuit from utility power to an emergency generatorin the event of a power failure.Emergency Electrical Power Systems Neither of the systems described in this section are required or necessary for safe facility operation or shutdown.There are several Emergency Electrical Power Systems, shown in Figure 8-2, which the Reactor Laboratory utilizes.
Those which are a part of the Mechanical Engineering building are powered by a diesel 800 kW 277/480 generator located in room B Mechanical Engineering.
The other Emergency Electrical Power Systems are two battery based Uninterruptible Power Supplies located within the Laboratory in rooms 1215 and B Mechanical Engineering.
8.2.1 Reactor Laboratory Uninterruptible Power Supply A 6000 VA Uninterruptible Power Supply (UPS) is connected between utility power and the reactor's control console. This system is installed solely to protect the reactors instrumentation from sudden power losses as well as dirty power. This UPS provides protection from surges, brownouts, noise, spikes, frequency variations, transients and harmonic distortion.
In the event of a loss of utility, the UPS is capable of providing up to 76 minutes of power to the reactor's control console and instrumentation.
This allows for a monitored shutdown and facilitates uninterrupted recording and monitoring.
A failsafe SCRAM and alarm connected to a non backed up power circuit ensures that control blades drop and the operator is informed in the event of a loss of utility.Even with a UPS failure the reactor is designed to SCRAM in the event of a loss of power.Therefore, this UPS is not required for a failsafe shutdown when power is lost and in the event of UWNR Safety Analysis Report Rev. 2Sept. * *
8.1.5 277 V AC Lighting Electrical Power Most Reactor Laboratory lighting is supplied by the same 277/480 VAC 3000 kVA transfonner and through the same 3000 A circuit breaker as the Laboratory's 277/480 3 phase electrical power. The lighting circuit has its own 200 A breaker and ground fault interrupter circuit. This circuit is connected through a cutoff switch (located in room B 1140 Mechanical Engineering) to Panel BlB5 (located in room BlB5 Mechanical Engineering).
Panel BlB5 has various breakered circuits which then power the various lights within the Reactor Laboratory and the Reactor Laboratory's support space. 8.1.6 277 V AC Backed Up Lighting Electrical Power A few selected lights within the Reactor Laboratory and the support space are powered by a separate 277 VAC circuit. These lights are always on and are connected through an Automatic Transfer Switch (ATS) which transfers the circuit from utility power to an emergency generator in the event of a power failure. 8.2 Emergency Electrical Power Systems Neither of the systems described in this section are required or necessary for safe facility operation or shutdown . There are several Emergency Electrical Power Systems, shown in Figure 8-2, which the Reactor Laboratory utilizes.
Those which are a part of the Mechanical Engineering building are powered by a diesel 800 kW 277/480 generator located in room Bl002A Mechanical Engineering.
The other Emergency Electrical Power Systems are two battery based Uninterruptible Power Supplies located within the Laboratory in rooms 1215 and BlB5E Mechanical Engineering.
8.2.1 . Reactor Laboratory Uninterruptible Power Supply A 6000 V A Uninterruptible Power Supply (UPS) is connected between utility power and the reactor's control console. This system is installed solely to protect the reactors instrumentation from sudden power losses as well as dirty power. This UPS provides protection from surges, brownouts, noise, spikes, frequency variations, transients and hannonic distortion.
In the event of a loss of utility, the UPS is capable of providing up to 76 minutes of power to the reactor's control console and instrumentation.
This allows for a monitored shutdown and facilitates uninterrupted recording and monitoring.
A failsafe SCRAM and alann connected to a non backed up power circuit ensures that control blades drop and the operator is infonned in the event of a loss of utility. Even with a UPS failure the reactor is designed to SCRAM in the event of a loss of power. Therefore, this UPS is not required for a failsafe shutdown when power is lost and in the event of UWNR Safety Analysis Report Rev. 2 8-3 Sept. 2008 a failure and loss of utility, the reactor will shutdown.
This is not required for the facility in a safe shutdown condition, even for extended periods of time.
Reactor Laboratory Support Space Uninterruptible Power Supply VA Uninterruptible Power Supply identical to the Reactor Laboratory's is located in the Laboratory's support space. This serves as both an installed spare for the Reactorand as a support space power backup. In the support space, this powers several outlets, allowing the support space's counting computers, sample changing equipment most importantly, the Stack Air Monitor and Continuous Air Monitor to operating or be shut down gracefully in the event of a utility power failure.
Mechanical Engineering Building Emergency Powermentioned in section above, the Mechanical Engineering building has an emergency backup generator and several emergency power systems. These systems include the backed up lights panel, which powers selected lights throughout the Reactor Laboratory and the Engineering Building, the backed up equipment panel, which powers the Reactor air compressor, smoke detector/fire alarm and ventilation.
As mentioned in Sectionthese systems are connected through automatic transfer switches which switch them to power in the event of a loss of utility power.addition, the Mechanical Engineering Buildings fire suppression system pump also has its own transfer switch and emergency power. This system includes the fire suppression in the Reactor Laboratory UWNR Safety Analysis Report Rev. 2 84Sp.20 8-4 Sept. a UPS failure and loss of utility, the reactor will shutdown.
This UPS is not required for maintaining the facility in a safe shutdown condition, even for extended periods of time. 8.2.2 Reactor Laboratory Support Space Uninterruptible Power Supply A 6000 V A Uninterruptible Power Supply (UPS) identical to the Reactor Laboratory's is located in the Laboratory's support space. This UPS serves as both an installed spare for the Reactor Laboratory UPS and as a support space power backup. In the support space, this UPS powers several outlets, allowing the support space's counting computers, sample changing equipment and, most importantly, the Stack Air Monitor (SAM) and Continuous Air Monitor (CAM) to continue operating or be shut down gracefully in the event of a utility power failure. 8.2.3 Mechanical Engineering Building Emergency Power As mentioned in section 8.1 above, the Mechanical Engineering building has an emergency backup generator and several emergency power systems. These systems include the backed up emergency lights panel, which powers selected lights throughout the Reactor Laboratory and the Mechanical Engineering Building, the backed up equipment panel, which powers the Reactor Laboratory's air compressor, smoke detector/fire alarm and ventilation.
As mentioned in Section 8.1, these systems are connected through automatic transfer switches which switch them to generator power in the event of a loss of utility power.
In addition, the Mechanical Engineering Buildings fire suppression system pump also has its own
automatic transfer switch and emergency power. This system includes the fire suppression I system in the Reactor Laboratory
UWNR Safety Analysis Report Rev. 2 8-4 Sept. 2008 to c5 u5 oFigure 8-1 Normal Electrical System UWNR Safety Analysis Report Rev. 2 8-5Sept. 2008
c ::=0 r/1 ;:t> ::s -'< '" r,;;' ::=0 (t) "0 0 ::t-::=0 (t) :< N 00 I Vl r/1 (t) N o o 00 .... = ., QO I .... Z 0 S -m -(t) 0 .-+ >-t ..... O r/1 '< '" .-+ (t) S 200A ) 3P 1 *
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AHU-5 EF-7 20VPPCRLAB SUPPORTATS CR600A 4 POLE 65,000 AICLIFE SAFETY SWITCHBOARD ATS LS 200APOLEL 25,000 AIC) 200A~200A' ) 000A PPE 277/480 VOLT 4W GENERATOR 277/480V Figure 8-2 Backup Electrical System UWNR Safety Analysis Report Rev. 2 8-6 Sept. 2008 ,----------------1 I ',-..::2.::.08::.,:Y.:..../.:....'
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LAB DJERGENCY r LIGHTS 1--------------, I 100A) I I 3P r I I I I I 100A ) I I 3P I I LIFE SAFETY I I SWITCHBOARD I _ _ _ _ _ _ _ _
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____ ) lOOOA b 3P GENERATOR 800KW 277 /480V 3i/J,4W Figure 8-2 Backup Electrical System UWNR Safety Analysis Report Rev. 2 8-6 Sept. 2008 * *
AUXILIARY Heating, Ventilation, and Air Conditioning Systems The heating, ventilation, and air condition system (hereafter called the ventilation system), is adedicated reactor laboratory system with no interaction with the rest of the building.
The system incorporates a fresh air supply air handling unit, two exhaust fans, filter banks, variable supply and exhaust air mixing boxes, and duct work. The system provides HVAC service to both the reactor laboratory and auxiliary support spaces. Figure shows the ventilation system flow schematic, while Figure and Figure 9-3 show the physical location of ventilation system components.
The ventilation system is designed to prevent the spread of airborne particulate radioactive material into occupied areas outside the Reactor Laboratory.
It removes particulates with high-efficiency filtration and assures that all releases of either gaseous and particulate activity are monitored and discharged at an elevated release point. Accidents which might result in discharge of radioactive material from the stack are discussed elsewhere in this report, and remarks may befound there indicating the concentrations which might be expected.
In addition, a portion of the ventilation system vents the beam ports, thermal column, and liquid waste holdup tank to assure that air flow is from the Reactor Laboratory into these facilities.
9.1.1 Air Handling Unit The ventilation system consists of a dedicated air handling unit (AHU-5) mounted in the fifth floor of the Mechanical Engineering building behind a security fence which only reactor staff can access. The unit supplies 9150 scfm 70'F fresh air at a relative humidity of 30% to the reactorlaboratory and auxiliary support space through 7 variable air volume boxes; 5 in the auxiliary support space, 1 in the reactor laboratory confinement, and 1 in the reactor control room. The unit is powered from a variable frequency drive which maintains duct static pressure at 1.0 inches of water gauge. The unit is interlocked with the exhaust fans to trip if neither exhaust fan is running.9.1.2 Exhaust Fans The ventilation system incorporates two roof mounted exhaust fans (EF-7 and EF-8), each individually capable of exhausting 9600 scfm air. Normally only one fan is running, but in emergency venting mode both fans may be run for increased exhaust dilution with 19200 scfmair. The reactor laboratory exhaust duct pressure is maintained at -1.5 inches of water gauge to maintain air flow from surrounding areas into the reactor laboratory.
If one exhaust fan fails, the other fan attempts to pickup. If both exhaust fans fail, an interlock will trip the air handling unit to prevent having a positive pressure in the reactor laboratory.
UWNR Safety Analysis Report Rev. 2Sept. * *
9 AuxiLIARY SYSTEMS 9.1 Heating, Ventilation, and Air Conditioning Systems The heating, ventilation, and air condition system (hereafter called the ventilation system), is a dedicated reactor laboratory system with no interaction with the rest of the building.
The system incorporates a fresh air supply air handling unit, two exhaust fans, filter banks, variable supply and exhaust air mixing boxes, and duct work. The system provides HV AC service to both the reactor laboratory and auxiliary support spaces. Figure 9-1 shows the ventilation system flow schematic, while Figure 9-2 and Figure 9-3 show the physical location of ventilation system components.
The ventilation system is designed to prevent the spread of airborne particulate radioactive material into occupied areas outside the Reactor Laboratory.
It removes particulates with efficiency filtration and assures that all releases of either gaseous and particulate activity are monitored and discharged at an elevated release point. Accidents which might result in discharge of radioactive material from the stack are discussed elsewhere in this report, and remarks may be found there indicating the concentrations which might be expected.
In addition, a portion of the ventilation system vents the beam ports, thermal column, and liquid waste holdup tank to assure that air flow is from the Reactor Laboratory into these facilities.
9.1.1 Air Handling Unit The ventilation system consists of a dedicated air handling unit (AHU-S) mounted in the fifth floor of the Mechanical Engineering building behind a security fence which only reactor staff can access. The unit supplies 91S0 scfm 70°F fresh air at a relative humidity of 30% to the reactor laboratory and auxiliary support space through 7 variable air volume boxes; S in the auxiliary support space, 1 in the reactor laboratory confinement, and 1 in the reactor control room. The unit is powered from a variable frequency drive which maintains duct static pressure at 1.0 inches of water gauge. The unit is interlocked with the exhaust fans to trip if neither exhaust fan is runmng. 9.1.2 Exhaust Fans The ventilation system incorporates two roof mounted exhaust fans (EF -7 and EF -8), each individually capable of exhausting 9600 scfm air. Normally only one fan is running, but in emergency venting mode both fans may be run for increased exhaust dilution with 19200 scfm air. The reactor laboratory exhaust duct pressure is maintained at -1.S inches of water gauge to maintain air flow from surrounding areas into the reactor laboratory.
If one exhaust fan fails, the other fan attempts to pickup. If both exhaust fans fail, an interlock will trip the air handling unit to prevent having a positive pressure in the reactor laboratory . UWNR Safety Analysis Report Rev. 2 9-1 Sept. 2008 Each fan has isolation dampers that close when the fan is not running to prevent short cycling the flow path. Both fans take a suction on a common header (located below the roof on thefifth floor Mechanical Engineering) to which all exhaust variable air volume boxes discharge to the filter bank. This provides diluting air from the auxiliary support space prior toreactor confinement air. In addition, the common header is equipped with duct work draws a suction on outside air through the reactor attic for further dilution. This outside air is not filtered.
This duct work includes an automatically controlled damper which is adjusted to maintain the exhaust duct pressure at inches of water gauge. This is necessary because exhaust fans do not have variable frequency drives and so must always run at capacity.An additional manually controlled damper is available for bringing in additional outside air into common header, but this damper is normally locked closed. The air from this common header is continuously monitored the stack air monitor to maintain a record of radioactivity discharged.
Flow rate indication is also displayed in the control room.
exhaust fans are at a height of meters above grade. These fans, manufactured incorporate a unique nozzle design which operates on a principle of internal and external stream dilution.
The nozzle entrains outside air with the primary exhaust stream to a substantially diluted exhaust stream. This enhanced flow stream then undergoes a
increase to increase stack outlet velocities which increases the effective stack height.
the analysis in Appendix neglects the increase in stack height for a bounding Filters ventilation exhaust is filtered just prior to entering the common header (before any outside air This main filter bank (F-2 1) consists of a 4x4 array consisting of pleated pre-filtersnuclear grade filters which are rated at efficiency for particles ofmicron and larger.
The pneumatic system fume hood also has its own basement filter bank(F-22), even though the fume hood exhaust is sent to the main ventilation exhaust system. This bank has 1 pre-filter followed filter located upstream of the fume hood exhaust fan The main filter bank (F-2 is located with the common header in the fifth Mechanical Engineering, . The fume hood exhaust basement filter bank (F-22) is located in the auxiliary support space in room Bi The pressure drop for the main filter bank is indicated in the control room. Local indications for pressure drop across the and across the filters are on the fifth floor Mechanical Engineering and the Exhaust Variable Air Volume Boxesshown in Figure the exhaust system incorporates variable air volume boxes. Four boxes are located in the auxiliary support space, three are located in the reactor laboratory for venting, and three are located in the reactor laboratory for emergency venting. the boxes in the auxiliary support space, one is designated for the pneumatic system fume hood.UWNR Safety Analysis Report Rev. 2Sept. Each fan has isolation dampers that close when the fan is not running to prevent short cycling the
exhaust flow path. Both fans take a suction on a common header (located below the roof on the fifth floor Mechanical Engineering) to which all exhaust variable air volume boxes discharge to through the filter bank. This provides diluting air from the auxiliary support space prior to discharging reactor confinement air. In addition, the common header is equipped with duct work which draws a suction on outside air through the reactor attic for further dilution.
This outside air is not filtered.
This duct work includes an automatically controlled damper which is adjusted to maintain the exhaust duct pressure at -1.5 inches of water gauge. This is necessary because the exhaust fans do not have variable frequency drives and so must always run at full capacity.
An additional manually controlled damper is available for bringing in additional outside air into the common header, but this damper is normally locked closed. The air from this common header is continuously monitored by the stack air monitor to maintain a record of radioactivity discharged.
Flow rate indication is also displayed in the control room. The exhaust fans are at a height of 26.5 meters above grade. These fans, manufactured by Strobic, incorporate a unique nozzle design which operates on a principle of internal and external exhaust stream dilution.
The nozzle entrains outside air with the primary exhaust stream to produce a substantially diluted exhaust stream. This enhanced flow stream then undergoes a . pressure increase to increase stack outlet velocities which increases the effective stack height. However, the analysis in Appendix A neglects the increase in stack height for a bounding analysis.
9.1.3 Filters All ventilation exhaust is filtered just prior to entering the common header (before any outside air dilution).
This main filter bank (F -21) consists of a 4x4 array consisting of 16 pleated pre":filters followed by 16 nuclear grade HEPA filters which are rated at 99.97% efficiency for particles of 1 0.1 micron and larger. The pneumatic system fume hood also has its own basement filter bank *1 (F-22), even though the fume hood exhaust is sent to the main ventilation exhaust system. This 1 filter bank has 1 pre-filter followed by 1 HEPA filter located upstream of the fume hood exhaust 1 booster fan (EF-13). The with the common header in the fifth 1 floor Mechanical Engineering,_.
The fume hood exhaust basement filter 1 bank (F-22) is located in the auxiliary support space in room Bl135A. The pressure drop for the 1 main filter bank is indicated in the control room. Local indications for pressure drop across the 1 pre-filters and across the HEPA filters are on the fifth floor Mechanical Engineering and the 1 basement.
1 1 I 1 1 I I 9.1.4 Exhaust Variable Air Volume Boxes As shown in Figure 9-1, the exhaust system incorporates 10 variable air volume boxes. Four boxes are located in the auxiliary support space, three are located in the reactor laboratory for normal venting, and three are located in the reactor laboratory for emergency venting. Of the four boxes in the auxiliary support space, one is designated for the pneumatic system fume hood. UWNR Safety Analysis Report Rev. 2 9-2 Sept. 2008 *
Each box is adjustable from 0% to 100% design flow. The normal confinement boxes have a combined rating of 2700scfm.
With this flow rate, assuming a confinement volume of 200Gm 3 , it would take 26.16 minutes (1 569s) to completely exhaust confinement.
9.1.5 Emergency Venting Mode Emergency venting mode is for use when it is desirable to rapidly change the air in the reactor laboratory to prevent spread of contamination to adjacent occupied areas. Use of emergencyventing mode is governed by the Emergency Plan, which states that the decision to operate in emergency venting mode should be reached by common consent of the Emergency Coordinator and the campus Health Physics organization.
Emergency venting is initiated by one of two large red push-buttons (with switch covers) located in the control room by the south door and in the first floor vestibule, room 1200J, by the east catwalk door. When either of these buttons are pressed, the second exhaust fan activates to double the flow-rate.
The three variable air volumeboxes designated for emergency venting mode are normally closed until emergency venting mode is activated by either of the switches. Because the negative pressure in reactor confinement would be extreme with both fans running, a makeup exhaust damper in the confinement roof opens in emergency venting mode to allow outside air to enter confinement through the reactor attic. This makeup exhaust damper is normally closed with a weather-proof seal.9.1.6 Beam Port and Thermal Column Ventilation System The Beam Port and Thermal Column Ventilation system is designed to sweep out the Ar-41 activity present in an experimental facility when the facility is opened. During ordinary operation the experimental facilities are closed and there is an essentially zero rate of discharge.
When a beam port flange or the thermal column door is opened there is a slug of activity discharged.
The average concentration discharged will, therefore, be extremely low due to dilution by the rest of the ventilation system and the fact that no activity is discharged most of the time. Section 11.1.1.1 of this report discusses the levels of activity discharged.
The Beam Port
& Thermal Column ventilation system consists of a booster exhaust fan (EF-17) mounted in the Reactor Attic (room 3110), which discharges into the main ventilation system exhaust duct work (before the main filter bank F-2 1), and a makeup damper located in theReactor Laboratory.
The booster exhaust fan is needed to provide sufficient suction on the Beam Port & Thermal Column ventilation system. The makeup damper is adjusted to maintain approximately 960 cfm at 1.5 inches suction pressure and an air velocity of about 40 feet per minute into all beam ports and the thermal column, should all be opened simultaneously.
Normal flow rate with the system sealed is about 450 cfm. The thermal column shielding door is weather-stripped to maintain a nearly airtight seal. An air-operated flapper valve at the end of the duct connected to the thermal column and beam port vents is normally open.
This maintains a slight negative pressure within the thermal column when the door is closed, but prevents the full static suction of the system from forcing air in through the thermal column door seals. When the thermal column door is opened, however, the flapper valve closes and full system suction is UWNR Safety Analysis Report Rev. 2Sept. * *
Each box is adjustable from 0% to 100% design flow. The normal confinement boxes have a combined rating of 2700scfm.
With this flow rate, assuming a confinement volume of2000m 3 , it would take 26.16 minutes (1569s) to completely exhaust confinement.
9.1.5 Emergency Venting Mode Emergency venting mode is for use when it is desirable to rapidly change the air in the reactor laboratory to prevent spread of contamination to adjacent occupied areas. Use of emergency venting mode is governed by the Emergency Plan, which states that the decision to operate in emergency venting mode should be reached by common consent of the Emergency Coordinator and the campus Health Physics organization.
Emergency venting is initiated by one of two large red push-buttons (with switch covers) located in the control room by the south door and in the first floor vestibule, room 12001, by the east catwalk door. When either of these buttons are pressed, the second exhaust fan activates to double the flow-rate.
The three variable air volume boxes designated for emergency venting mode are normally closed until emergency venting mode is activated by either of the switches.
Because the negative pressure in reactor confinement would be extreme with both fans running, a makeup exhaust damper in the confinement roof opens in emergency venting mode to allow outside air to enter confinement through the reactor attic. This makeup exhaust damper is normally closed with a weather-proof seal. 9.1.6 Beam Port and Thermal Column Ventilation System The Beam Port and Thermal Column Ventilation system is designed to sweep out the Ar-41 activity present in an experimental facility when the facility is opened. During ordinary operation the experimental facilities are closed and there is an essentially zero rate of discharge.
When a beam port flange or the thermal column door is opened there is a slug of activity discharged.
The average concentration discharged will, therefore, be extremely low due to dilution by the rest of the ventilation system and the fact that no activity is discharged most ofthe time. Section 11.1.1.1 of this report discusses the levels of activity discharged.
The Beam Port & Thermal Column ventilation system consists of a booster exhaust fan (EF -17) mounted in the Reactor Attic (room 3110), which discharges into the main ventilation system exhaust duct work (before the main filter bank F-21), and a makeup damper located in the Reactor Laboratory.
The booster exhaust fan is needed to provide sufficient suction on the Beam Port & Thermal Column ventilation system. The makeup damper is adjusted to maintain approximately 960 cfm at 1.5 inches suction pressure and an air velocity of about 40 feet per minute into all beam ports and the thermal column, should all be opened simultaneously.
Normal flow rate with the system sealed is about 450 cfm. The thermal column shielding door is weather-stripped to maintain a nearly airtight seal. An air-operated flapper valve at the end of the duct connected to the thermal column and beam port vents is normally open. This maintains a slight negative pressure within the thermal column when the door is closed, but prevents the full static suction of the system from forcing air in through the thermal column door seals. When the thermal column door is opened, however, the flapper valve closes and full system suction is UWNR Safety Analysis Report Rev. 2 9-3 Sept. 2008 impressed on the thermal column to cause the desired in-flow of air. In addition, a ball check valve within the thermal column vent prevents mixing of the air in the thermal column with the ventilation flow except when the door is not sealed. The facility ALARA program identified this feature as one which resulted in a significant reduction in activity discharged through the stack without increasing the Ar-4 1 dose within the laboratory.
Also incorporated into the system for ALARA considerations are ball-check valves in the vent line from each beam port. Because the pressure drop in the BP&TC Ventilation System duct causes lower pressures in beam ports closer to the exhaust fan, the common drain connection for the four beam ports allows air flow through the drain connections from higher to lower pressure beam ports. The check valve prevents flow into the beam port vents, thus preventing circulation through the drain system and substantially reducing the amount of Ar-41 discharged to the atmosphere.
Should a beam port rupture and with water, water would leak out of the flapper valve at the end of the Beam Port Thermal Column ventilation system on the shield step, which would still leave at least of water covering the core. The water would eventually spill onto the reactor outside of the reactor laboratory because of the height of the filter bank and exhaust fans.The vent of the waste holdup tank is also connected to the Beam Port and Thermal Column Ventilation system in order to assure any gaseous effluent from the holdup tank is discharged through a monitored release path.
In addition, the stack air activity monitor discharges the sample stream extracted from the stack into this ventilation system.
The pneumatic system fume hood includes its own exhaust booster fan and filter bank system, or it may be manually run independent of the pneumatic system if desired. Operation of the booster fan has a negligible impact on the main ventilation system exhaust flow-rates due to the small size of the booster fan relative to the main exhaust fans. The control room ventilation system panel includes indication that the booster fan is running and that it has power available to run if needed, as well as the pressure drop across the fume hood filter bank (F-22).U"R Safety Analysis Report Rev. 2 9-4 Sept. impressed on the thermal column to cause the desired in-flow of air. In addition, a ball check
valve within the thermal column vent prevents mixing of the air in the thermal column with the ventilation flow except when the door is not sealed. The facility ALARA program identified this feature as one which resulted in a significant reduction in activity discharged through the stack without increasing the Ar-41 dose within the laboratory.
Also incorporated into the system for ALARA considerations are ball-check valves in the vent line from each beam'port.
Because the pressure drop in the BP&TC Ventilation System duct causes lower pressures in beam ports closer to the exhaust fan, the common drain connection for the four beam ports allows air flow through the drain connections from higher to lower pressure beam ports. The check valve prevents flow into the beam port vents, thus preventing circulation through the drain system and substantially reducing the amount of Ar-41 discharged to the atmosphere.
Should a beam port rupture and fill with water, water would leak out of the flapper valve at the end of the Beam Port & Thermal Column ventilation system on the shield step, which would still leave at least 11 ft of water covering the core. The water would eventually spill onto the reactor laboratory floor and into the holdup tank; no water would leak out of the ventilation system outside of the reactor laboratory because of the height of the filter bank and exhaust fans. The vent of the waste holdup tank is also connected to the Beam Port and Thermal Column Ventilation system in order to assure any gaseous effluent from the holdup tank is discharged through a monitored release path. In addition, the stack air activity monitor discharges the sample stream extracted from the stack into this ventilation system. 9.1.7 Pneumatic System Fume Hood Exhaust The pneumatic system fume hood includes its own exhaust booster fan (EF-13) and filter bank (F-22) to help ensure fume hood air velocity is sufficient to protect the operator in the event of a sample breakage.
This booster fan exhausts into the common header of the main ventilation system .. The fume hood exhaust fan is automatically activated upon starting the pneumatic system, or it may be manually run independent of the pneumatic system if desired. Operation of the booster fan has a negligible impact on the main ventilation system exhaust flow-rates due to the small size of the booster fan relative to the main exhaust fans. The control room ventilation system panel includes indication that the booster fan is running and that it has power available to run if needed, as well as the pressure drop across the fume hood filter bank (F-22). UWNR Safety Analysis Report Rev. 2 9-4 Sept. 2008 *
LoU z>a c 0 o UtoUto U 0 Ox Figure Ventilation System Schematic UWNR Safety Analysis Report Rev. 2Sept.
e i'O \fJ (tl q i:l e:.. '< 'JJ ...... 'JJ i'O (1) '0 0 :4 N \0 I VI \fJ .g :-+ N o o 00 ... == ., \0 I <: (1) g ...... ,..... ...... 0 i:l \fJ '<
Sl.-Iction
'JJ 5th Floor level (t S \fJ 0 i:l" (1) S ...... 0 B1l35Cj Sl.Ipply BllJ5D Supply *
Ventilo. tion SysteM S"troblc ;::xho.us':
Fans Rev, 3 AHU-5 Filter Bank 6 Plea:tecl Pre-Filters 5 PocketfBog Filters Control ROOM Reactor Lnb Supply Supply BI135 B!l35A i BI135 ET Shop Supply MlJ.chine Shop Supply Si"jpply BI135 B!135B HPGe , Supply Supply B1135 NaJ/ABG Sl.Ipply B1215 Supply 31135 RO B1215 Supply S .... pply BI215 B1l35E Supply Supply !VAV Box Air MiXing Dlenul"1 Flow Rate Sensors I_ine ;or MonitJr Filter Bank ;--21 16 Pleated Pre-Filters 16 HEPA Filters Concrol NO MClk;:oup Air' FrOM f-------R'?Clct<Jr j "-Ie El"lergency Ve>ntlng Mode Dal"1pe>r Re>actor Lab ConFineMent NC EMergency Ne El"lergency NC EMerge>ncy Roon Exhnust ROOM [)(ho.ust Roon Exhaust E)(hnust [x:ho.ust Exhaust Rea.ctor lab Reactor Lab Beo.l"I Port Floor Roan Ex:haust BI135 MachinE' Shop EF-17 Ben!'! Port &. Te Exhaust FUME' Hoool Ex:ho.ust B1I3SB t'--__ Frlter Bank F-22 1 P1E"tlteo PrE'-Filte>r 1 HEPA Fil te>r Roan El(haust B1l35C EF-13 Roon E>:haust B1135 MachinE' Shop UWNR Safety Analysis Report Rev. 2Sept. * *
UWNR Safety Analysis Report Rev. 2 9-6 Sept. 2008 Figure 9-3 Ventilation System Cross Section Looking East UWNR Safety Analysis Report Rev. 2Sept. 2008*
Figure 9-3 Ventilation System Cross Section Looking East
UWNR Safety Analysis Report Rev. 2 9-7 Sept. 2008 Figure 9-4 Fuel Storage PositionsHandling and Storage of Reactor Fuel 9.2.1 Fuel HandlingUWNR Safety Analysis Report Rev. 2Sept.
Figure 9-4 Fuel Storage Positions
9.2 Handling and Storage of Reactor Fuel 9.2.1 Fuel Handling UWNR Safety Analysis Report Rev. 2 9-8 Sept. 2008
9.2.2 Fuel Storage New (unirradiated) reactor fuel can be handled manually and is stored in a steel safe. Thoseneeding further details on new fuel storage should consult the facility security plan.Sufficient storage room is provided for all on-site irradiated fuel. The fuel storage locations are indicated in Figure 9-4. All fuel storage facilities are designed to allow sufficient convective water flow to remove decay heat.9.2.3 Fuel Bundle Maintenance and Measurements Fuel bundle disassembly, assembly, and fuel element bow and elongation measurements are conducted underwater using a special tool (shown in Figure 9-5 with a dummy element installed) designed for both these purposes. This tool is used under about The fuel-element handling tool also operates the socket and crowsfoot wrenches.
For disassembly or assembly of fuel bundles, the maintenance and measuring tool holds the bundle, provides a reference plate for a crows-foot wrench, provides storage space for four individual fuel elements, and restrains the individual elements after they have been screwed out of the bottom end box. An air-operated clamping device reproducibly positions the bottom end box of the fuel bundle. The crows-foot wrench must be used for the initial loosening of each element (and for tightening each element to the specified torque upon reassembly).
Once the elements are loose enough to turn freely the top end fitting is removed so a socket wrench can be used on a hexagonal portion of the top fitting to completely unscrew the individual element from the bottom end box. The fuel element handling-tool then may be used to remove individual elements and place them in the storage positions.
While it is possible to disassemble and re-assemble the fuel bundles, it is tedious even with use of the written procedure and practice.
Disassembly or assembly of an element takes about 30 minutes.Required measurements of fuel element bow and elongation also are made with the maintenance and measurement tool. Because of the excessive time and added handling required to disassemble the bundle and measure each element in a separate measuring tool, the tool was UWNR Safety Analysis Report Rev. 2Sept. * *
9.2.2 Fuel Storage New (unirradiated) reactor fuel can be handled manually and is stored in a steel safe. Those needing further details on new fuel storage should consult the facility security plan. Sufficient storage room is provided for all on-site irradiated fuel. The fuel storage locations are indicated in Figure 9-4. All fuel storage facilities are designed to allow sufficient convective water flow to remove decay heat. 9.2.3 Fuel Bundle Maintenance and Measurements Fuel bundle disassembly, assembly, and fuel element bow and elongation measurements are conducted underwater using a special tool (shown in Figure 9-5 with a . de . for both these s. This tool is used under about_ The fuel-element handling tool also operates . the socket and crowsfoot wrenches.
For disassembly or assembly of fuel bundles, the maintenance and measuring tool holds the bundle, provides a reference plate for a crows-foot wrench, provides storage space for four individual fuel elements, and restrains the individual elements after they have been screwed out of the bottom end box. An air-operated clamping device reproducibly positions the bottom end box of the fuel bundle. The crows-foot wrench must be used for the initial loosening of each element (and for tightening each element to the specified torque upon reassembly).
Once the elements are loose enough to tum freely the top end fitting is removed so a socket wrench can be used on a hexagonal portion of the top fitting to completely unscrew the individual element from the bottom end box. The fuel element tool then may be used to remove individual elements and place them in the storage positions.
While it is possible to disassemble and re-assemble the fuel bundles, it is tedious even with use of the written procedure and practice.
Disassembly or assembly of an element takes about 30 minutes. Required measurements of fuel element bow and elongation also are made with the maintenance and measurement tool. Because ofthe excessive time and added handling required to disassemble the bundle and measure each element in a separate measuring tool, the tool was UWNR Safety Analysis Report Rev. 2 9-9 Sept. 2008 designed to make the measurements without disassembly.
Figure 9-6 shows the three sensors employed (a portion of the housing is removed in this view.) Each uses a differential transformeras a transducer to give a remote electrical output proportional to displacement of the sensors.The X and Y sensors employ spring-loaded aluminum wheels attached to the differential transformer cores. When the bundle is lowered into the tool the wheels are forced back and they then ride on the fuel element clad surface. These sensors are adjusted to give a zero signal for a standard fuel element dummy.The length sensor differential transformer is actuated by one lobe of a cam. A second lobe of this cam is rotated into contact with the top edge of the fuel element cladding by a leaf spring attached to the operating rod. The cam pivots out into the measuring position only when theoperating rod is fully withdrawn.
The length sensor is also adjusted to zero output for the standard fuel element dummy.
A readout device is positioned on the pool curb or reactor bridge and connected to the underwater portion of the tool.
Differential transformer core position is indicated by a meter for length measurements, and a recorder output is provided to the horizontal axis of an X-Y recorderfor the bow measurements.
Polarity is set so an increase in element length or a bow away from the center-line of the bundle gives a positive meter indication or recorder readout.The dummy fuel bundle has one dummy element exactly 0.100 inches longer than the other three elements.
This element also has a section in which the radius has been reduced by 0.060 inch, and a section in which the radius has been increased by 0.060 inch. By using this element the attenuation and zero controls on the readout box may be adjusted to give a calibrated readout of bow and length.After calibration of the tool, measurement can be made on standard fuel elements.
The standard setup used gives a 1 cm horizontal displacement for 0.060 inch transverse bend (bow) and ameter reading of length in thousandths of an inch deviation from the dummy element reference length. A trace is drawn for both the X and Y sensors while the length measurement meter reading is manually recorded on the form. A complete set of measurements for all four elements in a bundle can be completed in about twenty minutes.
UWNR Safety Analysis Report Rev. 2Sept. designed to make the measurements without disassembly.
Figure 9-6 shows the three sensors
employed (a portion of the housing is removed in this view.) Each uses a differential transformer as a transducer to give a remote electrical output proportional to displacement of the sensors. The X and Y sensors employ spring-loaded aluminum wheels attached to the differential transformer cores. When the bundle is lowered into the tool the wheels are forced back and they then ride on the fuel element clad surface. These sensors are adjusted to give a zero signal for a standard fuel element dummy. The length sensor differential transformer is actuated by one lobe of a cam. A second lobe of this cam is rotated into contact with the top edge of the fuel element cladding by a leaf spring attached to the operating rod. The cam pivots out into the measuring position only when the operating rod is fully withdrawn.
The length sensor is also adjusted to zero output for the standard fuel element dummy. A readout device is positioned on the pool curb or reactor bridge and connected to the underwater portion of the tool. Differential transformer core position is indicated by a meter for length measurements, and a recorder output is provided to the horizontal axis of an X -Y recorder for the bow measurements.
Polarity is set so increase in element length or a bow away from the center-line of the bundle gives a positive meter indication or recorder readout. The dummy fuel bundle has one dummy element exactly 0.100 inches longer than the other three
elements.
This element also has a section in which the radius has been reduced by 0.060 inch, and a section in which the radius has been increased by 0.060 inch. By using this element the attenuation and zero controls on the readout box may be adjusted to give a calibrated readout of bow and length. After calibration of the tool, measurement can be 'made on standard fuel elements.
The standard setup used gives a 1 cm horizontal displacement for 0.060 inch transverse bend (bow) and a meter reading of length in thousandths of an inch 'deviation from the dummy element reference length. A trace is drawn for both the X and Y sensors while the length measurement meter reading is manually recorded on the form. A complete set of measurements for all four elements in a bundle can be completed in about twenty minutes. UWNR Safety Analysis Report Rev. 2 9-10 Sept. 2008
Operating Rod Extension Socket Differential Transformers Clamping Device Figure Fuel Maintenance and Measuring Tool UWNR Safety Analysis Report Rev. 2Sept. * '. Figure 9-5 Fuel Maintenance and Measuring Tool
UWNR Safety Analysis Report Rev. 2 9-11 Sept. 2008 Bow Sensors Figure Bow and Elongation Sensors UWNR Safety Analysis Report Rev. 2 9-12 Sept. 2008*
Figure 9-6 Bow and Elongat,ion Sensors
UWNR Safety Analysis Report Rev. 2 9-12 Sept. 2008 Since the X and Y readouts are apart, the maximum possible bow will be the square root of the sum of the squares of the bows indicated direct measurement.
As long as neither measured bow exceeds inch, no calculations or other measurements are necessary. either bow measurement exceeds inch, then the square root of the sum of the squares of the measured bows must be calculated to determine whether or not this resultant is less than inch.the calculated number is less than inch, the element is within technical specifications.
Should the calculated bow exceed inch, the crowsfoot wrench may be used to rotate the element being measured so that the reading of one bow sensor is maximized and the true bow may be determined directly to see whether it exceeds technical specification limits.
The fire detection and protection systems in the laboratory meet the local and state requirements.
Interior doors from the Reactor Laboratory to the remaining parts of the building are also fire doors that meet local codes. walls between the reactor laboratory and the remainder of the building are masonry. An early warning smoke detection system within the laboratory is connected to the building fire alarm system. The fire alarm system alarms locally The laboratory is equipped with a sprinkler system and portable fire extinguishers which areregularly inspected and serviced the University Safety Department.
The Reactor Laboratory and auxiliary support space and offices are equipped with commercial telephones.
Two lines are available at the reactor control center. cellular phone is also kept in the control room except when being used for maintaining contact with the control room.An intercom system is installed in the Reactor Laboratory and auxiliary support space and.offices. This system provides two-way communications between stations and an all-call capability for paging.activities using radioactive and special nuclear materials covered under the reactor licensetake place within the Mechanical Engineering Building and a small portion of the Engineering Research Building in the rooms and areas indicated below:* Reactor Laboratory (Room 1215);* Auxiliary Support Space (Rooms B and B 1215, including sub-rooms);
UWNR Safety Analysis Report Rev. 2 913Sp.20Sept.
Since the X and Y readouts are 90° apart, the maximum possible bow will be the square root of the sum of the squares of the bows indicated by direct measurement.
As long as neither measured bow exceeds 0.088 inch, no calculations or other measurements are necessary.
If either bow measurement exceeds 0.088 inch, then the square root of the sum of the squares of the measured bows must be calculated to determine whether or not this resultant is less than 1/8 inch. If the calculated number is less than 1/8 inch, the element is within technical specifications.
Should the calculated bow exceed 1/8 inch, the crowsfoot wrench may be used to rotate the element being measured so that the reading of one bow sensor is maximized and the true bow may be determined directly to see whether it exceeds technical specification limits. 9.3 Fire Protection Systems and Programs The fire detection and protection systems in the laboratory meet the local and state requirements.
Interior doors from the Reactor Laboratory to the remaining parts of the building are also fire doors that meet local codes. All walls between the reactor laboratory and the remainder of the building are masonry. An early warning smoke detection system within the connected to the fire alarm . The fire alarm alarms The laboratory is equipped with a sprinkler system and portable fire extinguishers which are
regularly inspected and serviced by the University Safety Department.
9.4 Systems The Reactor Laboratory and auxiliary support space and offices are equipped with commercial telephones.
Two lines are available at the reactor control center. A cellular phone is also kept in the control room except when being used for maintaining contact with the control room. An intercom system is installed in the Reactor Laboratory and auxiliary support space and offices. This system provides two-way communications between stations and an all-call capability for paging. 9.5 Possession and Use of Byproduct, Source, and Special Nuclear Material All activities using radioactive and special nuclear materials covered under the reactor license take place within the Mechanical Engineering Building and a small portion of the Engineering Research Building in the rooms and areas indicated below:
Reactor Laboratory (Room 1215);
Auxiliary Support Space (Rooms Bl13S and B121S, including sub-rooms);
UWNR Safety Analysis Report Rev. 2 9-13 Sept. 2008 Radioactive and special nuclear material use outside these areas is conducted under Wisconsin Department of Health and Family Services license 25-1323-01, the University of Wisconsin Radioactive Materials License.Cover Gas Control in closed Primary Coolant Systems There is no cover gas control in the primary coolant system.Other Auxiliary Systems There are no other auxiliary systems required for safe reactor operation.References There are no references.UWNR Safety Analysis Report Rev. 2 9-14 Sept. Radioactive and special nuclear material use outside these areas is conducted under Wisconsin Department of Health and Family Services license 25-1323-01, the University of Wisconsin Radioactive Materials License. 9.6 Cover Gas Control in closed Primary Coolant Systems There is no cover gas control in the primary coolant system. 9.7 Other Auxiliary Systems There are no other auxiliary systems required for safe reactor operation.
9.8 References There are no references.
UWNR Safety Analysis Report Rev. 2 9-14 Sept. 2008 * *
Facilities are provided to permit use of radiation from the reactor in experimental work withoutendangering personnel. Facilities provided with this reactor include four beam ports, a thermal column, and pneumatic and hydraulic irradiation transfer systems. systems are designed to control radiation exposure to personnel using the facility as well as members of the public.Consideration is given to controlling the Ar-4 1 effluent from the experimental facilities in order to meet both the limits on releases to the environment and exposure of personnel within the laboratory.
10.2.1 Thermal Column The thermal column, Figure is a graphite-filled, horizontal penetration through thebiological shield which provides neutrons in the thermal energy range (about eV) for irradiation experiments.
The column, which is about feet long, is filled with about feet of graphite. small experimental air chamber between the face of the graphite and the thermal column door has conduits for service connections (air, water, electricity) to the biological shield face. Detectors for the safety channels and the LogN channel are located within the thermal column. The location of the thermal column is indicated in Figure 10-2.Personnel in the building are protected against gamma radiation from the column a dense concrete door which closes the column at the biological shield. The door moves on tracks set into the concrete floor perpendicular to the shield face.ventilation system maintains a low pressure within the thermnal column so that air flow is into the column when the door is open. The door is gasketed so that air flow is very small when the door is closed. When the door is opened, however, an air velocity of about 40 feet per minuteinto the column prevents the Ar-4 1 activity from diffusing into the Reactor Laboratory.
Sectioncontains further information on the ventilation system for the thermal column and beam ports.An annunciator is activated whenever the thermal column door is not fully closed. In addition, an area radiation monitor beside the thermal column door will give an alarm should the reactor be operated at a substantial power with the door open.UWNR Safety Analysis Report Rev. 2 101Sp.28Sept.
10 EXPERIMENTAL FACILITIES AND UTILIZATION 10.1 Summary Description Facilities are provided to permit use of radiation from the reactor in experimental work without endangering personnel.
Facilities provided with this reactor include four beam ports, a thermal column, and pneumatic and hydraulic irradiation transfer systems. All systems are designed to control radiation exposure to personnel using the facility as well as members of the public. Consideration is given to controlling the Ar-4l effluent from the experimental facilities in order to meet both the limits on releases to the environment and exposure of personnel within the laboratory.
10.2 Experimental Facilities 10.2.1 Thermal Column The thermal column, Figure 10-1, is a graphite-filled, horizontal penetration through the biological shield which provides neutrons in the thermal energy range (about 0.025 eV) for irradiation experiments.
The column, which is about 8 feet long, is filled with about 6 feet of graphite.
A small experimental air chamber between the face of the graphite and the thermal column door has conduits for service connections (air, water, electricity) to the biological shield face. Detectors for the safety channels and the LogN channel are located within the thermal
column. The location of the thermal column is indicated in Figure 10-2.
Personnel in the building are protected against gamma radiation from the column by a dense concrete door which closes the column at the biological shield. The door moves on tracks set into the concrete floor perpendicular to the shield face. A ventilation system maintains a low pressure within the thermal column so that air flow is into the column when the door is open. The door is gasketed so that air flow is very small when the door is closed. When the door is opened, however, an air velocity of about 40 feet per minute into the column prevents the Ar-4l activity from diffusing into the Reactor Laboratory.
Section 9.1 contains further information on the ventilation system for the thermal column and beam ports. An annunciator is activated whenever the thermal column door is not fully closed. In addition, an area radiation monitor beside the thermal column door will give an alarm should the reactor be operated at a substantial power with the door open . UWNR Safety Analysis Report Rev. 2 lO-l Sept. 2008 Q)U)~U) U Figure Thermal Column UWNR Safety Analysis Report Rev. 0 10-2 April 2000*
Figure 10-1 Thermal Column
UWNR Safety Analysis Report Rev. 0 10-2 April 2000 10.2.2 Beam Ports Four 6-inch beam ports penetrate the shield and provide fluxes of both fast and thermal neutrons for experimental use. Figure 10-2 indicates the positions of the beam ports with respect to the grid box and shield while Figure shows the construction of the beam ports.The ports are air-filled tubes, welded shut at the core ends and provided with water-tight coverson the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel. A shutter assembly, made of lead encased in aluminum, is opened for irradiations by a cable lifting device that extends to the pool curb. When closed, the shutter shields against gamma rays from the shut-down core, allowing experiments to be loaded and unloaded without excessive radiation exposure to personnel.
A drain line is attached to the bottom of the shutter housing, while a vent line attaches to the top of the shutter housing. All beam port drains combine before exiting the concrete shield, where a stop valve is provided. When beams of radiation are not being extracted, shielding plugs are installed in the outer end of each port, filling almost all of the volume within. These plugs, made of dense concrete in aluminum Thermal Columncasings, have spiral conduits for passage of instrument leads. These plugs completely stop the beam of radiation and minimize the production of Ar-41 in the beam ports.Sealed aluminum cans are installed in the in-pool portion of the beam ports unless the particular beam port experiment requires installation of a collimator or filter in that location.
These cans contain the Ar-41 produced and further minimize the release of Ar-41 activity.
Additional control of Ar-41 activity released is accomplished by bellows seals on the lifting cables and maintaining the valve on the common drain header closed.Since extremely high radiation levels could exist should the reactor be operated at substantial power levels with the shielding plugs removed, a beam port monitoring system is provided.
The system consists of radiation detectors mounted on the walls in line with each beam port and a read-out device at the console which gives an audible and visual alarm should a preset radiation level be exceeded.
The system is set to alarm at a radiation level equivalent to a dose rate of about 60 mrem/hour at the beam port openings.UWNR Safety Analysis Report Rev. 2Sept.
10.2.2 Beam Ports *
Four 6-inch beam ports penetrate the shield and provide fluxes of both fast and thermal neutrons for experimental use. Figure 10-2 indicates the positions of the beam ports with respect to the grid box and shield while Figure 10-3 shows the construction of the beam ports. The ports are air-filled tubes, welded shut at the core ends and provided with water-tight covers on the outer ends. The portions of the ports within the pool are made of aluminum, while the portions within the shield are steel. A shutter assembly, made of lead encased in aluminum, is opened for irradiations by a cable lifting device that extends to the pool curb. When closed, the shutter shields against gamma rays from the shut-down core, allowing experiments to be loaded and unloaded without excessive radiation exposure to personnel.
A drain line is attached to the bottom of the shutter housing, while a vent line attaches to the top of the shutter housing. All beam port drains combine before exiting the concrete shield, where a stop valve is provided . When beams of radiation are not being extracted, shielding plugs are installed in the outer end of each port, filling almost all of the volume within. These plugs, made of dense concrete in aluminum casings, have spiral conduits for passage of instrument leads. These plugs completely stop the production of Ar-41 in the beam ports. Figure 10-2 Location of Beam Ports & Thermal Column beam of radiation and minimize the Sealed aluminum cans are installed in the in-pool portion of the beam ports unless the particular beam port experiment requires installation of a collimator or filter in that location.
These cans contain the Ar-41 produced and further minimize the release of Ar-41 activity.
Additional control of Ar-41 activity released is accomplished by bellows seals on the lifting cables and maintaining the valve on the common drain header closed. Since extremely high radiation levels could exist should the reactor be operated at substantial power levels with the shielding plugs removed, a beam port monitoring system is provided.
The system consists of radiation detectors mounted on the walls in line with each beam port and a read-out device at the console which gives an audible and visual alarm should a preset radiation level be exceeded.
The system is set to alarm at a radiation level equivalent to a dose rate of about 60 mremlhour at the beam port openings . UWNR Safety Analysis Report Rev. 2 10-3 Sept. 2008 The beam port & thermal column ventilation system (Section 9.1) exhausts the beam ports through the vent pipes shown in Figure Vent pipes are connected to the ventilation system through a check valve which prevents back-flow into the vent and an isolation valve which may be closed should the beam port fill with water. With the beam port open, a linear flow velocity of about 40 feet per minute is maintained into the port opening, preventing diffusion of the airborne activity into the laboratory.
With the beam port closed Ar-41 is almost entirely contained within the beam port.UWNR Safety Analysis Report Rev. 10-4April 2000 The beam port & thennal column ventilation system (Section 9.1) exhausts the beam ports through the vent pipes shown in Figure 10-3. Vent pipes are connected to the ventilation system through a check valve which prevents back-flow into the vent and an isolation valve which may be closed should the beam port fill with water. With the beam port open, a linear flow velocity of about 40 feet per minute is maintained into the port opening, preventing diffusion of the airborne activity into the laboratory.
With the beam port closed Ar-41 is almost entirely contained within the beam port. UWNR Safety Analysis Report Rev. 0 10-4 April 2000 * *
Figure Beam Port UWNR Safety Analysis Report Rev. 0April 2000* *
Figure 10-3 Beam Port UWNR Safety Analysis Report Rev. 0 April 2000 10-5
Pneumatic Tubepneumatic tube is used to irradiate samples for a short time and when the sample must be processed immediately after irradiation, as in neutron act 'ivation analysis of short-lived The"rabbits" used in the system will convey samples up to 1-1/4 inches diameter and 5-1/2 inches long, although the gross weight of a sample is kept below 12 ounces. Although the polyethylene rabbits used in the system can withstand longer irradiations, this facility is usually used forshorter irradiations of small objects. The system operates as a closed ioop with CO 2 cover gas startup to remove any air which may have leaked in. An in-line CO 2 detector is used to monitor during the purge. Two isolation ball-valves are installed just outside of the shield which are closed unless the system is running. These valves serve as a barrier air leaking into the reactor portion of piping when not in use, as well as preventing loss of pool water if the internal piping should rupture.
reactor control room pneumatic system panel includes pneumatic system "System Start" and Stop" buttons, an "Emergency Return" button (which allows the reactor operator to rabbits to the fume hood if desired), and indicator lights for "System On," "System Purge In Progress" (for CO 2 purge), "Reactor Isolation Valves Open," and "Rabbit In Reactor."*
and controls except for the system start capability are duplicated at the pneumatic control center in the basement auxiliary support space immediately west of the Reactor Laboratory.
Automatic timing of irradiations is done at this control center, and rabbits are inserted, dispatched, and removed at this location.
An area monitor indicates radiation level and gives a visible and audible alarm should the radiation level exceed a preset level at the station.The preset level is selected according to the computed activity of the sample being irradiated.
pneumatic tube station is installed in a fume hood with a high efficiency filter to control any releases from sample failures. Sample activity is limited to a level which, should the samplerupture upon discharge from the system, will result in keeping the concentration exhausted belowCFR Part 20 limits for unrestricted areas when averaged over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for routine samples, or days for non-routine samples (requiring RSC approval).
The reactivity effect from a sample is restricted to less than 0.2% Tests run with water and cadmium samples indicate that sample reactivity effects will normally be less than Static reactivity measurements will be run for samples of fissionable material or particularlystrong absorbers such as some of the rare earths.Since the pneumatic tube penetrates the shield below water level, a leak in the tubing could drain pool. Spring-close air-open automatic ball valves located in the tubing just outside the shield close when the pneumatic system is turned off. Should these valves fail to close, water will not be lost from the system from a break inside the pool unless another break in the U"R Safety Analysis Report Rev. 2Sept. I* 10.2.3 Pneumatic Tube A pneumatic tube is used to irradiate samples for a short time and when the sample must be processed immediately after irradiation, as in neutron act}vation analysis of short-lived radioisotopes.
The currently installed pneumatic tube system conveys samples from the basement auxiliary support space to an irradiation position beside the core (Figure 10-4). The "rabbits" used in the system will convey samples up to 1-114 inches diameter and 5-112 inches long, although the gross weight of a sample is kept below 12 ounces. Although the polyethylene rabbits used in the system can withstand longer irradiations, this facility is usually used for shorter irradiations of small objects. The system operates as a closed loop with CO 2 cover gas controlling generation and discharge of Ar-41 activity.
The system is purged with CO 2 upon startup to remove any air which may have leaked in. An in-line CO 2 detector is used to monitor concentrations during the purge. Two isolation ball-valves are installed just outside of the biological shield which are closed unless the system is running. These valves serve as a barrier against air leaking into the reactor portion of piping when not in use, as well as preventing loss of pool water if the internal piping should rupture. The reactor control room pneumatic system panel includes pneumatic system "System Start" and "System Stop" buttons, an "Emergency Return" button (which allows the reactor operator to return rabbits to the fume hood if desired), and indicator lights for "System On," "System Purge In Progress" (for CO 2 purge), "Reactor Isolation Valves Open," and "Rabbit In Reactor." All
indications and controls except for the system start capability are duplicated at the pneumatic
tube control center in the basement auxiliary support space immediately west of the Reactor Laboratory.
Automatic timing of irradiations is done at this control center, and rabbits are inserted, dispatched, and removed at this location.
An area monitor indicates radiation level and gives a visible and audible alarm should the radiation level exceed a preset level at the station. The preset level is selected according to the computed activity of the sample being irradiated.
The pneumatic tube station is installed in a fume hood with a high efficiency filter to control any releases from sample failures.
Sample activity is limited to a level which, should the sample rupture upon discharge from the system, will result in keeping the concentration exhausted below 10 CFR Part 20 limits for unrestricted areas when averaged over a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for routine samples, or 30 days for non-routine samples (requiring RSC approval).
The reactivity effect from a sample is restricted to less than 0.2% p. Tests run with water and cadmium samples indicate that sample reactivity effects will normally be less than 0.01 % p. Static reactivity measurements will be run for samples of fissionable material or particularly strong absorbers such as some of the rare earths. Since the pneumatic tube penetrates the shield below water level, a leak in the tubing could drain the pool.
Spring-close air-open automatic ball valves located in the tubing just outside the shield automatically close when the pneumatic system is turned off. Should these valves fail to close, water will not be lost from the system from a break inside the pool unless another break in the UWNR Safety Analysis Report Rev. 2 10-6 Sept. 2008
tubing occurs outside the pool. Further, to drain more than 8 feet of water from the pool a siphon action would have to be set up. A siphon action is prevented by a solenoid valve controlled siphon breaker at the highest point in the system. The solenoid valves close when the pneumatic system is started. When the pneumatic system is off, the solenoid valves open and check valves will then allow air to enter the system if a siphon action starts. Normally these check valves prevent loss of cover gas from the system.The system is operated using a written check-list type procedure to assure that the built-in safe-guards remain effective.
Figure 10-4 Pneumatic Tube System Layout UWNR Safety Analysis Report Rev. 2Sept. * *
tubing occurs outside the pool.
Further, to drain more than 8 feet of water from the pool a siphon action would have to be set up. A siphon action is prevented by a solenoid valve controlled siphon breaker at the highest point in the system. The solenoid valves close when the pneumatic system is started. When the pneumatic system is off, the solenoid valves open and check valves will then allow air to enter the system if a siphon action starts. Normally these check valves prevent loss of cover gas from the system. The system is operated using a written check-list type procedure to assure that the built-in guards remain effective . Figure 10-4 Pneumatic Tube System Layout UWNR Safety Analysis Report Rev. 2 10-7 Sept. 2008 10.2.4 Grid Box Irradiation Facilities Irradiation of larger samples and most irradiations of more than twenty minutes duration are performed in irradiation facilities on the core periphery inside the grid box.Radiation baskets (Figure are 3-inch square aluminum containers which fit into the grid plate and may contain one or more samples. The bottom end boxes are similar to those ofreflector elements, thus positioning the devices in fixed positions relative to the core. Thesedevices may contain internal shelves or other positioning devices to position samples in fixed positions.
Figure shows a hydraulic irradiation tube (called a whale tube at the facility).
In thisfacility sample movement is powered by a separate pump located beneath the north side of the reactor bridge. The bottom ends of these tubes fit into the grid plate, and the top of the tube is fastened to the bridge structure to provide further support and prevent inadvertent movement.The motor for the pump is electrically paralleled to the diffuser pump and thus runs when the diffuser pump is in operation.
The pump takes its suction just below the pool surface and directs its flow to a jet pump near the bottom end of each tube, causing sufficient induced flow down the tube to move samples to the irradiation position and hold them in place. Samples which float return to the top of the tube where they are retained until removal by operating personnel.
Non-floating samples can be removed with a retriever tool, or they may be installed with a retrieving string or wire attached.
Flow direction and "sample in" indicators and controls are located at the pool top and control console.
Rupture of piping connected to the hydraulic irradiation facility will not result in loss of pool water due to its location within and immediately above the pool. Reactivity effects of samples are much smaller than those associated with installation and removal of conventional irradiation baskets with samples in them. The remarks regarding reactivity effects for samples in thepneumatic tube apply to this facility.UWNR Safety Analysis Report Rev. 2Sept. 10.2.4 Grid Box Irradiation Facilities Irradiation of larger samples and most irradiations of more than twenty minutes duration are performed in irradiation facilities on the core periphery inside the grid box. Radiation baskets (Figure 10-5) are 3-inch square aluminum containers which fit into the grid plate and may contain one or more samples. The bottom end boxes are similar to those of reflector elements, thus positioning the devices in fixed positions relative to the core. These devices may contain internal shelves or other positioning devices to position samples in fixed positions.
Figure 10-6 shows a hydraulic irradiation tube (called a whale tube at the facility).
In this facility sample movement is powered by a separate pump located beneath the north side of the reactor bridge. The bottom ends of these tubes fit into the grid plate, and the top of the tube is fastened to the bridge structure to provide further support and prevent inadvertent movement.
The motor for the pump is electrically paralleled to the diffuser pump and thus runs when the diffuser pump is in operation.
The pump takes its suction just below the pool surface and directs its flow to ajet pump near the bottom end of each tube, causing sufficient induced flow down the tube to move samples to the irradiation position and hold them in place. Samples which float return to the top of the tube where they are retained until removal by operating personnel. floating samples can be removed with a retriever tool, or they may be installed with a retrieving
string or wire attached.
Flow direction and "sample in" indicators and controls are located at the
pool top and control console. Rupture of piping connected to the hydraulic irradiation facility will not result in loss of pool water due to its location within and immediately above the pool. Reactivity effects of samples are much smaller than those associated with installation and removal of conventional irradiation baskets with samples in them. The remarks regarding reactivity effects for samples in the pneumatic tube apply to this facility.
UWNR Safety Analysis Report Rev. 2 10-8 Sept. 2008
ET Figure Radiation Basket UWNR Safety Analysis Report Rev. 0 10-9 April 2000
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Experiment Reviewbody of operating procedures is in place to assure that experiments are conducted in a manner that will ensure the protection of the public. Experiment review meets the requirements ofRegulatory Guide 2.21 and standard ANSI N401-1974/ANS-15.6 2 as modified by Regulatory Guide 2.4 UWNR 002, Experiment Standing Operating Instructions, defines several classes of experiments that are routinely conducted and states the limitations and precautions to be observed as well as the methodology to be used. Control Element Calibrations, reactivity coefficient measurements, in-core neutron flux distribution measurements and sample irradiation/isotope production experiments are specifically defined in these instructions.
Since sample irradiation and isotope production are major experimental activities at UWNR, several additional standard operating procedures and limitations are in effect for these activities.Limits on potential airborne radioactivity produced in the event of sample breakage are included in the procedures to assure releases will not exceed those considered in the safety analysis report or those permitted under technical specifications.
irradiation of fueled experiments is controlled so that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 Curies. UWNR 002 allows SRO approval of irradiations meeting the requirements of UWNR 131 up to the limits for routine approval of gas, dust, highly volatile material, and fissionable material stated on UWNR 130, Request For Isotope Production.
Approval for limits above those stated in UWNR 130 but below 10 CFR Part 20 limits when averaged over 30 days of dilution require approval by the Reactor Safety Committee.
Other written procedures in the UWNR 130 series, including sample packaging requirements for the different irradiation facilities and approvals, are in effect for operation of all experimental facilities.
Irradiation of material that is to be transferred to the campus broad radioisotope license requires both written and telephone approvals to assure that the recipient of the material is permitted possession and use of the material under that license.For other experiments the senior reactor operator (SRO) responsible for operation when the experiment is performed classifies the experiment as routine (previously approved and performed), modified routine (determined not to be significantly different from previously performed experiment), special (not previously approved, but within constraints of technicalspecifications), or special requiring NRC approval (involving technical specification changes or unreviewed safety questions).
Routine experiments may be approved by the SRO without further evaluation. For other experiments, the SRO evaluates the experiment in terms of its effect on reactor operation and the possibility and consequences of experiment failure, including consideration of chemicalreactions, physical integrity, design life, proper cooling, reactivity effects, and interaction with core components.
If the experiment is classified as modified routine then two SROs may approve operation of the experiment if it is determined and specified in a written record that the UWNR Safety Analysis Report Rev.
2 Sept. * *
10.3 Experiment Review A body of operating procedures is in place to assure that experiments are conducted in a manner that will ensure the protection of the public. Experiment review meets the requirements of Regulatory Guide 2.21 and standard ANSI N401-1974/ANS-15.6 2 as modified by Regulatory Guide 2.4 3. UWNR 002, Experiment Standing Operating Instructions, defines several classes of experiments that are routinely conducted and states the limitations and precautions to be observed as well as the methodology to be used. Control Element Calibrations, reactivity coefficient measurements, in-core neutron flux distribution measurements and sample irradiation/isotope production experiments are specifically defined in these instructions.
Since sample irradiation and isotope production are major experimental activities at UWNR, several additional standard operating procedures and limitations are in effect for these activities.
Limits on potential airborne radioactivity produced in the event of sample breakage are included in the procedures to assure releases will not exceed those considered in the safety analysis report or those permitted under technical specifications.
Irradiation of fueled experiments is controlled so that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 Curies. UWNR 002 allows SRO approval of irradiations meeting the requirements of UWNR 131 up to the limits for routine approval of gas, dust, highly volatile material, and fissionable material stated on UWNR 130, Request For Isotope Production.
Approval for limits above those stated in UWNR 130 but below 10 CFR Part 20 limits when averaged over 30 days of dilution require approval by the Reactor Safety Committee.
Other written procedures in the UWNR 130 series, including sample packaging requirements for the different irradiation facilities and approvals, are in effect for operation of all experimental facilities.
Irradiation of material that is to be transferred to the campus broad radioisotope license requires both and telephone approvals to assure that the recipient of the material is permitted possession and use of the material under that license. For other experiments the senior reactor operator (SRO) responsible for operation when the experiment is performed classifies the experiment as routine (previously approved and performed), modified routine (determined not to be significantly different from previously performed experiment), special (not previously approved, but within constraints of technical specifications), or special requiring NRC approval (involving technical specification changes or unreviewed safety questions) . Routine experiments may be approved by the SRO without further evaluation.
For other experiments, the SRO evaluates the experiment in terms of its effect on reactor operation and the possibility and consequences of experiment failure, including consideration of chemical reactions, physical integrity, design life, proper cooling, reactivity effects, and interaction with core components.
Ifthe experiment is classified as modified routine then two SROs may approve operation of the experiment if it is determined and specified in a written record that the UWNR Safety Analysis Report Rev. 2 10-11 Sept. 2008 hazards associated with the modified routine experiment are not significantly greater or different from those involved with the corresponding routine experiment.
If the experiment is determined to be a special experiment, an experiment review questionnaire (UWNR 030) which includes description of the experiment including materials inserted, thermodynamics, reactivity effects, radioactivity, shielding, instrumentation related to control panel instruments, administration, and procedures, as well as a safety analysis must be completed and reviewed.
Special experiments must be reviewed and approved by the Reactor Director and the Reactor Safety Committee.
Favorable evaluation of an experiment shall conclude that failure of the experiment will not lead directly to damage of reactor fuel or interference with movement of a control element. Special requiring NRC approval experiments require local approvals as well as approval by NRC.10.4 References
1. Regulatory Guide 2.2, Development of Technical Specifications for Experiments in Research Reactors, US Nuclear Regulatory Commission, November 1973 2. American National Standard ANSI N401-1974/ANS 15-6, Review of Experiments for Research Reactors, American Nuclear Society, November 19, 1974 3. Regulatory Guide 2.4, Review of Experiments for Research Reactors, U. S. Nuclear Regulatory Commission, May 1977 UWNR Safety Analysis Report Rev. 2 10-12 Sept. hazards associated with the modified routine experiment are not significantly greater or different
from those involved with corresponding routine experiment.
If the experiment is determined to be a special experiment, an experiment review questionnaire (UWNR 030) which includes description of the experiment including materials inserted, thermodynamics, reactivity effects, radioactivity, shielding, instrumentation related to control panel instruments, administration, and procedures, as well as a safety analysis must be completed and reviewed.
Special experiments must be reviewed and approved by the Reactor Director and the Reactor Safety Committee.
Favorable evaluation of an experiment shall conclude that failure ofthe experiment will not lead directly to damage of reactor fuel or interference with movement of a control element. Special requiring NRC approval experiments require local approvals as well as approval by NRC. 10.4 References
1. Regulatory Guide 2.2, Development of Technical Specifications for Experiments in Research Reactors, US Nuclear Regulatory Commission, November 1973 2. American National Standard ANSI N401-1974/ANS 15-6, Review of Experiments for Research Reactors, American Nuclear Society, November 19, 1974 3. Regulatory Guide 2.4, Review of Experiments for Research Reactors, U. S. Nuclear Regulatory Commission, May 1977 UWNR Safety Analysis Report Rev. 2 10-12 Sept. 2008 *
RADIATION PROTECTION PROGRAM WASTE This chapter deals with the program and procedures for dealing with radioactive materials, radiation, and radioactive waste management. Since the Nuclear Reactor Laboratory is a part of the University of Wisconsin-Madison the campus radiation safety regulations govern activity under the reactor license. Information on these campus regulations was promulgated under Wisconsin Department of Health and Family Services license 25-1323-01.
The Radiation Safety Regulations' are also available on the UW Radiation Safety website. This information is incorporated by reference as part of this Safety Analysis Report.The intent of the campus radiation safety program is to maintain radiation exposure toexperimenters, students, and the general public as low as reasonably achievable as well as below regulatory limits while using radiation and radioactivity for teaching and research purposes.
The implementation of the campus program within the activities of the Nuclear Reactor Laboratory has the same intent.
Radiation Protection The radiation protection program at the reactor facility, while conforming to the campus program, has some specific aspects that apply only to the reactor facility. For instance, the design of the experimental facilities, the reactor pool, and the reactor shield includes protective measures and devices which limit radiation exposures and release of radioactive material to theenvironment. Information on these aspects of the radiation control program is included in the sections of this report that describe that equipment.
General requirements, such as dosimeter use and records, certification of training, survey frequency, leak testing of sources, and overall ALARA program are discussed in the campus documentation.
The remaining portions of this chapter will deal with the issues specific to the reactor.11.1.1 Radiation Sources 11.1.1.1 Airborne Radiation Sources 11.1.1.1.1 Releases from abnormal reactor operations The fuel retains the fission products, with releases to the environment only if the fuel clad is breached.
This possibility is one of the accidents considered in Chapter 13 of this report in the analysis of the maximum hypothetical accident.
This event would result in maximum dose to personnel within the Reactor Laboratory and the maximum dose released to the environment.
The maximum occupational dose calculated is 10 millirem whole body, 1 rad to the lung, and 18.9 rads to the thyroid, while the maximum dose to persons in unrestricted areas will be less than 0.153 rem whole body and 1.019 rad to the thyroid.UWNR Safety Analysis Report Rev. 2Sept. 2008* *
11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT This chapter deals with the program and procedures for dealing with radioactive materials, radiation, and radioactive waste management.
Since the Nuclear Reactor Laboratory is a part of the University of Wisconsin-Madison the campus radiation safety regulations govern activity under the reactor license. Information on these campus regulations was promulgated under Wisconsin Department of Health and Family Services license 25-1323-01.
The Radiation Safety Regulations i are also available on the UW Radiation Safety website. This information is incorporated by reference as part of this Safety Analysis Report. The intent of the campus radiation safety program is to maintain radiation exposure to experimenters, students, and the general public as low as reasonably achievable as well as below regulatory limits while using radiation and radioactivity for teaching and research purposes.
The implementation of the campus program within the activities of the Nuclear Reactor Laboratory has the same intent. 11.1 Radiation Protection The radiation protection program at the reactor facility, while conforming to the campus program, has some specific aspects that apply only to the reactor facility.
For instance, the design of the experimental facilities, the reactor pool, and the reactor shield includes protective measures and devices which limit radiation exposures and release of radioactive material to the environment.
Information on these aspects of the radiation control program is included in the sections of this report that describe that equipment.
General requirements, such as dosimeter use and records, certification of training, survey frequency, leak testing of sources, and overall ALARA program are discussed in the campus documentation.
The remaining portions of this chapter will deal with the issues specific to the reactor. 11.1.1 Radiation Sources 11.1.1.1 Airborne Radiation Sources 11.1.1.1.1 Releases from abnormal reactor operations The fuel retains the fission products, with releases to the environment only if the fuel clad is breached.
This possibility is one of the accidents considered in Chapter 13 of this report in the analysis of the maximum hypothetical accident.
This event would result in maximum dose to personnel within the Reactor Laboratory and the maximum dose released to the environment.
The maximum occupational dose calculated is 10 millirem whole body, 1 rad to the lung, and 18.9 rads to the thyroid, while the maximum dose to persons in unrestricted areas will be less than 0.153 rem whole body and 1.019 rad to the thyroid . UWNR Safety Analysis Report Rev. 2 11-1 Sept. 2008 11.1.1.1.2 Releases from normal reactor operations Argon-41 is the only activity released in significant quantities during normal operations.
Calculations and measurements have been performed to determine production and release rates of the various activities that might be discharged due to normal operation.
The calculationmethod used for Ar-41 release is shown in Appendix A, Sections A and B.Due to the operation of the beam port and thermal column ventilating system and the laboratoryexhaust fan, the airborne activity levels in the laboratory are low. Some Ar-41 is produced in the dissolved air in the pool water as it passes through the reactor core and is released as the water iswarmed while passing through the core. Some of the resulting activity eventually reaches the pool surface where it is released to the laboratory atmosphere.
The concentration of Ar-41 in the air immediately above the pool surface during full-power operation reaches about one-third of the DAC for occupational exposure; as this air diffuses throughout the laboratory, the activity in the laboratory as a whole is at least a factor of 6 below the DAC. Therefore, further discussion will be concerned with the activity released to the atmosphere.
The maximum release rate of Ar-41 would occur with the reactor operating continuously at MW and all four beam ports and the thermal column open. Such operation is not reasonable, but it does establish an upper limit to the activity that might be discharged.
This maximum release rate is 13.3 iCi/sec, giving an Ar-41 concentration at the stack outlet of 2.94xl [tCi/ml. The EPA COMPLY program 2 indicates that the maximally exposed receptor would receive a dose of 0.6 mrem/year if all activity generated were discharged continuously.
The maximum concentration to which the public would be exposed (using Gifford's model as discussed in Appendix A and assuming a zero stack height) in this case would be about 3.31 xl -9[tCi/ml.As previously indicated, the above maximum value is for a situation not likely to occur duringoperation. The usual procedure is to have the experimental facilities in a no-flow condition if possible. Under no-flow conditions the beam port and thermal column ventilation system keeps the pressure in the experimental facilities lower than room pressure, and the activity produced in the facilities remains there and decays. The ALARA measures taken on the experimental facilities limits the typical release rate to about 10% of the production rate. Historically, in the year in which the maximum recorded Ar-41 release to the environment occurred (1999-2000 fiscal year), the COMPLY program indicated a resulting dose of 0.004 mrem/year.
One theoretically important consideration in the analysis of a reactor location is the effect on surrounding unrestricted areas of a spillage of radioactive materials.
A release of radioactive material might occur, for example, if a highly-volatile liquid were irradiated in the reactor for the production of isotopes.
If, while it was being transferred from the reactor to a cask, it were dropped and its container broken, the atmosphere within the Reactor Laboratory could become UWNR Safety Analysis Report Rev. 2 11-2 Sept. 11.1.1.1.2 Releases from normal reactor operations Argon-41 is the only activity released in significant quantities during normal operations. Calculations and measurements have been performed to determine production and release rates of the various activities that might be discharged due to normal operation.
The calculation method used for Ar-41 release is shown in Appendix A, Sections A and B. Due to the operation of the beam port and thermal column ventilating system and the laboratory exhaust fan, the airborne activity levels in the laboratory are low. Some Ar-41 is produced in the dissolved air in the pool water as it passes through the reactor core and is released as the water is warmed while passing through the core. Some of the resulting activity eventually reaches the pool surface where it is released to the laboratory atmosphere.
The concentration of Ar-41 in the air immediately above the pool surface during full-power operation reaches about one-third of the DAC for occupational exposure; as this air diffuses throughout the laboratory, the activity in the laboratory as a whole is at least a factor of 6 below the DAC. Therefore, further discussion will be concerned with the activity released to the atmosphere.
The maximum release rate of Ar-41 would occur with the reactor operating continuously at 1 MW and all four beam ports and the thermal column open. Such operation is not reasonable, but it does establish an upper limit to the activity that might be discharged.
This maximum release rate is 13.3 !lCi/sec, giving an Ar-41 concentration at the stack outlet of2.94x10-6 !lCi/ml. The
EPA COMPLY program 2 indicates that the maximally exposed receptor would receive a dose of
0.6 mremlyear if all activity generated were discharged continuously.
The maximum concentration to which the public would be exposed (using Gifford's model as discussed in Appendix A and assuming a zero stack height) in this case would be about 3.31x10-9
!lCi/ml. As previously indicated, the above maximum value is for a situation not likely to occur during operation.
The usual procedure is to have the experimental facilities in a no-flow condition if possible.
Under no-flow conditions the beam port and thermal column ventilation system keeps the pressure in the experimental facilities lower than room pressure, and the activity produced in the facilities remains there and decays. The ALARA measures taken on the experimental facilities limits the typical release rate to about 10% of the production rate. Historically, in the year in which the maximum recorded Ar-41 release to the environment occurred (1999-2000 fiscal year), the COMPLY program indicated a resulting dose of 0.004 mremlyear.
One theoretically important consideration in the analysis of a reactor location is the effect on surrounding unrestricted areas of a spillage of radioactive materials.
A release of radioactive material might occur, for example, if a highly-volatile liquid were irradiated in the reactor for the production of isotopes.
If, while it was being transferred from the reactor to a cask, it were dropped and its container broken, the atmosphere within the Reactor Laboratory could become UWNR Safety Analysis Report Rev. 2 11-2 Sept. 2008
conceivably contaminated; further, this atmosphere could conceivably be released to the surroundings in such a fashion as to present an exposure in unrestricted areas.For a typical solid or liquid spill, no special problems exist other than the direct radiation fromthe sample and cleaning up contamination.
Since the level of radiation will be known for each sample, adequate equipment for handling the sample will be available when the material isdischarged from the reactor. Equipment adequate for cleanup of spills will be kept available sothat spills can be dealt with immediately, lessening the possibility of spreading contamination to adjacent areas.The remainder of this section will deal with gases, highly volatile liquids, or powdered samples which might cause air-borne activity in the event of a spill. This problem is handled at Wisconsin by a combination of administrative and operational procedures.
For routine operations, a concerted effort will be made to keep the concentration of contaminants in theatmosphere released from the Reactor Laboratory well below the limits as stated in Table 2, Appendix B, 10 CFR Part 20, "Standards for Protection Against Radiation." Among theprocedures which will be followed to achieve this goal will be the double-encapsulating of materials to be exposed in the reactor in aluminum containers (for long exposure) or sealed polyethylene containers for exposures of less than 4 x 10"7 thermal neutrons/cm 2 with accompanying gamma ray and fast neutron fluxes. Only members of the reactor staff (or selected and trained individuals working under their supervision) will be permitted to handle these capsules within the Reactor Laboratory and the capsules will normally be opened only at appropriate locations outside the laboratory.
Further, a log book will be maintained of all material exposures.
However, because accidents can occur, the amount of radioactivity which will be generated in any one sample of material will be limited. Specifically, this amount of radioactivity will be limited for routine samples such that, should a container be broken and itscontents disperse in the air within the Reactor Laboratory, the concentrations discharged through the stack when averaged over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will not exceed concentrations of 10 CFR Part 20 Appendix B Table 2. In normal operation (single fan running), the ventilation system has a capacity of 9,600 scfm through its filters, therefore 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of dilution is 3.9x10" ml. For non-routine samples, RSC approval will be required, but the above limits will still apply assuming a 30 day dilution instead of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 30 days of dilution is 1.2x10i3 ml. These approvals will consider all other activity discharged, and will insure that the total stack discharge lies within permissible limits should the sample rupture.The pneumatic tube station is located in the Reactor Laboratory auxiliary support space and thusit is subject to the laboratory ventilation system. The station is installed within a fume hood having a face velocity of > 100 lfpm to protect the system operator in case of sample breakage.
The air discharged from the hood is passed through a dedicated high efficiency filter before connecting to the main laboratory ventilation exhaust.Although special packaging requirements are enforced to prevent breakage of pneumatic tube samples, such breakage may occur. Sample activity is limited as discussed above, assuming 24 UWNR Safety Analysis Report Rev. 2Sept. * *
conceivably contaminated; further, this atmosphere could conceivably be released to the surroundings in such a fashion as to present an exposure in unrestricted areas. For a typical solid or liquid spill, no special problems exist other than the direct radiation from the sample and cleaning up contamination.
Since the level of radiation will be known for each sample, adequate equipment for handling the sample will be available when the material is discharged from the reactor. Equipment adequate for cleanup of spills will be kept available so that spills can be dealt with immediately, lessening the possibility of spreading contamination to adjacent areas. The remainder of this section will deal with gases, highly volatile liquids, or powdered samples which might cause air-borne activity in the event of a spill. This problem is handled at Wisconsin by a combination of administrative and operational procedures.
For routine operations, a concerted effort will be made to keep the concentration of contaminants in the atmosphere released from the Reactor Laboratory well below the limits as stated in Table 2, Appendix B, 10 CFR Part 20, "Standards for Protection Against Radiation." Among the procedures which will be followed to achieve this goal will be the double-encapsulating of materials to be exposed in the reactor in aluminum containers (for long exposure) or sealed polyethylene containers for exposures ofless than 4 x 10 17 thermal neutrons/cm 2 with accompanying gamma ray and fast neutron fluxes. Only members of the reactor staff (or selected and trained individuals working under their supervision) will be permitted to handle these capsules within the Reactor Laboratory and the capsules will normally be opened only at appropriate locations outside the laboratory.
Further, a log book will be maintained of all material exposures.
However, because accidents can occur, the amount of radioactivity which will be generated in anyone sample of material will be limited. Specifically, this amount of radioactivity will be limited for routine samples such that, should a container be broken and its contents disperse in the air within the Reactor Laboratory, the concentrations discharged through the stack when averaged over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will not exceed concentrations of 10 CFR Part 20 Appendix B Table 2. In normal operation (single fan running), the ventilation system has a capacity of 9,600 scfm through its filters, therefore 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of dilution is 3.9xl 011 ml. For routine samples, RSC approval will be required, but the above limits will still apply assuming a 30 day dilution instead of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 30 days of dilution is 1.2x 1 0 13 ml. These approvals will consider all other activity discharged, and will insure that the total stack discharge lies within permissible limits should the sample rupture. The pneumatic tube station is located in the Reactor Laboratory auxiliary support space and thus it is subject to the laboratory ventilation system. The station is installed within a fume hood having a face velocity of> 100 lfpm to protect the system operator in case of sample breakage.
The air discharged from the hood is passed through a dedicated high efficiency filter before connecting to the main laboratory ventilation exhaust. Although special packaging requirements are enforced to prevent breakage of pneumatic tube samples, such breakage may occur. Sample activity is limited as discussed above, assuming 24 UWNR Safety Analysis Report Rev. 2 11-3 Sept. 2008 hours0.0232 days 0.558 hours 0.00332 weeks 7.64044e-4 months of normal ventilation dilution for routine samples, and 30 days of dilution for non-routine samples requiring RSC approval.
The fume-hood blower operates automatically whenever the pneumatic tube system is used. As with the other samples, the maximum activities generated for non-routine samples must have RSC approval, and only quantities considerably smaller are routinely approved.11.1.1.2 Liquid Radioactive Sources The only activity produced in liquid form in amounts sufficient to be a personnel exposure hazard is Nitrogen-16, which is produced in the reactor coolant as it passes through the reactor core when operating at power levels above 100kW. N-16 is controlled by use of the diffuser system (discussed in Section 5.6), which reduces the dose rate at the pool surface to 2 to 3 mrem/hour during full power operation.
If the diffuser system fails during full power operation the dose rate at the pool surface is less than 100 mremlhour.
Small quantities of liquid radioactive waste are generated by regeneration of the demineralizer and from liquids irradiated as part of sample irradiation.
The radiation level from such liquids isextremely low and does not produce radiation exposure hazards. Disposal of this material is addressed in section 11.2.3. Releases are made to the sewer system within 10 CFR Part 20 Appendix B Table 3 limits. Annual liquid releases have ranged from 0 to 10,000 gallons, with 3000 gallons being typical.11.1.1.3 Solid Radioactive Sources The major source of radiation and radioactivity is the fission product generation in the reactor fuel. Typical four-element fuel bundles will generate fields of 100 to more than 1000 R/hour in air at 3 feet if removed from the reactor pool.
As long as the fuel is contained within the pool filled with water this source of radiation dose presents no personnel hazard.
Loss of pool water is considered in Chapter 13, with the conclusion that the dose rates from pool water loss after long periods of operation could result in high radiation levels at the pool top (1200 R/hour one day after shutdown), but not so high that persons could not perform corrective actions to restore enough pool level to reduce the dose rate to tolerable levels (dose rate at the pool top level when shielded by the pool curb would be about 240 mrem/hour at the same decay time). Further, the pool is designed to preclude loss of pool water, and operation would not take place if there were any difficulty in maintaining pool level.Other possibilities of significant radiation exposure from solid radioactive material are the standard 20% enriched TRIGA core, samples irradiated for isotope production, reactorcomponents which have spent a long time near the core, and the reactor startup source.
All of these are small sources compared to fuel fission product activity in the operating core. Dose UWNR Safety Analysis Report Rev. 2 11-4 Sept. hours of normal ventilation dilution for routine samples, and 30 days of dilution for non-routine
samples requiring RSC approval.
The fume-hood blower operates automatically whenever the pneumatic tube system is used. As with the other samples, the maximum activities generated for non-routine samples must have RSC approval, and only quantities considerably smaller are routinely approved.
11.1.1.2 Liquid Radioactive Sources The only activity produced in liquid form in amounts sufficient to be a personnel exposure hazard is Nitrogen-16, which is produced in the reactor coolant as it passes through the reactor core when operating at power levels above 100kW. N-16 is controlled by use of the diffuser system (discussed in Section 5.6), which reduces the dose rate at the pool surface to 2 to 3 mremlhour during full power operation.
If the diffuser system fails during full power operation the dose rate at the pool surface is less than 100 mremlhour.
Small quantities of liquid radioactive waste are generated by regeneration of the demineralizer and from liquids irradiated as part of sample irradiation.
The radiation level from such liquids is extremely low and does not produce radiation exposure hazards. Disposal of this material is addressed in section 11.2.3. Releases are made to the sewer system within 10 CFR Part 20 Appendix B Table 3 limits. Annual liquid releases have ranged from 0 to 10,000 gallons, with 3000 gallons being typical. 11.1.1.3 Solid Radioactive Sources The major source of radiation and radioactivity is the fission product generation in the reactor fuel. Typical four-element fuel bundles will generate fields of 100 to more than 1000 Rlhour in air at 3 feet if removed from the reactor As long as the fuel is contained within the pool filled with water this source of radiation dose presents no personnel hazard. Loss of pool water is considered in Chapter 13, with the conclusion that the dose rates from pool water loss after long periods of operation could result in high radiation levels at the pool top (1200 Rlhour one day after shutdown), but not so high that persons could not perform corrective actions to restore enough pool level to reduce the dose rate to tolerable levels (dose rate at the pool top level when shielded by the pool curb would be about 240 mrem/hour at the same decay time). Further, the pool is designed to preclude loss of pool water, and operation would not take place if there were any difficulty in maintaining pool level. Other possibilities of significant radiation exposure from solid radioactive material are the standard 20% enriched TRIGA core, samples irradiated for isotope production, reactor components which have spent a long time near the core, and the reactor startup source. All of these are small sources compared to fuel fission product activity in the operating core. Dose UWNR Safety Analysis Report Rev. 2 11-4 Sept. 2008 *
rates from the old fuel are several orders of magnitude lower than those from the operating core.Sample handling equipment and procedures and use of aluminum for almost all structure near the core reduce exposure rates from samples and activated materials to levels which generate no significant personnel hazard during operation or maintenance of the reactor. For example, the shim-safety blades, reflector elements, and transient control rod have maximum radiation levels of a few R/hour at contact after a week of reactor shutdown.
Activity produced during irradiations is calculated before the irradiations are performed and equipment and procedures are in place to deal with the activity after the irradiation is completed.
11.1.2 Radiation Protection Program 11.1.3 ALARA Program Note: These two sections are combined.The University Radiation Safety Regulations 1 are written to incorporate ALARA principles andpractices. The Nuclear Reactor Laboratory policies and procedures reflect the commitment to ALARA principles.
An annual ALARA review is conducted jointly by campus Safety Department health physics staff and the Reactor Laboratory staff with a report of the results of the review being submitted to the Reactor Director and the Reactor Safety Committee.
11.1.4 Radiation Monitoring and Surveying The campus regulations 1 specify requirements on monitoring and surveying.
Procedures for reactor operation reflect these requirements. Installed radiation and air activity monitors are described in Section 7.7 of this report. Area radiation surveys are conducted each month, including checks for contamination and particulate air activity.
Sample irradiation procedures and forms require checks of radiation level each time a sample is removed from an irradiation facility.
Experiment reviews and approvals require radiation surveys for new experiments and modifications of experiments.
11.1.5 Radiation Exposure Control and DosimetryThe campus regulations' specify requirements on radiation control and dosimetry, and the SafetyDepartment administers the dosimetry program. TLD dosimeters are used for operating personnel and experimenters using the laboratory on a regular basis, and electronic dosimeters are used and records are maintained for tour groups and visitors.Experiment approval requires that no Very High Radiation Areas are created external to the experiment shielding.
Some experiments have shield cavities large enough for personnel entry, however, and higher radiation levels can exist inside the shield.
Should an experiment design be approved with a Very High Radiation Level within the experiment shield, protective measures will be in place that will reduce radiation levels to no more than a high radiation area if access is UWNR Safety Analysis Report Rev. 2 11-5 Sept. * *
rates from the old fuel are several orders of magnitude lower than those from the operating core. Sample handling equipment and procedures and use of aluminum for almost all structure near the core reduce exposure rates from samples and activated materials to levels which generate no significant personnel hazard during operation or maintenance of the reactor. For example, the shim-safety blades, reflector elements, and transient control rod have maximum radiation levels of a few Rlhour at contact after a week of reactor shutdown.
Activity produced during irradiations is calculated before the irradiations are performed and equipment and procedures are in place to deal with the activity after the irradiation is completed.
11.1.2 Radiation Protection Program 11.1.3 ALARA Program Note: These two sections are combined.
The University Radiation Safety Regulations!
are written to incorporate ALARA principles and practices.
The Nuclear Reactor Laboratory policies and procedures reflect the commitment to ALARA principles.
An annual ALARA review is conducted jointly by campus Safety Department health physics staff and the Reactor Laboratory staff with a report of the results of the review being submitted to the Reactor Director and the Reactor Safety Committee.
11.1.4 Radiation Monitoring and Surveying The campus regUlations!
specify requirements on monitoring and surveying.
Procedures for reactor operation reflect these requirements.
Installed radiation and air activity monitors are described in Section 7.7 of this report. Area radiation surveys are conducted each month, including checks for contamination and particulate air activity.
Sample irradiation procedures and forms require checks of radiation level each time a sample is removed from an irradiation facility.
Experiment reviews and approvals require radiation surveys for new experiments and modifications of experiments.
11.1.5 Radiation Exposure Control and Dosimetry The campus regulations!
specify requirements on radiation control and dosimetry, and the Safety Department administers the dosimetry program. TLD dosimeters are used for operating personnel and experimenters using the laboratory on a regular basis, and electronic dosimeters are used and records are maintained for tour groups and visitors.
Experiment approval requires that no Very High Radiation Areas are created external to the experiment shielding.
Some experiments have shield cavities large enough for personnel entry, however, and higher radiation levels can exist inside the shield. Should an experiment design be approved with a Very High Radiation Level within the experiment shield, protective measures will be in place that will reduce radiation levels to no more than a high radiation area if access is UWNR Safety Analysis Report Rev. 2 11-5 Sept. 2008 attempted.
If a High Radiation Area is created inside or outside of the shielding, access will be controlled and posted in accordance with 10 CFR 20.1601.Radiation doses received by visitors and tour groups are so low that they routinely cannot be measured; the maximum dose rate allowed for any tours is 0.5 mrem/hour and for any non-radiation workers is 2.0 mrem/hour.
Visitors, who are radiation workers not part of the campus dosimetry program such as visiting researchers, are allowed higher dose rates, but rarely do they exceed 2.0 mrem/hour due to ALARA practices. No student dosimeter has ever received ameasurable exposure from reactor operation. Occupational exposures of operations and maintenance personnel have historically been very low, seldom exceeding 0.5 Rem TEDE in a year and usually below 100 mrem/year.
11.1.6 Contamination Control The campus regulations 1 specify requirements on Contamination Control. As noted in section 11.1.4, monthly contamination surveys are conducted.
Laboratory policy is that no detectable removable contamination is allowed; any contamination discovered is immediately decontaminated.
For routine cleaning, the laboratory has cleaning equipment which is dedicated to use in thelaboratory area, and custodial personnel use this equipment in order to prevent the possibility of spreading unidentified contamination.
Floor sweepings are surveyed for radioactivity before disposal.11.1.7 Environmental Monitoring Environmental TLD monitors are used and evaluated on a quarterly basis. The dosimeters are distributed around the engineering campus so that they surround the Reactor Laboratory.
At the present time more than 25 points are monitored.
Effluent concentrations are measured at the point of release.11.2 Radioactive Waste Management The campus regulations 1 specify requirements for dealing with radioactive waste on campus.The Reactor Laboratory follows the campus regulations.
11.2.1 Radioactive Waste Management Program This is a campus program.UWNR Safety Analysis Report Rev. 2Sept. attempted.
If a High Radiation Area is created inside or outside of the shielding, access will be
controlled and posted in accordance with 10 CFR 20.1601. Radiation doses received by visitors and tour groups are so low that they routinely cannot be measured; the maximum dose rate allowed for any tours is 0.5 mremlhour and for any radiation workers is 2.0 mremlhour.
Visitors, who are radiation workers not part of the campus dosimetry program such as visiting researchers, are allowed higher dose rates, but rarely do they exceed 2.0 mremlhour due to ALARA practices.
No student dosimeter has ever received a measurable exposure from reactor operation.
Occupational exposures of operations and maintenance personnel have historically been very low, seldom exceeding 0.5 Rem TEDE in a year and usually below 100 mrem/year.
11.1.6 Contamination Control The campus regulations 1 specify requirements on Contamination Control. As noted in section 11.1.4, monthly contamination surveys are conducted.
Laboratory policy is that no detectable removable contamination is allowed; any contamination discovered is immediately decontaminated.
For routine cleaning, the laboratory has cleaning equipment which is dedicated to use in the laboratory area, and custodial personnel use this equipment in order to prevent the possibility of spreading unidentified contamination.
Floor sweepings are surveyed for radioactivity before
disposal.
11.1.7 Environmental Monitoring Environmental TLD monitors are used and evaluated on a quarterly basis. The dosimeters are distributed around the engineering campus so that they surround the Reactor Laboratory.
At the present time more than 25 points are monitored.
Effluent concentrations are measured at the point of release. 11.2 Radioactive Waste Management The campus regulations 1 specify requirements for dealing with radioactive waste on campus. The Reactor Laboratory follows the campus regulations.
11.2.1 Radioactive Waste Management Program This is a campus program. UWNR Safety Analysis Report Rev. 2 11-6 Sept. 2008
11.2.2 Radioactive Waste Control This is a campus-wide program. Liquid waste from beam port drains, pool overflow, laboratoryfloor drains, radioactive sink, and demineralizer regeneration is stored in a 2000-gallon holdup tank (see Chapter 5, section 5.5), and other liquid radioactive wastes generated in the laboratory are collected in local containers.
Filled local containers may be dumped into the holdup tank.11.2.3 Release of Radioactive Waste Solid radioactive waste is transferred to the Safety Department for disposal.Liquid wastes can be transferred to the Safety Department, but most are placed into the holdup tank. The Reactor Laboratory occasionally discharges liquid waste from the holdup tank to the sewer system. All discharges are filtered so that no particulate activity above 0.5 micron size isdischarged. Sampling, analysis, and release of the holdup tank contents are governed by a written procedure that assures releases are within 10 CFR Part 20 Appendix B Table 3 limits and that the pH is within local limits for discharge to the sewer.
References
1. Radiation Safety Regulations, University of Wisconsin-Madison, Revision 2, January 1997.2. U. S. Environmental Protection Agency, COMPLY Program Rev. 2, October 1989 UWNR Safety Analysis Report Rev. 2Sept. * *
11.2.2 Radioactive Waste Control This is a campus-wide program. Liquid waste from beam port drains, pool overflow, laboratory floor drains, radioactive sink, and demineralizer regeneration is stored in a 2000-gallon holdup tank (see Chapter 5, section 5.5), and other liquid radioactive wastes generated in the laboratory are collected in local containers.
Filled local containers may be dumped into the holdup tank. 11.2.3 Release of Radioactive Waste Solid radioactive waste is transferred to the Safety Department for disposal.
Liquid wastes can be transferred to the Safety Department, but most are placed into the holdup tank. The Reactor Laboratory occasionally discharges liquid waste from the holdup tank to the sewer system. All discharges are filtered so that no particulate activity above 0.5 micron size is discharged.
Sampling, analysis, and release of the holdup tank contents are governed by a W1,"itten procedure that assures releases are within 10 CFR Part 20 Appendix B Table 3 limits and that the pH is within local limits for discharge to the sewer. 11.3 References
1. Radiation Safety Regulations, University of Wisconsin-Madison, Revision 2, January 1997. 2. U. S. Environmental Protection Agency, COMPLY Program Rev. 2, October 1989 UWNR Safety Analysis Report Rev. 2 11-7 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2Sept.
This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 2 11-8 Sept. 2008
12.1.1 Structure is a chart indicating the operating organization.
Position responsibilities and authorities are summarized in the following sections.12.1.2 Responsibility 12.1.2.1 University Radiation Safety CommitteeTo exercise its prerogatives (as a campus-wide committee appointed the Chancellor of the University of Wisconsin-Madison Campus to review all activities on campus which involve the use of radiation) in reviewing all activities related to the Reactor Laboratory.
2. To advise the Reactor Director of all studies and/or actions taken with regard to the Reactor Laboratory.To overrule the Reactor Director where necessary in carrying out its function.4. To supply health physics services to the University.
12.1.2.2 University Radiation Safety (Part of University Department of Environment, Health and Safety) 1 To assist the University Radiation Safety Committee conducting inspections, making recommendations, maintaining records, and establishing procedures foremergency operations, waste disposal, etc.2. To provide similar inspections and service functions to the Reactor Safety Committee.
Chair, Engineering Physics Department 1 Responsible for the reactor facilities licenses and charter.2. To appoint the Reactor Director.
Safety Analysis Report Rev. 2 121Sp.08 12-1 Sept. * *
12 CONDUCT OF OPERATIONS 12.1 Organization 12.1.1 Structure Figure 12-1 is a chart indicating the operating organization.
Position responsibilities and authorities are summarized in the following sections.
12.1.2 Responsibility 12.1.2.l 1. 2. 3 . 4. 12.1.2.2 1. 2. 12.1.2.3 1. 2 . University Radiation Safety Committee To exercise its prerogatives (as a campus-wide committee appointed by the Chancellor of the University of Wisconsin-Madison Campus to review all activities on campus which involve the use of radiation) in reviewing all activities related to the Reactor Laboratory.
To advise the Reactor Director of all studies and/or actions taken with regard to the Reactor Laboratory.
To overrule the Reactor Director where necessary in carrying out its function.
To supply health physics services to the University.
University Radiation Safety (Part of University Department of Environment, Health and Safety) To assist the University Radiation Safety Committee by conducting inspections, making recommendations, maintaining records, and establishing procedures for emergency operations, waste disposal, etc. To provide similar inspections and service functions to the Reactor Safety Committee.
Chair, Engineering Physics Department Responsible for the reactor facilities licenses and charter. To appoint the Reactor Director.
UWNR Safety Analysis Report Rev. 2 12-1 Sept. 2008 To approve all policy decisions and all basic regulations, basic instructions, andbasic procedures governing the use and operation of the reactor and related facilities.
2. "To designate the Reactor Supervisor and other Senior operators.To take cognizance of all recommendations and actions the University Radiation Safety Committee (which relate to the reactor facility) and the Reactor Safety Committee.
4. To appoint qualified members to the Reactor Safety Committee as necessary.
Reactor Safety Committee 1 Review and approval of new experiments utilizing the reactor facilities;
2. Review and approval of all proposed changes to the facility, procedures, license, and technical specifications;Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications;
4. Review of abnormal performance of plant equipment and operating anomalies having safety significance; and
5. Review of unusual or reportable occurrences and incidents which are reportable under CFR Part 20 and CFR Part 0.Review of audit reports.Review of violations of technical specifications, license, or procedures and orders having safety significance.
1 To initiate and enforce policies, administrative rules, regulations, and operating procedures relating to the Reactor Laboratory, subject to the appropriate approvals of the Reactor Safety Committee, the University Radiation Safety Committee, andthe Reactor Director.UWNR Safety Analysis Report Rev. 2 122Sp.08 12-2 Sept. 12.1.2.4 1. 2. 3. 4. 12.1.2.5 1. 2. 3. 4. 5. 6. 7. 12.1.2.6 1. Reactor Director To approve all policy decisions and all basic regulations, basic instructions, and basic procedures governing the use and operation of the reactor and related facilities.
!To designate the Reactor Supervisor and other Senior operators.
To take cognizance of all recommendations and actions by the University Radiation Safety Committee (which relate to the reactor facility) and the Reactor Safety Committee.
To appoint qualified members to the Reactor Safety Committee as necessary.
Reactor Safety Committee Review and approval of new experiments utilizing the reactor facilities; Review and approval of all proposed changes to the facility, procedures, license, and technical specifications; Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications; Review of abnormal performance of plant equipment and operating anomalies having safety significance; and Review of unusual or reportable occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50. Review of audit reports. Review of violations of technical specifications, license, or procedures and orders having safety significance.
Reactor Supervisor To initiate and enforce policies, administrative rules, regulations, and operating procedures relating to the Reactor Laboratory, subject to the appropriate approvals of the Reactor Safety Committee, the University Radiation Safety Committee, and the Reactor Director.
UWNR Safety Analysis Report Rev. 2 12-2 Sept. 2008 * * *
2. To ensure that all activities within the Reactor Laboratory are in accordance with prior approvals from the appropriate committees or from the Reactor Director.3. The Reactor Supervisor shall have authority to authorize experiments and/or procedures which have been approved by the Reactor Safety Committee.
He will prepare specific detailed procedures based on the general procedures approved by the Committee.
4. To see that all proper records are kept.To maintain a Senior Operator's License.
6. To appoint Reactor Operators.
7. The Reactor Supervisor or another Senior operator shall be in charge of the Reactor Laboratory at all times (although not necessarily physically present). The individual in charge, if physically present, shall be responsible for prompt execution of emergency procedures.
The Reactor Supervisor or another Senior operator will be present at the facility during fuel manipulation, reactor start-up and approach to power, and recovery from unscheduled scrams and shut-downs, and shall be available on call at other times during reactor operation.
8. To be responsible for safety in the Reactor Laboratory, including responsibility for health physics matters.
9. To advise and prepare information for the committees concerned with the Reactor Laboratory, and to present such information to the committees 12.1.2.7 Senior Operators (alternate Supervisors)
1. To accept responsibility for safe and efficient operation of the Reactor Laboratorywhen designated by the Reactor Supervisor.
2. To maintain a Senior Operator's License.
12.1.2.8 Reactor operators
1. To hold a Reactor operator's License.
2. To conform to all rules and regulations for operation of the reactor.3. A reactor operator will be present at the control console at all times when the reactor is in operation.
UWNR Safety Analysis Report Rev. 2Sept. * *
2 . To ensure that all activities within the Reactor Laboratory are in accordance with prior approvals from the appropriate committees or from the Reactor Director.
3. The Reactor Supervisor shall have authority to authorize experiments and/or procedures which have been approved by the Reactor Safety Committee.
He will prepare specific detailed procedures based on the general procedures approved by the Committee. . 4. To see that all proper records are kept. 5. To maintain a Senior Operator's License. 6. To appoint Reactor Operators.
7. The Reactor Supervisor or another Senior operator shall be in charge of the Reactor Laboratory at all times (although not necessarily physically present).
The individual in charge, if physically present, shall be responsible for prompt execution of emergency procedures.
The Reactor Supervisor or another Senior operator will be present at the facility during fuel manipulation, reactor start-up and approach to power, and recovery from unscheduled scrams and shut-downs, and shall be available on call at other times during reactor operation . 8. 9. 12.1.2.7 1. 2. 12.1.2.8 1. 2. 3. To be responsible for safety in the Reactor Laboratory, including responsibility for health physics matters. To advise and prepare information for the committees concerned with the Reactor Laboratory, and to present such information to the committees Senior Operators (alternate Supervisors)
To accept responsibility for safe and efficient operation of the Reactor Laboratory when designated by the Reactor Supervisor.
To maintain a Senior Operator's License. Reactor operators To hold a Reactor operator's License. To conform to all rules and regulations for operation of the reactor. A reactor operator will be present at the control console at all times when the reactor is in operation . UWNR Safety Analysis Report Rev. 2 12-3 Sept. 2008
4. To monitor laboratoryactivities from a health-physics standpoint.
12.1.3 Staffing The minimum staffing when the reactor is not secured shall be: 1. A licensed reactor operator in the control room (if senior operator licensed, may also be the person required in 3 below): 2. A second designated person present at the facility complex able to carry out prescribed written instructions.
3. A designated senior reactor operator shall be readily available at the facility or on call.A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator.A licensed senior reactor operator shall be present at the facility for: 1. Initial startup and approach to power.2. All fuel handling or control-element manual manipulations.
3. Relocation of any in-core experiment with a reactivity worth greater than 0.7%AK/K.4. Recovery from unplanned or unscheduled shutdown or significant power reduction.
12.1.4 Selection and Training of Personnel The selection and training of operations personnel meets or exceeds the requirements of ANSI/ANS-15.4-1988 Sections 4-6 The operator training program includes sufficient radiation safety training to meet the requirements of 10 CFR Part 19 and the campus Radiation Safety Regulations. The operator training program is a two step process. First the candidate must take an elective four credit-hour course with a formal training manual, homework, and practical exercises (On the Job Training) included.
It includes the equivalent of 0.5 weeks of reactor fundamentals, 1.25 weeks of systems coverage, 0.5 weeks of systems observation, and 0.8 weeks of control room operations administered over the period of one semester.
The course iscompletely described on the UW Nuclear Reactor web page.U NRSaey nayisReot ev 1-4Spt 20 UWNR Safety Analysis Report Rev. 2 12-4 Sept. 4. To monitor laboratory activities from a health-physics standpoint.
12.1.3 Staffing The minimum staffing when the reactor is not secured shall be: 1. A licensed reactor operator in the control room (if senior operator licensed, may also be the person required in 3
2. A second designated person present at the facility complex able to carry out prescribed written instructions.
3. A designated senior reactor operator shall be readily available at the facility or on call. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator.
A licensed senior reactor operator shall be present at the facility for: 1. Initial startup and approach to power. , 2. All fuel handling or control-element manual manipulations.
3. Relocation of any in-core experiment with a reactivity worth greater than 0.7% LlKlK. 4. . Recovery from unplanned or unscheduled shutdown or significant power reduction.
12.1.4 Selection and Training of Personnel The selection and training of operations personnel meets or exceeds the requirements of ANSI! ANS-15.4-1988 Sections 4-6 1. The operator training program includes sufficient radiation safety training to meet the requirements of 10 CFR Part 19 and the campus Radiation Safety Regulations.
The operator training program is a two step process. First the candidate must take an elective four credit-hour course with a formal training manual, homework, and practical exercises (On the Job Training) included.
It includes the equivalent of 0.5 weeks of reactor fundamentals, 1.25 weeks of systems coverage, 0.5 weeks of systems observation, and 0.8 weeks of control room operations administered over the period of one semester.
The course is completely described on the UW Nuclear Reactor web page. UWNR Safety Analysis Report Rev. 2 12-4 Sept. 2008 l * *
After completing the course, successful candidates are selected to participate in the candidacyprogram. Under this program, candidates perform approximately 12 additional weeks of control room operations and other non-licensed duties, including preventative maintenance and health physics surveys.The operator proficiency maintenance program (re-qualification program) fully meets the requirements of 10 CFR Part 55 and is formalized as a facility procedure, UWNR 004 2, which received NRC approval upon initial implementation and is reviewed annually by the facility operating organization along with other facility procedures.
The program includes written, oral, and performance testing as well as emergency procedure drills and classes on changes inexperiments, facility equipment, and procedures.
12.1.5 Radiation Safety Radiation safety aspects of facility operation are routinely performed by members of the reactor operating staff, including routine radiation and contamination surveys and sampling of water and air samples. The campus radiation safety organization (see chapter 11), established to oversee all activities involving ionizing radiation on campus, is part of the University Department of Environment, Health and Safety, and thus is an independent organization which reports to the central campus administration.
The radiation safety organization has the authority to interdict or terminate radiation safety related activities conducted under the reactor license.UWNR Safety Analysis Report Rev. 2Sept. *
After completing the course, successful candidates are selected to participate in the candidacy program. Under this program, candidates perform approximately 12 additional weeks of control room operations and other non-licensed duties, including preventative maintenance and health physics surveys. The operator proficiency maintenance program (re-qualification program) fully meets the requirements of 10 CFR Part 55 and is formalized as a facility procedure, UWNR 004 2, which received NRC approval upon initial implementation and is reviewed annually by the facility operating organization along with other facility procedures.
The program includes written, oral, and performance testing as well as emergency procedure drills and classes on changes in experiments, facility equipment, and procedures.
12.1.5 Radiation Safety Radiation safety aspects of facility operation are routinely performed by members of the reactor operating staff, including routine radiation and contamination surveys and sampling of water and air samples. The campus radiation safety organization (see chapter 11), established to oversee all activities involving ionizing radiation on campus, is part ofthe University Department of Environment, Health and Safety, and thus is an independent organization which reports to the central campus administration.
The radiation safety organization has the authority to interdict or terminate radiation safety related activities conducted under the reactor license . UWNR Safety Analysis Report Rev. 2 12-5 Sept. 2008 BOARD OF REGENTS CHANCELLOR-MADISON CAMPUS UNIVERSITY OF WISCONSIN RADIATION SAFETY COMMITTEE University Safety Department Radiation Safety Office CHAIRMAN ENGINEERING PHYSICS DEPARTMENT (ANSI/ANS-15.4 Level 1)REACTOR SAFETY COMMITTEE REACTOR DIRECTOR (ANSI/ANS-15.4 Level 2)
Level (ANSI/ANS-1 5.1 Level 3)REACTOR OPERATORS (RO)(ANSI/ANS-1 5.1 Level 4)Figure 12-1Organization Chart UWNR Safety Analysis Report Rev. 2Sept. BOARD OF REGENTS
CHANCELLOR.:
MADISON CAMPUS UNIVERSITY OF WISCONSIN m RADIATION SAFETY COMMITTEE
.. :: .. if!;JJ,;l?j,/
.. ",,' I University Safety Department Radiation Safety Office CHAIRMAN ENGINEERING PHYSICS DEPARTMENT (ANSI/ANS-15.4 Level 1) REACTOR SAFETY COMMITTEE
,--:-* REACTOR DIRECTOR (ANSI/ANS-15.4 Level 2) REACTOR SUPERVISOR (SRO) (ANSI/ANS-15.1 Level 3) AL TERNATE SUPERVISORS (SRO) (ANSI/ANS-15.1 Level 3) REACTOR OPERATORS (RO) (ANSI/ANS-15.1 Level 4) Figure 12-1 Organization Chart UWNR Safety AnalysisJReport Rev. 2 12-6 Sept. 2008
12.2 Review and Audit Activities 12.2.1 Composition and Qualifications The Reactor Safety Committee is appointed by the Reactor Director. Minimum committee size is six members, one of whom is a health physicist from the University Radiation Safety office.Other members are faculty and staff of the university selected based on expertise to assure that the following disciplines are represented:
1. Reactor Physics -Nuclear Engineering
2. Mechanical Engineering
-Heat transfer and fluid mechanics
3. Metallurgy/Materials
4. Instruments and Control SystemsChemistry and Radio-chemistry
6. Radiation Safety Reactor operations staff is not precluded from membership on the committee as long as such members do not reach a majority of a quorum for voting. The health physics personnel who perform the monthly audits and inspections are invited to the meetings, but are not necessarily members of the committee.
12.2.2 Charter and Rules The Reactor Safety Committee operates with a written charter which specifies the manner inwhich business is conducted.
The charter includes rules on meeting frequency (at least annually),voting rules, agenda, quorums, use of subcommittees, minutes, and methods and content of submissions to the committee.
Provisions for use of telephone polls or subcommittees for approval of items not requiring a formal meeting are also a part of the charter.12.2.3 Review Function The reactor director or designee reviews all written operating procedures at least annually.Results of this review, along with suggested procedure revisions, are submitted to the Reactor Safety Committee for approval, or re-affirmation if no changes are deemed necessary.
The review responsibilities of the Reactor Safety Committee shall include, but are not limited to, the following:
UWNR Safety Analysis Report Rev. 2Sept; * *
12.2 Review and Audit Activities 12.2.1 Composition and Qualifications The Reactor Safety Committee is appointed by the Reactor Director.
Minimum committee size is six members, one of whom is a health physicist from the University Radiation Safety office. Other members are faculty and staff of the university selected based on expertise to assure that the following disciplines are represented:
1. Reactor Physics -Nuclear Engineering
2. Mechanical Engineering
-Heat transfer and fluid mechanics
3. Metallurgy/Materials
4. Instruments and Control Systems 5. Chemistry and Radio-chemistry
6. Radiation Safety Reactor operations staff is not precluded from membership on the committee as long as such members do not reach a majority of a quorum for voting. The health physics personnel who perform the monthly audits and inspections are invited to the meetings, but are not necessarily members of the committee.
12.2.2 Charter and Rules The Reactor Safety Committee operates with a written charter which specifies the manner in which business is conducted.
The charter includes rules on meeting frequency (at least annually), voting rules, agenda, quorums, use of subcommittees, minutes, and methods and content of submissions to the committee.
Provisions for use of telephone polls or subcommittees for approval of items not requiring a formal meeting are also a part of the charter. 12.2.3 Review Function The reactor director or designee reviews all written operating procedures atleast annually.
Results of this review, along with suggested procedure revisions, are submitted to the Reactor Safety Committee for approval, or re-affirmation if no changes are deemed necessary.
The review responsibilities of the Reactor Safety Committee shall include, but are not limited to, the following:
UWNR Safety Analysis Report Rev. 2 12-7 Sept. 2008
1. Review and approval of new experiments utilizing the reactor facilities.
2. Review and approval of all proposed changes to the facility, procedures, license, and technical specifications.
3. Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications.
4. Review of abnormal performance of plant equipment and operating anomalies having safety significance.
5. Review of unusual or reportable occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50.6. Review of audit reports.
7. Review of violations of technical specifications, license, or procedures and orders having safety significance.
12.2.4 Audit Function A Health Physicist from the University Radiation Safety Office represents the University Radiation Safety Committee and conducts an inspection of the facility at least monthly to assure compliance with the regulations of 10 CFR Part 20. The services and inspection function of the Radiation Safety Office are also used by the Reactor Safety Committee, with the scope of the audit extended to cover license, technical specification, and procedure adherence.
ProceduresWritten operating procedures are used to assure the safety of operation of the reactor. Procedure use does not preclude the use of independent judgement and action should the situation require such. Operating procedures are in effect for the following items: 1. Testing and calibration of reactor operating instrumentation and controls, controlrod drives, area radiation monitors, and air particulate monitors.2. Reactor startup, operation, and shutdown.3. Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support.
4. Fuel element and experiment loading or unloading.
UWNR Safety Analysis Report Rev. 2Sept. 1. Review and approval of new experiments utilizing the reactor facilities.
2. Review and approval of all proposed changes to the facility, procedures, license, and technical specifications.
3. Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications.
4. Review of abnormal performance of plant equipment and operating anomalies having safety significance.
5. Review of unusual or reportable occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50. 6. Review of audit reports. 7. Review of violations of technical specifications, license, or procedures and orders having safety significance.
12.2.4 Audit Function
A Health Physicist from the University Radiation Safety Office represents the University
Radiation Safety Committee and conducts an inspection of the facility at least monthly to assure compliance with the regulations of 10 CFR Part 20. The services and inspection function of the Radiation Safety Office are also used by the Reactor Safety Committee, with the scope of the audit extended to cover license, technical specification, and procedure adherence.
12.3 Procedures Written operating procedures are used to assure the safety of operation of the reactor. Procedure use does not preclude the use of independent judgement and action should the situation require such. Operating procedures are in effect for the following items: 1. Testing and calibration of reactor operating instrumentation and controls, control rod drives, area radiation monitors, and air particulate monitors.
2. Reactor startup, operation, and shutdown.
3. Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support. 4. Fuel element and experiment loading or unloading.
UWNR Safety Analysis Report Rev. 2 12-8 Sept. 2008
Control rod removal or replacement.Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety.Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes.Civil disturbances on or near the facility site.Substantive changes to the above procedures may be made only with the approval of the Reactor Safety Committee.
Temporary changes to the procedures that do not change their original intent may be made with the approval of two SROs. such temporary changes are documented and subsequently reviewed the Reactor Safety Committee.
12.4 Required Actions In the event a safety limit is exceeded:The reactor shall be shut down and reactor operation shall not be resumed until authorized the NRC.2. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section of the technical specifications.report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
This report shall be submitted to the Reactor Safety Committee (RSC) for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.reportable occurrence is defined as any of the following that occur during reactor operation:
1 Operation with any safety system setting less conservative than specified in the technical specifications.
2. Operation in violation of a Limiting Condition for Operation listed in the Technical Specifications.
UWNR Safety Analysis Report Rev. 2 129Sp.08Sept. * *
5 . 6. Control rod removal or Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety. 7. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes. 8. Civil disturbances on or near the facility site. Substantive changes to the above procedures may be made only with the approval of the Reactor Safety Committee.
Temporary changes to the procedures that do not change their original intent may be made with the approval of two SROs. All such temporary changes are documented and subsequently reviewed by the Reactor Safety Committee.
12.4 Required Actions In the event a safety limit is exceeded:
1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC . 2. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section 6.7 of the technical specifications.
3. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
This report shall be submitted to the Reactor Safety Committee (RSC) for review and then submitted to the NRC when authorization is sought to resume operation of the reactor. A reportable occurrence is defined as any of the following that occur during reactor operation:
1. Operation with any safety system setting less conservative than specified in the technical specifications.
2. Operation in violation of a Limiting Condition for Operation listed in the Technical Specifications . UWNR Safety Analysis Report Rev. 2 12-9 Sept. 2008
3. Operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of performing its intended safety function.4. Any unanticipated or uncontrolled change in reactivity greater than 0.7% AK/K, excluding reactor trips from a known cause.5. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of a condition which could result in operation of the reactor outsidethe specified safety limits.6. Abnormal and significant degradation in reactor fuel or cladding which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.In the event of an reportable occurrence as defined in the Technical Specifications, the following actions shall be taken: 1. The reactor shall be shut down.2. The Reactor Director or designated alternate shall be notified and corrective action taken with respect to the operations involved.3. The Director or designated alternate shall notify the Chairman of the Reactor Safety Committee.
4. A report shall be made to the Reactor Safety Committee which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.A report shall be made to the NRC.
ReportsReports will be made to NRC in accordance with the following:
1. An annual report covering the activities of the reactor facility during the previous calendar year shall be submitted (in writing to U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC 20555) within six months following the end of each calendar year, providing the following information:
UWNR Safety Analysis Report Rev.
2 12-10 Sept. 3. Operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of performing its intended safety function.
4. Any unanticipated or uncontrolled change in reactivity greater than 0.7%
excluding reactor trips from a known cause. 5. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of a condition which could result in operation of the reactor outside the specified safety limits. , 6. Abnormal and significant degradation in reactor fuel or cladding which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both. In the event of an reportable occurrence as defined in the Technical Specifications, the following actions shall be taken: 1. The reactor shall be shut down. 2. The Reactor Director or designated alternate shall be notified and corrective action taken with respect to the operations involved.
3. The Director or designated alternate shall notify the Chairman of the Reactor Safety Committee.
4. A report shall be made to the Reactor Safety Committee which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
5. A report shall be made to the NRC. 12.5 Reports Reports will be made to NRC in accordance with the following:
1. An annual report covering the activities of the reactor facility during the previous calendar year shall be submitted (in writing to U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC 20555) within six months following the end of each calendar year, providing the following information:
UWNR Safety Analysis Report Rev. 2 12-10 Sept. 2008 * * *
a. brief narrative summary of operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and results of surveillance tests and inspections.Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality.The number of emergency shutdowns and inadvertent scrams, including reasons therefor.Discussion of the major maintenance operations performed during the period, including-the effect, if any, on the safety of the operation of the*reactor and the reasons for any corrective maintenance required.
e. brief description, including a summary of the safety evaluations of changes in the facility or in the procedures and of tests and experiments carried pursuant to Section of CFR Part summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control, of the licensee as measured at or prior to the point of such release or discharge.Liquid effluents (summarized on a monthly basis)Liquid radioactivity discharged during the reporting period.Tabulated as follows: (a) Total estimated radioactivity released (in curies).The isotopic composition if greater than 1 x microcuries/cc for fission and activation products.(c) Total radioactivity (in curies), released nuclide, during the reporting period based on representative isotopic analysis.Average concentration at point of release (in microcuries/c) during the reporting period and the fraction of the applicable limit in CFR 20.UWNR Safety Analysis Report Rev. 2 1-1Sp.20 12-11 Sept. * *
a . A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections.
b. Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality.
c. The number of emergency shutdowns and inadvertent scrams, including reasons therefor.
d. Discussion of the major maintenance operations performed during the period, including-the effect, if any, on the safety of the operation of the _ reactor and the reasons for any corrective maintenance required.
e. A brief description, including a summary of the safety evaluations of changes in the facility or in the procedures and of tests and experiments carried pursuant to Section 50.59 of 10 CFR Part 50. f . A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge.
1. Liquid effluents (summarized on a monthly basis) (1) Liquid radioactivity discharged during the reporting Tabulated as follows: (a) Total estimated radioactivity released (in curies). (b) The isotopic composition if greater than 1 x 10-7 microcuries/cc for fission and activation products. (c) Total radioactivity (in curies), released by nuclide, during the reporting period based on representative isotopic analysis. (d) Average concentration at point of release (in microcuries/cc) during the reporting period and the fraction of the applicable limit in 10 CFR 20 . UWNR Safety Analysis Report Rev. 2 12-11 Sept. 2008 (2) Total volume (in gallons) of effluent water (including diluent) during periods of release.ii. Gaseous Waste (summarized on a monthly basis)(1) Radioactivity discharged during the reporting period (in curies) for: (a) Gases.(b) Particulates with half lives greater than eight days.The estimated activity (in curies) discharged during thereporting period, by nuclide, for all gases and particulates based on representative isotopic analysis and the fraction of the applicable 10 CFR 20 limits for these values.iii. Solid Waste (1) The total amount of solid waste packaged (in cubic feet).(2) The total activity involved (in curies).(3) The dates of shipment and disposition (if shipped off site).g. A summary of radiation exposures received by facility personnel and visitors, including dates and time of significant exposures and a summary of the results of radiation and contamination surveys performed within the facility.h. A description of any environmental surveys performed outside the facility.2. A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555 with a copy to the NRC compliance inspector assigned to the facility) upon receipt of a new facility license or an amendment tothe license authorizing an increase in reactor power level describing the measured values of the operating conditions or characteristics of the reactor under the new conditions including:
a. An evaluation of facility performance to date in comparison with design predictions and specifications.
UWNR Safety Analysis Report Rev. 2 12-12 Sept. (2) Total volume (in gallons) of effluent water (including diluent) during periods of release. 11. Gaseous Waste (summarized on a monthly basis) (1) Radioactivity discharged during the reporting period (in curies) for: (a) Gases. (b) Particulates with half lives greater than eight days. The estimated activity (in curies) discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis and the fraction of the applicable 10 CFR 20 limits for these values. 111. Solid Waste (1) The total amount of solid waste packaged (in cubic feet). (2) The total activity involved (in curies). (3) The dates of shipment and disposition (if shipped off site). g. A summary of radiation exposures received by facility personnel and visitors, including dates and time of significant exposures and a summary of the results of radiation and contamination surveys performed within the facility.
h. A description of any environmental surveys performed outside the facility.
2. A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555 with a copy to the NRC compliance inspector assigned to the facility) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating conditions or characteristics of the reactor under the new conditions including:
a. An evaluation of facility performance to date in comparison with design predictions and specifications.
UWNR Safety Analysis Report Rev. 2 12-12 Sept. 2008 * * *
b. A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.3. A report of any of the following not later than the following day by telephone or similar conveyance to the NRC Headquarters Operation Center, and followed by a written report describing the circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the NRC inspector assigned to the facility: a. Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure.b. Any violation of a safety limit.c. Any reportable occurrences.
4. A written report within 30 days in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, of: a. Permanent changes in facility organization at Reactor Director or Department Chair level.b. Any significant change in the transient or accident analysis as described in the Safety Analysis Report.
Records The following records are retained for a period of at least five years or for the life of the component involved if less than five years.1. Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc. which shall be maintained for a period of at least one year)2. Principal maintenance activities
3. Reportable occurrences
4. Surveillance activities required by the Technical Specifications UWNR Safety Analysis Report Rev. 2Sept. * *
b . A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.
3. A report of any of the following not later than the following day by telephone or similar conveyance to the NRC Headquarters Operation Center, and followed by a written report describing the circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555, with a copy to the NRC inspector assigned to the facility:
a. Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure.
b. Any violation of a safety limit. c. Any reportable occurrences.
4. A written report within 30 days in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555, of: a. b. 12.6 Records Permanent changes in facility organization at Reactor Director or Department Chair level. Any significant change in the transient or accident analysis as described in the Safety Analysis Report. The following records are retained for a period of at least five years or for the life of the component involved if less than five years. 1. Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc. which shall be maintained for a period of at least one year) 2. Principal maintenance activities
3. Reportable occurrences
4. Surveillance activities required by the Technical Specifications UWNR Safety Analysis Report Rev. 2 12-13 . Sept. 2008 Reactor facility radiation and contamination surveys where required by applicable regulations
6. Experiments performed with the reactor 7. Fuel inventories, receipts, and shipments 8. Approval of changes in operating procedures
9. Records of meeting and audit reports of the review and audit group.
Operator qualification and re-qualification records will be retained for at least one cycle of the re-qualification program.The following records will be retained for the lifetime of the reactor facility. (Note: Retention of annual reports which contain the information in items 1. and 2. are considered as suitable records for those items.)1. Gaseous and liquid radioactive effluents released to the environs, 2. Offsite environmental monitoring surveys required by technical specifications
3. Radiation exposures for all personnel monitored 4. Updated, corrected, and as-built drawings of the facility 12.7 Emergency Planning The Emergency Plan for the University of Wisconsin Nuclear Reactor was prepared to meet the requirements of ANSI/ANS 15.16-1978 as amplified by Nureg-0849 This plan was submitted to NRC for review in May 21,1980, with subsequent revisions in October 25, 1982, and May 17, 1984. By letter dated July 25, 1984 NRC indicated that the plan met the requirements referenced above. The plan was again modified and submitted to NRC on May 16, 1990, with supporting information submaitted on August 12, 1990. NRC notification that the revision was acceptable was received in a letter dated April 26, 1991. The plan was again modified (Revision
4) to reflect the changes in section number and nomenclature of 10 CFR Part 20 and submitted to NRC on February 17, 1994 and April 22, 1994.
This version is the current version in use at the facility.The Emergency Plan indicates response capabilities for emergency conditions arising in connection with operation of the reactor.
It includes identification of various precursor conditions (loss of electrical power, fires, reactor pool leaks, riots, etc) and the consequences for various independent or simultaneous precursor.
The plan includes the event classification system. The dose to which people could be exposed under various conditions is indicated, as are the actionsUWNR Safety Analysis Report Rev. 2 12-i4 Sept. 5. Reactor facility radiation and contamination surveys where required by applicable regulations
6. Experiments performed with the reactor 7. Fuel inventories, receipts, and shipments
8. Approval of changes in operating procedures
9. Records of meeting and audit reports of the review and audit group. Operator qualification and re-qualification records will be retained for at least one cycle of the qualification program. The following records will be retained for the lifetime of the reactor facility. (Note: Retention of annual reports which contain the information in items 1. and 2. are considered as suitable records for those items.) 1. Gaseous and liquid radioactive effluents released to the environs 2. Offsite environmental monitoring surveys required by technical specifications
3. Radiation exposures for all personnel monitored
4. Updated, corrected, and as-built drawings of the facility 12.7 Emergency Planning The Emergency Plan for the University of Wisconsin Nuclear Reactor was prepared to meet the requirements of ANSI/ANS 15.16-1978 3 as amplified by Nureg-0849
4. This plan was submitted to NRC for review in May 21,1980, with subsequent revisions in October 25, 1982, and May 17, 1984. By letter dated July 25,1984 NRC indicated that the plan met the requirements referenced above. The plan was again modified and submitted to NRC on May 16, 1990, with supporting . information subinitted on August 12, 1990. NRC notification that the revision was acceptable was received in a letter dated April 26, 1991. The plan was again modified (Revision
4) to reflect the changes in section number and nomenclature of 10 CFR Part 20 and submitted to NRC on February]
7, 1994 and April 22, 1994. This version is the current version in use at the facility.
The Emergency Plan indicates response capabilities for emergency conditions arising in connection with operation of the reactor. It includes identification of various precursor conditions (loss of electrical power, fires, reactor pool leaks, riots, etc) and the consequences for various independent or simultaneous precursor.
The plan includes the event classification system. The dose to which people could be exposed under various conditions is indicated, as are the actions UWNR Safety Analysis Report Rev. 2 12-14 Sept. 2008 * *
that can taken to minimize the consequences of the emergency.
Detailed emergency implementing procedures have been developed and are referenced in the plan.Primary responsibility for emergency planning and response is given to the Reactor Director.
Delegation of responsibility and authority in the absence of the Reactor Director is specified.
The Emergency Plan and implementing procedures are reviewed annually to assure that any required changes are incorporated into the plan.
Security Planning The facility physical security plan, UWvNR was initially submitted to NRC on October The security plan was revised and submitted again on June and was found to meet the applicable requirements, including the format of Regulatory Guide 5.59 The plan will require revision as a result of this Safety Analysis Report, since some figures from the previous Safety Analysis Report are included reference. These changes will be made during the usual annual reviews of the plan once this SAR becomes the document referenced in the reactor operating license.The security plan indicates the measures provided to protect special nuclear material, including details of the protective equipment and police agencies, and is thus withheld from publicdisclosure. The Reactor Director is responsible for administering the security program andassuring that it is updated as required.
Quality Assurance Since no construction permit is sought in the application for renewal of the license for the University of Wisconsin Nuclear Reactor, no description of a quality assurance program for the design and construction of the structures, systems, and components of the facility is included.This section describes the Quality Assurance program that is in place to govern safe operation and modification of the facility. This program meets the applicable requirements of Regulatory Guide 2.5 and The Reactor Director has responsibility for the quality assurance activities, and thus has the authority to identify problems, to initiate corrective actions, and to insure that corrective actions are performed.
He exercises oversight assuring that operating and maintenance procedures include specific requirements to assure that modification, maintenance, and calibration of safety-related systems are performed in a manner that maintains the quality and reliability of equipment.
Further, experiment reviews use written requirements to assure that installation and operation of the experiment does not degrade the performance of safety equipment.
Modification of safety-related equipment is planned and reviewed using formalwritten checklist-type procedures that assure that equipment continues to meet the original specifications.
Most of the reactor equipment in use in the facility does not have formal documentation because it was built before the requirements were in effect. This equipment is Safety Analysis Report Rev. 2 1-5Sp.20 Sept. * *
that can be taken to minimize the consequences ofthe emergency.
Detailed emergency implementing procedures have been developed and are referenced in the plan. Primary responsibility for emergency planning and response is given to the Reactor Director.
Delegation of responsibility and authority in the absence of the Reactor Director is specified.
The Emergency Plan and implementing procedures are reviewed annually to assure that any required changes are incorporated into the plan. 12.8 Security Planning The facility physical security plan, UWNR 003, was initially submitted to NRC on October 18, 1988. The security plan was revised and submitted again on June 17,1991 and was found to meet the applicable requirements, including the format of Regulatory Guide 5.59 5. The plan will require revision as a result of this Safety Analysis Report, since some figures from the previous Safety Analysis Report are included by reference.
These changes will be made during the usual annual reviews of the plan once this SAR becomes the document referenced in the reactor operating license. The security plan indicates the measures provided to protect special nuclear material, including details of the protective equipment and police agencies, and is thus withheld from public disclosure.
The Reactor Director is responsible for administering the security program and assuring that it is updated as required . 12.9 Quality Assurance Since no construction permit is sought in the application for renewal of the license for the . University of Wisconsin Nuclear Reactor, no description of a quality assurance program for the design and construction of the structures, systems, and components of the facility is included.
This section describes the Quality Assurance program that is in place to govern safe operation and modification of the facility.
This program meets the applicable requirements of Regulatory Guide 2.5 6 and ANSIIANS-15.8-1995 7 The Reactor Director has responsibility for the quality assurance activities, and thus has the authority to identify problems, to initiate corrective actions, and to insure that corrective actions are performed.
He exercises QA oversight by assuring that operating and maintenance procedures include specific requirements to assure that modification, maintenance, and calibration of safety-related systems are performed in a manner that maintains the quality and reliability of equipment.
Further, experiment reviews use written requirements to assure that installation and operation of the experiment does not degrade the performance of safety equipment.
Modification of safety-related equipment is planned and reviewed using formal written checklist-type procedures that assure that equipment continues to meet the original specifications.
Most ofthe reactor equipment in use in the facility does not have formal QA documentation because it was built before the QA requirements were in effect. This equipment is UWNR Safety Analysis Report Rev. 2 12-15 Sept. 2008 covered under the provisions of section 4 of ANSI/ANS-15.8.
Several instruments are replacements for the vacuum-tube electronics originally provided by the reactor manufacturer, General Electric Company. This replacement equipment was designed, built, and tested to meet the original specifications stated in the equipment manuals provided with the General Electric equipment.
After-maintenance checks, alignment, and calibration of the replacement equipment still assures the equipment meets the original equipment specifications.Procedures include schedules of equipment maintenance and calibration, and provide records that such functions have been completed.
Calibration procedures include requirements that critical equipment and instruments used in the calibrations are themselves currently calibrated (when appropriate).
12.10 Operator Training and Requalification Operator training and Requalification programs are briefly described in section 12.1.4. The Requalification plan at UWNR is published as a standard procedure (UWNR 004, "University of Wisconsin Nuclear Reactor Operator Proficiency Maintenance Program")
which was submitted to NRC on October 24, 1973 and revised on February 7, 1974. By letter dated March 29, 1974 we were notified by NRC that the program meets the requirements of Section 50.54(i-1) of 10 CFR Part 50 and Appendix A of 10 CFR Part 55. Since the program is a numbered procedure it is reviewed by management on an annual basis.12.11 Startup Plan The facility has been in routine operation for many years, so a startup plan is not included in this Safety Analysis Report for license renewal.12.12 Environmental Reports On January 23, 1974 the AEC staff concluded in a memorandum addressed to D. Skovholt and signed by R. Miller, "that there will be no significant environmental impact associated with the licensing of research reactors or critical facilities designed to operate at power levels of 2 Mwt or lower and that no environmental impact statements are required to be written for the issuance of construction permits or operating licenses for such facilities." Since this Safety Analysis Report is written in support of extending the license expiration date for an additional 20 years, no changes in land and water use are contemplated.
Emissions of radioactive materials or other effluents will not change as a result of extending the license term.
References
1. Standard ANSI/ANS-15.4-1988, Selection and Training of Personnel For ResearchReactors, American Nuclear Society, June 9, 1988 ANSI Approval UWNR Safety Analysis Report Rev. 2 Sept. covered under the provisions of section 4 of ANSI/ANS-15.8.
Several instruments are
replacements for the vacuum-tube electronics originally provided by the reactor manufacturer, General Electric Company. This replacement equipment was designed, built, and tested to meet the original specifications stated in the equipment manuals provided with the General Electric equipment.
After-maintenance checks, alignment, and calibration of the replacement equipment still assures the equipment meets the original equipment specifications.
Procedures include schedules of equipment maintenance and calibration, and provide records that such functions have been completed.
Calibration procedures include requirements that critical equipment and instruments used in the calibrations are themselves currently calibrated (when appropriate
). 12.10 Operator Training and Requalification Operator training and Requalification programs are briefly described in section 12.1.4. The Requalification plan at UWNR is published as a standard procedure (UWNR 004, "University of Wisconsin Nuclear Reactor Operator Proficiency Maintenance Program")
which was submitted to NRC on October 24, 1973 and revised on February 7, 1974. By letter dated March 29, 1974 we were notified by NRC that the program meets the requirements of Section 50.54(i-l) of 10 CFR Part 50 and Appendix A of 10 CFR Part 55. Since the program is a numbered procedure it is reviewed by management on an annual basis. 12.11 Startup Plan The facility has been in routine operation for many years, so a startup plan is not included in this Safety Analysis Report for license renewal. 12.12 Environmental Reports On January 23, 1974 the AEC staff concluded in a memorandum addressed to D. Skovholt and signed by D. R. Miller, "that there will be no significant environmental impact associated with the licensing of research reactors or critical facilities designed to operate at power levels of 2 Mwt or lower and that no environmental impact statements are required to be written for the issuance of construction permits or operating licenses for such facilities." Since this Safety Analysis Report is written in support of extending the license expiration date for an additional 20 years, no changes in land and water use are contemplated.
Emissions of radioactive materials or other effluents will not change as a result of extending the license term. 12.13 References
1. Standard ANSIIANS-15.4-1988, Selection and Training of Personnel For Research Reactors, American Nuclear Society, June 9, 1988 ANSI Approval UWNR Safety Analysis Report Rev. 2 12-16 Sept. 2008 * *
2. UWNR 004, Operator Proficiency Maintenance Program 3. ANSI/ANS 15.16-1978, "Emergency; Planning for Research Reactors", ANS, LaGrangePark, Illinois, 1978 4. NUREG-0849, "Standard Review Plan for the Review and Evaluation of Emergency Plans for Research and Test Reactors", USNRC, October 1983 5. Regulatory Guide 5.59, Revision 1, "Standard Format and Content for A Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or LowStrategic Significance, US Nuclear Regulatory Commission, February 1983 6. Regulatory Guide 2.5, Revision 0-R, :Quality Assurance Program Requirements for Research Reactors, October 1977 7. ANSI-15.8-1995, "quality assurance program requirements for research reactors", ANS, La Grange Park, Illinois, 1995 UWNR Safety Analysis Report Rev. 2 Sept. * *
2. UWNR 004, Operator Proficiency Maintenance Program 3. ANSIIANS 15.16-1978, "Emergency; Planning for Research Reactors", ANS, LaGrange Park, Illinois, 1978 4. NUREG-0849, "Standard Review Plan for the Review and Evaluation of Emergency Plans for Research and Test Reactors", USNRC, October 1983 5. Regulatory Guide 5.59, Revision 1, "Standard Format and Content for A Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance, US Nuclear Regulatory Commission, February 1983 6. Regulatory Guide 2.5, Revision O-R, :Quality Assurance Program Requirements for Research Reactors, October 1977 7. ANSI-15.8-1995, "quality assurance program requirements for research reactors", ANS, La Grange Park, Illinois, 1995 UWNR Safety Analysis Report Rev. 2 12-17 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2 12- Sept.
This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 2 12-18 Sept. 2008
NUREG -1537 divides accident analysis into initiating and consequences sections.
In this report the two sections are combined; that is, the analyses of the consequences of accidents are grouped with the accident-initiating events and scenarios in order to reduce duplication which would otherwise occur. In addition, a better appreciation of the likelihood and consequences of accidents is afforded.
The sections are numbered to correspond to the numbering system of NUREG-1537.
Accident Analysis Initiating Events and Determination of Consequences Note: NUREG-1537 specifies separate sections for Initiating Events and Determination ofConsequences. Both of these sections have been combined for clarity.
NUREG/CR-2387 2 reports an independent study of accidents in TRIGA-type reactors and concludes "The only potential for offsite exposure appears to be from a fuel-handling accidentthat, based on highly conservative assumptions, would result in dose equivalents of 1 mrem to the total body from noble gases and
__ 1.2 rem to the thyroid from radioiodines." Notwithstanding that conclusion, the following sections reiterate the analysis done for a license amendment allowing the University of Wisconsin Nuclear Reactor to operate with FLIP or mixed Standard-FLIP cores, using values specific to the UWNR reactor location and characteristics.
13.1.1 Maximum Hypothetical Accident The maximum hypothetical accident for UWNR is postulated as damage to a fuel element resulting in failure of the fuel cladding.
It is postulated that this damage occurs after a very long time of operation at 125% of full power (the power level limiting safety system setting) and that it occurs in the fuel element with the highest power density possible in permitted UWNR mixed core fuel loadings.
In a compact 9-Bundle FLIP core the highest power density is 18.1 kW at 1000 kW; the corresponding number for the 15-bundle FLIP core is 17.2 kW, while the value for the currently-used all FLIP core is only 15.2 kW. Continuous operation at the power level scram setpoint is highly unlikely, but for this hypothetical computation the power level in themaximally exposed fuel element of the 9-bundle core is assumed to be 23 kW (1.25 times 18.1 kW rounded to the next highest number).The likelihood of a major fuel element cladding failure is considered small. The elements mustmeet rigid quality control standards; pool water quality is carefully controlled; and much care is taken in handling fuel. Such clad failures are, however, possible and the remainder of this section is concerned with the consequences of such a failure.The release of radioactivity by corrosion and leaching by the pool water has been measured at Gulf General Atomic. About 100 micrograms of U-ZrH per square centimeter of exposed fuel surface per day is released for shutdown conditions. This release is easily controlled by isolating UWNR Safety Analysis Report Rev. 2Sept.
13 ACCIDENT ANALYSIS NUREG -1537 1 divides accident analysis into initiating and consequences sections.
In this report the two sections are combined; that is, the analyses of the consequences of accidents are grouped with the accident-initiating events and scenarios in order to reduce duplication which would otherwise occur. In addition, a better appreciation of the likelihood and consequences of accidents is afforded.
The sections are numbered to correspond to the numbering system of NUREG-1537.
13.1 Accident Analysis Initiating Events and Determination of Consequences Note: NUREG-1537 specifies separate sections for Initiating Events and Determination of Consequences.
Both of these sections have been combined for clarity. NUREG/CR-2387 2 reports an independent study of accidents in TRIGA-type reactors and concludes "The only potential for offsite exposure appears to be from a fuel-handling accident that, based on highly conservative assumptions, would result in dose equivalents of .s;; 1 mrem to the total body from noble gases and .s;; 1.2 rem to the thyroid from radioiodines." Notwithstanding that conclusion, the following sections reiterate the analysis done for a license amendment allowing the University of Wisconsin Nuclear Reactor to operate with FLIP or mixed FLIP cores, using values specific to the UWNR reactor location and characteristics.
l3.1.1 Maximum Hypothetical Accident
The maximum hypothetical accident for UWNR is postulated as damage to a fuel element resulting in failure of the fuel cladding.
It is postulated that this damage occurs after a very long time of operation at 125% of full power (the power level limiting safety system setting) and that it occurs in the fuel element with the highest power density possible in permitted UWNR mixed core fuel loadings.
In a compact 9-Bundle FLIP core the highest power density is 18.1 kWat 1000 kW; the corresponding number for the 15-bundle FLIP core is 17.2 kW, while the value for the currently-used all FLIP core is only 15.2 kW. Continuous operation at the power level scram setpoint is highly unlikely, but for this hypothetical computation the power level in the maximally exposed fuel element of the 9-bundle core is assumed to be 23 kW (1.25 times 18.l kW rounded to the next highest number). The likelihood of a major fuel element cladding failure is considered small. The elements must meet rigid quality control standards; pool water quality is carefully controlled; and much care is taken in handling fuel. Such clad failures are, however, possible and the remainder of this section is concerned with the consequences of such a failure. The release of radioactivity by corrosion and leaching by the pool water has been measured at Gulf General Atomic. About 100 micrograms of U -ZrH per square centimeter of exposed fuel surface per day is released for shutdown conditions.
This release is easily controlled by isolating UWNR Safety Analysis Report Rev. 2 13-1 Sept. 2008 the leaking element in a container provided for that purpose. The gaseous and highly volatile fission products that have collected in the space between fuel and cladding would be the activity contributing to personnel hazards.13.1.1.1 Fission Product Inventory in Fuel Element The quantity of these volatile and gaseous fission products was determined by the use of Perkins and King 3 data. Column B of Table 13.1 indicates the fission product activities in the fuel element exposed to the maximum power density.13.1.1.2 Fission Product Release Fraction The release of fission products from U-ZRH fuel elements has been extensively studied by Gulf General Atomic and others.
The results of this work indicate that the release of fission product gases into the gap between fuel and cladding is given by the following relationship:
FR=1.5E-5
+ 3.6E3exp(-1.34E4/T) where T is the maximum fuel temperature
('K) in the element during normal operation.
The maximum fuel temperature in a fuel element operated in the steady-state mode at 23 KW will be less than 440 Calculations of release fraction however, are based on 600 in order to assure a conservative result.The release fraction corresponding to 600 is 7.9 E-4. Applying this fraction to the total inventory of the fuel element as given in column B of Table, 03.1 gives the released activity as shown in column C of the table.For the purpose of further calculations, it is assumed that all gaseous fission products are released to the room air whether the pool is filled with water or not. For soluble volatiles, calculations assume all activity is absorbed in pool water for calculations of pool water activity (column D).For calculations of air activity, the assumption is made that 10% of the volatiles escape with thepool filled with water (columns E and F) and 100% escape with the pool empty.
13.1.1.3 Activity in Pool Water If 100% of the soluble fission products are absorbed in the pool water, the resulting activity 'level will be 0.075 [tCi/ml. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the level would be reduced by radioactive decay to about 0.012 p.Ci/ml. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the activity decay rate would be chiefly determined by the half life (8.05 days). The demineralizer will remove most of this activity, giving a radiation dose rate of about 88 mrem/hr at one meter after the activity is deposited in the resins. The resins can be dumped to an underground storage pit or the underground liquid waste holdup tank where the activity will decay without hazard to personnel.
UWNR Safety Analysis Report Rev. 2Sept. the leaking element in a container provided for that purpose. The gaseous and highly volatile
fission products that have collected in the space between fuel and cladding would be the activity contributing to personnel hazards. 13.1.1.1 Fission Product Inventory in Fuel Element The quantity of these volatile and gaseous fission products was determined by the use of Perkins and King 3 data. Column B of Table 13.1 indicates the fission product activities in the fuel element exposed to the maximum power density. 13.1.1.2 Fission Product Release Fraction The release of fission products from U-ZRH fuel elements has been extensively studied by Gulf General Atomic and others. The results of this work indicate that the release of fission product gases into the gap between fuel and cladding is given by the following relationship:
FR=1.5E-5
+ 3.6E3exp(-1.34E4/T) where T is the maximum fuel temperature (DK) in the element during normal operation.
The maximum fuel temperature in a fuel element operated in the steady-state mode at 23 KW will be less than 440 DC. Calculations of release fraction however, are based on 600 DC in order
to assure a conservative result. The release fraction corresponding to 600 DC is 7.9 E-4. Applying this fraction to the total inventory of the fuel element as given in column B of Table' t3.1 gives the released activity as , shown in column C of the table. For the purpose of further calculations, it is assumed that all gaseous fission products are released to the room air whether the pool is filled with water or not. For soluble volatiles, calculations assume all activity is absorbed in pool water for calculations of pool water activity (column D). For calculations of air activity, the assumption is made that 10% of the volatiles escape with the pool filled with water (columns E and F) and 100% escape with the pool empty. 13.1.1.3 Activity in Pool Water If 100% of the soluble fission products are absorbed in the pool water, the resulting activity level will be 0.075 IlCi/ml. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the level would be reduced by radioactive decay to about 0.012 IlCi/ml. After 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the activity decay rate would be chiefly determined by the 1-131 halflife (8.05 days). The demineralizer will remove most of this activity, giving a radiation dose rate of about 88 mrem/hr at one meter after the activity is deposited in the resins. The resins can be dumped to an underground storage pit or the underground liquid waste holdup tank where the activity will decay without hazard to personnel.
UWNR Safety Analysis Report Rev. 2 13-2 Sept. 2008
13.1.1.4 Fission Product Release to Air within the Reactor Laboratory It is estimated that it would take an individual five minutes to evacuate confinement and an additional five minutes to evacuate the building.
However for conservatism, it is assumed that the individual is exposed at the highest concentration, namely that found in confinement, for the full ten minutes. Calculations were performed in Appendix A to determine (1) the dose rate due to gamma emitters uniformly dispersed throughout the volume of the reactor lab, (2) the dose tothe lungs from beta emitters for an individual remaining in the laboratory for ten minutes; and (3)the dose to the thyroid of an individual remaining in the room ten minutes. For the latter calculations, it is assumed that 10% of the iodine radioisotopes escape from the pool water. In addition to these calculations, a computation of the number of DAC hours indicates that a person present in the room for 10 minutes after the release would receive less than the annual limit on intake for occupational exposure.(1) Whole body exposure due to gamma emitters The amount of insoluble volatiles released to the room would be 5.89 Ci. If this activity isdistributed uniformly in the laboratory volume, the resulting concentration would be 2.95E-3 I.Ci/cm 3.The resulting maximum dose rate is calculated to be 60 mrem/hr. An individual remaining in the laboratory for 10 minutes after a release would receive a whole body dose of 10 mrem.(2) Dose to the lungs The lung is the critical organ when considering the effects of inhaling the insoluble volatiles from a ruptured fuel element. The beta emitting nuclides become more important than those emitting gamma rays since all the decay energy is absorbed in lung tissue. The calculation outlined in theappendix indicates the lung exposure for an individual remaining in the laboratory for 10 minutes after a clad rupture to be 1.0 rad.(3) Thyroid dose The thyroid dose to a person in the reactor room was calculated assuming that he remained in the laboratory for 10 minutes after the fission product release. If the pool water is not lost and 10%of the halogens released escape into the atmosphere, the concentrations of the various iodine isotopes would be as presented in Table 13.1. In a ten minute period the lungs would be exposed to the iodine isotope activities shown in Table 13.2. As before, it was assumed that the"standard man" breathes 1.25 m 3/active-hour and his lungs hold 3 liters of air. A conservativecalculation results in a dose to the thyroid of 18.9 rads. Although all doses were calculated based on an individual remaining in the laboratory for ten minutes, emergency procedures require immediate evacuation after scramming the reactor, and re-entry to the area is made using a powered air purifying respirator. Actual doses in the event of the accident would be a factor of 10 less than calculated, considering reasonable evacuation times.UWNR Safety Analysis Report Rev. 2Sept. * *
13.1.1.4 Fission Product Release to Air within the Reactor Laboratory It is estimated that it would take an individual five minutes to evacuate confinement and an additional five minutes to evacuate the building.
However for conservatism, it is assumed that the individual is exposed at the highest concentration, namely that found in confinement, for the full ten minutes. Calculations were performed in Appendix A to determine (1) the dose rate due to gamma emitters uniformly dispersed throughout the volume of the reactor lab, (2) the dose to the lungs from beta emitters for an individual remaining in the laboratory for ten minutes; and (3) the dose to the thyroid of an individual remaining in the room ten minutes. For the latter calculations, it is assumed that 10% ofthe iodine radioisotopes escape from the pool water. In addition to these calculations, a computation of the number ofDAC hours indicates that a person present in the room for 10 minutes after the release would receive less than the annual limit on intake for occupational exposure.
(1) Whole body exposure due to gamma emitters The amount of insoluble volatiles released to the room would be 5.89 Ci. If this activity is distributed uniformly in the laboratory volume, the resulting concentration would be 2.95E-3 IlCi/cm 3* The resulting maximum dose rate is calculated to be 60 mrem/hr. An individual remaining in the laboratory for 10 minutes after a release would receive a whole body dose of 10 mrem . (2) Dose to the lungs The lung is the critical organ when considering the effects of inhaling the insoluble volatiles from a ruptured fuel element. The beta emitting nuclides become more important than those emitting gamma rays since all the decay energy is absorbed in lung tissue. The calculation outlined in the appendix indicates the lung exposure for an individual remaining in the laboratory for 10 minutes after a clad rupture to be 1.0 rad. (3) Thyroid dose The thyroid dose to a person in the reactor room was calculated assuming that he remained in the laboratory for 10 minutes after the fission product release. If the pool water is not lost and 10% of the halogens released escape into the atmosphere, the concentrations of the various iodine isotopes would be as presented in Table 13.1. In a ten minute period the lungs would be exposed to the iodine isotope activities shown in Table 13.2. As before, it was assumed that the "standard man" breathes 1.25 m 3/active-hour and his lungs hold 3 liters of air. A conservative calculation results in a dose to the thyroid of 18.9 rads. Although all doses were calculated based on an individual remaining in the laboratory for ten minutes, emergency procedures require immediate evacuation after scramming the reactor, and re-entry to the area is made using a powered air purifYing respirator.
Actual doses in the event of the accident would be a factor of 10 less than calculated, considering reasonable evacuation times . UWNR Safety Analysis Report Rev. 2 13-3 Sept. 2008 TABLE 13.1 FISSION PRODUCT RELEASE FROM CLAD RUPTURE A B C D E ISOTOPE SATURATED RELEASED AMOUNT AMOUNT 1N F LABORATORY CONCENTRATION (ItCi/ml)Br 82 83 84'85'87 TOTAL Br I '130m 131 132 133 134 135'136 TOTAL I Kr 83m 85m 85 87 88'89 TOTAL Kr INVENTORY INVENTORY (Ci) (Ci)30 0.024 105 0.083 194 0.153 253 0.200 600 0.473 0.933 200 0.158 563 0.446 855 0.677 1282 1.015 1554 1.230 1185 0.938 602 0.477 4.941 WATER (Ci)0.024 0.083 0.154 0.200 0.475 0.158 0.446 0.677 1.015 1.230 0.938 0.477 AIR (Ci)G H Part 20 CONCENTRATION TABLE 1 DISCHARGED
([.tCi/ml)
DAC (-tC i/ml)I Part 20 TABLE 2 MAX.EFFLUENT CONCENTRATION (4tCi/ml)6E-09 9E-08 8E-08 IE-09 1E-09 J RATIO COL. H/I 105 253 51 486 699 855 0.002 0.008 0.015 0.020 0.047 0.093 0.016 0.045 0.068 0.102 0.123 0.094 0.048 0.494 0.084 0.200 0.040 0.385 0.555 0.669 1.935 0.004 0.025 1.015 0.277 0.984 0.938 0.707 3.950 1.2E-06 4.2E-06 7.7E-06 1.OE-05 2.4E-05 7.9E-06 2.2E-05 3.4E-05 5.1E-05 6.2E-05 4.7E-05 2.4E-05 4.2E-05 1.OE-04 2.OE-05 1.9E-04 2.8E-04 3.4E-04 2.OE-06 1.6E-05 5. 1E-04 1.4E-04 4.9E-04 4.7E-04 3.5E-04 2E-06 3E-05 2E-05 1E-07 1E-07 1E-07 2E-08 3E-06 1E-07 2E-05 7E-07 1E-07 IE-02 2E-05 1E-04 5E-06 2E-06 1E-07 4E-04 1E-04 1E-04 9E-06 IE-07 4E-06 1.46E-10 4.95E-10 9.23E-10 1.20E-09 2.85E-09 9.45E-10 2.68E-09 4.06E-09 6.09E-09 7.38E-09 5.63E-09 2.86E-09 4.95E-10 1.20E-08 2.40E-09 2.3 1E-08 3.32E-08 4.06E-08 2.36E-10 6.09E-08 1.67E-08 5.91E-08 5.63E-08 4.24E-08 0.084 0.200 0.040 0.386 0.556 0.669 1.935 1E-09 2E-10 2E-08 IE-09 6E-08 6E-09E-09 1E-07 7E-07 2E-08 9E-09 1E-09 2E-06 6E-07 5E-07 4E-08 7E-08 1E-09 2E-08 0.0244 0.0055 0.0115 1.2038 2.8463 0.9450 13.33875 0.2031 6.0863 0.1230 0.9375 2.8575 9.9E-06 0.1204 0.0034 1.1531 3.6875 40.6125 0.0001 0.0025 0.1217 0.4163 0.8438 56.2500 2.1206 Xe 131m 5 0.004 133m 31 0.025 133 1282 1.015 135m 0.2771243 0.984'137 1185 0.938 138 894 0.707 TOTAL Xe 3.950'indicates generic <2 hr half-life values used UWNR Safety Analysis Report Rev. 2 13-4 Sept. * *
TABLE 13.1 FISSION PRODUCT RELEASE FROM CLAD RUPTURE A B C D E F G H J ISOTOPE SATURATED RELEASED AMOUNT IN AMOUNT IN LABORATORY Part 20 CONCENTRATION Part 20 TABLE 2 RATIO INVENTORY INVENTORY WATER(Ci)
AIR (Ci) CONCENTRATION TABLE 1 DISCHARGED
(!lCi/ml)
MAX. EFFLUENT COL. HlI (Ci) (Ci) (!lCi/ml)
DAC CONCENTRATION
(!lCi/ml)
(!lCi/ml)
Br 82 30 0.024 0.024 0.002 1.2E-06 2E-06 1.46E-10 6E-09 0.0244 83 105 0.083 0.083 0.008 4.2E-06 3E-05 4.95E-10 9E-08 0.0055 84 194 0.153 0.154 0.015 7.7E-06 2E-05 9.23E-10 8E-08 0.0115 '85 253 0.200 0.200 0.020 1.0E-05 1E-07 1.20E-09 1E-09 1.2038 '87 600 0.473 0.475 0.047 2.4E-05 1E-07 2.85E-09 1E-09 2.8463 TOTALBr 0.933 0.093 I '130m 200 0.158 0.158 0.016 7.9E-06 1E-07 9.45E-10 1E-09 0.9450 131 563 0.446 0.446 0.045 2.2E-05 2E-08 2.68E-09 2E-10 13.3875 132 855 0.677 0.677 0.068 3.4E-05 3E-06 4.06E-09 2E-08 0.2031 133 1282 1.015 1.015 0.102. 5.1E-05 1E-07 6.09E-09 1E-09 6.0863 134 1554 1.230 1.230 0.123 6.2E-05 2E-05 7.38E-09 6E-08 0.1230 135 1185 0.938 0.938 0.094 4.7E-05 7E-07 5.63E-09 6E-09 0.9375 '136 602 0.477 0.477 0.048 2.4E-05 1E-07 2.86E-09 1E-09 2.8575 TOTAL I 4.941 0.494 Kr 83m 105 0.084 0.084 4.2E-05 1E-02 4.95E-10 5E-05 9.9E-06 85m 253 0.200 0.200 1.0E-04 2E-05 1.20E-08 1E-07 0.1204 85 51 0.040 0.040 2.0E-05 1E-04 2.40E-09 7E-07 0.0034 87 486 0.386 0.385 1.9E-04 5E-06 2.31E-08 2E-08 1.1531 88 699 0.556 0.555 2.8E-04 2E-06 332E-08 9E-09 3.6875 '89 855 0.669 0.669 3.4E-04 1E-07 4.06E-08 1E-09 40.6125 TOTALKr 1.935 1.935 Xe 131m 5 0.004 0.004 2.0E-06 4E-04 2.36E-10 2E-06 0.0001 133m 31 0.025 0.025 1.6E-05 1E-04 1.50E-09 6E-07 0.0025 133 1282 1.015 1.015 5.1E-04 1E-04 6.09E-08 5E-07 0.1217 135m 350 0.277 0.277 1.4E-04 9E-06 1.67E-08 4E-08 0.4163 135 1243 0.984 0.984 4.9E-04 1E-05 5.91E-08 7E-08 0.8438 '137 1185 0.938 0.938 4.7E-04 lE-07 5.63E-08 lE-09 56.2500 138 894 0.707 0.707 3.5E-04 4E-06 4.24E-08 2E-08 2.1206 TOTALXe 3.950 3.950 , indicates generic <2 hr half-life values used UWNR Safety Analysis Report Rev. 2 13-4 Sept. 2008 Continuation of Table 13.1SELECTED RELEASE TOTALS Halogen Gamma Emitters 5.2 Ci Halogen Beta Emitters 5.8 Ci Total Halogens 5.87 Ci Insoluble Gamma Emitters 3.52 Ci Insoluble Beta Emitters 5.50 Ci Total Insoluble Volatiles 5.89 Ci UWNR Safety Analysis Report Rev. 0 April 2000* Continuation of Table 13.1 SELECTED RELEASE TOTALS Halogen Gamma Emitters Halogen Beta Emitters Total Halogens Insoluble Gamma Emitters Insoluble Beta Emitters Total Insoluble Volatiles 5.2 Ci 5.8 Ci 5.87 Ci 3.52 Ci 5.50 Ci 5.89 Ci UWNR Safety Analysis Report Rev. 0 *
13-5 April 2000
Release of Fission Products to Unrestricted Areas Columns H, I, and J of Table 13.1 are concerned with the exposure of personnel outside the restricted area. Calculations were performed as indicated in Appendix A. The maximumconcentrations which might be expected in unrestricted areas were calculated under theassumption that venting took place in the time required for the ventilation system to make one complete change in the laboratory (1569 seconds).
Wind velocity was assumed to be the lowest average for any month (3.54 m/s).The total dose to personnel in the unrestricted area is independent of whether the ventilation system is operating in normal mode (one exhaust fan running) or emergency venting mode (both exhaust fans running);
the concentration would be considerably higher in the case of emergency venting mode, but the period of exposure would be proportionally shorter. It is also emphasized that the total exposure figure is a maximum to be expected at any point other than within the areas evacuated in the event of an accidental release.The total of the ratios of instantaneous individual concentrations to 10 CFR Part 20 Appendix B Table 2 maximum air concentrations for discharge (sum of column J from Table 13.1) is calculated to be 134.0, where the maximum air concentration values are for unrestricted areas, 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months per week (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, 7 days per week). When averaged over a year's time ofventilation (instead of just 1569 seconds), the resulting average concentration is 0.007 of the maximum indicated by 10 CFR Part 20 for non-occupational exposure in unrestricted areas.Even with the effluent discharge from normal operation (see Section 11 .1.1) the total concentration to which personnel might be exposed is below the 10 CFR Part 20 limits.A more conservative calculation which assumes zero stack height (see Appendix A) was performed.
This analysis is applicable to a situation in which the laboratory ventilation systemfails and the release takes place through building leaks. For purposes of comparison, it was againassumed that the release occurred in the time required for the ventilation system to make an air change in the laboratory (1569 seconds).
The effect of this analysis is to multiply the values in columns H and I by a factor of 2.6, giving a resulting average concentration (yearly average) of 0.018 times 10 CFR Part 20 Appendix B Table 2 limits.Finally, an additional calculation was performed assuming 100% release of Br and I and the more conservative calculation (zero stack height) of atmospheric dilution.
The resulting summation of ratio of concentrations to release limit in this case would be 1039. Averaged over a year's time, the resulting concentration (yearly average) is 0.052 times the 10 CFR Part 20 Appendix B Table 2 limits, still within the permissible release concentration when averaged over a period of one year.As indicated in the table, releases are in all cases less than the 10 CFR Part 20 Appendix B Table 2 limits when averaged over one year. As a backup check to assure that these calculations were conservative, cases equivalent to the maximum hypothetical acident were entered into the EPA UWNR Safety Analysis Report Rev. 2Sept.'2008 13.1.1.5 Release of Fission Products to Unrestricted Areas Columns H, I, and J of Table 13.1 are concerned with the exposure of personnel outside the restricted area. Calculations were performed as indicated in Appendix A. The maximum concentrations which might be expected in unrestricted areas were calculated under the assumption that venting took place in (he time required for the ventilation system to make one complete change in the laboratory (1569 seconds).
Wind velocity was assumed to be the lowest average for any month (3.54 m/s). The total dose to personnel in the unrestricted area is independent of whether the ventilation system is operating in normal mode (one exhaust fan running) or emergency venting mode (both exhaust fans running);
the concentration would be considerably higher in the case of emergency venting mode, but the period of exposure would be proportionally shorter. It is also emphasized that the total exposure figure is a maximum to be expected at any point other than within the areas evacuated in the event of an accidental release. The total of the ratios of instantaneous individual concentrations to 10 CFR Part 20 Appendix B Table 2 maximum air concentrations for discharge (sum of column J from Table 13.1) is calculated to be 134.0, where the maximum air concentration values are for unrestricted areas, 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months per week (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day, 7 days per week). When averaged over a year's time of ventilation (instead of just 1569 seconds), the resulting average concentration is 0.007 of the maximum indicated by 10 CFR Part 20 for non-occupational exposure in unrestricted areas. Even with the effluent discharge from normal operation (see Section 11.1.1) the total concentration to which personnel might be exposed is below the 10 CFR Part 20 limits. A more conservative calculation which assumes zero stack height (see Appendix A) was performed.
This analysis is applicable to a situation in which the laboratory ventilation system fails and the release takes place through building leaks. For purposes of comparison, it was again assumed that the release occurred in the time required for the ventilation system to make an air change in the laboratory ( 1569 seconds).
The effect of this analysis is to multiply the values in columns H and I by a factor of2.6, giving a resulting average concentration (yearly average) of 0.018 times 10 CFR Part 20 Appendix B Table 2 limits. Finally, an additional calculation was performed assuming 100% release ofBr and I and the more conservative calculation (zero stack height) of atmospheric dilution.
The resulting summation of ratio of concentrations to release limit in this case would be 1039 . Averaged over a year's time, the resulting concentration (yearly average) is 0.052 times the 10 CFR Part 20 AppendixB Table 2 limits, still within the permissible release concentration when averaged over a period of one year. As indicated in the table, releases are in all cases less than the 10 CFR Part 20 Appendix B Table 2 limits when averaged over one year. As a backup check to assure that these calculations were conservative, cases equivalent to the hypothetical acident were entered into the EPA UWNR Safety Analysis Report Rev. 2 13-6 Sept. 2008 *
COMPLY program 4 at level 4. Six of the radioisotopes in Table 13.1 are not in the COMPLY listing of radioisotopes (BR-85, Br-87, 1-130m, 1-136, Kr-89, and Xe-137) and are thus not included in the COMPLY results (these all have half-lives less than 10 minutes).
Further, COMPLY does not permit zero height releases from a building, so the stack height was input as 1 meter. For the release with the ventilation system operable and the pool filled, COMPLY indicated an annually averaged dose of 1.8 mrem, with 1.4 mrem due to Iodine. For the release with pool water lost and the ventilation system inoperable (assumed stack height of 1 meter) theCOMPLY program indicated an annual dose of 14.2 mrem and 13.7 mrem from Iodine.
Table 13.2 Maximum Exposures In Unrestricted Areas from Maximum Hypothetical Accident Assumed Failures Total Body Dose Thyroid Dose Fraction of Part 20 Annual LimitsFuel clad leak with normal operation of ventilation system; pool filledFuel clad leak with failure of ventilation system; pool filled Fuel clad leak with failure of ventilation system and concurrent loss of pool water 0.006 rem 0.0 10 rad 0.007 0.084 rem 0.102 rad 0.018 0.153 rem rad 0.052 UWNR Safety Analysis Report Rev. 2Sept. * *
COMPLY program 4 at level 4. Six of the radioisotopes in Table 13.1 are not in the COMPLY listing of radioisotopes (BR-85, Br-87, I-130m, 1-136, Kr-89, and Xe-137) and are thus not included in the COMPLY results (these all have half-lives less than 10 minutes).
Further, COMPL Y does not permit zero height releases from a building, so the stack height was input as 1 meter. For the release with the ventilation system operable and the pool filled, COMPLY indicated an annually averaged dose of 1.8 mrem, with 1.4 mrem due to Iodine. For the release with pool water lost and the ventilation system inoperable (assumed stack height of 1 meter) the COMPLY program indicated an annual dose of 14.2 mrem and 13.7 mrem from Iodine. Table 13.2 Maximum Exposures In Unrestricted Areas from Maximum Hypothetical Accident Assumed Failures Total Body Thyroid Fraction of Dose Dose Part 20 Annual Limits Fuel clad leak with normal operation of 0.006 rem 0.010 rad 0.007 ventilation system; pool filled Fuel clad leak with failure of ventilation 0.084 rem 0.102 rad 0.018 system; pool filled Fuel clad leak with failure of ventilation 0.153 rem 1.019 rad 0.052 system and concurrent loss of pool water UWNR Safety Analysis Report Rev. 2 13-7 Sept. 2008
Insertion of Excess Reactivity The worst case result of insertion of excess reactivity would be insertion of the maximum allowed experiment reactivity worth or ejection of the transient rod (1.4% AK/K) while the reactor is operating at maximum steady-state power.Calculations 5 performed by Gulf General Atomic indicate that a peak temperature of 1150 inFLIP fuel will not produce a stress in the fuel clad in excess of the ultimate yield strength.Further, TRIGA fuel with a H/Zr ratio of at least 1.65 has been pulsed to temperatures of about 1150 without any damage to the clad 6.In a mixed FLIP-Standard TRIGA core the peak temperatures in FLIP fuel are much higher than in standard fuel due to the peaking of the power distribution near water gaps. For this reason the subsequent analysis in this section is concerned with internal temperatures in FLIP fuel elements.A worst case core arrangement is considered, in which a FLIP element is located adjacent to a 3-inch square water gap. The power density in the FLIP element is at the maximum permissible value based on consideration of the loss of coolant accident (23 KW when the core is operating at 1 MW). The core is operating at the power level scram point of 1.25 MW, and the transient control rod is fired to initiate a pulse.Pulses of 2.1% AK/K fired in standard TRIGA at this facility have had energy releases of less than 20 MW seconds.
FLIP and mixed cores have been operated with maximum reactivity insertions for pulses reduced to 1.4%AK/K because of the shorter prompt neutron lifetime in FLIP fueled cores. Typical 1.4% reactivity pulses in an all-FLIP core have energy releases of only about 14 MW seconds, while mixed cores have slightly lower releases.
Computations will be done for 20 MW second release.The limitation of experiment reactivity to 1.4% AK/K will insure that reactivity insertions from experiment removal or failure will insure that such an accident will result in consequences no worse than those considered here.Firing the transient rod while at full power is prevented by interlocks and administrative requirements.
Removal of an experiment while operating at full power would not result in a reactivity insertion rate as large as that resulting from firing the transient rod, and the most likely result of experiment removal under the conditions assumed would be a reactor scram from power level, and fuel temperature trips. Further, experiments having worths approaching 1.4% A K/K are fastened to prevent inadvertent removal, and administrative restrictions do not allow such manipulations while the reactor is in operation.
The predicted conditions establish an upper limit for a reactivity accident.UWNR Safety Analysis Report Rev. 2Sept. 13.1.2 Insertion of Excess Reactivity The worst case result of insertion of excess reactivity would be insertion of the maximum allowed experiment reactivity worth or ejection of the transient rod (1.4% LlKlK) while the reactor is operating at maximum steady-state power. Calculations 5 performed' by Gulf General Atomic indicate that a peak temperature of 1150 °C in FLIP fuel will not produce a stress in the fuel clad in excess of the ultimate yield strength.
Further, TRIGA fuel with a H/Zr ratio of at least 1.65 has been pulsed to temperatures of about 1150 °C without any damage to the clad 6* In a mixed FLIP-Standard TRIGA core the peak temperatures in FLIP fuel are much higher than in standard fuel due to the peaking of the power distribution near water gaps. For this reason the subsequent analysis in this section is concerned with internal temperatures in FLIP fuel elements.
A worst case core arrangement is considered, in which a FLIP element is located adjacent to a 3-inch square water gap. The power density in the FLIP element is at the maximum permissible value based on consideration of the loss of coolant accident (23 KW when the core is operating at 1 MW). The core is operating at the power level scram point of 1.25 MW, and the transient control rod is fired to initiate a pulse. Pulses of 2.1 % LlKlK fired in standard TRIGA at this facility have had energy releases of less
than 20 MW seconds. FLIP and mixed cores have been operated with maximum reactivity
insertions for pulses reduced to 1.4%LlKlK because of the shorter prompt neutron lifetime in FLIP fueled cores. Typical 1.4% reactivity pulses in an all-FLIP core have energy releases of only about 14 MW secohds, while mixed cores have slightly lower releases.
Computations will be done for 20 MW second release. The limitation of experiment reactivity to 1.4% LlKlK will insure that reactivity insertions from experiment removal or failure will insure that such an accident will result in consequences no worse than those considered here. Firing the transient rod while at full power is prevented by interlocks and administrative requirements.
Removal of an experiment while operating at full power would not result in a reactivity insertion rate as large as that resulting from firing the transient rod, and the most likely result of experiment removal under the conditions assumed would be a reactor scram from power level, and fuel temperature trips. Further, experiments having worths approaching 1.4% Ll KlK are fastened to prevent inadvertent removal, and administrative restrictions do not allow such manipulations while the reactor is in operation.
The predicted conditions establish an upper limit for a reactivity accident.
UWNR Safety Analysis Report Rev. 2 13-8 Sept. 2008
Fuel Temperatures from Operation at the Scram Point Calculations for the SAR 7 of the Puerto Rico Reactor resulted in the information presented in the lower curve in Figure This curve shows the fuel temperature distribution at the axial centerline in a FLIP fuel element operating at conditions of slightly higher power density than that assumed here. The Puerto Rico case is an element operating at a power density in the maximum element of 1.4 times the average of 22.3 KW/element.
The axial peaking factor is 1.3. Calculations done for UWNR considered the case of an element operating at 23 KW timesthe ratio of 1.25 of scram setting/licensed power level, with the same axial peaking factor of 1.3.Using these numbers, the fuel centerline and average temperatures will be lower in the UWNR core, but the temperature at the outer surface of the fuel would be approximately the same in both cases.F,,.
/A/
OVS/rY 5'Ec. V030 50 t t P I 70Figure Fuel Temperature Distribution in a Fuel Element UWNR Safety Analysis Report Rev. 2Sept. * *
13 .1.2.l Fuel Temperatures from Operation at the Scram Point Calculations for the SAR 7 of the Puerto Rico Reactor resulted in the information presented in the lower curve in Figure 13-1 . This curve shows the fuel temperature distribution at the axial centerline in a FLIP fuel element operating at conditions of slightly higher power density than that assumed here. The Puerto Rico case is an element operating at a power density in the maximum element of 1.4 times the average of22.3 KW/element.
The axial peaking factor is 1.3. Calculations done for UWNR considered the case of an element operating at 23 KW times the ratio of 1.25 of scram setting/licensed power level, with the same axial peaking factor of 1.3. Using these numbers, the fuel centerline and average temperatures will be lower in the UWNR core, but the temperature at the outer surface of the fuel would be approximately the same in both cases. 200 f-I FtlEL TEMP. 11/ MAX. POWER !)fNSITY fLtMF.NT P[l=o)RE . .f AFTER ,4 20 Mill stc. ?tJLSE FRJM /.25' Mw AFT[R PI)LS[ EfFOR[ 1'/)1..5£
/
__ __ __ __ __ '0 10 ZJ 30 40 50 GO 70 80 90 100 10 OF OtJTS//)E" FVEL RAOIIJS Figure 13-1 . Fuel Temperature Distribution in a Fuel Element UWNR Safety Analysis Report Rev. 2 13-9 Sept. 2008 13.1.2.2 Temperature after Pulse Firing a pulse while at the scram point would cause the reactor to scram from power level and fuel temperature scrams. The entire pulse energy release is used, however, in the following analysis.The temperature distribution in the fuel element immediately after a 20 MW second pulse is plotted as the top curve in Figure The peaking factor within a FLIP element adjacent to a 3-inch square water-gap is 2.49, and an axial peaking factor of 1.3 is used as in the steady state conditions.
The energy deposited in the element under consideration is calculated using the same peaking factor (power in maximum element/ power in average element in core) which resulted in the 23 KW steady state level.The maximum adiabatic temperature reached in the element will occur at the outer surface of the fuel element adjacent to the water-gap.
This maximum temperature would be 1133 C, slightly below the safety limit of 1150 C.Although such an event is considered highly unlikely, it would not cause fuel damage or release of fission products from the reactor.After these computations were completed, NRC requested that the accident be re-evaluated for the permitted cores under the technical specifications that were proposed for UWNR. The major changes were limiting the minimum FLIP content to be 9 fuel bundles (35 elements, since the transient rod is located within the 9 central fuel bundles in the UWNR core).The re-analysis was based on a lower power in maximally exposed fuel element (18.1 kW instead of 22.3 kW) and a limitation of the reactivity insertion from AK/K to 1.4% AK/K.First, the temperature of the fuel in the maximum element will be lower at the beginning of the pulse by about 70'C. Second, the use of a compact array of nine (9) FLIP bundles reduces the possible peaking factor within a FLIP element from the 2.49 value used in the original calculation to a value of 2.03 for a FLIP element beside the transient rod guide tube (this is the position with highest power density in the core.) Finally, reduction of allowable pulsed reactivity insertion from 2.1% AK/K to 1.4% AK/K will substantially reduce the energy generation in a pulse, while the limitation of experiment worth to 1.4% AK/K will provide similar safeguards for experiment failure or removal. Measurements performed on the Puerto Rico Nuclear Center, TRIGA-FLIP reactor indicated that a pulse insertion of 1.4% AK/K resulted in a maximum fuel temperature rise of approximately 400 and measurements at Wisconsin confirmed that prediction.
Consideration of all these differences shows a peak fuel temperature of about 450 lower thanthat indicated above. It is therefore concluded that fuel damage would occur in neither case, but with a much larger safety margin in the more restrictive case considered here.UWNR Safety Analysis Report Rev. 2 Sept. 13.1.2.2 Temperature after Pulse Firing a pulse while at the scram point would cause the reactor to scram from power level and fuel temperature scrams. The entire pulse energy release is used, however, in the following analysis.
The temperature distribution in the fuel element immediately after a 20 MW second pulse is plotted as the top curve in Figure 13-1. The peaking factor within a FLIP element adjacent to a 3-inch square water-gap is 2.49, and an axial peaking factor of 1.3 is used as in the steady state conditions.
The energy deposited in the element under consideration is calculated using the same peaking factor (power in maximum element/ power in average element in core) which resulted in the 23 KW steady state level. The maximum adiabatic temperature reached in the element will occur at the outer surface of the fuel element adjacent to the water-gap.
This maximum temperature would be 1133 C, slightly below the safety limit of 1150 C. Although such an event is considered highly unlikely, it would not cause fuel damage or release of fission products from the reactor. After these computations were completed, NRC requested that the accident be re-evaluated for
the permitted cores under the technical specifications that were proposed for UWNR. The major
changes were limiting the minimum FLIP content to be 9 fuel bundles (35 elements, since the transient rod is located within the 9 central fuel bundles in the UWNR core). The re-analysis was based on a lower power in maximally exposed fuel element (18.1 kW instead of22.3 kW) and a limitation of the reactivity insertion from 2.1%
to 1.4%
First, the temperature of the fuel in the maximum element will be lower at the beginning of the ) . pulse by about 70°C. Second, the use of a compact array of nme (9) FLIP bundles reduces the possible peaking factor within a FLIP element from the 2.49 value used in the original calculation to a value of2.03 for a FLIP element beside the transient rod guide tube (this is the position with highest power density in the core.) Finally, reduction of allowable pulsed reactivity insertion from 2.1 %
to 1.4%
will substantially reduce the energy generation in a pulse, while the limitation of experiment worth to 1.4%
will provide similar safeguards for experiment failure or removal. Measurements performed on the Puerto Rico Nuclear Center TRIGA-FLIP reactor indicated that a pulse insertion of 1.4% t.KlK resulted in a maximum fuel temperature rise of approximately 400°C 8, and measurements at Wisconsin confirmed that prediction.
Consideration of all these differences shows a peak fuel temperature of about 450 °C lower than that indicated above. It is therefore concluded that fuel damage would occur in neither case, but with a much larger safety margin in the more restrictive case considered here. UWNR Safety Analysis Report Rev. 2 13-10 Sept. 2008
13.1.3 Loss of Coolant Although there is little likelihood of complete loss of water from the reactor pool, an analysis is made to demonstrate that such loss will not damage reactor fuel.13.1.3.1 Possible Means of Water Loss The pool is contained within the thick reinforced concrete reactor shield which will maintain its integrity under the most severe earthquake that would be expected in this area.The only credible scenario for draining the pool would be a sheared and open beam port. For analysis in the next section, the time to drain the pool is estimated.
The pool water level is routinely maintained at least 20.875ft above the top of the core (this is approximately the low pool level alarm point) which corresponds to 21.5ft above the fuel center. However, the limiting assumption is made that the water level is only 19ft above the top of the core, or 19.625ft above the fuel center. The pool has a surface area of 89.13ft 2.The inner beam port diameter is 0.5ft, and the beam ports are at core center. Using these values, the following equation 9 can be used to estimate the time required to drain the pool.where: td time to drain pool to height h (sec)g = acceleration due to gravity (32.174ft/s 2)cross-sectional area of pool surface (ft 2)Ao = cross-sectional area of drain opening (ft 2)Cd = discharge coefficient (0.6)h= initial height of water above drain opening h = final height of water above drain opening (ft)The calculated drain time is 836s. A sheared and open beam port could drain the water level to mid-core height, but water would still be in contact with the fuel and would prevent excessive temperatures.
The 8-inch stainless steel pipes built into the pool walls for possible future use in a forcedconvection cooling system are flange sealed on the outer ends. In addition, one of these pipes has a loop and a siphon breaker extending well above the core so that a rupture cannot lower pool level below the core. The other pipe is flange sealed inside the pool and penetrates the shield wall well above the core.
Rupture of either of these lines will not uncover the core.Rupture of the piping in the demineralizer could cause only slight water loss due to location of the outlet lines from the pool and a check valve at the demineralizer outlet.UWNR Safety Analysis Report Rev. 2 Sept. * *
13.1.3 Loss of Coolant Although there is little likelihood of loss of water from the reactor pool, an analysis is made to demonstrate that such loss will not damage reactor fuel. 13.1.3.1 Possible Means of Water Loss The pool contained within the thick reinforced concrete reactor shield which will maintain its integrity under the most severe earthquake that would be expected in this area. The only credible scenario for draining the pool would be a sheared and open beam port. For analysis in the next section, the time to drain the pool is estimated.
The pool water level is routinely maintained at least 20.875ft above the top of the core (this is approximately the low pool level alarm point) which corresponds to 21.5ft above the fuel center. However, the limiting assumption is made that the water level is only 19ft above the top of the core, or 19.625ft above the fuel center. The pool has a surface area of 89. 13ft2. The inner beam port diameter is 0.5ft, and the beam ports are at core center. Using these values, the following equation 9 can be used to estimate the time required to drain the pool. where: td = time to drain pool to height h (sec) g = acceleration due to gravity (32. 1 74ft1s 2) Ap = cross-sectional area of pool surface (fe) Ao = cross-sectional area of drain opening (ft2) Cd = discharge coefficient (0.6) ho = initial height of water above drain opening (ft) h = final height of water above drain opening (ft) The calculated drain time is 836s. A sheared and open beam port could drain the water level to mid-core height, but water would still be in contact with the fuel and would prevent excessive temperatures.
The 8-inch stainless steel pipes built into the pool walls for possible future use in a forced convection cooling system are flange sealed on the outer ends. In addition, one of these pipes has a loop and a siphon breaker extending well above the core so that a rupture cannot lower pool level below the core. The other pipe is flange sealed inside the pool and penetrates the shield wall well above the core. Rupture of either of these lines will not uncover the core. Rupture of the piping in the dernineralizer could cause only slight water loss due to location of the outlet lines from the pool and a check valve at the demineralizer outlet. UWNR Safety Analysis Report Rev. 2 13-11 Sept. 2008
Radiation Levels in Confinement Due to Unshielded Core Calculations of radiation levels at various points in the Reactor Laboratory were made assumingcontinuous operations at 1.02MW. The fission product inventory was determined by the use of the ORIGEN2 computer code version 2.1, using the PWRUS cross-section library. A single fuel pin was simulated at the power level of the hottest rod, with an assumed exposure of 500MWd.The resulting gamma source term was multiplied by the number of fuel elements in the core and then divided by the pin power peaking factor to calculate the core gamma source term at various decay times. The decay time of 836s represents the time required to drain the pool (see section 13.1.3.1).
Dose rates from direct and scattered radiation were modeled using the MCNP5 computer code. Results of the calculations are given in Table 13.3.Table 13.3Calculated Radiation Dose Rates in Confinement After Pool Water is Lost Time After Beam Port Console Pool Curb Pool TopShutdown Floor Over Core Behind Curb(R/hr) (R/hr) (R/hr) (R/hr) 836 seconds 4.89 6.72 7,430 28.1 1 day 1.23 1.69 1,720 7.07 1 week 0.682 0.938 992 3.92 1 month 0.353 0.485 492 2.02 These levels are not too high to allow emergency repairs to be made. Facility emergencyprocedures cover the situation of pool water loss.13.1.3.3 Radiation Levels in Unrestricted Areas Due to Unshielded Core The calculations from the previous section were also made for the floor non-restricted classroom to the west of the reactor. This classroom would be subject to the highest dose rate field of any non-restricted area due to the elevation above the biological shield and its line-of-sight with the reactor core.UWNR Safety Analysis Report Rev. 2 Sept. 13.1.3.2 Radiation Levels in Confinement Due to Unshielded Core Calculations of radiation levels 'at various points in the Reactor Laboratory were made assuming I continuous operations at 1.02MW. The fission product inventory was determined by the use of I the ORIGEN2 computer code version 2.1, using the PWRUS cross-section library. A single fuel I pin was simulated at the power level of the hottest rod, with an assumed exposure of 500MWd. I The resulting gamma source term was multiplied by the number of fuel elements in the core and I then divided by the pin power peaking factor to calculate the core gamma source term at various I decay times. The decay time of 836s represents the time required to drain the pool (see section I 13.1.3.l).
Dose rates from direct and scattered radiation were modeled using the MCNP5 I computer code. Results of the calculations are given in Table 13.3. Table 13.3 Calculated Radiation Dose Rates in Confinement After Pool Water is Lost Time After Beam Port Console Pool Curb Pool Top Shutdown Floor Over Core Behind Curb (Rlhr) (Rlhr) (R/hr) . (Rlhr) 836 seconds 4.89 6.72 7,430 28.1 1 day 1.23 1.69 1,720 7.07 1 week 0.682 0.938 992 3.92 1 month 0.353 0.485 492 2.02 These levels are not too high to allow emergency repairs to be made. Facility emergency procedures cover the situation of pool water loss. 13.1.3.3 Radiation Levels in Unrestricted Areas Due to Unshielded Core The calculations from the previous section were also made for the 3 rd floor non-restricted classroom to the west of the reactor. This classroom would be subject to the highest dose rate field of any non-restricted area due to the elevation above the biological shield and its sight with the reactor core. UWNR Safety Analysis Report Rev. 2 13-12 Sept. 2008 * *
Table 13.4Calculated Radiation Dose Rates in Non-Restricted Area After Pool Water is Lost Time After Shutdown 3 rd Floor Classroom (R/hr)836 seconds 4.14 1 day 0.764 1 week 0.482 1 month 0.210 The calculated dose rate to the 3 rd floor non-restricted classroom is significant, but in the event of a loss of coolant accident the building evacuation alarm would alert people to evacuate these classrooms before the core was completely uncovered.
In order to estimate the integrated dose received by a member of the public during the evacuation, the MCNP5 model of the unshielded core was modified to include partial water shielding at several time steps. The core gamma source term was also modified to simulate an appropriate level of decay from full power. The integrated dose to the 3 rd floor classroom was calculated at various times during the pool waterloss and is shown in Figure 200 150 100 50 Core uncovered 5 minute building evacuation complete Bridge ARM initiatesevacuation alarm0 Figure 0 100 200 300 400 500 600 700 800 Time Since Start of 3 rd Floor Integrated Dose vs. Time during LOCA UWNR Safety Analysis Report Rev. 2 Sept. * *
Table 13.4 Calculated Radiation Dose Rates in Non-Restricted Area After Pool Water is Lost Time After 3'd Floor Classroom Shutdown (Rlhr) 836 seconds 4.14 1 day 0.764 1 week 0.482 1 month 0.210 The calculated dose rate to the 3 rd floor non-restricted classroom is significant, but in the event of a loss of coolant accident the building evacuation alarm would alert people to evacuate these classrooms before the core was completely uncovered.
In order to estimate the integrated dose received by a member of the public during the evacuation, the MCNP5 model of the unshielded core was modified to include partial water shielding at several time steps. The core gamma source term was also modified to simulate an appropriate level of decay from full power. The integrated dose to the 3 rd floor classroom was calculated at various times during the pool water loss and is shown in Figure 13-2. 250 200 E (I) ... .§. ------, Core i uncovered 7 \ I / 150 (I) I/) 0 0 "0 (I) 100 5 minute building / evacuation complete -cu ... CI (I) -c: / Bridge ARM initiates evacuation alarm 50 I V o o 100 200 300 400 500 600 Time Since Start of LOCA (sec) Figure 13-2 3 rd Floor Integrated Dose vs. Time during LOCA UWNR Safety Analysis Report Rev. 2 13-13 700 I ! , i , ! I i I i 800 900 Sept. 2008 A 5 minute evacuation time from the sounding of the evacuation alarm is assumed. If the initiation of the pool water loss was observed then the building evacuation would be manually activated at the start of the LOCA, but if the accident was unobserved then the pool water would drain to approximately 7.4 ft above the core in 300 s before tripping the bridge area radiation monitor, which would in turn automatically initiate the building evacuation alarm. Therefore, the hypothetical member of the public that remains in the floor classroom for 5 minutes following the automatic initiation of the building evacuation alarm (300 s after start of the LOCA) wouldreceive an integrated dose of about 13 mrem. This is less than the 100mrem limit (10 CFR 20.1301).
Realistic doses would be far less than this, because the preceding analysis does nottake into account time spent in hallways and stairwells (where the dose rate is much lower)during the evacuation.
Because the time spent in the high dose rate field in the floor classroom would be far less than 5 minutes, the integrated dose would be substantially lower dueto the majority of the dose being received in the final minute as shown in Figure 13.1.3.4 Fuel Temperature After Loss of Pool Water Calculations performed at Texas A & M University have treated the loss of coolant accident in detail, based on reactor shutdown 15 minutes before the core is uncovered. At Wisconsin, the pool level scram would cause automatic shutdown much sooner, as the A & M calculation is based on pool drainage by rupture of a 10-inch line. Other pertinent parameters of the two facilities are identical.
The calculations employed the Gulf computer code TAC for calculation of system temperatures.
The results of these calculations (Pages 25-31 of Texas A & M University Nuclear Science Center Amendment II to the Safety Analysis Report, November 1, 1972 submitted under Docket for License R-83) indicate that for a maximum power density of less than 21 kW/element for standard fuel and 23 kW/element for FLIP fuel, loss of coolant water would not result in fuel clad failure and release of fission products.13.1.4 Loss of Coolant Flow Not applicable; natural convection cooling 13.1.5 Mishandling or Malfunction of Fuel Reference 1 states that this condition produces the maximum consequence to the public. This accident is therefore included as the maximum hypothetical accident (Section 13.1/2.1), when combined with failure of the ventilation system and loss of pool water. The effect of a fuel clad failure with normal pool level and with the ventilation system operating normally has no significant effect on the public.UWNR Safety Analysis Report Rev. 2 13-14 Sept. A 5 minute evacuation time from the sounding of the evacuation alarm is assumed. If the initiation of the pool water loss was observed then the building evacuation would be manually activated at the start of the LOCA, but if the accident was unobserved then the pool water would drain to approximately 7.4 ft above the core in 300 s before tripping the bridge area radiation monitor, which would in tum automatically initiate the building evacuation alarm. Therefore, the hypothetical member ofthe public that remains in the yd floor classroom for 5 minutes following the automatic initiation of the building evacuation alarm (300 s after start of the LOCA) would receive an integrated dose of about 13 mrem. This is less than the 100mrem limit (10 CFR 20.1301).
Realistic doses would be far less than this, because the preceding analysis does not take into account time spent in hallways and stairwells (where the dose rate is much lower) during the evacuation.
Because the time spent in the high dose rate field in the yd floor classroom would be far less than 5 minutes, the integrated dose would be substantially lower due to the majority of the dose being received in the final minute as shown in Figure 13-2. 13.1.3.4 Fuel Temperature After Loss of Pool Water Calculations performed at Texas A & M University have treated the loss of coolant accident in detail, based on reactor shutdown 15 minutes before the core is uncovered.
At Wisconsin, the pool level scram would cause automatic shutdown much sooner, as the A & M calculation is based 'on pool drainage by rupture of a 10-inch line. Other pertinent parameters of the two facilities are identical.
The calculations employed the Gulf computer code TAC for calculation of system temperatures.
The results of these calculations (Pages 25-31 of Texas A & M University Nuclear Science Center Amendment II to the Safety Analysis Report, November 1, 1972 submitted under Docket for License R-83) indicate that for a maximum power density ofless than 21 kWlel,ement for standard fuel and 23 k W lelement for FLIP fuel, loss of coolant water would not result in fuel clad failure and release of fission products.
13.1.4 Loss of Coolant Flow Not applicable; natural convection cooling 13.1.5 Mishandling or Malfunction of Fuel Reference 1 states that this condition produces the maximum consequence to the public. This accident is therefore included as the maximum hypothetical accident (Section 13.1/2.1), when combined with failure of the ventilation system and loss of pool water. The effect of a fuel clad failure with normal pool level and with the ventilation system operating normally has no significant effect on the public. UWNR Safety Analysis Report Rev. 2 13-14 Sept. 2008 * *
Experiment Malfunction Experiment reactivity worth and composition are controlled and limited so that experiment failure will not insert a step change in reactivity greater than 1.4% AK/K (fixed experiments; movable experiments are limited to AK/K. Procedure for experiment review includes consideration of chemical and explosive hazards to the reactor. Any experiment containingfissionable material is limited so that production of gaseous and volatile fission products results in releases lower than that considered in section Therefore, experiment malfunction will not result in consequences more severe than those listed in other parts of this chapter.
Loss of Normal Electrical Power Loss of normal electrical power will cause the reactor to shut down. It will not result in any release of radioactive material or increase the dose to the population.
Emergency core cooling engineered safety systems are not required.
The maximum hypothetical accident analysis does include loss of the ventilation system in the analysis, thus effectively including loss of electrical power.External Events Since the safety of a TRIGA reactor is so strongly a function of the fuel composition and characteristics, none of the usual external event initiators will cause any effect on the public. As stated in chapter 2 of this report, floods and hurricanes are an insignificant threat to the safety of the reactor (Section 2.4), with the year flood causing nearby Lake Mendota to expand only feet, not threatening the laboratory.
Further, should the water level within the room rise above ground level it would not affect the safety of the reactor. Tornados do occur in the Midwest, but damage to the concrete shield which protects the reactor core is not credible.
The seismicity of the area is extremely low, with the estimated year peak ground acceleration to a seismic event is less than (Section 2.5.4). Since no engineered structures other than the reactor shield are required to provide protection to the reactor, such an acceleration will have no effect on the reactor. Likewise, though aircraft collisions with the building are not impossible, they are unlikely due to the location of flight paths. Further, such impacts will not breach the concrete shield at core level. Impacting the outer walls of the building will not result in radiationbeing released.
UWNR Safety Analysis Report Rev. 2 1-5Sp.20 Sept. * *
13.1.6 Experiment Malfunction Experiment reactivity worth and composition are controlled and limited so that experiment failure will not insert a step change in reactivity greater than 1.4% LlKlK (fixed experiments; movable experiments are limited to 0.7% LlKlK. Procedure for experiment review includes consideration of chemical and explosive hazards to the reactor. Any experiment containing fissionable material is limited so that production of gaseous and volatile fission products results in releases lower than that considered in section 13.1.1. Therefore, experiment malfunction will not result in consequences more severe than those listed in other parts of this chapter. 13.1.7 Loss of Normal Electrical Power Loss of normal electrical power will cause the reactor to shut down. It will not result in any release of radioactive material or increase the dose to the population.
Emergency core cooling engineered safety systems are not required.
The maximum hypothetical accident analysis does include loss of the ventilation system in the analysis, thus effectively including loss of electrical power. 13.1.8 External Events Since the safety of a TRIGA reactor is so strongly a function of the fuel composition and characteristics, none of the usual external event initiators will cause any effect on the public. As stated in chapter 2 of this report, floods and hurricanes are an insignificant threat to the safety of the reactor (Section 2.4), with the 100 year flood causing nearby Lake Mendota to expand by only 30 feet, not threatening the laboratory.
Further, should the water level within the room rise above ground level it would not affect the safety of the reactor. Tornados do occur in the Midwest, but damage to the concrete shield which protects the reactor core is not credible.
The seismicity of the area is extremely low, with the estimated 50 year peak ground acceleration to a seismic event is less than 0.01 g (Section 2.5.4). Since no engineered structures other than the reactor shield are required to provide protection to the reactor, such an acceleration will have no effect on the reactor. Likewise, though aircraft collisions with the building are not impossible, they are unlikely due to the location of flight paths. Further, such impacts will not breach the concrete shield at core level. Impacting the outer walls of the building will not result in radiation being released . UWNR Safety Analysis Report Rev. 2 13-15 Sept. 2008 13.1.9 Mishandling or Malfunction of Equipment An analysis was made of the possibility that loss of water from the reactor or from the radioactive liquid waste storage tanks could affect the city water supply, and a negative result was obtained as indicated below. Madison obtains its drinking water supply from several wells drilled into the Cambrian sandstone described above. The location of these wells is shown on Figure 2-12, and they supply the University as well as the city. All of these wells are cased from ground level into the sandstone so as to keep out water from the glacial deposit. The closest well to the reactor site is about 2,000 feet southeast.
The Reactor Laboratory floor drain empties into the hold tank. Should the entire contents of the pool be let out into the room, however, some water could escape into the sewer system through a drain thimble into which waste water is pumped from the hold tank. There are four methods by which water may leave the reactor room: (1) by flow pumped from the radioactive waste storage tanks through the elevated drain thimble provided for emptying the tanks;(2) directly into the drain thimble should the pool be completely ruptured, thus reaching a level high enough to overflow into the thimbles or escape from the laboratory and enter floor drains in surrounding areas;(3) by loss through the floor and into the ground; and (4) by rupture of the liquid radioactive waste storage tank directly into the soil underthe laboratory floor.Analysis for cases (1) and (2)In so far as the first two discharge paths are concerned, the flow through the drain thimble or floor drains empties into a sanitary sewer main. From there it would travel through mains via a pumping station to the main sewage plant, located south of and outside the corporate limits of the city.
From there, the sewage travels through mains an additional five miles to the south before it empties into an open ditch. On the way, any UWNR Safety Analysis Report Rev. 2 Sept. 13.1.9 Mishandling or Malfunction of Equipment An analysis was made of the possibility that loss of water from the reactor or from the radioactive liquid waste storage tanks could affect the city water supply, and a negative result was obtained as indicated below. Madison obtains its drinking water supply from several wells drilled into the Cambrian sandstone described above. The location of these wells is shown on Figure 2-12, and they supply the University as well as the city. All of these wells are cased from . ground level into the sandstone so as to keep out water from the glacial deposit. The closest well to the reactor site is about 2,000 feet southeast.
The Reactor Laboratory floor drain empties into the hold tank. Should the entire contents of the e pool be let out into the room, however, some water could escape into the sewer system through a e. drain thimble into which waste water is pumped from the hold tank. There are four methods by which water may leave the reactor room: (1 ) by flow pumped from the radioactive waste storage tanks through the elevated drain thimble provided for emptying the tanks; (2) directly into the drain thimble should the pool be completely ruptured, thus reaching a level high enough to overflow into the thimbles or escape from the laboratory and enter floor drains in surrounding areas; (3) by loss through the floor and into the ground; and (4) by rupture of the liquid radioactive waste storage tank directly into the soil under the laboratory floor. Analysis for cases (1) and (2) In so far as the first two discharge paths are concerned, the flow through the drain thimble or floor drains empties into a sanitary sewer main. From there it would travel through mains via a pumping station to the main sewage plant, located south of and outside the corporate limits of the city. From there, the sewage travels through mains an additional five miles to the south before it empties into an open ditch. On the way, any UWNR Safety Analysis Report Rev. 2 13-16 Sept. 2008 e water from the reactor would become considerably diluted since the minimum flow-rate into the ditch is 7,000 gpm whereas the probable maximum rate of entry into the floor drain would not be more than 10 to 100 gpm. These facts, coupled with the fact thatstringent administrative precautions will be taken to ensure that water contaminated beyond established tolerance levels is not released to the drain, tend to preclude that the city water supply could be adversely affected by this method.Analysis for case (3)The possibility that the city water supply could be affected via the third method (3) is also negligible. The base of the reactor is about 8 feet below ground level, and water cannot be dissipated via surface run-off. Since the walls of the building surrounding the reactor are made of concrete up to the ground level, significant water loss through the floor could result only if the concrete was breached.
In fact, it would appear that the only mechanism by which contaminated water could enter the soil would be the result of an earthquake sufficiently severe to rupture both the reactor tank and shield, as well as the floor of the building, at a time when the reactor pool water was radioactive beyond tolerance levels.Such a set of coincidental occurrences is considered extremely remote. Further, even if it did occur, there is no assurance that the water supply would be adversely affected.
For example, the nearest city well is about 2,000 feet from the reactor site, and it has been estimated by a ground water specialist that water would flow through the sandstone from the reactor to the well at not more than 0.1 foot per day. Thus, as long as 55 years might be required for the reactor water to reach the well.Analysis for case (4)Should the radioactive waste storage tanks rupture, a similar analysis to that in case (3)indicates no adverse effect on the well. Furthermore, the quantities of water likely to be lost are small and activities are expected to be low enough that no hazard exists.Summary and Conclusions None of the accidents considered here will result in consequences to the public health and safety.Even the maximum hypothetical accident does not result in releases of radioactivity in excess of 10 CFR Part 20 limits when averaged over a year.
References
1. NUREG- 1537 Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, USNRC, February 1996.2. Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors, NUREG/CR-2387, PNL-4028, April 1982.UWNR Safety Analysis Report Rev.
2 Sept. * *
water from the reactor would become considerably diluted since the minimum flow-rate into the ditch is 7,000 gpm whereas the probable maximum rate of entry into the floor drain would not be more than 10 to 100 gpm. These facts, coupled with the fact that stringent administrative precautions will be taken to ensure that water contaminated beyond established tolerance levels is not released to the drain, tend to preclude that the city water supply could be adversely affected by this method. Analysis for case (3) The possibility that the city water supply could be affected via the third method (3) is also negligible.
The base of the reactor is about 8 feet below ground level, and water cannot be dissipated via surface run-off. Since the walls ofthe building surrounding the reactor are made of concrete up to the ground level, significant water loss through the floor could result only if the concrete was breached.
In fact, it would appear that the only mechanism by which contaminated water could enter the soil would be the result of an earthquake sufficiently severe to rupture both the reactor tank and shield, as well as the floor of the building, at a time when the reactor pool water was radioactive beyond tolerance levels. Such a set of coincidental occurrences is considered extremely remote. Further, even if it did occur, there is no assurance that the water supply would be adversely affected.
For example, the nearest city well is about 2,000 feet from the reactor site, and it has been estimated by a ground water specialist that water would flow through the sandstone from the reactor to the well at not more than 0.1 foot per day. Thus, as long as 55 years might be required for the reactor water to reach the well. Analysis for case (4) Should the radioactive waste storage tanks rupture, a similar analysis to that in case (3) indicates no adverse effect on the well. Furthermore, the quantities of water likely to be lost are small and activities are expected to be low enough that no hazard exists. 13.2 Summary and Conclusions None of the accidents considered here will result in consequences to the public health and safety. Even the maximum hypothetical accident does not result in releases of radioactivity in excess of 10 CFR Part 20 limits when averaged over a year. 13.3 References
1. NUREG-1537 Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, USNRC, February 1996. 2. Credible Accident Analyses for TRIGA and TRIGA-fueled Reactors, NUREG/CR-2387, PNL-4028, April 1982 . UWNR Safety Analysis Report Rev. 2 13-17 Sept. 2008
3. J. F. Perkins and R. W. King, "Energy Release from the Decay of Fission Products", Nuclear Science and Engineering, 3, 726 (1958).4. COMPLY V1.5d, U.S. Environmental Protection Agency, Office of Radiation and Indoor air, Washington DC 20460, October 1989."Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor", GA-9064, Gulf General Atomic, Jan. 5, 1970.6. "Annular Core Pulse Reactor", General Dynamics, General Atomic Division Report GACD 6977, Supplement 2, 9/30/66.7. Safeguards Summary Report for the TRIGA-FLIP Reactor at Puerto Rico Nuclear Center, Report PRNC 123, Revision C, November 11, 1969.8. Docket 50-120, Change No. 11 to the Technical Specifications Facility License R-83, Texas A & M University, Section 3.2 Basis.9. Hunsaker and Rightmire, "Engineering Applications of Fluid Mechanics," McGraw-Hill Book Company, 1947 (pp. 69-70).UWNR Safety Analysis Report Rev. 2 Sept. 3. J. F. Perkins and R. W. King, "Energy Release from the Decay of Fission Products", Nuclear Science and Engineering, 3, 726 (1958). 4. COMPL Y V1.5d, U.S. Environmental Protection Agency, Office of Radiation and Indoor air, Washington DC 20460, October 1989. 5. "Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor", GA-9064, Gulf General Atomic, Jan. 5, 1970. 6. "Annular Core Pulse Reactor", General Dynamics, General Atomic Division Report GACD 6977, Supplement 2,9/30/66.
c 7. Safeguards Summary Report for the TRIGA-FLIP Reactor at Puerto Rico Nuclear Center, Report PRNC 123, Revision C, November 11, 1969. 8. Docket 50-120, Change No. 11 to the Technical Specifications Facility License R-83, Texas A & M University, Section 3.2 Basis. 9. Hunsaker and Rightmire, "Engineering Applications of Fluid Mechanics," McGraw-Hill Book Company, 1947 (pp. 69-70). UWNR Safety Analysis Report Rev. 2 13-18 Sept. 2008 * *
14 ScopeThis section of the SAR for license renewal of the University of Wisconsin NuclearReactor constitutes the proposed Technical Specifications for that facility as required by 10 CFR 50.36. This document includes the basis to support the selection and significance of the specifications. Each basis is included for information purposesonly, and is not part of the Technical Specifications in that it does not constitute requirements or limitations which the licensee must meet in order to meet the specifications. Dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values due to construction and manufacturing tolerances or normal degree of accuracy or of instrument readings.These specifications are re-formatted from the technical specifications in force in 1999. Changes reflect only changes required by name changes or to include information not in the original technical specifications.
In addition, certain additions required by NUREG-1537 are included.
All substantive changes were denoted by redlining in Rev 0, but currently only changes since the last revision are redlined (indicated by vertical line in margin).
These technical specifications continue toinclude use of TRIGA-FLIP and the original LEU TRIGA fuels, either separately or in mixed cores.1.2 Format Content and section numbering is in accordance with section 1.2.2 of ANSI/ANSDefinitions The terms used herein are explicitly defined to ensure uniform interpretation of theTechnical Specifications.
UWNR Safety Analysis Report Rev. 2 14-1 Sept. * *
14 TECHNICAL SPECIFICATIONS TS 1 INTRODUCTION TS 1.1 Scope This section of the SAR for license renewal of the University of Wisconsin Nuclear Reactor constitutes the proposed Technical Specifications for that facility as required by 10 CFR 50.36. This document includes the basis to support the selection and significance of the specifications.
Each basis is included for information purposes only, and is not part of the Technical Specifications in that it does not constitute requirements or limitations which the licensee must meet in order to meet the specifications.
Dimensions, measurements, and other numerical values given in these specifications may differ slightly from actual values due to construction and manufacturing tolerances or normal degree of accuracy or of instrument readings.
These specifications are re-formatted from the technical specifications in force in 1999. Changes reflect only changes required by name changes or to include information not in the original technical specifications.
In addition, certain additions required by NUREG-1537 are included.
All substantive changes were denoted by redlining in Rev 0, but currently only changes since the last revision are redlined (indicated by vertical line in margin). These technical specifications continue to include use of TRIGA-FLIP and the original LEU TRIGA fuels, either separately or in mixed cores. TS 1.2 Format Content and section numbering is in accordance with section 1.2.2 of ANSI! ANS 15.1. TS 1.3 Definitions The terms used herein are explicitly defined to ensure uniform interpretation of the Technical Specifications . UWNR Safety Analysis Report Rev. 2 14-1 Sept. 2008
Reactor Operating Conditions CRITICAL: The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 125'F.PULSE MODE (PU)Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.REACTOR SECURED: The reactor is secured when: 1. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material present in the reactor to attain criticality upon optimum available conditions of moderation and reflection, or 2. The following conditions exist: a. The reactor is shut down, b. The console key switch is.in the "off' position and the key isremoved from the console and under the control of a licensed operator or stored in a locked storage area, and c. No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms,or insertion or withdrawal of in-core experiments with a reactivity worth exceeding 0.7% AK/K.REACTOR SHUTDOWN:The reactor is shut down when the reactor is subcritical by least 0.7% Ak/k of reactivity.
REACTOR OPERATION:Reactor operation is any condition wherein the reactor is not secured.UWNR Safety Analysis Report Rev. 2 14-2 Sept. TS 1.3.1 Reactor Operating Conditions COLD CRITICAL:
The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures both below 12soF. PULSE MODE (PU) Pulse mode operation shall mean any operation of the reactor with the mode selector switch in the pulse position.
REACTOR SECURED: The reactor is secured when: 1. Either there is insufficient moderator available in the reactor to attain criticality or there is insufficient fissile material present in the reactor to attain criticality upon optimum available conditions of moderation and reflection, or 2. The following conditions exist: a. The reactor is shut down, b. The console key switch is.in the "off' position and the key is removed from the console and under the control of a licensed operator or stored in a locked storage area, and c. No work is in progress involving in-core fuel handling or refueling operations, maintenance of the reactor or its control mechanisms, or insertion or withdrawal of in-core experiments with a reactivity worth exceeding 0.7%
REACTOR SHUTDOWN:
The reactor is shut down when the reactor is subcritical by least 0.7%
of reactivity.
REACTOR OPERATION:
Reactor operation is any condition wherein the reactor is not secured. UWNR Safety Analysis Report Rev. 2 14-2 Sept. 2008 * ( *
REPORTABLE OCCURRENCE:
A reportable occurrence is any of the following that occur during reactor operation:
1. Operation with any safety system setting less conservative than specified in the technical specifications;
2. Operation in violation of a Limiting Condition for Operation listed in Section 3;3. Operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of performing its intended safety function;4. Any unanticipated or uncontrolled change in reactivity greater than 0.7% AK/K, excluding reactor trips from a known cause;An observed inadequacy in the implementation of eitheradministrative or procedural controls, such that the inadequacy couldhave caused the existence or development of a condition which could result in operation of the reactor outside the specified safety limits; and 6. Abnormal and significant degradation in reactor fuel or cladding which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both.SHUTDOWN MARGIN: Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting. from any permissible operatingcondition (assuming the most reactive scrammable control element and any non-scrammable control elements remain full out), and the reactor will remain subcritical without further operator action.SQUARE WAVE MODE (SW)Square wave mode operation shall mean any operation of the reactor with the mode selector switch in the square wave position.STEADY STATE MODE (SS)Steady state mode operation shall mean operation of the reactor with themode selector switch in the manual or automatic positions.
UWNR Safety Analysis Report Rev. 2 14-3 Sept. * *
REPORTABLE OCCURRENCE:
A reportable occurrence is any of the following that occur during reactor operation:
1. Operation with any safety system setting less conservative than specified in the technical specifications;
2. Operation in violation of a Limiting Condition for Operation listed in Section 3; 3. Operation with a required reactor or experiment safety system component in an inoperative or failed condition which could render the system incapable of performing its intended safety function;
4. Any unanticipated or uncontrolled change in reactivity greater than 0.7% LlKlK, excluding reactor trips from a known cause; 5. An observed inadequacy in the implementation of either administrative or procedural controls, such that the inadequacy could have caused the existence or development of a condition which could result in operation of the reactor outside the specified safety limits; and 6. Abnormal and significant degradation in reactor fuel or cladding which could result in exceeding prescribed radiation exposure limits of personnel or environment, or both. SHUTDOWN MARGIN:* Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems, starting.
from any permissible operating condition (assuming the most reactive scrammable control element and any non-scrammable control elements remain full out), and the reactor will remain subcritical without further operator action. SQUARE WAVE MODE (SW) Square wave mode operation shall mean any operation of the reactor with the mode selector switch in the square wave position.
STEADY STATE MODE (SS) Steady state mode operation shall mean operation of the reactor with the mode selector switch in the manual or automatic positions . UWNR Safety Analysis Report Rev. 2 14-3 Sept. 2008
Reactor Experiments and Irradiation EXPERIMENT:
Experiment shall mean: 1. Any apparatus, device or material which is not a normal part of the reactor core or experimental facility, or 2. Any activity external to the biological shield using a beam of radiation emanating from the reactor core, or 3. Any operation designed to measure reactor parameters or characteristics, or any activity external to the biological shield using a beam of radiation emanating from the reactor core: Classification of experiments shall be: 1. Routine experiments.
Routine experiments are those which have previously been performed at the facility.2. Modified routine experiments. Modified routine experiments are those which have not been performed previously but are similar to theroutine experiments in that the hazards are neither greater nor significantly different than those for the corresponding routine experiments.
3. Special experiments.
Special experiments are those which are not routine or modified experiments.
EXPERIMENTAL FACILITIES:
Experimental facilities shall mean beam ports, including extension tubes with shields, thermal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems and any other in-pool irradiation facilities.
IRRADIATION:
Irradiation shall mean the insertion of any device or material that is not anormal part of the core or experimental facilities into an experimental facility so that the device.or material is exposed to a significant amount of the radiation available in that irradiation facility.UWNR Safety Analysis Report Rev. 2 14-4 Sept. TS 1.3.2 Reactor Experiments and Irradiation . EXPERIMENT:
Experiment shall mean: 1. Any apparatus, device or material which is not a normal part of the reactor core or experimental facility, or 2. Any activity external to the biological shield using a beam of radiation emanating from the reactor core, or 3. Any operation designed to measure reactor parameters or characteristics, or any activity external to the biological shield using a beam of radiation emanating from the reactor core: Classification of experiments shall be: 1. Routine experiments.
Routine experiments are those which have previously been performed at the facility.
2. Modified routine experiments.
Modified routine experiments are those
which have not been performed previously but are similar to the
routine experiments in that the hazards are neither greater nor significantly different than those for the corresponding routine experiments.
3. Special experiments.
Special experiments are those which are not routine or modified experiments.
EXPERIMENTAL FACILITIES:
Experimental facilities shall mean beam ports, including extension tubes with shields, thermal columns with shields, vertical tubes, through tubes, in-core irradiation baskets, irradiation cell, pneumatic transfer systems and any other in-pool irradiation facilities.
IRRADIATION:
Irradiation shall mean the insertion of any device or material that is not a normal part of the core or experimental facilities into an experimental facility so that the device or material is exposed to a significant amount of the radiation available in that irradiation facility.
UWNR Safety Analysis Report Rev. 2 14-4 Sept. 2008
EXPERIMENT Any experiment not meeting the criteria of a secured experiment.
EXPERIMENT:secured experiment shall mean any experiment that is held firmly in place a mechanical device or gravity, that is not readily removable from the reactor, and that requires one of the following actions to permit removal:Removal of mechanical fasteners 2. Use of underwater handling toolsMoving of shield blocks or beam port containers.
Reactor Components CORE core lattice position is that region in the core (approximately over a grid hole. It may occupied a fuel bundle, an experiment or experimental facility, or a reflector element.fuel bundle is a cluster of three or four fuel elements secured in a square array a top handle and a bottom grid plate adaptor.fuel element is a single TRIGA fuel rod of either standard or FLIP type.FLIP CORE:FLIP core is an arrangement of TRIGA-FLIP fuel in the reactor grid plate.FLIP FLIP fuel is TRIGA fuel that contains a nominal weight percent of uranium with a 11 5 enrichment of about and erbium as burnable poison.UWNR Safety Analysis Report Rev. 2 1-5Sp.20 14-5 Sept. * *
NON-SECURED EXPERIMENT Any experiment not meeting the criteria of a secured experiment.
SECURED EXPERIMENT:
A secured experiment shall mean any experiment that is held firmly in place by a mechanical device or by gravity, that is not readily removable from the reactor, and that requires one of the following actions to permit removal: 1. Removal of mechanical fasteners
2. Use of underwater handling tools 3. Moving of shield blocks or beam port containers.
TS 1.3.3 Reactor Components CORE LATTICE POSITION:
A core lattice position is that region in the core (approximately 3" by 3") over a grid hole. It may be occupied by a fuel bundle, an experiment or experimental facility, or a reflector element. FUEL BUNDLE: A fuel bundle is a cluster of three or four fuel elements secured in a square array by a top handle and a bottom grid plate adaptor. FUEL ELEMENT: A fuel element is a single TRIGA fuel rod of either standard or FLIP type. FLIP CORE: A FLIP core is an arrangement ofTRIGA-FLIP fuel in the reactor grid plate. FLIP FUEL: FLIP fuel is TRIGA fuel that contains a nominal 8.5 weight percent of uranium with a 235U enrichment of about 70% and erbium as burnable pOlson . UWNR Safety Analysis Report Rev. 2 14-5 Sept. 2008 INSTRUMENTED ELEMENT: An instrumented element is a special fuel element in which thermocouples are embedded for the purpose of measuring fuel temperatures during reactor operation.
MIXED CORE: A mixed core is an arrangement of standard TRIGA fuel elements and FLIP fuel elements with at least 35 TRIGA-FLIP fuel elements located in a central region of the core.OPERATIONAL CORE: An operational core may be a standard core, mixed core, or FLIP core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion have been determined to satisfy the requirements of the Technical Specifications.
REGULATING BLADE: The regulating blade is a low worth control blade that need not have scram capability.
Its position may be varied manually or by the servo-controller.
SHIM-SAFETY BLADE: A shim-safety blade is a control blade having an electric motor drive and scram capabilities.
Its position may be varied manually or by the servo-controller.
STANDARD TRIGA FUEL: Standard TRIGA fuel is TRIGA fuel that contains a nominal 8.5 weight percent of uranium with a 235 U enrichment of less than 20% and no burnable poison.STANDARD CORE: A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate.UWNR Safety Analysis Report Rev. 2 14-6 Sept. INSTRUMENTED ELEMENT: An instrumented element is a special fuel element in which thennocouples are embedded for the purpose of measuring fuel temperatures during reactor operation.
MIXED CORE: A mixed core is an arrangement of standard TRIGA fuel elements and FLIP fuel elements with at least 35 TRIGA-FLIP fuel elements located in a central region of the core. OPERATIONAL CORE: An operational core may be a standard core, mixed core, or FLIP core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion have been determined to satisfy the requirements ofthe Technical Specifications.
REGULATING BLADE: The regulating blade is a low worth control blade that need not have scram capability.
Its position may be varied manually or by the servo-controller.
SHIM-SAFETY BLADE: A shim-safety blade is a control blade having an electric motor drive and scram capabilities.
Its position may be varied manually or by the controller. STANDARD TRIGA FUEL: Standard TRIGA fuel is TRIGA fuel that contains a nominal 8.5 weight percent of uranium with a 235U enrichment of less than 20% and no burnable poison. STANDARD CORE: A standard core is an arrangement of standard TRIGA fuel in the reactor grid plate. UWNR Safety Analysis Report Rev. 2 14-6 Sept. 2008 * *
TRANSIENT ROD: The transient rod is a control rod with scram capabilities that can berapidly ejected from the reactor core to produce a pulse. Its position may be varied manually or by the servo-controller.
It may have a voided orsolid aluminum follower.1.3.4 Reactor Instrumentation:
CHANNEL CALIBRATION:A channel calibration consists of comparing a measured value from the measuring channel with a corresponding known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to known values of the measured variable.CHANNEL CHECK:A channel check is a qualitative verification of acceptable performance by observation of channel behavior.CHANNEL TEST: A channel test is the introduction of a signal into the channel to verify that it is operable.EXPERIMENT SAFETY SYSTEMS:
Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information whichrequires manual protective action to be initiated.
LIMITING SAFETY SYSTEM SETTINGS: Limiting safety system settings are settings for automatic protectivedevices related to those variables having significant safety functions.
MEASURED VALUE: The measured value is the magnitude of that variable as it appears on the output of a measuring channel.UWNR Safety Analysis Report Rev. 2 14-7 Sept. * *
TS 1.3.4 TRANSIENT ROD: The transient rod is a control rod with scram capabilities that can be rapidly ejected from the reactor core to produce a pulse. Its position may be varied manually or by the servo-controller.
It may have a voided or solid aluminum follower.
Reactor Instrumentation:
CHANNEL CALIBRATION:
A channel calibration consists of comparing a measured value from the measuring channel with a corresponding known value of the parameter so that the measuring channel output can be adjusted to respond with acceptable accuracy to known values of the measured variable.
CHANNEL CHECK: A channel check is a qualitative verification of acceptable performance by observation of channel behavior.
CHANNEL TEST: A channel test is the introduction of a signal into the channel to verify that it is operable.
EXPERIMENT SAFETY SYSTEMS: Experiment safety systems are those systems, including their associated input circuits, which are designed to initiate a scram for the primary purpose of protecting an experiment or to provide information which requires manual protective action to be initiated.
LIMITING SAFETY SYSTEM SETTINGS:
Limiting safety system settings are settings for automatic protective devices related to those variables having significant safety functions.
MEASURED VALUE: The measured value is the magnitude of that variable as it appears on the output of a measuring channel. UWNR Safety Analysis Report Rev. 2 14-7 Sept. 2008 MEASURING CHANNEL: A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.OPERABLE: A system, device, or component shall be considered operable when it is capable of performing its intended functions in a normal manner.REACTOR SAFETY SYSTEMS: Reactor safety systems are those systems, including their associated input circuits, which are designed to initiate a reactor scram for the primary purpose of protecting the reactor or to provide information which requires manual protective action to be initiated.
SAFETY CHANNEL: A safety channel is a measuring channel in the reactor safety system.SAFETY LIMITS: Safety limits are limits on important process variables which are found tobe necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
UWNR Safety Analysis Report Rev. 2 14-8 Sept. MEASURING CHANNEL: A measuring channel is the combination of sensor, interconnecting cables or lines, amplifiers, and output device which are connected for the purpose of measuring the value of a variable.
OPERABLE:
A system, device, or component shall be considered operable when it is capable of performing its intended functions ih a normal manner. REACTOR SAFETY SYSTEMS: Reactor safety systems are those systems, including their associated input circuits, which are designed to initiate a reactor scram for the primary purpose of protecting the reactor or to provide information which requires manual protective action to be initiated.
SAFETY CHANNEL: A safety channel is a measuring channel in the reactor safety system. SAFETY LIMITS: Safety limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.
UWNR Safety Analysis Report Rev. 2 14-8 Sept. 2008 * *
2 SAFETY LIMITS SAFETY SYSTEM 2.1 Safety Limits Applicability This specification applies to fuel element temperature and steady-state reactor power level.Objective The objective is to define the maximum fuel element temperature and reactor power level that can be permitted with confidence that no fuel element cladding failure will result.Specifications
1. The temperature in a TRIGA-FLIP fuel element shall not exceed 1150'C under any conditions of operation.
2. The temperature of a standard TRIGA fuel element shall not exceed 1000°C under any conditions of operation.
3. The reactor steady-state power level shall not exceed 1500 kW under any conditions of operation.
Basis A loss of integrity of the fuel element cladding could arise from a buildup ofexcessiye pressure between the fuel moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by air, fission productgases, and hydrogen from dissociation of the fuel moderator.
The magnitude of this;pressure is determined by the fuel moderator temperature and the ratio of hydrogen to zirconium in the alloy.The safety limit for the TRIGA-FLIP fuel element is based on data which indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconiumhydride will remain below the ultimate stress provided the temperature does not exceed 1150 0 C and the fuel cladding is water cooled'.The safety limit for the standard TRIGA fuel is based on data including the large amount of experimental evidence obtained during high performance reactor tests of this fuel. These data indicate that the stress in the cladding (due to hydrogen pressure UWNR Safety Analysis Report Rev. 2 14-9 Sept. * '.
TS 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TS 2.1 Safety Limits Applicability This specification applies to fuel element temperature and steady-state reactor power level. Objective The objective is to define the maximum fuel element temperature and reactor power level that can be permitted with confidence that no fuel element cladding failure will result. Specifications
1. The temperature in a TRIGA-FLIP fuel element shall not exceed 1150°C under any conditions of operation.
2. The temperature of a standard TRIGA fuel element shall not exceed 1000°C under any conditions of operation.
3. The reactor steady-state power level shall not exceed 1500 kW under any conditions of operation.
A loss of integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel moderator and the cladding if the fuel temperature exceeds the safety limit. The pressure is caused by air, fission product gases, and hydrogen from dissociation of the fuel moderator.
The magnitude of this: pressure is determined by the fuel moderator temperature and the ratio of hydrogen to zirconium in the alloy. The safety limit for the TRIGA-FLIP fuel element is based on data which indicate that the stress in the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided the temperature does not exceed 1150°C and the fuel cladding is water cooled 1. The safety limit for the standard TRIGA fuel is based on data including the large amount of experimental evidence obtained during high performance reactor tests of this fuel. These data indicate that the stress in the cladding (due to hydrogen pressure UWNR Safety Analysis Report Rev. 2 14-9 Sept. 2008 from the dissociation of zirconium hydride) will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1000 0 C and the fuel cladding is water cooled'.It has been shown by experience that operation of TRIGA reactors at a power level of 1500 kW will not result in damage to the fuel. Several reactors of this type have operated successfully for several years at power levels up to 1500kW. It has been shown by analysis and by measurements on other TRIGA reactors that a power level of 1500 kW corresponds to a peak fuel temperature of approximately 600'C. Thus a Safety Limit on power level of 1500 kW provides an ample margin of safety for operation.2.2 Limiting Safety System Settings Applicability This specification applies to the scram setting which prevents the safety limit from being reached.Objective The objective is to prevent the safety limits from being reached.Specifications
1. The limiting safety system setting for fuel temperature shall be 400'C (750'F) as measured in an instrumented fuel element. For a mixed core, the instrumented element shall be located in the region of the core containing FLIP type elements.2. The limiting safety system setting for reactor power level shall be 1.25 MW.Basis The first limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded.
A setting of 400'C provides a safety margin of 750'C for FLIP type fuel elements and a margin of 600'C for standard TRIGA fuel elements.
A part of the safety margin is used to account for the difference between the true and measured temperatures resulting from the actual location of the thermocouple.
If the thermocouple element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. If the UWNR Safety Analysis Report Rev. 2 14-10 Sept. from the dissociation of zirconium hydride) will remain below the ultimate stress provided that the temperature of the fuel does not exceed 1000 0 e and the fuel cladding is water cooled 2
It has been shown by experience that operation of TRIGA reactors at a power level of 1500 kW will not result in damage to the fuel. Several reactors of this type have operated successfully for several years at power levels up to 1500kW. It has been shown by and by measurements on other TRIGA reactors that a power level of 1500 kW corresponds to a peak fuel temperature of approximately 600 o e. Thus a Safety Limit on power level of 1500 k W provides an ample margin of safety for operation.
TS 2.2 Limiting Safety System Settings Applicability This specification applies to the scram setting which prevents the safety limit from being reached. Objective The objective is to prevent the safety limits from being reached. Specifications
1. The limiting safety system setting for fuel temperature shall be 400 0 e (750°F) as measured in an instrumented fuel element. Fora mixed core, the instrumented element shall be located in the region of the core containing FLIP type elements.
2. The li.!lliting safety system setting for reactor power level shall be 1.25 MW. The first limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded.
A setting of 400 0 e provides a safety margin of750oe for FLIP type fuel elements and a margin of600oe for standard TRIGA fuel elements.
A part of the safety margin is used to account for the difference between the true and measured temperatures resulting from the actual location of the thermocouple.
If the thermocouple element is located in the hottest position in the core, the difference between the true and measured temperatures will be only a few degrees since the thermocouple junction is at the mid-plane of the element and close to the anticipated hot spot. If the UWNR Safety Analysis Report Rev. 2 14-10 Sept. 2008 ( " .' *
thermocouple element is located in a region of lower temperature, such as on the periphery of the core, the measured temperature will differ by a greater amount from that actually occurring at the core hot spot. Calculations and measurements made atthe facility with all permitted mixed core arrangements indicate that, for this case, thetrue temperature at the hottest location in the core will differ from the measured temperature by no more than a factor of two. Thus, when the temperature in the thermocouple elements reaches the trip setting of 400'C, the true temperature at the hottest location would be no greater than 800'C providing a margin to the safety limit of at least 200'C for standard fuel elements and 350'C for FLIP type elements.
These margins are ample to account for the remaining uncertainty in the accuracy of the fueltemperature measurement channel and any overshoot in reactor power resulting from a reactor transient during steady state mode operation. For a mixed core (i.e., onecontaining both standard and FLIP type elements), the requirement that the instrumented element be located in the FLIP region of the core provides an even greater margin of safety since the peak to average power ratio within that region will be smaller than over an entire core composed of elements of the same type.Calculations and measurements for this and similar TRIGA reactors indicate at 1.25 MW, the peak fuel temperature in the most limiting core loading permitted under section 3 of these specifications (9 FLIP bundles) will be less than 600'C so that the second limiting power level setting provides an ample safety margin to accommodate errors in power level measurement and anticipated operational transients.
In the pulse mode of operation, the first limiting safety system setting will apply.However, the power level channels do not provide protection in pulse mode, and the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).
The limit on transient rod worth in another specification limits the generated power so that fuel temperatures reached in a transient are smallerthan those from full power operation. This transient rod worth limit is less than the reactivity required for steady-state full power. If the transient rod fails to automatically drop after the pulse, fuel temperature reached due to energy generationin the tail of a pulse will be less than that at full power. Only in the case of operationoutside the permitted parameters of core composition and reactivity limitations would fuel temperature safety system actuation be needed to provide protection in any operating mode.UWNRSafty nalsis epot Rv. 14-1 Spt.200 U"R Safety Analysis Report Rev. 2 14-11 Sept. * *
thermocouple element is located in a region of lower temperature, such as on the periphery of the core, the measured temperature will differ by a greater amount from that actually occurring at the core hot spot. Calculations and measurements made at the facility with all permitted mixed core arrangements indicate that, for this case, the true temperature at the hottest location in the core will differ from the measured temperature by no more than a factor of two. Thus, when the temperature in the thermocouple elements reaches the trip setting of 400°C, the true temperature at the hottest location would be no greater than 800°C providing a margin to the safety limit of at least 200°C for standard fuel elements and 350°C for FLIP type elements.
These margins are ample to account for the remaining uncertainty in the accuracy of the fuel temperature measurement channel and any overshoot in reactor power resulting from a reactor transient during steady state mode operation.
For a mixed core (i.e., one containing both standard and FLIP type elements), the requirement that the instrumented element be located in the FLIP region of the core provides an even greater margin of safety since the peak to average power ratio within that region will be smaller than over an entire core composed of elements of the same type. Calculations and measurements for this and similar TRIGA reactors indicate at 1.25 MW, the peak fuel temperature in the most limiting core loading permitted under section 3 of these specifications (9 FLIP bundles) will be less than 600°C so that the second limiting power level setting provides an ample safety margin to accommodate errors in power level measurement and anticipated operational transients . In the pulse mode of operation, the first limiting safety system setting will apply. However, the power level channels do not provide protection in pulse mode, and the temperature channel will have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds).
The limit on transient rod worth in another specification limits the generated power so that fuel temperatures reached in a transient are smaller than those from full power operation.
This transient rod worth limit is less than the reactivity required for steady-state full power. Ifthe transient rod fails to automatically drop after the pulse, fuel temperature reached due to energy generation in the tail of a pulse will be less than that at full power. Only in the case of operation outside the permitted parameters of core composition and reactivity limitations would fuel temperature safety system actuation be needed to provide protection in any operating mode . UWNR Safety Analysis Report Rev. 2 14-11 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2 14-12 Sept.
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UWNR Safety Analysis Report Rev. 2 14-12 Sept. 2008
FOR OPERATIONReactor Core Parameters ApplicabilityThese specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods. They apply for all modes of operation.
Objective The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit will not be exceeded.
Excess Reactivity Specifications The excess reactivity shall not exceed 5.6% Ak/k.BasisAs shown in chapter 4 of the SAR, this amount of excess reactivity will provide the capability to operate the reactor at full power with experiments in place. The primary limitation providing reactivity safety, however, is the shutdown margin requirement discussed in the next specification.Shutdown Margin Specifications The reactor shall not be operated unless the shutdown margin provided by control rods shall be greater than 0.2% Ak/k with: 1. the highest worth non-secured experiment in its most reactive state, 2. the highest worth control element and the regulating blade (if not scrammable) fully withdrawn, and 3. the reactor in the cold condition without xenon.UWNR Safety Analysis Report Rev. 2 14-13 Sept. * *
TS 3 LIMITING CONDITIONS FOR OPERATION TS 3.1 Reactor Core Parameters Applicability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods. They apply for all modes of operation.
Objective The objective is to assure that the reactor can be shut down at all times and to assure that the fuel temperature safety limit will not be exceeded.
TS 3.1.1 Excess Reactivity TS 3.1.2 Specifications The excess reactivity shall not exceed 5.6% Liklk. As shown in chapter 4 ofthe SAR, this amount of excess reactivity will provide the capability to operate the reactor at full power with experiments in place. The primary limitation providing reactivity safety, however, is the shutdown margin requirement discussed in the next specification.
Shutdown Margin Spec ifi cati ons The reactor shall not be operated unless the shutdown margin provided by control rods shall be greater than 0.2% Liklk with: 1. the highest worth non-secured experiment in its most reactive state, 2. the highest worth control element and the regulating blade (if not scrammable) fully withdrawn, and 3. the reactor in the cold condition without xenon . UWNR Safety Analysis Report Rev. 2 14-13 Sept. 2008 BasisThe value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control element should remain in the fully withdrawn position.
If the regulating blade is not scrammable, its worth is not used in determining the shutdown reactivity.
Pulse Limits Specifications
1. The reactivity to be inserted for pulse operation shall be determined and mechanically limited such that the reactivity insertion will not exceed 1.4% Ak/k.2. Pulses shall not be initiated at power levels exceeding 1 kilowatt.Basis Measurements performed on the Puerto Rico Nuclear Center TRIGA-FLIP reactor indicated that a pulse insertion of reactivity of 1.4% A resulted in a maximum temperature rise of approximately 400'C. With an ambient water temperature of approximately 1 00°C, the maximum fuel temperature would be approximately 500'C resulting in a safety margin of 500'C for standard fuel and 650'C for FLIP type fuel. Tests done on the mixed and all-FLIP cores indicate that the fuel temperatures measured and calculated for the core arrangements allowed by these specifications do not exceed 683°C in the worst case allowed.The temperature rise from pulse initiation is in addition to the temperature in the fuel at the time the pulse is initiated.
Limiting the initial power level to 1kW assures that excessive temperatures will not be reached.These margins allow amply for uncertainties due to the accuracy of measurement or location of the instrumented fuel element or due to the extrapolation of data from the PRNC reactor.UWNR Safety Analysis Report Rev. 2 14-14 Sept. The value of the shutdown margin assures that the reactor can be shut down from any operating condition even if the highest worth control element should remain in the fully withdrawn position.
If the regulating blade is not scrammable, its worth is not used in determining the shutdown reactivity.
TS 3.1.3 Pulse Limits Specifications
1. The reactivity to be inserted for pulse operation shall be determined and mechanically limited such that the reactivity insertion will not exceed 1.4%
2. Pulses shall not be initiated at power levels exceeding 1 kilowatt . Measurements performed on the Puerto Rico Nuclear Center TRIGA-FLIP
reactor indicated that a pulse insertion of reactivity of 1.4% klk resulted in a
maximum temperature rise of approximately 400°C. With an ambient water temperature of approximately 100°C, the maximum fuel temperature would be approximately 500°C resulting in a safety margin of 500°C for standard fuel and 650°C for FLIP type fuel. Tests done on thy mixed and all-FLIP cores 3,4,5,6 indicate that the fuel temperatures measured and calculated for the core arrangements allowed by these specifications do not exceed 683°C in the worst case allowed. The temperature rise from pulse initiation is in addition to the temperature in the fuel at the time the pulse is initiated.
Limiting the initial power level to I kW assures that excessive temperatures will not be reached. These margins allow amply for uncertainties due to the accuracy of measurement or location of the instrumented fuel element or due to the extrapolation of data from the PRNC reactor. UWNR Safety Analysis Report Rev. 2 14-14 Sept. 2008
3.1.4 Core Configurations Applicability This specification applies to the configuration of fuel and in-core experiments.
ObjectiveThe objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessivepower densities will not be produced.Specifications
1. The core shall be an arrangement of TRIGA uranium-zirconium hydridefuel-moderator bundles positioned in the reactor grid plate.2. The TRIGA core assembly may be standard, FLIP, or a combination, thereof (mixed core) provided that any FLIP fuel be comprised of at least thirty-five (35) fuel elements, located in a contiguous, central region.3. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.4. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water.5. Fuel shall not be inserted or removed from the core unless the reactor is subcritical by more than the calculated worth of the most reactive fuel assembly.6. Control elements shall not be manually removed from the core unless thecore has been shown to be subcritical with all control elements in the full out position.UWNR Safety Analysis Report Rev. 2 14-15 Sept. * *
TS 3.1.4 Core Configurations Applicability This specification applies to the configuration of fuel and in-core experiments.
Objective The objective is to assure that provisions are made to restrict the arrangement of fuel elements and experiments so as to provide assurance that excessive power densities will not be produced.
Specifications
1. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator bundles positioned in the reactor grid plate. 2. The TRIGA core assembly may be standard, FLIP, or a combination, thereof (mixed core) provided that any FLIP fuel be comprised of at least thirty-five (35) fuel elements, located in a contiguous, central region. 3. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.
4. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water. 5. Fuel shall not be inserted or removed from the core unless the reactor is subcritical by more than the calculated worth of the most reactive fuel assembly.
6. Control elements shall not be manually removed from the core unless the core has been shown to be subcritical with all control elements in the full out position . UWNR Safety Analysis Report Rev. 2 14-15 Sept. 2008 Basis 1. Standard TRIGA cores have been in use for years and their characteristics are well documented. The Puerto Rico Nuclear Center and the Gulf Mark III all-FLIP cores have operated and their characteristics are available.
Gulf has also performed a series of experiments using standard and FLIP fuel in mixed cores and a mixed core has been used successfully in theTexas A&M University TRIGA reactor. In addition, studies performed at Wisconsin for a variety of mixed core arrangements indicate that such cores with mixed loadings would safely satisfy all operational requirements (SAR Chapters 4 and 6).2. In mixed cores, it is necessary to arrange FLIP elements in a contiguous, central region of the core to control flux peaking and power generation peak values in individual elements.3. Vacant core lattice positions will contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core. They will be permitted only on the periphery of the core to prevent power perturbations in regions of high power density.4. The core will be assembled in the reactor grid plate which is located in a pool of light water. Water in combination with graphite reflectors can beused for neutron economy and the enhancement of experimental facilityradiation requirements.
5-6. Manual manipulation of core components will be allowed only when a single manipulation can not result in inadvertent criticality.
Reactivity Coefficients Does not apply to TRIGA and TRIGA-FLIP reactors.UWNR Safety Analysis Report Rev. 2 14-16 Sept. 1. Standard TRIGA cores have been in use for years and their characteristics are well documented.
The Puerto Rico Nuclear Center and the Gulf Mark III all-FLIP cores have operated and their characteristics are available.
Gulf has also performed a series of experiments using standard and FLIP fuel in mixed cores and a mixed core has been used successfully in the Texas A&M University TRIGA reactor. In addition, studies performed at Wisconsin for a variety of mixed core arrangements indicate that such cores with mixed loadings would safely satisfy all operational requirements (SAR Chapters 4 and 6). 2. In mixed cores, it is necessary to arrange FLIP elements in a contiguous, central region of the core to control flux peaking and power generation peak values in individual elements.
3. Vacant core lattice positions will contain experiments or an experimental facility to prevent accidental fuel additions to the reactor core. They will be permitted only on the periphery of the core to prevent power perturbations in regions of high power density.
4. The core will be assembled in the reactor grid plate which is located in a
pool oflight water. Water in combination with graphite reflectors can be used for neutron economy and the enhancement of experimental facility radiation requirements.
5-6. Manual manipulation of core components will be allowed only when a single manipulation can not result in inadvertent criticality.
TS 3.1.5 Reactivity Coefficients Does not apply to TRIGA and TRIGA-FLIP reactors.
UWNR Safety Analysis Report Rev. 2 14-16 Sept. 2008
Fuel Parameters Applicability This specification applies to the dimensional and structural integrity of the fuel elements.ObjectiveThe objective is to assure that the reactor will not be operated with defective fuel elements installed.
SpecificationsThe reactor shall not be operated with damaged fuel except for purposes ofidentifying the damaged fuel. A fuel element shall be considered damaged and must be removed from the core if: 1. In measuring the transverse bend, its sagitta 7 exceeds 0.125 inch over the length of the cladding;
2. In measuring the elongation, its length of the cladding exceeds its original length by 0.125 inch; and 3. A clad defect exists as indicated by detection of release of fission products.4. The fuel has not been visually inspected within the previous 15 months.The burnup of uranium-235 in the UzrH fuel matrix shall not exceed 50 percent of the initial concentration.
8'9 Basis The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Experience with TRIGA reactors has shown that fuelelement bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow through the top grid plate.UWNR Safety Analysis Report Rev.
2 14-17 Sept. * *
TS 3.1.6 Fuel Parameters Applicability This specification applies to the dimensional and structural integrity of the fuel elements.
Objective The objective is to assure that the reactor will not be operated with defective fuel elements installed.
Specifications The reactor shall not be operated with damaged fuel except for purposes of identifying the damaged fuel. A fuel element shall be considered damaged and must be removed from the core if: 1. In measuring the transverse bend, its sagitta 7 exceeds 0.125 inch over the length of the cladding;
2. In measuring the elongation, its length of the cladding exceeds its original length by 0.125 inch; and 3. A clad defect exists as indicated by detection of release of fission products.
4. The fuel has not been visually inspected within the previous 15 months. 5. The bumup ofuranium-235 in the UzrH fuel matrix shall not exceed 50 percent of the initial concentration.
8 ,9 The limit of transverse bend has been shown to result in no difficulty in disassembling the core. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching.
Experience with TRIGA reactors has shown that fuel element bowing that could result in touching has occurred without deleterious effects. The elongation limit has been specified to assure that the cladding material will not be subjected to stresses that could cause a loss of integrity in the fuel containment and to assure adequate coolant flow through the top grid plate . UWNR Safety Analysis Report Rev. 2 14-17 Sept. 2008 Reactor Control and Safety SystemsOperable Control Rods ApplicabilityThis specification applies to the number of operable control elements that must exist in order to operate the reactor.Objective The objective of this requirement is to insure that the reactor may be shut down from any condition of operation.
Specifications The reactor shall not be operated unless at least three control elements are functioning and scrammable.
Basis In most cores the limits on shutdown margin actually dictate the number of operable control elements required.
Non-pulsing cores do not require presence of a transient control rod if the shutdown margin requirements are met by the control blades.UWNR Safety Analysis Report Rev. 2 14-18 Sept. TS 3.2 Reactor Control and Safety Systems TS 3.2.1 Operable Control Rods Applicability This specification applies to the number of operable control elements that must exist in order to operate the reactor. Objective The objective of this requirement is to insure that the reactor may be shut down from any condition of operation.
Specifications The reactor shall not be operated unless at least three control elements are functioning and scrammable.
In most cores the limits on shutdown margin actually dictate the number of operable control elements required.
Non-pulsing cores do not require presence of a transient control rod if the shutdown margin requirements are met by the control blades. UWNR Safety Analysis Report Rev. 2 14-18 Sept. 2008 * *
Reactivity Insertion Rates (Scram time)ApplicabilityThis specification applies to the time required for the scrammable control elements to be fully inserted from the instant that a safety channel variablereaches the Safety System Setting.ObjectiveThe objective is to achieve prompt shutdown of the reactor to prevent fuel damage.Specifications The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control element reaches its fully inserted position shall not exceed 2 seconds.Basis This specification assures that the reactor will be promptly shut down when a scram signal is initiated.
Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor.
Other Pulsed Operation Limitations Limitations other than those on core configuration and pulsed reactivity insertion limits are not required on this reactor.UWNR Safety Analysis Report Rev. 2 14-19 Sept. * *
TS 3.2.2 Reactivity Insertion Rates (Scram time) TS 3.2.3 Applicability This specification applies to the time required for the scrammable control elements to be fully inserted from the instant that a safety channel variable reaches the Safety System Setting. Objective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage. Specifications The scram time measured from the instant a simulated signal reaches the value of the LSSS to the instant that the slowest scrammable control element reaches its fully inserted position shall not exceed 2 seconds. This specification assures that the reactor will be promptly shut down when a scram signal is initiated.
Experience and analysis have indicated that for the range of transients anticipated for a TRIGA reactor, the specified scram time is adequate to assure the safety of the reactor. Other Pulsed Operation Limitations Limitations other than those on core configuration and pulsed reactivity insertion limits are not required on this reactor. UWNR Safety Analysis Report Rev. 2 14-19 Sept. 2008 3.2.4 Reactor Safety System ApplicabilityThis specification applies to the reactor safety system channels.Objective The objective is to specify the minimum number of reactor safety channelsthat must be operable for safe operation.
SpecificationsThe reactor shall not be operated unless the safety channels described in Table 3.2.4 are operable.Table 3.2.4 Reactor Safety System Channels Number operable Safety Channel Setpoint and Function in specified mode SS SW PU Fuel Temperature Scram if fuel temperature exceeds >400'C in 1 1 1 the fuel temperature safety channel. In the event of loss of all available fuel thermocouples and inability to obtain a replacement instrumented fuel element, operation may continue in any operational core if the linear power level scram points are reduced to full power.
Linear Power Level Scram if power > 125% full power 2 2 -Manual Scram Manually initiated scram 1 1 1 Preset Timer Transient rod scram 15 seconds or less after 1 pulse Reactor water level Scram if < 19 feet above top of core 1 1 1 High Voltage Monitor Scram on loss of high voltage to neutron and 1 1 1 gamma ray power level instrument detectorsUWNR Safety Analysis Report Rev. 2 14-20 Sept. TS 3.2.4 Reactor Safety System Applicability This specification applies to the reactor safety system channels.
Objective The objective is to specify the minimum number of reactor safety channels that must be operable for safe operation.
Specifications The reactor shall not be operated unless the safety channels described in Table 3.2.4 are operable.
bl Ta e 3.2.4 Reactor Safety System Channels Number operable Safety Channel Setpoint and Function in specified mode SS SW PU Fuel Temperature Scram if fuel temperature exceeds >400°C in 1 1 1 the fuel temperature safety channel. In the event of loss of all available fuel thermocouples and inability to obtain a replacement instrumented fuel element, operation may continue in any operational core if the linear power level scram points are reduced to 110% full power. Linear Power Level Scram if power > 125% full power 2 2 -Manual Scram Manually initiated scram 1 1 1 Preset Timer Transient rod scram 15 seconds or less after --1 pulse Reactor water level Scram if < 19 feet above top of core 1 1 1 High Voltage Monitor Scram on loss of high voltage to neutron and 1 1 1 gamma ray power level instrument detectors UWNR Safety Analysis Report Rev. 2 14-20 Sept. 2008 * *
BasisThe fuel temperature and power scrams provide protection to ensure that the reactor is shut down before the safety limit on fuel temperature is reached.The exception is required because FLIP fuel is no longer manufactured.
If a core has been tested to meet the definition of an operational core the power level scrams provide adequate protection to assure the LCO of fuel temperature is not exceeded.The manual scram allows the operator a means of rapid shutdown in the event of unsafe or abnormal conditions.
The preset timer assures reduction of reactor power to a low level after a pulse.The reactor pool water level scram assures shutdown of the reactor in the event of a serious leak in the primary system or pool.The high voltage monitor prevents operation of the reactor with other systems inoperable due to failure of the detector high voltage supplies.UWNR Safety Analysis Report Rev. 2 14-21 Sept. * *
The fuel temperature and power scrams provide protection to ensure that the reactor is shut down before the safety limit on fuel temperature is reached. The exception is required because FLIP fuel is no longer manufactured.
If a core has been tested to meet the definition of an operational core the power level scrams provide adequate protection to assure the LCO of fuel temperature is not exceeded.
The manual scram allows the operator a means of rapid shutdown in the event of unsafe or abnormal conditions.
The preset timer assures reduction of reactor power to a low level after a pulse. The reactor pool water level scram assures shutdown of the reactor in the event of a serious leak in the primary system or pool. The high voltage monitor prevents operation of the reactor with other systems inoperable due to failure of the detector high voltage supplies . UWNR Safety Analysis Report Rev. 2 14-21 Sept. 2008
Interlocks Applicability This section applies to the interlocks which inhibit or prevent control elementwithdrawal or reactor startup.
Objective The objective of these interlocks is to prevent operation under unanalyzed or imprudent conditions.
Specifications The reactor shall not be operated in the indicated modes unless the interlocks in Table 3.2.5 are operable.Table 3.2.5 Interlocks Number operable Channel Setpoint and Function in specified mode SS SW PU Log Count Rate Prevent control element withdrawal when 1 1 1 neutron count rate < 2 per second Transient Rod Control Prevent application of air to fire transient rod 1 0 0 unless drive is at IN limit.Log N Power Level Prevent application of air to fire transient rod 1 1 1when power level is above kW and transient rod is not full in.Pulse Mode Control Prevents withdrawal of control blades while 0 0 1 in pulse mode.BasisThe Log count rate interlock does not allow control element withdrawal unless the neutron count rate is high enough to assure proper instrument' response during reactor startup.
The Transient Rod Control interlock prevents inadvertent addition of excessive amounts or reactivity in steady-state modes.The Log N interlock prevents firing of the transient rod at power levels above 1.0 kW if the transient rod drive is not in the full down position.
This UWNR Safety Analysis Report Rev. 2 14-22 Sept. TS 3.2.5 Interlocks Channel Log Count Rate Applicability This section applies to the interlocks which inhibit or prevent control element withdrawal or reactor startup. Objective The objective ofthese interlocks is to prevent operation under unanalyzed or imprudent conditions.
Specifications The reactor shall not be operated in the indicated modes unless the. interlocks in Table 3.2.5 are operable.
Table 3.2.5 Interlocks Number operable Setpoint and Function in specified mode SS SW PU Prevent control element withdrawal when 1 1 1 neutron count rate < 2 per second Transient Rod Control Prevent application of air to fire transient rod 1 0 0 unless drive is at IN limit. Log N Power Level Prevent application of air to fire transient rod 1 1 1 when power level is above 1 kW and transient rod is not full in. Pulse Mode Control Prevents withdrawal of control blades while 0 0 1 in pulse mode. The Log count rate interlock does not allow control element withdrawal unless the neutron count rate is high enough to assure proper instrument' response during reactor startup. The Transient Rod Control interlock prevents inadvertent addition of excessive amounts or reactivity in steady-state modes. *
The Log N interlock prevents firing of the transient rod at power levels above
1.0 kW if the transient rod drive is not in the full down position.
This UWNR Safety Analysis Report Rev. 2 14-22 Sept. 2008 effectively prevents inadvertent pulses which might cause fuel temperature to exceed the safety limit on fuel temperature.
The pulse mode control blade withdrawal interlock prevents reactivity addition in pulse mode other than by firing the transient rod.
Backup Shutdown Mechanisms Backup shutdown mechanisms are not required for this reactor.
Bypassing Channels ApplicabilityThis specification applies to the interlocks in Table 3.2.5.Objective The objective is to indicate the conditions in which an interlock may be bypassed.SpecificationsThe Log Count Rate interlock in Table 3.2.5 may be bypassed: 1. During fuel loading in order to allow control element withdrawal necessary for the fuel loading procedure or 2. When Log Power Level and Linear Power Level channels are on-scale.Basis During early stages of fuel loading the count-rate on the source range channel will be below the interlock setpoint.
The bypass allows control element movements necessary for loading fuel with control elements partially withdrawn and for performing inverse multiplication determinations of controlelement worth and core reactivity status. Once the other power indications are available the startup count rate channel is no longer required, so the interlock no longer serves any purpose.UWNR Safety Analysis Report Rev. 2 14-23 Sept. * *
TS 3.2.6 effectively prevents inadvertent pulses which might cause fuel temperature to exceed the safety limit on fuel temperature . The pulse mode control blade withdrawal interlock prevents reactivity addition in pulse mode other than by firing the transient rod. Backup Shutdown Mechanisms Backup shutdown mechanisms are not required for this reactor. TS 3.2.7 Bypassing Channels Applicability This specification applies to the interlocks in Table 3.2.5. Objective The objective is to indicate the conditions in which an interlock may be bypassed.
Specifications The Log Count Rate interlock in Table 3.2.5 may be bypassed:
1. During fuel loading in order to allow control element withdrawal necessary for the fuel loading procedure or 2. When Log Power Level and Linear Power Level channels are on-scale.
During early stages of fuel loading the count-rate on the source range channel will be below the interlock setpoint.
The bypass allows control element movements necessary for loading fuel with control elements partially withdrawn and for performing inverse multiplication determinations of control element worth and core reactivity status. Once the other power indications are available the startup count rate channel is no longer required, so the interlock no longer serves any purpose . UWNR Safety Analysis Report Rev. 2 14-23 Sept. 2008
Control Systems and Instrumentation Required for Operation ApplicabilityThis specification applies to the information which must be available to the reactor operator during reactor operation.
Objective The objective is to require that sufficient information is available to the operator to assure safe operation of the reactor.Specifications The reactor shall not be operated unless measuring channels listed in Table 3.2.8 are operable.Table 3.2.8 Instrumentation and Controls Required for Operation Number operable Channel Function in specified mode SS SW PU Fuel Temperature Input for fuel temperature scram. In the 1 1 1 event of loss of all available fuel thermocouples and inability to obtain a replacement operation may continue in any operational core.Linear Power Level Input for safety system power level scram j2 2 0 Log Power Level Wide range power indication, permissive for 1 1 0 initiation of Pulse Mode Startup Log Count Rate Wide range power indication, permissive for 0 control element withdrawalPulsing Power Level Pulse power level indication 0 0 1* Required during startup only until the Log Power Level and Linear Power Level channels are on-scale Basis Fuel temperature indicated at the control console gives continuous informationon the process variable which has a specified safety limit. The exception is required because FLIP fuel is no longer manufactured.
If a core has been tested to meet the definition of an operational core the power level scramsUWNR Safety Analysis Report Rev. 2 14-24 Sept.
TS 3.2.8 Control Systems and Instrumentation Required for Operation Channel Applicability This specification applies to the information which must be available to the reactor operator during reactor operation.
Objective The objective is to require that sufficient information is available to the operator to assure safe operation ofthe reactor. Specifications The reactor shall not be operated unless measuring channels listed in Table 3.2.8 are operable.
T bl 328 I a e .. f nstrumenta IOn an dC I R . d £ 0 f ontro s eqUlre or 'pera IOn Number operable Function in specified mode SS SW PU Fuel Temperature Input for fuel temperature scram. In the 1 1 1 , event of loss of all available fuel thermocouples and inability to obtain a replacement operation may continue in any operational core. Linear Power Level Input for safety system power level scram 2 2 0 Log Power Level Wide range power indication, permissive for 1 1 0 initiation of Pulse Mode Startup Log Count Rate Wide range power indication, permissive for 1* 1* 0 control element withdrawal Pulsing Power Level Pulse power level indication 0 0 1 Required during startup only until the Log Power Level and Linear Power Level channels are on-scale Fuel temperature indicated at the control console gives continuous information on the process variable which has a specified safety limit. The exception is required because FLIP fuel is no longer manufactured.
If a core has been tested to meet the definition of an operational core the power level scrams UWNR Safety Analysis Report Rev. 2 14-24 Sept. 2008 * *
provide adequate protection to assure the LCO of fuel temperature is not exceeded.The power level monitors assure that reactor power level is adequately monitored for all modes of operation.Reactor Pool Water Systems Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water.Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding and to prevent damage to in-pool components by corrosion.
Specifications
1. The reactor core shall be cooled by natural convective water flow.2. The pool water inlet pipe to the demineralizer shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped with a check valve and siphon breaker to preventinadvertent draining of the pool.3. Diffuser and other auxiliary systems pumps shall be located no more than feet below the top of the reactor pool.
4. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool.
5. A pool level alarm shall indicate loss of coolant if the pool level drops one foot or less below normal level.6. The reactor shall not be operated if the conductivity of the pool water exceeds 5 micromhos/cm
(<0.2 MegOhm-cm) when averaged over a period of one week.7. The reactor shall not be operated if the radioactivity of pool water exceeds the limits of 10 CFR Part 20 Appendix B Table 3 for radioisotopes with half-lives>24 hours.
UWNR Safety Analysis Report Rev. 2 14-25 Sept. * *
provide adequate protection to assure the LCO of fuel temperature is not exceeded . The power level monitors assure that reactor power level is adequately monitored for all modes of operation.
TS 3.3 Reactor Pool Water Systems Applicability This specification applies to the pool containing the reactor and to the cooling of the core by the pool water. Objective The objective is to assure that coolant water shall be available to provide adequate cooling of the reactor core and adequate radiation shielding and to prevent damage to in-pool components by corrosion.
Specifications
l. The reactor core shall be cooled by natural convective water flow . 2. The pool water inlet pipe to the demineralizer shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped with a check valve and siphon breaker to prevent inadvertent draining of the pool. 3. Diffuser and other auxiliary systems pumps shall be located no more than 15 feet below the top of the reactor pool. 4. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool. 5. A pool level alarm shall indicate loss of coolant if the pool level drops one foot or less below normal level. 6. The reactor shall not be operated if the conductivity of the pool water exceeds 5 micrornhos/cm
<<0.2 MegOhm-cm) when averaged over a period of one week. 7. The reactor shall not be operated if the radioactivity of pool water exceeds the limits of 10 CFR Part 20 Appendix B Table 3 for radioisotopes with half-lives
>24 hours . UWNR Safety Analysis Report Rev. 2 14-25 Sept. 2008 Basis 1. This specification is based on thermal and hydraulic calculations which show that the TRIGA-FLIP core can operate in a safe manner at power levels up to 2,700 kW with natural convection flow of the coolant water. A comparison of operation of the TRIGA-FLIP and standard TRIGA Mark III has shown operation to be safe for the above power level. Thermal and hydraulic characteristics of mixed cores are essentially the same asthat for TRIGA-FLIP and standard cores.2. The inlet pipe to the demineralizer is positioned so that a siphon action will drain less than 15 feet of water. The outlet pipe from the demineralizer discharges into a pipe entering the bottom of the pool through a check valve which prevents leakage from the pool by reverse flow from pipe ruptures or improper operation of the demineralizer valve manifold.
In addition, the pipe has a loop equipped with a siphon breaker which prevents loss of pool water.3. In the event of pipe failure and siphoning of pool water, the pool water level will drop no more than 15 feet from the top of the pool.4. Other pipes which enter the pool have siphon breakers which prevent pool drainage.
Valves are provided for pneumatic tube system lines and primary cooling system pipe. Other piping installed in the pool has blind flanges permanently installed.
5. Loss of coolant alarm, after one foot of loss, requires corrective action. This alarm is observed in the reactor control room and outside the reactor building.6. The conductivity limit assures that materials within the pool will not be degraded and that the radioactivity of the pool water will be minimized.
7. Analyses in section 12.2.9 of the Safety Analysis Report show that limiting the activity to this level will not result in any person being exposed to concentrations greater than those permitted by 10 CFR Part 20.3.4 Confinement Applicability These specifications apply to the room housing the reactor and the ventilation system controlling that room.Objective The objective is to provide restrictions on release of airborne radioactive materials to the environs. UWNR Safety Analysis Report Rev. 2 14-26 Sept. Basis 1. This specification is based on thermal and hydraulic calculations which show that
the TRIGA-FLIP core can operate in a safe manner at power levels up to 2,706 kW with natural convection flow of the coolant water. A comparison of operation of the TRIGA-FLIP and standard TRIGA Mark III has shown operation to be safe for the above power level. Thermal and hydraulic characteristics of mixed cores are essentially the same as that for TRIGA-FLIP and standard cores .. 2. The inlet pipe to the demineralizer is positioned so that a siphon action will drain less than 15 feet of water. The outlet pipe from the demineralizer discharges into a pipe entering the bottom of the pool through a check valve which prevents leakage from the pool by reverse flow from pipe ruptures or improper operation of the demineralizer valve manifold.
In addition, the pipe has a loop equipped with a siphon breaker which prevents loss of pool water. 3. In the event of pipe failure and siphoning of pool water, the pool water level will drop no more than 15 feet from the top of the pool. 4. Other pipes which enter the pool have siphon breakers which prevent pool drainage.
Valves are provided for pneumatic tube system lines and primary cooling system pipe. Other piping installed in the pool has blind flanges permanently installed.
5. Loss of coolant alarm, after one foot of loss, requires corrective action. This alarm is observed in the reactor control room and outside the reactor building.
6. The conductivity limit assures that materials within the pool will not be degraded and that the radioactivity of the pool water will be minimized.
7. Analyses in section 12.2.9 of the Safety Analysis Report show that limiting the activity to this level will not result in any person being exposed to conceritrations greater than those pennitted by 10 CFR Part 20. TS 3.4 Confinement Applicability These specifications apply to the room housing the reactor and the ventilation system controlling that room. Objective
The objective is to provide restrictions on release of airborne radioactive materials to
the environs.
UWNR Safety Analysis Report Rev. 2 14-26 Sept. 2008 -
Specifications
1. The reactor shall be housed in a closed room designed to restrict leakage. The minimum free volume shall be 2,000 cubic meters.2. All air or other gas exhausted from the reactor room and associated experimental facilities shall be released to the environment a minimum of 26.5 meters above ground level.Basis Calculations in Chapter 13 of the Safety Analysis Report show that exposure of occupants of the Laboratory can be kept below 10 CFR Part 20 limits for occupational exposure under accident conditions if the room volume is 2,000 Calculations in Chapter 13 of the SAR based on release of radioactive effluent at ground level show that concentrations of radioactive materials are within limits of 10 CFR Part 20 for non-restricted areas during the accidents considered. Further calculations based on release at the stack height show a further reduction by a factor of 2.6 due to operation of the ventilation system and release of effluent at a height of 26.5m.Ventilation Systems Applicability This specification applies to the operation of the reactor laboratory ventilation system.Objective The objective is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation.
SpecificationsThe reactor shall not be operated unless the laboratory ventilation system is in operation, except for periods of time not to exceed two days, to permit repairs of the system.Basis It is shown in the SAR Chapter 11 that Argon-41 release at zero stack height results in concentrations less than the concentrations permitted for non-restricted areas.Further, the calculations indicate that operation of the ventilation system reduces the concentration to which the public would be exposed by a factor of 10 below this limit.Exposures in the event of a fuel element cladding leak are also calculated based on UWNR Safety Analysis Report Rev. 2 14-27 Sept. * *
Specifications
1. The reactor shall be housed in a closed room designed to restrict leakage. The minimum free volume shall be 2,000 cubic meters. 2. All air or other gas exhausted from the reactor room and associated experimental facilities shall be released to the environment a minimum of 26.5 meters above ground level. Calculations in Chapter 13 of the Safety Analysis Report show that exposure of occupants of the Laboratory can be kept below 10 CFR Part 20 limits for occupational exposure under accident conditions if the room volume is 2,000 m 3* Calculations in Chapter 13 of the SAR based on release of radioactive effluent at ground level show that concentrations of radioactive materials are within limits of 10 CFR Part 20 for non-restricted areas during the accidents considered.
Further calculations based on release at the stack height show a further reduction by a factor of2.6 due to operation of the ventilation system and release of effluent at a height of26.5m. TS 3.5 Ventilation Systems Applicability This specification applies to the operation of the reactor laboratory ventilation system. Objective The objective is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation.
Specifications The reactor shall not be operated unless the laboratory ventilation system is in operation, except for periods of time not to exceed two days, to permit repairs of the system. It is shown in the SAR Chapter 11 that Argon-41 release at zero stack height results in less than the concentrations permitted for non-restricted areas. Further, the calculations indicate that operation of the ventilation system reduces the concentration to which the public would be exposed by a factor of 10 below this limit . Exposures in the event of a fuel element cladding leak are also calculated based on UWNR Safety Analysis Report Rev. 2 14-27 Sept. 2008 non-operation of the ventilation system. Therefore, operation of the reactor with the ventilation system shut down in order to make repairs assures the degree of control on which the calculations are, based.3.6 Emergency Power Emergency power systems are not required for this facility.TS Radiation Monitoring Systems and Effluents TS 3.7.1 Monitoring Systems ApplicabilityThis specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.
Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor.Specifications The reactor shall not be operated unless the radiation monitoring channels listed in Table 3.7.1 are operable.Table 3.7.1 Radiation Monitoring SystemsRadiation Monitoring Function Number Channels*Area Radiation Monitor Monitor radiation levels within the reactor room 3 Exhaust Gas Radiation Monitor radiation levels in the exhaust air stack 1 Monitor Exhaust Particulate Radiation Monitor radiation levels in the exhaust air stack Monitor Environmental Radiation TLD dosimeters evaluated on a quarterly basis 4 Monitors record exposure in area surrounding the stack* For periods of time for maintenance to the radiation monitoring channels, the intent of this specification will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be kept under visual observation.
UWNRSafety Analysis Report Rev. 2 14-28 Sept.
non-operation of the ventilation system. Therefore, operation of the reactor with the ventilation system shut down in order to make repairs assures the degree of control on which the calculations are. based. TS 3.6 Emergency Power Emergency power systems are not required for this facility.
TS 3.7 Radiation Monitoring Systems and Effluents TS 3.7.1 Monitoring*Systems Applicability This specification applies to the radiation monitoring information which must be available to the reactor operator during reactor operation.
Objective The objective is to assure that sufficient radiation monitoring information is available to the operator to assure safe operation of the reactor. Specifications The reactor shall not be operated unless the radiation monitoring channels listed in Table 3.7.1 are operable.
a e .. a la IOn om onng )ys ems T bl 37 1 R d* f M *t* S t Radiation Monitoring Function Number Channels
Area Radiation Monitor Monitor radiation levels within the reactor room 3 Exhaust Gas Radiation Monitor radiation levels in the exhaust air stack 1 Monitor , Exhaust Particulate Radiation Monitor radiation levels in the exhaust air stack 1 Monitor Environmental Radiation TLD dosimeters evaluated on a quarterly basis 4 Monitors record exposure in area surrounding the stack For periods of time for maintenance to the radiation monitoring channels, the intent of this specification will be satisfied if they are replaced with portable gamma sensitive instruments having their own alarms or which shall be kept under visual observation.
UWNRBafety Analysis Report Rev. 2 14-28 Sept. 2008 * *
Basis The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings. The environmental monitors are placed in areas immediately surrounding the reactor laboratory to record actual dose that would have been delivered to a person continually present in the area.3.7.2 Effluent (Argon-41) Discharge Limit Applicability This specification applies to the concentration of Ar-41 which may be discharged from the facility.
Objective The objective is to insure that the health and safety of the public are not endangered by the discharge of Ar-41.Specifications The concentration of Ar-41 in the effluent gas from the facility, as diluted by atmospheric air in the lee of the facility as a result of the turbulent wake effect, shall not exceed lxi 0'8 [Ci/ml averaged over one year.Basis 10 CFR Part 20 Appendix B, Table specifies a limit of 1 x ptCi/ml for Ar-41. Chapter 11 and Appendix A of the SAR substantiates a release level of 3.3E-9 p.Ci/ml for a 3.54 meters/second (lowest monthly average) wind speed if all Ar-41 produced were continuously discharged.
The dilution factor bywhich emitted material is diluted is 2.5E-4 [tCi/ml per Ci/second discharged.
UWNR Safety Analysis Report Rev. 2 14-29Sept. 2008
*
The radiation monitors provide information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surroundings.
The environmental monitors are placed in areas immediately surrounding the reactor laboratory to record actual dose that would have been delivered to a person continually present in the area. TS 3.7.2 Effluent (Argon-41)
Discharge Limit Applicability This specification applies to the concentration of Ar-41 which may be discharged from the facility.
Objective The objective is to insure that the health and safety ofthe public are not endangered by the discharge of Ar-41. Specifications The concentration of Ar-41 in the effluent gas from the facility, as diluted by atmospheric air in the lee of the facility as a result of the turbulent wake effect, shall not exceed 1 x 10-8 !-lCi/ml averaged over one year. 10 CFR Part 20 Appendix B, Table II specifies a limit of 1 x 10-8 !-lCi/ml for Ar-41. Chapter 11 and Appendix A of the SAR substantiates a release level of 3.3E-9 !-lCi/ml for a 3.54 meters/second (lowest monthly average) wind speed if all Ar-41 produced were continuously discharged.
The dilution factor by which emitted material is diluted is 2.5E-4 !-lCi/ml per Ci/second discharged.
UWNR Safety Analysis Report Rev. 2 14-29 Sept. 2008 Experiments Applicability W This specification applies to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure.Reactivity Limits Specifications The reactor shall not be operated unless the following conditions governing experiments exist: 1. The reactivity worth of any single non-secured experiment shall not exceed 0.7% A 2. The reactivity worth of any single secured experiment shall not exceed 1.4% A k/k.Basis 1. This specification is intended to provide assurance that the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were to be suddenly inserted (SAR Chapter 13).2. The maximum worth of a single experiment is limited so that its removal from the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit.Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.
SAR accident analysis includes a sudden addition of 1.4% A k/k from firing the transient control rod while operating at the power level scram point, a more severe transient than that which could result from removal of a fixed experiment with the same reactivity worth.UWNR Safety Analysis Report Rev. 2 14-30 Sept. TS 3.8 Experiments Applicability This specification applies to experiments installed in the reactor and its experimental facilities.
Objective The objective is to prevent damage to the reactor or excessive release of radioactive materials in the event of an experiment failure. TS 3.8.1 Reactivity Limits Specifications The reactor shall not be operated unless the following conditions governing experiments exist: 1. The reactivity worth of any single non-secured experiment shall not exceed 0.7% !:::. kJk. 2. The reactivity worth of any single secured experiment shall not exceed 1.4%!:::.
kJk. 1. This specification is intended to provide assurance that the worth of a single unfastened experiment will be limited to a value such that the safety limit will not be exceeded if the positive worth of the experiment were to be suddenly inserted (SAR Chapter 13). 2. The maximum worth of a single experiment is limited so that its removal from the cold critical reactor will not result in the reactor achieving a power level high enough to exceed the core temperature safety limit. Since experiments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power increase such that the reactor protective systems would act to prevent high power levels from being attained.
SAR accident analysis includes a sudden addition of 1.4%!:::.
k/k from firing the transient control rod while operating at the power level scram point, a more severe transient than that which could result from removal of a: fixed experiment with the same reactivity worth. UWNRSafety Analysis Report Rev. 2 14-30 Sept. 2008 * *
Materials Specifications
1. Explosive materials, such as gunpowder, TNT, nitroglycerin, or PETN, in quantities greater than 25 milligrams shall not be irradiated in the reactoror experimental facilities. Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.
2. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B of 10 CFR Part 20.3. In calculations pursuant to 2 above, the following assumptions shall be used: a. If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape.
b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of these vapors can escape.c. For materials whose boiling point is above 1307F and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape.d. An atmospheric dilution factor of 2.5 x ýiCi/ml per Ci/s for gaseous discharges from the facility.4. Each fueled experiment shall be controlled such that the total inventory ofiodine isotopes 131 through 135 in the experiment is no greater than 1.5 curies.UWNR Safety Analysis Report Rev. 2 14-31 Sept. TS 3.8.2 Materials
Specifications

1. Explosive materials, such as gunpowder, TNT, nitroglycerin, or PETN, in quantities greater than 25 milligrams shall not be irradiated in the reactor or experimental facilities.
Explosive materials in quantities less than 25 milligrams may be irradiated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.
2. Experiment materials, except fuel materials, which could off-gas, sublime, volatilize, or produce aerosols under (1) normal operating conditions of the experiment or reactor, (2) credible accident conditions in the reactor, or (3) possible accident conditions in the experiment shall be limited in activity such that if 100% of the gaseous activity or radioactive aerosols produced escaped to the reactor room or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the limit of Appendix B of 10 CFR Part 20. 3. In calculations pursuant to 2 above, the following assumptions shall be used: a. If the effluent from an experimental facility exhausts through a holdup tank which closes automatically on high radiation level, at least 10% of the gaseous activity or aerosols produced will escape. b. If the effluent from an experimental facility exhausts through a filter installation designed for greater than 99% efficiency for 0.3 micron particles, at least 10% of these vapors can escape. c. For materials whose boiling point is above 130°F and where vapors formed by boiling this material can escape only through an undisturbed column of water above the core, at least 10% of these vapors can escape. d. An atmospheric dilution factor of2.5 x 10-4 !lCi/ml per Ci/s for gaseous discharges from the facility.
4. Each fueled experiment shall be controlled such that the total inventory of iodine isotopes 131 through 135 in the experiment is no greater than 1.5 cunes . UWNR Safety Analysis Report Rev. 2 14-31 Sept. 2008 Basis 1. This specification is intended to prevent damage to reactor components resulting from failure of an experiment involving explosive materials.
2-3. These specifications are intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary of the UWNR. The dilution factor is based on computations reported in Chapter 11 and Appendix A of the Safety Analysis Report.4. The 1.5 curie limitation on iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than that allowed by 10 CFR Part 20 for an unrestricted area.
Experiment Failure and Malfunctions Specifications If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection of the capsule shall be performed to determine the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Reactor Director or his designated alternate and determined to be satisfactory before operation of the reactor is resumed.Basis Operation of the reactor with a failed capsule is prohibited to prevent damageto the reactor fuel or structure.
Failure of a capsule must be investigated to assure no damage has or will occur.Facility Specific LCOs There are no facility specific LCOs at this facility.UWNR Safety Analysis Report Rev. 2 14-32 Sept. 1. This specification is intended to prevent damage to reactor components resulting from failure of an'experiment involving explosive materials.
2-3. These specifications are intended to reduce the likelihood that airborne activities in excess of the limits of Appendix B of 10 CFR Part 20 will be released to the atmosphere outside the facility boundary of the UWNR. The dilution factor is based on computations reported in Chapter 11 and Appendix A of the Safety Analysis Report. 4. The 1.5 curie limitation on iodine 131 through 135 assures that in the event of failure ofa fueled experiment leading to total release of the iodine, the exposure dose at the exclusion area boundary will be less than . that allowed by 10 CFR Part 20 for an unrestricted area. TS 3.8.3 Experiment Failure and Malfunctions Specifications
If a capsule fails and releases material which could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection of the capsule shall be performed to determine the consequences and need for
corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Reactor Director or his designated alternate and determined to be satisfactory before operation of the reactor is resumed. Operation of the reactor with a failed capsule is prohibited to prevent damage to the reactor fuel or structure.
Failure of a capsule must be investigated to assure no damage has or will occur. TS 3.9 Facility Specific LeOs There are no facility specific LCOs at this facility.
UWNR Safety Analysis Report Rev. 2 14-32 Sept. 2008
4 In accordance with section 4.0 of Standard ANSI/ANS-15.1, the following terms for average surveillance intervals shall allow, for operational flexibility only, maximum times between surveillance intervals as indicated below unless otherwise specified within the specification.
Five-year interval not to exceed six years.* Biennial interval not to exceed two and one-half years.* Annual interval not to exceed 15 months.* Semiannual interval not to exceed seven and one-half months.* Quarterly interval not to exceed four months.* Monthly interval not to exceed six weeks.* Weekly interval not to exceed ten days* Daily interval must be done within the calendar day.Scheduled surveillances, except those specifically required when the reactor is shut down, may be deferred during shutdown periods, but be completed prior to subsequent reactor startup unless operation is required for the performance of the surveillance.
Scheduledsurveillances which cannot be performed with the reactor operating may be deferred until a planned reactor shutdown.
If the reactor is not operational in a particular mode, surveillances required specifically for that mode may be deferred until the reactor becomes operational in that mode.General Applicability This specification applies to the surveillance requirements of any system related to reactor safety.Objective The objective is to verify the proper operation of any system related to reactor safety aftermaintenance or modification of the system.
Specifications Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or tospecifications approved by the Reactor Safety Committee. A system shall not be considered operable until after it is successfully tested.UWNR Safety Analysis Report Rev. 2 14-33Sept. 2008* *
TS4 SURVEILLANCE REQUIREMENTS In accordance with section 4.0 of Standard ANSIIANS-15.1, the following terms for average surveillance intervals shall allow, for operational flexibility only, maximum times between surveillance intervals as indicated below unless otherwise specified within the specification.
Five-year interval not to exceed six years.
Biennial interval not to exceed two and one-half years.
Annual interval not to exceed 15 months.
Semiannual interval not to exceed seven and one-half months.
Quarterly interval not to exceed four months.
Monthly interval not to exceed six weeks.
Weekly interval not to exceed ten days
Daily interval must be done within the calendar day. Scheduled surveillances, except those specifically required when the reactor is shut down, may be deferred during shutdown periods, but be completed prior to subsequent reactor startup unless operation is required for the performance of the surveillance.
Scheduled surveillances which cannot be performed with the reactor operating may be deferred until a planned reactor shutdown.
If the reactor is not operational in a particular mode, surveillances required specifically for that mode may be deferred until the reactor becomes operational in that mode . General Applicability This specification applies to the surveillance requirements of any system related to reactor safety. Objective The objective is to verify the proper operation of any system related to reactor safety after maintenance or modification of the system. Specifications Any additions, modifications, or maintenance to the ventilation system, the core and its associated support structure, the pool or its penetrations, the pool coolant system, the rod drive mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to which the systems were originally designed and fabricated or to specifications approved by the Reactor Safety Committee.
A system shall not be considered operable until after it is successfully tested . UWNR Safety Analysis Report Rev. 2 14-33 , Sept. 2008 Basis This specification relates to changes in reactor systems which could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.4.1 Reactor Core Parameters Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of reactor core parameters.
Objective The objective is to verify the core parameters which are directly related to reactor safety.Specifications
1. Excess reactivityExcess reactivity shall be determined at least annually and after changes in either the core, in-core experiments, or control elements for which the predicted change in reactivity exceeds the absolute value of the specifiedshutdown margin.
2. Shutdown margin The shutdown margin shall be determined at least annually and after changesin either the core, in-core experiments, or control elements.
3. Pulse limitsThe reactor shall be pulsed semiannually to compare fuel temperature measurements (if an operating fuel thermocouple is available) and peak power levels with those of previous pulses of the same reactivity value.4. Core configurationEach planned change in core configuration shall be determined to meet the requirements of Sections 3.1(4) and 5.3 of these specifications before the core is loaded.5. Reactivity Coefficients Power defect and pulsing characteristics shall be measured during startup testing of cores containing different fuel compositions and compared to predictions in the Safety Analysis Report.UWNR Safety Analysis Report Rev. 2 14-34 Sept. This specification relates to changes in reactor systems which could directly affect the safety of the reactor. As long as changes or replacements to these systems continue to meet the original design specifications, then it can be assumed that they meet the presently accepted operating criteria.
TS 4.1 Reactor Core Parameters Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of reactor core parameters.
Objective The objective is to verify the core parameters which are directly related to reactor safety. Specifications
1. Excess reactivity Excess reactivity shall be determined at least annually and after changes in either the core, in-core experiments, or control elements for which the predicted change in reactivity exceeds the absolute value of the specified shutdown margin. 2. Shutdown margin The shutdown margin shall be determined at least annually and after changes in either the core, in-core experiments, or control elements.
3. Pulse limits The reactor shall be pulsed semiannually to compare fuel temperature measurements (if an operating fuel thermocouple is available) and peak power levels with those of previous pulses of the same reactivity value. 4. Core configuration Each planned change in core configuration shall be determined to meet the requirements of Sections 3.1 (4) and 5.3 of these specifications before the core is loaded. 5. Reactivity Coefficients Power defect and pulsing characteristics shall be measured during startup testing of cores containing different fuel compositions and compared to predictions in the Safety Analysis Report. ' UWNR Safety Analysis Report Rev. 2 14-34 Sept. 2008 * * *
6. Fuel Parameters
a. All fuel elements shall be inspected visually for damage or deterioration annually.b. Uninstrumented fuel elements which have been resident in the core during the previous year shall be measured for length and sagitta annually. Fuel elements shall not be added to a core unless a measurement of length and sagitta has been completed within the previous fifteen months.c. Fuel elements in the hottest assumed location, as well as representative elements in each of the rows, shall be measured for possible damage in the event there is indication that the Limiting Safety System Setting may havebeen exceeded.
Basis 1-2. Annual measurements, coupled with measurements made after changes thatcan affect reactivity values provide adequate assurance that core behavior will be as analyzed.
The reactivity values in FLIP fuel change very slowly with fuel burnup.3. Semiannual verifications assure no changes in behavior are resulting from fuel characteristic changes.4. Checking contemplated core configurations against requirements will prevent inadvertent loading of cores which do not meet power peaking restraints imposed by composition restrictions.
5. Measurements made during core startup testing are sufficient to assure core behavior will be as analyzed.6. Annual inspection of the TRIGA fuel has been shown adequate to assure fuel element integrity through a long history of standard operation.
UWNR Safety Analysis Report Rev. 2 14-35 Sept. * *
6. Fuel Parameters
a. All fuel elements shall be inspected visually for damage or deterioration annually.
b. Uninstrumented fuel elements which have been resident in the core during the previous year shall be measured for length and sagitta annually.
Fuel elements shall not be added to a core unless a measurement of length and sagitta has been completed within the previous fifteen months. c. Fuel elements in the hottest assumed location, as well as representative elements in each of the rows, shall be measured for possible damage in the event there is indication that the Limiting Safety System Setting may have been exceeded.
1-2. Annual measurements, coupled with measurements made after changes that can affect reactivity values provide adequate assurance that core behavior will be as analyzed.
The reactivity values in FLIP fuel change very slowly with fuel bumup. 3. Semiannual verifications assure no changes in behavior are resulting from fuel characteristic changes . 4. Checking contemplated core configurations against requirements will prevent inadvertent loading of cores which do not meet power peaking restraints imposed by composition restrictions.
5. Measurements made during core startup testing are sufficient to assure core behavior will be as analyzed.
6. Annual inspection of the TRIGA fuel has been shown adequate to assure fuel element integrity through a long history of standard operation . UWNR Safety Analysis Report Rev. 2 14-35 Sept. 2008 4.2 Reactor Control and Safety Systems Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of the control and safety systems.Objective The objective is to verify the performance and operability of those systems and components which are directly related to reactor safety.Specifications
1. Reactivity worth of control elements The reactivity worth of control elements shall be determined upon substantiative changes in core composition or arrangement and annually thereafter.
2. Control element withdrawal and insertion speeds Control element drive withdrawal and insertion speeds shall be measured annually and following maintenance to the control element or the control element drive mechanism.
3. Transient Rod and Associated Mechanism The transient rod drive cylinder and associated air supply system shall be inspected, cleaned, and lubricated as necessary annually.4. Scram times of control and safety elements The scram time for all scrammable control elements shall be measured annually and following maintenance to the control elements or their drives.5. Scram and Power Measuring Channels a. A channel test of each Reactor Safety System measuring channel in Table 3.2.4 items (1) through (4) and the interlocks in Table 3.2.5 required for the intended modes of operation shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before each day's operation or prior to each operation extending more than one day.b. A channel test of items (5) and,(6) in Table 3.2.4 shall be performed semi-annually.6. Operability Tests This concern is covered by the General Surveillance criterion at the beginning of this section.UWNR Safety Analysis Report Rev. 2 14-36 Sept. TS 4.2 Reactor Control and Safety Systems Applicability These specifications apply to the surveillance requirements for measurements, tests, and calibrations of the control and safety systems. Objectiv,e The objective is to verify the performance and operability of those systems and components which are directly related to reactor safety. Specifications
1. Reactivity worth of control elements \ The reactivity worth of control elements shall be determined upon substantiative changes in core composition or arrangement and annually thereafter.
2. Control element withdrawal and insertion speeds Control element drive withdrawal and insertion speeds shall be measured annually and following maintenance to the control element or the control element drive mechanism.
3. Transient Rod and Associated Mechanism The transient rod drive cylinder and associated air supply system shall be inspected, cleaned, and lubricated as necessary annually.
4. Scram times of control and safety elements The scram time for all scrammable control elements shall be measured annually and following maintenance to the control elements or their drives. 5. Scram and Power Measuring Channels
a. A channel test of each Reactor Safety System measuring channel in Table 3.2.4 items (1) through (4) and the interlocks in Table 3.2.5 required for the intended modes of operation shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before each day's operation or prior to each operation extending more than one day. b. A channel test of items (5) and,(6) in Table 3.2.4 shall be performed annually.
6. Operability Tests This concern is covered by the General Surveillance criterion at the beginning of this section. UWNR Safety Analysis Report Rev. 2 14-36 Sept. 2008 * *
Thermal Power Calibration-Forced Convection Not applicable to this reactorThermal Power Calibration-Natural ConvectionChannel Calibration shall be made of the power level monitoring channelsthe calorimetric method upon substantiative changes in core compositionor arrangement and annually thereafter.Control Element Inspection The control elements shall be visually inspected for deterioration biennially.
Basis 1 Control element worths change slowly unless the core arrangement is changed, so annual measurement is sufficient to assure safety.2. Control element insertion or withdrawal speeds are fixed the motor design and thus do not change except for extreme binding conditions within the drive.Transient rod drive and air supply includes filtration and 'lubrication, so an annualcheck coupled with pre-startup checks is sufficient to assure operabilty.
4. Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.The items through 4 in the table are essential safety equipment and thus should be checked frequently, even though no failures have been observed checkout in nearly years of operation. Frequent testing is unnecessary for item a simple float switch which is very unlikely to fail, and has performed for nearly years without a failure. Testing item the high voltage monitor scram, results in changing the voltage to the neutron detectors.
This introduces step changes into the signal circuits of the measuring channels which can lead to long recovery times and a significant increase in failures of the measuring channels.
Further, since the checkout of the linear safety channels is a source check, if high voltage were lost that check would not be possible if the voltage had been lost.The general requirement for checks of equipment operability after maintenance or modification of systems will reveal any loss of safety functions due to the maintenance or modification.The power level channel calibration will assure that the reactor will be operated atthe proper power levels.Annual checks in other TRIGA reactors and for nearly years in this reactorhave been sufficient to insure no failures due to deterioration.
UWNR Safety Analysis Report Rev. 2 1-7Sp.20 14-37 Sept. * *
7. Thermal Power Calibration-Forced Convection Not applicable to this reactor 8. Thermal Power Calibration-Natural Convection A Channel Calibration shall be made of the power level monitoring channels by the calorimetric method upon substantiative changes in core composition or arrangement and annually thereafter.
9. Control Element Inspection The control elements shall be visually inspected for deterioration biennially.
1. Control element worths change slowly unless the core arrangement is changed, so annual measurement is sufficient to assure safety. 2. Control element insertion or withdrawal speeds are fixed by the motor design and thus do not change except for extreme binding conditions within the drive. 3. Transient rod drive and air supply includes filtration and,lubrication, so an annual check coupled with pre-startup checks is sufficient to assure operabilty.
4. Measurement of the scram time on an annual basis is a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.
5. The items 1 through 4 in the table are essential safety equipment and thus should be checked frequently, even though no failures have been observed by checkout in nearly 50 years of operation.
Frequent testing is unnecessary for item 5, a simple float switch which is very unlikely to fail, and has performed for nearly 50 years without a failure. Testing item 6, the high voltage monitor scram, results in changing the voltage to the neutron detectors.
This introduces step changes into the signal circuits of the measuring channels which can lead to long recovery times and a significant increase in failures of the measuring channels.
Further, since the checkout of the linear safety channels is a source check, if high voltage were lost that check would not be possible if the voltage had been lost. 6. The general requirement for checks of equipment operability after maintenance or modification of systems will reveal any loss of safety functions due to the maintenance or modification.
8. The power level channel calibration will assure that the reactor will be operated at the proper power levels. 9. Annual checks in other TRIGA reactors and for nearly 50 years in this reactor have been sufficient to insure no failures due to deterioration.
UWNR Safety Analysis Report Rev. 2 14-37 Sept. 2008 4.3 Coolant SystemsApplicability This specification applies to the reactor pool water.
Objective The objective is to assure the water quality and radioactivity is within the defined limits Specifications The pool water conductivity and radioactivity shall be measured quarterly.
Basis Pool water conductivity is continuously monitored, but would be manually monitored on a quarterly basis if the instruments failed. Radioactivity is indirectly monitored by an area radiation monitor near the demineralizer bed, so gross activity increases would be detected immediately. Experience with TRIGA reactors indicates the earliest detection of fuel clad leaks is usually from airborne activity, rather than pool water activity.
The quarterly measurement can identify specific radionuclides.4.4 Confinement No surveillances are required.4.5 Ventilation Systems ApplicabilityThis specification applies to the building confinement ventilation system.Objective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment.
Specifications It shall be verified quarterly and following repair or maintenance that the ventilation system is operable.UWNR Safety Analysis Report Rev. 2 14-38 Sept. TS 4.3 Coolant Systems Applicability This specification applies to the reactor pool water. Objective The objective is to assure the water quality and radioactivity is within the defined limits Specifications The pool water conductivity and radioactivity shall be measured quarterly.
Pool water conductivity is continuously monitored, but would be manually monitored on a quarterly basis if the instruments failed. Radioactivity is indirectly monitored by an area radiation monitor near the demineralizer bed, so gross activity increases would be detected immediately.
Experience with TRIGA reactors indicates the earliest detection of fuel clad leaks is usually from airborne activity, rather than pool water(activity.
The quarterly measurement can identify specific radionuclides.
TS 4.4 Confinement No surveillances are required.
TS 4.5 Ventilation Systems Applicability This specification applies to the building confinement ventilation system. Objective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment.
Specifications It shall be verified quarterly and following repair or maintenance that the ventilation system is operable.
UWNR Safety Analysis Report Rev. 2 14-38 Sept. 2008 * *
Basis Over 30 years of experience with the previous ventilation system has demonstrated that testing the system quarterly is sufficient to assure the proper operation of the system and control of the release of radioactive material.
The new ventilation system is expected to exceed the reliability of the previous system so quarterly testing is still appropriate.4.6 Emergency Electrical Power Systems Not Applicable.4.7 Radiation Monitoring Systems and Effluents4.7.1 Radiation Monitoring Systems Applicability This specification applies to the surveillance requirements for the arearadiation monitoring equipment and the stack air monitoring system.Objective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.
Specifications The radiation monitoring and stack monitoring systems shall be calibrated annually and shall be verified to be operable by monthly source checks or channel tests.Basis Experience has shown that monthly verification of area radiation monitor operability and setpoints in conjunction with the downscale-failure feature of the instrument is adequate to assure operability.
Annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span. Annual calibrations and monthly source or channel checks of the stack particulate and gaseous monitors, along with the high or low flow alarms associated with the monitorassure operability and accuracy.
UWNR Safety Analysis Report Rev. 2 14-3.9 Sept. * *
Over 30 years of experience with the previous ventilation system has demonstrated that testing the system quarterly is sufficient to assure the proper operati'on of the system and control of the release of radioactive material.
The new ventilation system is expected to exceed the reliability ofthe previous system so quarterly testing is still appropriate.
TS 4.6 Emergency Electrical Power Systems Not Applicable.
TS 4.7 Radiation Monitoring Systems and Effluents TS 4.7.1 Radiation Monitoring Systems Applicability This specification applies to the surveillance requirements for the area radiation monitoring equipment and the stack air monitoring system. Objective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.
Specifications The radiation monitoring and stack monitoring systems shall be calibrated annually and shall be verified to be operable by monthly source checks or channel tests . Experience has shown that monthly verification of area radiation monitor operability and setpoints in conjunction with the downscale-failure feature of the instrument is adequate to assure operability.
Annual calibration is adequate to correct for any variation in the system due to a change of operating characteristics over a long time span. Annual calibrations and monthly source or channel checks of the stack particulate and gaseous monitors, along with the high or low flow alarms associated with the monitor assure operability and accuracy . UWNR Safety Analysis Report Rev. 2 14-39 Sept. 2008 4.7.2 Effluents Applicability This specification applies to gaseous and liquid discharges from the reactor laboratory.
Objective The objective is to assure that ALARA and 10 CFR Part 20 limits are observed.Specifications Liquid radioactive waste discharged to the sewer system shall be sampled for radioactivity to assure levels are below applicable limits before discharge.
Results of the measurements shall be recorded and reported in the Annual Report.The total annual release of gaseous radioactivity to the environment shall be recorded and reported in the Annual Report.Basis Liquid waste releases are batch releases, so the liquid can be sampled before release. Air activity discharged is continuously recorded and the integrated release is reported.4.8 Experiments No surveillances are required.4.9 Facility-Specific Surveillance Not applicable.
There is no facility-specific surveillance.
UWNR Safety Analysis Report Rev. 2 14-40 Sept. 2008 TS 4.7.2 Effluents
Applicability This specification applies to gaseous and liquid discharges from the reactor laboratory.
Objective The objective is to assure that ALARA and 10 CFR Part 20 limits are observed.
Specifications Liquid radioactive waste discharged to the sewer system shall be sampled for radioactivity to* assure levels are below applicable limits before discharge.
Results of the measurements shall be recorded and reported in the Annual Report. The total annual release of gaseous radioactivity to the environment shall be recorded and reported in the Annual Report. Liquid waste releases are batch releases, so the liquid can be sampled before release. Air activity discharged is continuously recorded and the integrated release is reported.
TS 4.8 Experiments No surveillances are required.
TS 4.9 Facility-Specific Surveillance Not applicable.
There is no facility-specific surveillance. ( UWNR Safety Analysis Report Rev. 2 14-40 Sept. 2008 * *
Site and Facility Description Specifications
1. The reactor shall be housed in a closed room designed to restrict leakage. The minimum free volume shall be 2,000 cubic meters.
2. All air or other gas exhausted from the reactor room and the Beam Port and Thermal Column Ventilation System shall be released to the environment a minimum of 26.5 meters above ground level.5.2 Reactor Coolant System Specifications
1. The reactor core shall be cooled by natural convective water flow.
2. The pool water inlet pipe to the demineralizer shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped with a check valve to prevent inadvertent draining of the pool.3. Diffuser and other auxiliary systems pumps shall be located no more than 15 feet below the top of the reactor pool.4. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool.5. A pool level alarm shall indicate loss of coolant if the pool level drops approximately one foot below normal level.UWNR Safety Analysis Report Rev. 2 14-41 Sept. * *
TS 5 DESIGN FEATURES TS 5.1 Site and Facility Description Specifications
1. The reactor shall be housed in a closed room designed to restrict leakage. The minimum free volume shall be 2,000 cubic meters. 2. All air or other gas exhausted from the reactor room and the Beam Port and Thermal Column Ventilation System shall be released to the environment a minimum of 26.5 meters above ground level. TS 5.2 Reactor Coolant System Specifications
1. The reactor core shall be cooled by natural convective water flow. 2. The pool water inlet pipe to the demineralizer shall not extend more than 15 feet into the top of the reactor pool when fuel is in the core. The outlet pipe from the demineralizer shall be equipped with a check valve to prevent inadvertent draining of the pool. 3. Diffuser and other auxiliary systems pumps shall be located no more than 15 feet below the top of the reactor pool. 4. All other piping and pneumatic tube systems entering the pool shall have siphon breakers and valves or blind flanges which will prevent draining more than 15 feet of water from the pool. 5. A pool level alarm shall indicate loss of coolant if the pool level drops approximately one foot below normal level. UWNR Safe,ty Analysis Report Rev. 2 14-41 Sept. 2008 TS Reactor Core and Fuel Specifications
1. TRIGA-FLIP Fuel The individual unirradiated FLIP fuel elemenfs shall have the following characteristics:
a. Uranium content: maximum of 9 Wt-% enriched to nominal 70% Uranium 235.b. Hydrogen-to-zirconium atom ratio (in the ZrH,): nominal 1.6 H atoms to 1.0 Zr atoms.c. Natural erbium content (homogeneously distributed):
nominal 1.5 Wt-%.d. Cladding:
304 stainless steel, nominal 0.020 inch thick.e. Identification:
Top pieces of FLIP fuel bundles will have characteristic markings to allow visual identification of FLIP fuel employed in mixed cores.2. Standard TRIGA fuel The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:
a. Uranium content: maximum of 9.0 Wt-% enriched to a nominal 20%Uranium 235.b. Hydrogen-to-zirconium atom ratio (in the ZrHx): nominal 1.7 H atoms to 1.0 Zr atoms.c. Cladding:
304 stainless steel, nominal 0.020 inch thick.UWNR Safety Analysis Report Rev. 2 14-42 Sept. TS 5.3 Reactor Core and Fuel Specifications
1. TRIGA-FLIP Fuel The individual unirradiated FLIP fuel elements shall have the following characteristics:
a. Uranium content: maximum of 9 Wt-% enriched to nominal 70% Uranium 235. b. Hydrogen-to-zirconium atom ratio (in the ZrHJ: nominal 1.6 H atoms to 1.0 Zr atoms. c. Natural erbium content (homogeneously distributed):
nominal 1.5 Wt-%. d. Cladding:
304 stainless steel, nominal 0.020 inch thick. e. Identification:
Top pieces of FLIP fuel bundles will have characteristic markings to allow visual identification of FLIP fuel employed in mixed cores. j 2. Standard TRIGA fuel The individual unirradiated standard TRIGA fuel elements shall have the following characteristics:
a .. Uranium content: maximum of9.0 Wt-% enriched to a nominal 20% Uranium 235. b. Hydrogen-to-zirconium atom ratio (in the ZrHJ: nominal 1.7 H atoms to 1.0 Zr atoms. c. Cladding:
304 stainless steel, nominal 0.020 inch thick. UWNR Safety Analysis Report Rev. 2 14-42 Sept. 2008 * *
5.4 Reactor Core Specifications
1. The core shall be an arrangement of TRIGA uranium-zirconium hydride fuel-moderator bundles positioned in the reactor grid plate.2. The TRIGA core assembly may be standard, FLIP, or a combination thereof (mixed core), provided that any FLIP fuel be comprised of at least thirty-five (35)fuel elements, located in a contiguous, central region.3. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.4. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water.5.5 Control Elements Specifications
1. The safety blades shall be constructed of boral plate and shall have scram capability.
2. The regulating blade shall be constructed of stainless steel.3. The transient rod shall contain borated graphite or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad.
The transientcontrol rod shall have scram capability and may incorporate an aluminum or air follower.TS 5.6 Fissionable Material Storage Specifications
1. All fuel elements shall be stored in a geometrical array where the value of k-effective is less than 0.8 for all conditions of moderation.
2. Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.
UWNR Safety Analysis Report Rev. 2 14-43 Sept. * *
TS 5.4 Reactor Core Specifications
1. The core shall be an arrangement ofTRIGA uranium-zirconium hydride moderator bundles positioned in the reactor grid plate. 2. The TRIGA core assembly may be standard, FLIP, or a combination thereof (mixed core), provided that any FLIP fuel be comprised of at least thirty-five (35) fuel elements, located in a contiguous, central region. 3. The reactor shall not be operated with a core lattice position vacant except for positions on the periphery of the core assembly.
4. The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water. TS 5.5 Control Elements Specifications
1. The safety blades shall be constructed of boral plate and shall have scram capability . 2. The regulating blade shall be constructed of stainless steel. 3. The transient rod shall contain borated graphite or boron and its compounds in a solid form as a poison in an aluminum or stainless steel clad. The transient control rod shall have scram capability and may incorporate an aluminum or air follower.
TS 5.6 Fissionable Material Storage Specifications
1. All fuel elements shall be stored in a geometrical array where the value effective is less than 0.8 for all conditions of moderation.
2. Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values . UWNR Safety Analysis Report Rev. 2 14-43 Sept. 2008 This page is intentionally left blank.UWNR Safety-Analysis Report Rev. 2 14-44 Sept.
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UWNR Safety.Analysis Report Rev. 2 14-44 Sept. 2008 ADMINISTRATIVE Organization Structure The reactor facility shall be an integral part of the Engineering Physics Department of the College of Engineering of the University of Wisconsin-Madison. The reactor shall be related to the University structure as shown in Figure 14-1.The Radiation Safety office performs audit functions for both the Radiation Safety Committee and the Reactor Safety Committee and reports to both committees as well as to the Reactor Director.
Responsibility The Reactor Director is responsible for all activities at the facility, including licensing, security, emergency preparedness, and maintaining radiation exposures as low as reasonably achievable.The reactor facility shall be under the direct control of a Reactor Supervisor designated by the Reactor Director.
The Reactor Supervisor shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, procedures, and the requirements of the Radiation Safety Committee and the Reactor Safety Committee.
UWNR Safety Analysis Report Rev. 2 14-45 Sept. * *
TS 6. ADMINISTRATIVE CONTROLS TS 6.1 Organization TS 6.1.1 Structure TS 6.1.2 The reactor facility shall be an integral part of the Engineering Physics Department of the College of Engineering of the University of Madison. The reactor shall be related to the University structure as shown in Figure 14-1. The Radiation Safety office performs audit functions for both the Radiation Safety Committee and the Reactor Safety Committee and reports to both committees as well as to the Reactor Director.
Responsibility The Reactor Director is responsible for all activities at the facility, including licensing, security, emergency preparedness, and maintaining radiation exposures as low as reasonably achievable.
The reactor facility shall be under the direct control of a Reactor Supervisor designated by the Reactor Director.
The Reactor Supervisor shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, procedures, and the requirements of the Radiation Safety Committee and the Reactor Safety Committee . UWNR Safety Analysis Report Rev. 2 14-45 Sept. 2008 BOARD OF UNIVERSITY OF WISCONSIN RADIATION SAFETY COMMITTEE University Safety Department Radiation Safety Office CHAIRMAN ENGINEERING PHYSICS DEPARTMENT (ANSI/ANS-15.4 Level 1),REACTOR SAFETY COMMITTEEREACTOR DIRECTOR (ANSI/ANS-15.4 Level 2)REACTOR SUPERVISOR (SRO)(ANSI/ANS-1 5.1 Level 3)ALTERNATE SUPERVISORS (SRO)(ANSI/ANS-15.1 Level 3)REACTOR OPERATORS (RO)(ANSI/ANS-15.1 Level 4)Figure 14-1 Organization Chart UWNR Safety Analysis Report Rev. 2 14-46 Sept. 2008BOARD OF REGENTS CHANCELLOR-MADISON CAMPUS
UNIVERSITY OF WISCONSIN RADIATION SAFETY COMMITTEE University Safety Department Radiation Safety Office CHAIRMAN ENGINEERING PHYSICS DEPARTMENT (ANSI/ANS-1S.4 Level 1 ) ,REACTOR SAFETY COMMITIEE f---* REACTOR DIRECTOR (ANSI/ANS-1S.4 Level 2) REACTOR SUPERVISOR (SRO) (ANSI/ANS-1S.1 Level 3) ALTERNATE SUPERVISORS (SRO) (ANSI/ANS-1S.1 Level 3) REACTOR OPERATORS (RO) (ANSI/ANS-1S.1 Level 4) Figure 14-1 Organization Chart
UWNR Safety Analysis Report Rev. 2 14-46 Sept. 2008
Staffing 1. The minimum staffing when the reactor is not secured shall be: a. A licensed reactor operator in the control room (if senior operatorlicensed, may also be the person required in c).b. A second designated person present at the facility capable of carrying out prescribed written instructions.
c. A designated senior reactor operator shall be readily available at the facility or on call.2. A list of reactor facility personnel by name and telephone number shall bereadily available in the control room for use by the operator.3. A licensed senior reactor operator shall be present at the facility for:
a. Initial startup and approach to power.b. All fuel handling or control-element relocations.
c. Relocation of any in-core experiment with a reactivity worth greater than 0.7% AK/K.d. Recovery from unplanned or unscheduled shutdown or significant power reduction.6.1.4 Selection and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of ANSI/ANS-15.4-1988 Sections 4-6.Review and Audit There shall be a Reactor Safety Committee which shall review and audit reactor operations to assure that the facility is operated in a manner consistent with public safety and within the conditions of the facility license.UWNR Safety Analysis Report Rev. 2 14-47 Sept. TS 6.1.3 Staffing
1. The minimum staffing when the reactor is not secured shall be: *
a. A licensed reactor operator in the control room (if senior operator licensed, may also be the person required in c). b. A second designated person present at the facility capable of carrying out prescribed written instructions.
c. A designated senior reactor operator shall be readily available at the facility or on call. 2. A list of reactor facility personnel by name and telephone number shall be readily available in the control room for use by the operator.
3. A licensed senior reactor operator shall be present at the facility for: a. Initial startup and approach to power. b. All fuel handling or control-element relocations.
c. Relocation of any in-core experiment with a reactivity worth greater than 0.7%
d. Recovery from unplanned or unscheduled shutdown or significant power reduction.
TS 6.1.4 Selection and Training of Personnel The selection, training, and requalification of operations personnel shall meet or exceed the requirements of ANSIIANS-15.4-1988 Sections 4-6. TS 6.2 Review and Audit There shall be a Reactor Safety Committee which shall review and audit reactor operations to assure that the facility is operated in a manner consistent with public safety and within the conditions of the facility license . UWNR Safety Analysis Report Rev. 2 14-47 Sept. 2008
Composition and Qualifications The Committee shall be composed of a least six members, one of whom shall be a Health, Physicist from the University of Wisconsin Safety Department Radiation Safety Office. The Committee shall collectively possess expertise in the following disciplines:
1. Reactor Physics;2. Heat transfer and fluid mechanics;
3. Metallurgy
4. Instruments and Control Systems;5. Chemistry and Radio-chemistry;
6. Radiation Safety.
Charter and Rules The Committee shall meet at least annually.The Committee shall formulate written standards regarding the activities of the full committee; minutes, quorum, telephone polls for approvals not requiring a formal meeting, and subcommittees.
Review Function The responsibilities of the Reactor Safety Committee shall include, but are not limited to, the following:Review and approval of experiments utilizing the reactor facilities;
2. Review and approval of all proposed changes to the facility, procedures, license, and technical specifications;
3. Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications;
4. Review of abnormal performance of plant equipment and operating anomalies having safety significance; and UWNR Safety Analysis Report Rev. 2 14-48 Sept. TS 6.2.1 Composition and Qualifications , The Committee shall be composed of a least six members, one of whom shall be a Health, Physicist from the University of Wisconsin Safety Department Radiation Safety Office. The Committee shall collectively possess expertise in the following disciplines:
1. Reactor Physics; 2. Heat transfer and fluid mechanics;
3. Metallurgy A. Instruments and Control Systems; 5. Chemistry and Radio-chemistry;
6. Radiation Safety. TS6.2.2 Charter and Rules The Committee shall meet at least annually.
The Committee shall formulate written standards regarding the activities of the full committee; minutes, quorum, telephone polls for approvals not requiring a formal meeting, and subcommittees.
TS 6.2.3 Review Function The responsibilities of the Reactor Safety Committee shall include, but are not limited to, the following:
1. Review and approval of utilizing the reactor facilities;
2. Review and approval of all proposed changes to the facility, procedures, license, and technical specifications;
3. Determination of whether a proposed change, test or experiment would constitute an unreviewed safety question or a change in Technical Specifications;
4. Review of abnormal performance of plant equipment and operating anomalies having safety significance; and UWNR Safety Analysis Report Rev. 2 14-48 Sept. 2008 * * *
5. Review of unusual or reportable occurrences and incidents which arereportable under 10 CFR Part 20 and 10 CFR Part 50.6. Review of audit reports.7. Review of violations of technical specifications, license, or procedures andorders having safety significance.6.2.4 Audit FunctionHealth Physicist from the University of Wisconsin Safety Department Radiation Safety Office shall represent the University Radiation Safety Committee and shall conduct an inspection of the facility at least monthly toassure compliance with the regulations of 10 CFR Part 20. The services and inspection function of the Health Physics Office shall also be available to the Reactor Safety Committee, and will extend the scope of the audit to cover license, technical specification, and procedure adherence.
The committee shall audit operation and operational records of the facility.
If the committee chooses to use the staff of the Health Physics organization for the audit function, the reports of audit results will be distributed to the committee and included as an agenda item for committee meetings.Reactor staff shall perform annual reviews of the requalification program, the security plan, and the emergency plan and its implementing procedures.Radiation Safety The Reactor Laboratory shall meet the requirements of the University Radiation Safety Regulations as submitted for the University Broad License, License Number25-1323-01 and is subject to the authority of the state license.The Reactor Director shall have responsibility for maintaining radiation exposures as low as reasonably achievable and for implementation of laboratory procedure for insuring compliance with 10 CFR Part 20 regulations.
UWNR Safety Analysis Report Rev. 2 14-49 Sept. * *
5. Review of unusual or reportable occurrences and incidents which are reportable under 10 CFR Part 20 and 10 CFR Part 50 . 6. Review of audit reports. 7. Review of violations of technical specifications, license, or procedures and orders having safety significance.
TS 6.2.4 Audit .Function A Health Physicist from the University of Wisconsin Safety Department Radiation Safety Office shall represent the University Radiation Safety Committee and shall conduct an inspection of the facility at least monthly to assure compliance with the regulations of 10 CFR Part 20. The services and inspection function of the Health Physics Office shall also be available to the Reactor Safety Committee, and will extend the scope of the audit to cover license, technical specification, and procedure adherence.
The committee shall audit operation and operational records of the facility.
If the committee chooses to use the staff of the Health Physics organization for the audit function, the reports of audit results will be distributed to the committee and included as an agenda item for committee meetings . Reactor staff shall perform annual reviews of the requalification program, the security plan, and the emergency plan and its implementing procedures.
TS 6.3 Radiation Safety The Reactor Laboratory shall meet the requirements ofthe University Radiation Safety Regulations as submitted for the University Broad License, License Number 25-1323-01 and is subject to the authority of the state license. The Reactor Director shall have responsibility for maintaining radiation exposures as low as reasonably achievable and for implementation of laboratory procedure for insuring compliance with 10 CFR Part 20 regulations . UWNR Safety Analysis Report Rev. 2 14-49 Sept. 2008 6.4 Procedures Written operating procedures shall be adequate to assure the safety of operation of the reactor, but shall not preclude the use of independent judgement and action should the situation require such. Operating procedures shall be in effect for the following items: 1. Testing and calibration of reactor operating instrumentation and controls, control rod drives, area radiation monitors, and air particulate monitors;2. Reactor startup, operation, and shutdown;3. Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support;4. Fuel element and experiment loading or unloading;Control rod removal or replacement;
6. Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety;7. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes; and 8. Civil disturbances on or near the facility site.Substantive changes to the above procedures shall be made only with the approval of the Reactor Safety Committee.
Temporary changes to the procedures that do not change their original intent may be made by the Senior Operator in control or designated alternate. All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Committee.
UWNR Safety Analysis Report Rev. 2 14-50 Sept. TS 6.4 Procedures Written operating procedures shall be adequate to assure the safety of operation ofthe
reactor, but shall not preclude the use of independent judgement and action should the situation require such. Operating procedures shall be in effect for the following items: 1. Testing and calibration of reactor operating instrumentation and controls, control rod drives, area radiation monitors, and air particulate monitors;
2. Reactor startup, operation, and shutdown;
3. Emergency and abnormal conditions, including provisions for evacuation, reentry, recovery, and medical support; 4. Fuel element and experiment loading or unloading;
5. Control rod removal or replacement;
6. Routine maintenance of the control rod drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety; 7. Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms and abnormal reactivity changes; and 8. Civil disturbances on or near the facility site. Substantive changes to the above procedures shall be made only with the approval of the Reactor Safety Committee.
Temporary changes to the procedures that do not change their original intent may be made by the Senior Operator in control or designated alternate.
All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Committee.
UWNR Safety Analysis Report Rev. 2 14-50 Sept. 2008 *
Experiment Review and Approval 1. Routine experiments may be performed at the discretion of the senior operator responsible for operation without the necessity of further review or approval.2. Prior to performing any experiment which is not a routine experiment, the proposed experiment shall be evaluated by the senior operator responsible for operation.
The senior operator shall consider the experiment in terms of its effect on reactor operation and the possibility and consequences of its failure, including where significant, consideration of chemical reactions, physical integrity, design life, proper cooling, interaction with core components, reactivity effects, and interactions with reactor instrumentation.
3. Modified routine experiments may be performed at the discretion of the senioroperator responsible for operation without the necessity of further review or approval provided that the evaluation performed in accordance with Section 6.5(2) results in a determination that the hazards associated with the modifiedroutine experiment are neither greater nor significantly different than those involved with the corresponding routine experiment which shall be referenced.
4. No special experiment shall be performed until the proposed experiment has been reviewed and approved by the Reactor Safety Committee.
5. Favorable evaluation of an experiment shall conclude that failure of the experiment will not lead directly to damage of reactor fuel or interference with movement of a control element.UWNR Safety Analysis Report Rev. 2 14-51-.Sept. * *
TS 6.5 Experiment Review and Approval 1. Routine experiments may be performed at the discretion of the senior operator responsible for operation without the necessity of further review or approval.
2. Prior to performing any experiment which is not a routine experiment, the proposed experiment shall be evaluated by the senior operator responsible for operation.
The senior operator shall consider the experiment in terms of its effect on reactor operation and the possibility and consequences of its failure, including where significant, consideration of chemical reactions, physical integrity, design life, proper cooling, interaction with core components, reactivity effects, and interactions with reactor instrumentation.
3. Modified routine experiments may be performed at the discretion of the senior operator responsible for operation without the necessity of further review or approval provided that the evaluation performed in accordance with Section 6.5(2) results in a determination that the hazards associated with the modified routine experiment are neither greater nor significantly different than those involved with the corresponding routine experiment which shall be referenced.
4. No special experiment shall be performed until the proposed experiment has been reviewed and approved by the Reactor Safety Committee.
5. Favorable evaluation of an experiment shall conclude that failure ofthe experiment will not lead directly to damage of reactor fuel or interference with movement of a control element. UWNR Safety Analysis Report Rev. 2 14-51--Sept. 2008 Required Actions Action to be Taken in Case of Safety Limit Violation In the event a safety limit is exceeded: 1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.2. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section 6.7 of these specifications, and 3. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
This report shall be submitted to the Reactor Safety Committee(RSC) for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.
Action to be Taken in the Event of an Occurrence of the Type Identified in 6.7.2(1)b., and 6.7.2(1)c.
In the event of an reportable occurrence (1.3.1) the following actions shall be taken: 1. The reactor shall be shut down.2. The Director or designated alternate shall be notified and corrective action taken with respect to the operations involved, 3. The Director or designated alternate shall notify the Chairman of the Reactor Safety Committee, 4. A report shall be made to the Reactor Safety Committee which shall include an analysis of the cause of the occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence, and 5. A report shall be made to the NRC in accordance with Section 6.7.2 of these specifications.UWNR Safety Analysis Report Rev. 2 14-52 Sept. 2008 TS 6.6 Required Actions TS 6.6.1 Action to be Taken in Case of Safety Limit Violation In the event a safety limit is exceeded:
1. The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC. 2. An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Committee, and reports shall be made to the NRC in accordance with Section 6.7 of these specifications, and 3. A report shall be prepared which shall include an analysis of the causes and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence.
This report shall be submitted to the Reactor Safety Committee (RSC) for review and then submitted to the NRC when authorization is sought to resume operation of the reactor. TS 6.6.2 Action to be Taken in the Event of an Occurrence of the Type Identified in 6.7.2(I)b., and 6.7.2(I)c.
In the event of an reportable occurrence (1.3.1) the following actions shall be taken: 1. The reactor shall be shut down. 2. The Director or designated alternate shall be notified and corrective action taken with respect to the operations involved, 3. The Director or designated alternate shall notify the Chairman of the Reactor Safety Committee, 4. A report shall be made to the Reactor Safety Committee which shall include an analysis of the cause ofthe occurrence, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence, and 5. A report shall be made to the NRC in accordance with Section 6.7.2 of these specifications.
UWNR Safety Analysis Report Rev. 2 14-52 Sept. 2008 * *
Reports Operating Reports 1. An annual report covering the activities of the reactor facility during the previous calendar year shall be submitted (in writing to U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC 20555) within six months following the end of each calendar year, providing the following information:
a. A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections;
b. Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output sinceinitial criticality;
c. The number of emergency shutdowns and inadvertent scrams, including reasons therefor;
d. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required;
e. A brief description, including a summary of the safety evaluations of changes in the facility or in the procedures and of tests and experiments carried pursuant to Section 50.59 of 10 CFR Part 50;f. A summary of radiation exposures received by facility personnel andvisitors, including dates and time of significant exposures and a summary of the results of radiation and contamination surveys performed within the facility; and g. A description of any environmental surveys performed outside the facility.UWNR Safety Analysis Report Rev. 2 14-53ýSept. * *
TS 6.7 Reports TS 6.7.1 Operating Reports 1. An annual report covering the activities of the reactor facility during the previous calendar year shall be submitted (in writing to U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, DC 20555) within six months following the end of each calendar year, providing the following information:
a. A brief narrative summary of (1) operating experience (including experiments performed), (2) changes in facility design, performance characteristics, and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections;
b. Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was critical, and the cumulative total energy output since initial criticality;
c. The number of emergency shutdowns and inadvertent scrams, including reasons therefor;
d. Discussion of the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required;
e. A brief description, including a summary of the safety evaluations of changes in the facility or in the procedures and of tests and experiments carried pursuant to Section 50.59 of 10 CFR Part 50; f. A summary of radiation exposures received by facility personnel and visitors, including dates and time of significant exposures and a summary of the results of radiation and contamination surveys performed within the facility; and g. A description of any environmental surveys performed outside the facility . UWNR Safety Analysis Report Rev. 2 14-53. Sept. 2008
h. A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge; (1) Liquid Effluents (summarized on a monthly basis)'Liquid radioactivity discharged during the reporting period tabulated as follows: (a) Total estimated radioactivity released (in curies).(b) The isotopic composition if greater than 1 x microcuries/cc for fission and activation products.(c) Total radioactivity (in curies), released by nuclide, during the reporting period based on representative isotopic analysis.(d) Average concentration at point of release (in microcuries/cc) during the reporting period and the fraction of the applicable limit in 10 CFR Part 20.(e) Total volume (in gallons) of effluent water (including diluent) during periods of release.(2) Exhaust Effluents (summarized on a monthly basis)
Radioactivity discharged during the reporting period (in curies) for: (a) Gases.(b) Particulates with half lives greater than eight days.(c) The estimated activity (in curies) discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis and the fraction of the applicable 10 CFR Part 20 limits for these values.(3) Solid Waste (a) The total amount of solid waste packaged (in cubic feet).(b) The total activity involved (in curies).(c) The dates of shipment and disposition (if shipped off site). UWNR Safety Analysis Report Rev. 2 14-54 Sept. h. A summary of the nature and amount radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or prior to the point of such release or discharge; (1) Liquid Effluents (summarized on a monthly basis)" Liquid radioactivity discharged during the reporting period tabulated as follows: (a) Total estimated radioactivity released (in curies) .. (b) The isotopic composition if greater than 1 x 10-7 microcuries/cc for fission and activation products. (c) Total radioactivity (in curies), released by nuclide, during the reporting period based on representative isotopic analysis. (d) Average concentration at point of release (in microcuries/cc) during the reporting period and the fraction of the applicable limit in 10 CFR Part 20. ** (e) Total volume (in gallons) of effluent water (including diluent) during periods of release. * (2) Exhaust Effluents (summarized on a monthly basis) Radioactivity discharged during the reporting period (in curies) for: (a) Gases. (b) Particulates with half lives greater than eight days. (c) The estimated activity (in curies) discharged during the reporting period, by nuclide, for all gases and particulates based on representative isotopic analysis and the fraction of the applicable 10 CFR Part 20 limits for these values. (3) Solid Waste (a) The total amount of solid waste packaged (in cubic feet). (b) The total activity involved (in curies). ( c) The dates of shipment and disposition (if shipped off site). UWNR Safety Analysis Report Rev. 2 14-54 Sept. 2008 *
2. A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating conditions or characteristics of the reactor under the new conditions including:
a. An evaluation of facility performance to date in comparison with design predictions and specifications, and b. A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.TS 6.7.2 Special Reports 1. There shall be a report of any of the following not later than the following day by telephone or similar conveyance to the NRC HeadquartersOperation Center, and followed by a written report describing the circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555: a. Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure;b. Any violation of a safety limit; and
c. Any reportable occurrences as defined in Section 1.3.1 of these specifications.
2. A written report within 30 days in writing to the U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555 of: a. Permanent changes in facility organization at Reactor Director or Department Chair level.b. Any significant change in the transient or accident analysis as described in the Safety Analysis Report;UWNR Safety Analysis Report Rev. 2 14-55 Sept. * *
TS 6.7.2 2. A report within 60 days after completion of startup testing of the reactor (in writing to the U.S. Nuclear Regulatory Commission, Attn: Document Control Desk, Washington, D.C. 20555) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level describing the measured values of the operating conditions or characteristics of the reactor under the new conditions including:
a. An evaluation of facility performance to date in comparison with design predictions and specifications, and b. A reassessment of the safety analysis submitted with the license application in light of measured operating characteristics when such measurements indicate that there may be substantial variance from prior analysis.
Special Reports 1. There shall be a report of any of the following not later than the following day by telephone or similar conveyance to the NRC Headquarters Operation Center, and followed by a written report describing the circumstances of the event and sent within 14 days to U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555: a. Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure;
b. Any violation of a safety limit; and c. Any reportable occurrences as defined in Section 1.3.1 of these specifications.
2. A written report within 30 days in writing to the U.S. Nuclear Regulatory commission, Attn: Document Control Desk, Washington, D.C. 20555 of: a. Permanent changes in facility organization at Reactor Director or Department Chair level. b. Any significant change in the transient or accident analysis as described in the Safety Analysis Report; UWNR Safety Analysis Report Rev. 2 14-55 Sept. 2008 Records Records to be Retained for a Period of at least Five Years or for the Life of the Component Involved if Less than Five Years 1. Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc. which shall be maintained for a period of at least one year), 2. Principal maintenance activities, 3. Reportable occurrences, 4. Surveillance activities required by the Technical Specifications, 5. Reactor facility radiation and contamination surveys where required by applicable regulations, 6. Experiments performed with the reactor, 7. Fuel inventories, receipts, and shipments, 8. Approved changes in operating procedures, 9. Records of meeting and audit reports of the review and audit group.Records to Retained for at Least One Cycle Operator qualification and re-qualification records.
Records to be Retained for the Lifetime of the Reactor Facility Annual reports which contain the information in items 1 and 2 may be used as records for those items.
1. Gaseous and liquid radioactive effluents released to the environs, 2. Offsite environmental monitoring surveys required by technical specifications, 3. Radiation exposures for all personnel monitored, 4. Updated, corrected, and as-built drawings of the facility.UWNR Safety Analysis Report Rev. 2 14-56 Sept. TS 6.8 Records TS 6.8.1 Records to be Retained for a Period of at least Five Years or for the Life of the Component Involved if Less than Five Years 1. Normal reactor facility operation (but not including supporting documents such as checklists, log sheets, etc. which shall be maintained for a period of at least one year), 2. Principal maintenance activities, 3. Reportable occurrences, 4. Surveillance activities required by the Technical Specifications, 5. Reactor facility radiation and contamination surveys where required by applicable regulations, 6. Experiments performed with the reactor, 7. Fuel inventories, receipts, and shipments, 8. Approved changes in operating procedures, 9. Records of meeting and audit reports of the review and audit group. TS 6.8.2 Records to be Retained for at Least One Cycle Operator qualification and re-qualification records. TS 6.8.3 Records to be Retained for the Lifetime of the Reactor Facility Annual reports which contain the information in items I and 2 may be used as records for those items. 1. Gaseous and liquid radioactive effluents released to the environs, 2. Offsite environmental monitoring surveys required by technical specifications, 3. Radiation exposures for all personnel monitored, 4. Updated, corrected, and as-built drawings of the facility.
UWNR Safety Analysis Report Rev. 2 14-56 Sept. 2008 * *
TS 7 REFERENCES
1. GA-9064, pages 3-1 to 3-23 2. GA-9064, pages 3-1 to 3-23 3. NE Memo No. 4, Report on Refueling the University of Wisconsin Nuclear Reactor, R. J.Cashwell, March 1968, University of Wisconsin Department of Nuclear Engineering
4. Core Test Program, UWNR Mixed TRIGA-FLIP Core (9 FLIP), R. J. Cashwell, July 1974, University of Wisconsin Department of Nuclear Engineering
5. Core Test Propgram, UWNR Mixed TRIGA-FLIP Core (15 FLIP), R. J. Cashwell, February 1978, University of Wisconsin Department of Nuclear Engineering
6. Core Test Program, All FLIP Core, R. J. Cashwell, January 1980, University of Wisconsin Department of Nuclear Engineering
7. "Sagitta" refers to the bow of the element and means the maximum excursion of the clad surface from a chord connecting the two ends of the clad surface.8. Simnad and West, 1986 9. NUREG-1282 UWNR Safety Analysis Report Rev. 2 14-57 Sept. TS 7 REFERENCES
1. GA-9064, pages 3-1 to 3-23 2. GA-9064, pages 3-1 to 3-23 *
3. NE Memo No.4, Report on Refueling the University of Wisconsin Nuclear Reactor, R. J. Cashwell, March 1968, University of Wisconsin Department of Nuclear Engineering
4. Core Test Program, UWNR Mixed TRIGA-FLIP Core (9 FLIP), R. J. Cashwell, July 1974, University of Wisconsin Department of Nuclear Engineering
5. Core Test Propgram, UWNR Mixed TRIGA-FLIP Core (15 FLIP), R. J. Cashwell, February 1978, University of Wisconsin Department of Nuclear Engineering
6. Core Test Program, All FLIP Core, R. J. Cashwell, January 1980, University of Wisconsin Department of Nuclear Engineering
7. "Sagitta" refers to the bow of the element and means the maximum excursion of the clad surface from a chord connecting the two ends of the clad surface. 8. Simnad and West, 1986 9. NUREG-1282 UWNR Safety Analysis Report Rev. 2 14-57 Sept. 2008 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 2 14-58 Sept.
This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 2 14-58 Sept. 2008
Financial Ability to Construct a Non-Power Reactor Not applicable for renewal application.
Financial Ability to Operate a Non-Power Reactor The Reactor Laboratory is a part of the Engineering Physics Department of the College of Engineering at the University of Wisconsin-Madison.
The teaching mission of the laboratory takes precedence over the research and service missions; for this reason the primary fiscalsupport for operation of the facility is from the state-funded university budget. Reactor personnel have instructional duties, such as teaching courses, setting up laboratory experiments, and assuring that equipment used in the teaching laboratories is operable.
It thus becomes somewhat difficult to allocate funding precisely to just the operation of the reactor. In the following information no attempt is made to separate the instructional component of the budget from the reactor operations part of the budget. For instance, the Reactor Director is teaching two courses during the Spring 2000 semester, but his entire salary is included in the budget below.Recently the operating budget of the Reactor Laboratory has been considerably higher than what would be predicted from the past history of the facility.
This is due in large part to providing replacement .of long-term employees who are approaching retirement age with younger workers in time to allow adequate training of the new employees.
The total salary expenditures will be reduced significantly after the retirements actually take place, but the estimate below uses the current funding level.The total operating budget of the laboratory is $388,000.
Of this total, state instructional funding covers $299,000, with the remaining expenses split between grants ($48,000) and incomegenerated by reactor services ($41,000).
By expense category, the funds are spent for salary($253,000), fringe benefits ($77,300), supplies and expense ($24,000), and capital equipment
($33,000).
These numbers do not include the infrastructure provided by the University such as electricalpower, heating, janitor service, and health physics coverage.
Further, the fringe benefits included above are not specifically billed to the department in which the employee works for instructionalfunding, but come from a campus-wide fund, while the fringe benefits for salaries supported by non-instructional funding are charged to the fund paying the salary.
The instructional funding is appropriated by the state. The administration of the university has been very supportive of the reactor facility and continuation of the Nuclear Engineering curriculum.
The fact that the application for renewal of the license is signed by the administration indicates this support.UWNR Safety Analysis Report Rev.
0April 2000* *
15 FINANCIAL QUALIFICATIONS 15.1 Financial Ability to Construct a Non-Power Reactor Not applicable for renewal application.
15.2 Financial Ability to Operate a Non-Power Reactor The Reactor Laboratory is a part of the Engineering Physics Department of the College of Engineering at the University ofWisconsin:-Madison.
The teaching mission of the laboratory takes precedence over the research and service missions; for this reason the primary fiscal support for operation of the facility is from the state-funded university budget. Reactor personnel have instructional duties, such as teaching courses, setting up laboratory experiments, and assuring that equipment used in the teaching laboratories is operable.
It thus becomes somewhat difficult to allocate funding precisely to just the operation of the reactor. In the following information no attempt is made to separate the instructional component of the budget from the reactor operations part of the budget. For instance, the Reactor Director is teaching two courses during the Spring 2000 semester, but his entire salary is included in the budget below. Recently the operating budget of the Reactor Laboratory has been considerably higher than what would be predicted from the past history of the facility.
This is due in large part to providing replacement .of long-term employees who are approaching retirement age with younger workers in time to allow adequate training of the new employees.
The total salary expenditures will be reduced significantly after the retirements actually take place, but the estimate below uses the current funding level. The total operating budget ofthe laboratory is $388,000.
Ofthis total, state instructional funding covers $299,000, with the remaining expenses split between grants ($48,000) and income generated by reactor services ($41,000).
By expense category, the funds are spent for salary($253,000), fringe benefits ($77,300), supplies and expense ($24,000), and capital equipment
($33,000).
Thyse numbers do not include the infrastructure provided by the University such as electrical power, heating, janitor service, and health physics coverage.
Further, the fringe benefits included above are not specifically billed to the department in which the employee works for instructional funding, but come from a campus-wide fund, while the fringe benefits for salaries supported by non-instructional funding are charged to the fund paying the salary. The instructional funding is appropriated by the state. The administration of the university has been very supportive of the reactor facility and continuation of the Nuclear Engineering curriculum.
The fact that the application for renewal of the license is signed by the administration indicates this support . UWNR Safety Analysis Report Rev. 0 15-1 April 2000 Much of the capital equipment funding in recent years has come from the DOE program toupdate the instrumentation and experimental equipment on non-power reactors.
In addition, a local utility company has provided matching grants to support instruction in traditional "fission nuclear engineering", and this has contributed to both the supply and expense and capitalequipment budget.
The combination of funding opportunities has resulted in the reactor and associated laboratories being in excellent condition.
The reactor can continue to operate without the outside grant income, but it would not allow for upgrading the equipment as has been donefor the last 10 years. However, the present instrumentation and control systems are capable of continuing to operate for another 20 years with no loss of function.Funds from services provided to persons outside the university have been steady for the last 10 years, with some increases in the last three years. It is expected that this funding will continue for at least the next five years.
Financial Ability to Decommission the Facility By letter dated July 19, 19901, the University responded to CFR Part showing the expected cost of decommissioning the reactor. At that time estimated decommissioning cost was$1,200,000 in the year 2000.Early in 1999 the computations on which the funding plan was made were updated to extend the time of decommissioning to 2020 and to incorporate more recent figures on the cost of disposal of radioactive debris from the decommissioning. The estimate is again based on placing the facility in condition for unrestricted release three years after cessation of operations, now estimated as June 30, 2020. The result of this revised estimate is a decommissioning cost ofbetween $3,000,000 and $8,000,000, with the wide range of cost based primarily upon the uncertainty in disposal cost. Whereas the disposal cost was a minor part of the total decommissioning effort costs for the original computation, it is now by far the' largest component of the total cost.
Nevertheless, as a state agency, the funding plan remains to obtain the fundingwhen necessary.
15.4 Reference 1. Letter to USNRC Document Control Desk from R. J. Cashwell under Docket 50.156dated July 19,1990, with attachment signed on behalf of the Board of Regents.UWNR Safety Analysis Report Rev. 0April 2000 Much of the capital equipment funding in recent years has come from the DOE program to
update the instrumentation and experimental equipment on non-power reactors.
In addition, a local utility company has provided matching grants to support instruction in traditional "fission nuclear engineering", and this has contributed to both the supply and expense*and capital equipment budget. The combination of funding opportunities has resulted in the reactor and associated laboratories being in excellent condition.
The reactor can continue to operate without the outside grant income, but it would not allow for upgrading the equipment as has been done for the last 10 years. However, the present instrumentation and control systems are capable of continuing to *operate for another 20 years with no loss of function.
Funds from services provided to persons outside the university have been steady for the last 10 years, with some increases in the last three years. It is expected that this funding will continue for at least the next five years. 15.3 Financial Ability to the Facility By letter dated July 19, 1990 1; the University responded to 10 CFR Part 50.75 showing the expected cost of decommissioning the reactor. At that time estimated decommissioning cost was $1,200,000 in the year 2000. . Early in 1999 the computations on which the funding plan was made were updated to extend the time of decommissioning to 2020 and to incorporate more recent figures on the cost of disposal of radioactive debris from the decommissioning.
The estimate is again based on placing the
facility in condition for unrestricted release three years after cessation of operations, now estimated as June 30, 2020. The result of this revised estimate is a decommissioning cost of between $3,000,000 and $8,000,000, with the wide range of cost based primarily upon the uncertainty in disposal cost. Whereas the disposal cost was a minor part of the total decommissioning effort costs for the original computation, it is now by far the'largest component of the total cost. Nevertheless, as a state agency, the funding plan remains to obtain the funding when necessary.
15.4 Reference
1. Letter to USNRC Document Control Desk from R. J. Cashwell under Docket 50.156 dated July 19,1990, with attachment signed on behalf Board of Regents. UWNR Safety Analysis Report Rev. 0 15-2 April 2000
OTHER Prior Use of Reactor Components There are no components in use at the University of Wisconsin Nuclear Reactor Laboratory that have had prior use at any other facility or organization.
It is conceivable that prior use components could be integrated into Reactor Laboratory systems at some future time.Appropriate analysis and reviews of component replacement will be conducted in accordance toapplicable standards, regulations and facility procedures and licensed technical specifications.
Medical Use of Non-Power Reactors The University of Wisconsin Nuclear Reactor Laboratory is not engaged nor licensed to conduct any activities for medical use of the facility. Future medical use of the Reactor Laboratory would be conducted pursuant to appropriate license applications and approvals as authorized by the Atomic Energy Act of 1954 as amended.UWNR Safety Analysis Report Rev. 0April 2000 e e* e 16 OTHER LICENSE CONSIDERATIONS 16.1 Prior Use of Reactor Components There are no components in use at the University of Wisconsin Nuclear Reactor Laboratory that have had prior use at any other facility or organization.
It is conceivable that prior use components could be integrated into Reactor Laboratory systems at some future time. Appropriate analysis and reviews of component replacement will be conducted in accordance to applicable standards, regulations and facility procedures and licensed technical specifications.
16.2 Medical Use of Non-Power Reactors The University of Wisconsin Nuclear Reactor Laboratory is not engaged nor licensed to conduct any activities for medical use of the facility.
Future medical use of the Reactor Laboratory would be conducted pursuant to appropriate license applications and approvals as authorized by the Atomic Energy Act of 1954 as amended. UWNR Safety Analysis Report Rev. 0 16-1 April 2000 This page intentionally left blank.UWNR Safety Analysis Report Rev. 0April 2000* ( This page intentionally left blank. *
UWNR Safety Analysis Report Rev. 0 16-2 April 2000 Appendix Calculation Methods for Atmospheric Release of RadioactivityModels Used for Calculations in Sections 11.1.1.1.2 and For Sutton's diffusion model, the maximum concentration (Xma) at any point downwind is given as:(Reference 1)where j is mean wind speed in meters/secondQ is release rate in Ci/secondis stack height in meters For the generalized Gaussian Plume Model, the maximum concentration is given by the same equation (Reference 2, equation 8).For calculations in this report, the following values are used:= lowest monthly average = 3.54 meters/secondstack height above ground = 26.5 meters.Using these numbers, equation (1) reduces to (2) Xmax= 9.42E-5 Q, where Xma is in I.Ci/ml Reference 2 presents a method applicable to release from buildings with zero stack height to approximate release from leaks in a containment structure.
The relation given, as equation 4, is: (3) X= Q az+CA)iJ'where X, Q, and [i are defined as above,is an empirical constant with a value between 0.5 to 2, andis the minimum building cross section.The a terms are concerned with atmospheric dispersion, which will be neglected in this analysis, which will result in the equation;Q UWNR Safety Analysis Report Rev. 2Sept. 2008* *
Appendix A Calculation Methods for Atmospheric Release of Radioactivity A. Models Used for Calculations in Sections 11.1.1.1.2 and 13.1.1.5 For Sutton's diffusion model, the maximum concentration (Xmax) at any point downwind is given as: (1) X = 2Q max -h 2 e1tj.L (Reference I) where is mean wind speed in meters/second Q is release rate in Ci/second h is stack height in meters For the generalized Gaussian Plume Model, the maximum concentration is given by the same equation (Reference 2, equation 8). For calculations in this report, the following values are used: = lowest monthly average = 3.54 meters/second h = stack height above ground = 26.5 meters. Using these numbers, equation (1) reduces to (2) Xmax = 9.42E-5 Q, where Xmax is in j.LCi/ml Reference 2 presents a method applicable to release from buildings with zero stack height to approximate release from leaks in a containment structure.
The relation given, as equation 4, is: where X, Q, and j.L are defined as above, C is an empirical constant with a value between 0.5 to 2, and A is the minimum building cross section. The a terms are concerned with atmospheric dispersion, which will be neglected in this analysis, which will result in the equation; CAj.L UWNR Safety Analysis Report Rev. 2 A-I Sept. 2008 Inserting values for the UWNR facility used in the safety analysis for FLIP fuel conversion, and using a value of 1 for C yields: (5) , or (1)(12,200ft 2)(9.29E-2m 2)(3.54m/sec)
(6) X=2.49E-4 Q with X in units of pLCi/ml B. Sample Calculations Supporting Section 11.1.1.1.2 The maximum release rate for Ar-41 activity is 13.3 pCi/second. Using the ventilation system rated flow-rate of 9600 scfm, this activity is diluted to 2.94E-6 pCi/ml at the stack outlet. The resulting maximum concentration downwind, assuming the stack height, is calculated to be, from equation (2)(7) Xmax = (13.3E-6)(9.42E-5)
= 1.25E-9 ptCi/ml.If calculated using equation (6) (which assumes zero stack height with building wake dilution), the resulting value is (8) X- (13.3E-6)(2.49E-4)
= 3.3 1E-9 pCi/ml It is obvious that the two methods used to calculate the above values cannot both be applicable.
Since the reactor is not operated when the ventilation system is not in operation, the value in equation (7) is more realistic, but the more conservative value in equation (8) is used in the text.Calculations Supporting Section 13.1.1.4 Whole Body Exposure The activity concentration of the insoluble volatiles in the reactor room air was determined by dividing released activity by room volume.(9) _
= Since 3.7E4 dps = 1 IpCi, 109 y/sec-cm 3 The maximum dose rate is calculated by assuming the room is equivalent to a hemisphere with a radius of 782 cm. In addition, the average gamma energy is 0.7 MeV, the attenuation coefficientfor air is 3.5E-5 cm', and the flux-to-dose conversion factor is 4.2E4 y/cm 2/cm 2-mr/min Using the relationship (10) = 30S(1 -exp(-RY,))
where CY2 UWNR Safety Analysis Report Rev. 2Sept. Inserting values for the UWNR facility used in the safety analysis for FLIP fuel conversion, and
using a value of 1 for C yields: Q (5) X-* . ,or (1 )(12,200ft 2)(9.29E-2m 2ij(2)(3.54m/sec)
(6) X=2.49E-4 Q with X in units of\lCi/ml B. Sample Calculations Supporting Section 11.1.1.1.2 The maximum release rate for Ar-41 activity is 13.3 \lCi/second.
Using the ventilation system rated flow-rate of9600 scfm, this activity is diluted to 2.94E-6 \lCi/ml at the stack outlet. The resulting maximum concentration downwind, assuming the stack height, is calculated to be, from equation (2) (7) Xmax = (13.3E-6)(9.42E-5)
= 1.25E-9 \lCi/ml. If calculated using equation (6) (which assumes zero stack height with building wake dilution), the resulting value is (8) X= (13.3E-6)(2.49E-4)
= 3.31E-9 \lCi/ml It is obvious that the two methods used to calculate the above values cannot both be applicable.
Since the reactor is not operated when the ventilation system is not in operation, the value in equation (7) is more realistic, but the more conservative value in equation (8) is used in the text. C. Calculations Supporting Section 13.1.1.4 (1), Whole Body Exposure The activity concentration of the insoluble volatiles in the reactor room air was determined by dividing released activity by room volume. (9) A V 5.89E6 \lCi = 2.95E-3 2.00E9 em 3 Since 3.7E4 dps = 1 \lCi, A/V= 109 y/sec-cm 3 The maximum dose rate is calculated by assuming the room is equivalent to a hemisphere with a radius of782 cm. In addition, the average gamma energy is 0.7 MeV, the attenuation coefficient for air is 3.5E-5 cm-\ and the flux-to-dose conversion factor is 4.2E4 y/cm2/cm2-mr/min Using the relationship (10) DR = 30S(1 -exp( -where UWNR Safety Analysis Report Rev. 2 A-2 Sept. 2008 ' *
= dose rate in mr/hr S = Volumetric source strength in y/sec-cm 2= outer radius of hemisphereattenuation coefficient for air, yields a dose rate of 60 mr/hour.Calculations Supporting Section 13.1.1.4 (2), Dose to the Lungs The dose to the lungs was calculated by first assuming uniform dispersal of the released volatiles in the laboratory volume, giving a concentration of (11) _ 3 Since the "standard man" breathes 1.25 cubic meters of air per active hour, he would breathe approximately 0.21 cubic meters in 10 minutes, the assumed evacuation time. If this number is increased to 0.30 cubic meters to allow for excitement and stress, then his lungs would be exposed to an activity of (12) A (.75E-3 = 825 V The dose to the lungs is then calculated from the following expression to be 1 Rad.(13)
= where m = Activity exposure (825 ItCi)= Conversion factors P/sec-VQCi)(1.6E-6 = lung retention factor (0.125 is customary) m = mass of lungs (1000 grams)= fraction of total activity= energy of beta for nuclide i (MeV)= radioactive decay constant biological release constant (6.7E-8 sec-)t time of exposure (assumed infinite)E. Calculations Supporting Section and Table Release rate, Q, for an isotope is the total quantity released to air (column of Table 13.1, Chapter 13) divided by the assumed release time. The release time used in further calculations is the time for the ventilation system (room air and beam port and thermal column exhaust systems) to make a complete change of air in the Reactor Laboratory.
UWNR Safety Analysis Report Rev. 2Sept. * *
DR = dose rate in mr/hr S = Volumetric source strength in y/sec-cm 2 R = outer radius of hemisphere = attenuation coefficient for air, yields a dose rate of 60 mrlhour. D. Calculations Supporting Section 13.1.1.4 (2), Dose to the Lungs The dose to the lungs was calculated by first assuming uniform dispersal of the released volatiles in the laboratory volume, giving a concentration of (11) A = 5.54E6 = 2. 75E -3 3 . V 2.00E9 em 3 Since the "standard man" breathes 1.25 cubic meters of air per active hour, he would breathe approximately 0.21 cubic meters in 10 minutes, the assumed evacuation time. If this number is increased to 0.30 cubic meters to allow for excitement and stress, then his lungs would be exposed to an activity of (12) A V = (.75E-3 (3E5 em 3) = 825 V The dose to the lungs is then calculated from the following expression to be 1 Rad . ACR F. E. "/ (13) Dose (Rad) =--L _I _I (1 -e ), where m /=1 Aj A = Activity exposure (825 C = Conversion factors (3.7E4 erg/MeV) 1 OOerg/gm -rad R = lung retention factor (0.125 is customary) m = mass oflungs (1000 grams) F j = fraction of total activity E j = energy of beta for nuclide i (Me V) Aj = radioactive decay constant + biological release constant (6.7E-8 sec-I) t = time of exposure (assumed infinite)
E. Calculations Supporting Section 13.1.1.5 and Table 13.1 Release rate, Q, for an isotope is the total quantity released to air (column E of Table 13.1, Chapter 13) divided by the assumed release time. The release time used in further calculations is the time for the ventilation system (room air and beam port and thermal column exhaust systems) to make a complete change of air in the Reactor Laboratory . UWNR Safety Analysis Report Rev. 2 A-3 Sept. 2008 (14) T, 2000 m 3 35.31 60 sec 1569 2700 m 3 1 min Using the generalized Gaussian Plume Model (equation (2)), and demonstrating with data for Br-83, concentrations released to unrestricted areas (Chapter 13, Table 1 Column H) are calculated as shown below:
X (0.0083Ci)(9.42E-5)
= )
The remaining isotope values are calculated in the same manner.The activity release was also evaluated through use of equation (6). This calculation would be applicable to release of the activity through the building walls with the ventilation system not operating.
Again using Br-83 as an example, and assuming the same release time as in the previous calculation (16)
= (0.0083Ci)(2.49E-4) 1.32E-9 This value is a factor of 2.6 greater than that evaluated by the Gaussian Plume Model.
Allsimilar values in Chapter 13, Table 1, Columns H and I may be multiplied by this factor for a more conservative case. This calculation was done for the previous Safety Analysis Report using as the building dimensions only the minimum dimensions of the reactor laboratory, a room within the Mechanical Engineering Building. This considerably underestimates the "wake effect" of the actual building.
The current analysis uses a value appropriate for the renovated Mechanical Engineering Building (cross-sectional area of 12,200 ft 2).References 1 Meteorology and Atomic Energy, U. S. Dept of Commerce Weather Bureau, Govt.Printing Office, Washington, DC July 1955 3. F. A. Gifford, Jr, Atmospheric Dispersion Calculations Using the Generalized GaussianPlume Model, Nuclear Safety, December 1960 4. Calculation of Distance Factors for Power and Test Reactors, (TID-14844), USAEC, March 23, 1962 UWNR Safety Analysis Report Rev. 2 A-4 Sept. (14) T = 2000 m 3 35.31 ft3 60 sec = 1569 seconds release 2700 scfm 1 m 3 1 min Using the generalized Gaussian Plume Model (equation (2)), and demonstrating with data for Br-83, concentrations released to unrestricted areas (Chapter 13, Table 1 Column H) are calculated as shown below: (15) X = (0.0083Ci)(9.42E-5)
= 4.98E-1O Cilml Br-83 1569 1..1. The remaining isotope values are calculated in the same manner. The activity release was also evaluated through use of equation (6). This calculation would be applicable to release of the activity through the building walls with the ventilation system not operating.
Again using Br-83 as an example, and assuming the same release time as in the previous calculation (16) X = (0.0083Ci)(2.49E-4)
= 1.32E-9 Cilml Br-83 1569 1..1. This value is a factor of 2.6 greater than that evaluated by the Gaussian Plume Model. All similar values in Chapter 13, Table 1, Columns H and I may be multiplied by this factor for a
more conservative case. This calculation was done for the previous Safety Analysis Report using
as the building dimensions only the minimum dimensions of the reactor laboratory, a room within the Mechanical Engineering Building.
This considerably underestimates the "wake effect" of the actual building.
The current analysis uses a value appropriate for the renovated Mechanical Engineering Building (cross-sectional area of 12,200 ft2). References 1 Meteorology and Atomic Energy, U. S. Dept of Commerce Weather Bureau, Govt. Printing Office, Washington, DC July 1955 3. F. A. Gifford, Jr, Atmospheric Dispersion Calculations Using the Generalized Gaussian Plume Model, Nuclear Safety, December 1960 4. Calculation of Distance Factors for Power and Test Reactors, (TID-14844), USAEC, March 23, 1962 UWNR Safety Analysis Report Rev. 2 A-4 Sept. 2008 '
Appendix B Supporting Documents Task Order No. 2 Under Master Task Agreement No. C96-175937 LMfTCO FORM PROC-ISI8b Page 1 TASK ORDER 2 UNDER MASTER TASK AGREEMENT C96-175937 LOCKHEED MARTIN IDAHO TECHNOLOGIES COMPANY (LMITCO)2525 Fremont Avenue P. Box 1625, Idaho Falls, ID 83415-3521 CONTRACT To: University of Wisconsin-MadisonResearch Administration 750 University Avenue Madison, WI 53706-1490 Effective Date: August26, 1999 Completion Date: November 1, 2001 To: Tom Handland Ph: R.I. Cashwell This Task Order No. 2 is awarded tot 1. Transfer Reactor Fuel Assistance Subcontract No. C87-101251-002 to the new Master Task Agreement No. C96-175937 as Task No. 2.2. Extend the period of performance to November 2001. This extension is retroactive to November 1998.3. Confim the Statement of Work and modification thereto remain unchanged.
4. Assignment: September 1999, LMITCO's prime contract with DOE will expire.
Thus, pursuant to the article in the General Provisions entitled "assignment"" this Task Order is assigned to Bechtel BWXT Idaho, LLC, under its DOE Prime Contract No. DE-AC07-1DI3727, effective October 1, 1999.Procurement Agent Lynda Keller Telephone:
(208) 526-5597 ( Cost: $0.00 Ship via: N/A F.O.B/Trans.:
/A Cash Terms: Net 0 Days Lynda Keller LMITCOBox 1625 Idaho Falls, ID 83415-3521 Signed: Date Lockheed Martin Idaho Teehnologles Company Title: Procurement Agent Signed: Subcoulkraet6s Oftiat William J. Vance, Dean Title:
&p .-Return one signed copy of this Task Order to Lynda Keller.UWNR Safety Analysis Report Rev. 0 B-1 April 2000* *
Appendix B Supporting Documents Task Order No.2 Under Master Task Agreement No. C96-17S937 IMTCOfORM PROC-18I1b 01/99 Page 1 TASK ORDER NO.2 UNDER MASTER TASK AGREEMENT NO. C96-175937 LOCKHEED MARTIN IDAHO TECHNOLOGIES COMPANY (LMlTCO} 2525 Fremont Avenue P. O. 80x1625. Idaho Falls, ID 83415-3521 OPERATING UNDER U. S. GOVERNMENT COIITRACT NO. DE*AC07*94lDU221 To: University of Wisconsin-Madison Research Administration 750 University Avenue Madison, WI 53706-1490 To: Tom Handland PI: R. J. Cashwell This Task Order No.2 is awarded to! Effective Date: August 26, 1999 Completion Date: November I, 2001 1. Tl'llllsfer Reactor Fuel Assistance Subcontract No. C87-IOI1S 1-002 to the new Master Task AgRCment No. C96-175937 as TasIr;Order.No.
2. 2. Extend the period ofpcrformance to November 1,200 I. This extension is retroactive to November 1. 1998. 3. 4. ConflIlJl the Statement of Work and modification thereto remain unchanged.
Assignment:
On September 30, 1999, LMITCO's prime contract with DOE will expire. Thus, pursuant tathe article in the General Provisions entitled "assignment"" this Task Order is assigned to Bechtel 8WXT Idaho. LLC, under its DOE Prime Contract No. DE-AC07-IDI3727.
effective October 1. 1999. Proc:urement Agent Lynda Keller Cost: SO.OO Ship via: N/A 1"!Ins Add ..... : Lynda Kdler LMlTCO . P. O. BoX 1625 Idaho fall!, m 83415-J'21 Cash Terms: Net 0 Days Signed:
__
Subcontractor's Otl\clal ' Date William J. Vance, Assistant Dean Title: Ruaatch I!I Sponsored Return Dno signed copy of chis Task Order to Lynda Keller. UWNR Safety Analysis Report Rev. 0 B-1 April 2000 This page is intentionally left blank.UWNR Safety Analysis Report Rev. 0 B3-2 April 2000.' This page is intentionally left blank. *
UWNR Safety Analysis Report Rev. 0 B-2 April 2000

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