ML18025B745
| ML18025B745 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/04/1981 |
| From: | Congel F, Willis C NRC |
| To: | |
| Shared Package | |
| ML18025B739 | List: |
| References | |
| NUDOCS 8203120332 | |
| Download: ML18025B745 (72) | |
Text
I C~
el S'TATUS OF NRC RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION ACTIVITIES*
Charles A. Willis 8 Frank J.
Congel U. S. Nuclear Regulatory Commission Abstract The current NRC position on radiological effluent technical specifications (RETS) was established in 1978 (HUREG-0472 and -0473).
Progress in imple-menting current RETS requi rements was interrupted by the accident at Three Mile Island.
Since then the RETS requirements have been implemented at the operating license stage.
Efforts to implement current RETS requirements for operating reactors are now being resumed using a
new approach.
Operating reactors are being asked to meet the.intent, not the letter, of the model RETS requi rements.
This flexible approach increases the time required for review so contractor support was found necessary.
The current schedule calls for action on all operating reactors within a year.
Scope of the Radiolo ical Effluent Technical S ecifications (RETS)
The RETS are the technical specifications that deal with radioactive waste management systems and with environmental monitoring.
The RETS are commonly
~ but mistakenly called "Appendix I". Tech Specs.
This mi'snomer has been a
source of considerable misunderstanding despite repeated clarifications.
The model RETS did grow out of the 1970 requirements for keeping releases
~ "as low as rgpsonably achievable" (ALARA) which were 1'ater quantified in "Appendix I"'~'.
These regulations did not directly impose.limits on. releases.
but. instead required Tech Specs that would limit releases.
This approach was intended to provide the flexibility necessary to ensure that the constraints would indeed be reasonable (as well as low).
Thus the RETS are closely asso-ciatedd wi th "Appendix I" but as other waste management system problems arose other. appropri ate provi sions were added to the RETS.
The model RETS 'include a number of requirements not directly related to keep-ing releases ALARA.
Specific "non-Appendix I" issues include the following.
1.
2.
3.
4.
5.
Control of explosive gas mixtures Curie content of waste decay tanks Activity in BWR main condenser offgas Curie content of certain outdoor liquid radwaste tanks Level-measuring devices in liquid radwaste tanks "For presentation at the Atomic Industrial. Forum Conference on NEPA and Nuclear Regula'tion, October 4-7,
- 1981, Washington, D.C.
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7.
8.
Monitoring potentially contaminated liq'uid effluents Automatic flow termination in some liquid effluent lines Participation in an interlaboratory comparison program for radi ochemical analyses 9.
Process control program for solid radwaste management
- 10. 'onitoring and control of in-process radwastes 11'.
Venting/purging Mark I I containment drywel 1.
'Mhile it can be argued that the non-Appendix I matters should have been handled some other way, they are in fact integral parts of the RETS.
Re ulations on Keepin Releases ALARA On December 3,
- 1970, two new sectj~~s
( 10 CFR 50.34a and 50.36a) were added to
. the reactor licensing regulations In essence, these new requirements are to keep releases ALARA.
Section 50.34a requires designing plants to keep releases ALARA and Section 50.36a requires Tech Specs for keeping releases ALARA during operations.
Section '50.36a explicitly requires Tech Specs in the following four areas.
1.
Releases shall comply with the Section 20.106 limits.
2.
Procedures shall be establi.shed and followed for operating radwaste treaW'ent systems.
3.
Radwaste systems shall be maintained and used.
. 4.
'emiannual release reports shall be prepared and submitted to the
~
NRC.
Section 50.36a also'equires that the licensee be gui ded by the following considerations in developing operating procedures.
- 1. 'eleases, on the y~qrage, should be a small fraction of the limits of Section 20e106<
2.
8est efforts shall be exerted to keep releases ALARA.
3.
Appendix I provides numerical guidance on limiting conditions for operation for effluents.
In addition to the NRC regulations, the EPA has established a regulation
'40 CFR.190) requiring that offsite doses from uranium fuel cycle facilities (including reactors) be ALARA.
The EPA requirements are somewhat different from the NRC requirements.
Nevertheless, Frank Congel (NUREG-0543) has shown that meeting the Appendix I "$~~jgn objectives" provides reasonable assurance
'of compliance with 40 CFR 190<
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~Adi Appendix I provides numerical guides for ALARA design objectives and for limiting conditions for..operation.
Four. desi n objectives. are specified in Section II B ~
Liquid effluents fro~ each reactor each year shall not expose any indi-vidual to more than 3 mrem to the. total body or 10'mrem to any critical organ.
Gaseous effluents from each reactor each year (a) shall not produce, at any occupiable off-site location, air doses greater than 10 mrad'amma and 20 mrad beta or alternately, (b) shall not expose any individual to more than 5 mrem to the total body or 15 mrem to the skin.
\\
C.
'Airborne iodine and particulate effluents from each reactor each year shall not expose, any individual to more than 15 mrem to any critical organ.
D.
Ooses shall be further reduced as"much as practical up to the expendi-ture of S1000 per person-rem saved.
Appendix I also specifies Limitin Conditions for'peration (LCO)..*
A.
If releases during any. calendar quarter cause doses exceeding half an annual design objective, the licensee shall investigate, correct and report to the HRC in 30 days.
.S.
Surveil lance programs shal 1'e establi shed to monitor, rel eases, monitor the environment and identify changes in land use.
The annex to Appendix I, usually refered to as "RM 50-2", provides an alter-naiive to the Appendix I cost/benefit design objective.
The RM 50-2 criteria" were proposed by the staff. and used prior to the promulgation of Appendix I.
The One advantage of the RN 50-2 criteria is that they allow the cost-benefit criterion to be replaced by annual release criterih (5 Ci per reactor in liquid and 1 Ci per reactor of airborne iodine-131),
{a) if the dose criteria are changed from "per reactor" to "per site",
and {b) if liquid dose criterion is changed. to 5 mrem/yr to any critical organ. or the total body.
This option is most likely to be of value to licensees who had Tech Specs based on 'these cri-teria before the promulgation of Appendix I.
"C ear y the Appendix I LCOs play a role different from that of the Tech Specs on essential safety systems.
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'Early Pro ress in I lementation Appendix I required licensees to file, by June 3, 1976; information needed for evaluation of means for keeping releases ALAPA and proposed ALARA Tech Specs.
However, in February
- 1976, the HRC Staff issued guidance recommending that proposals to modify Tech Specs be deferred until the HRC completed the model RETS.
A draft version of the model RETS was transmitted to licensees in Hay 1976 but this draft was subsequently recalled.
On Hay 16,
- 1978, the HRC's Regulatory Requirements Review ("rachet") Committee approved the model RETS.
Copies were sent to licensees in July 1978, along with a request to submit proposed site specific RETS on a staggered schedule over a
6 month period.
Licensees responded with requests for clarifications and extensions.
The Atomic Industrial Forum (AIF) formed a Task Force to comment on the model RETS.
HRC Staff members f~rst met with the AIF Task Force on June 27, 1978.
The model RETS were subsequently revised to reflect comments from the AIF and others.
The principal changes were the removal frcm the model RETS of. much material concerning dose calculations and the addi tion of the "off-site dose calculation manual" (ODCM) to contain this material.
The revised model RETS were sent to licensees on November 15 and 16, 1978.
This transmittal p)eo 'included guidance on preparation of 'the RETS and the ODCH (HUREG-0133)', as'well as a
new schedule for responses, again
'taggered over a
6 month period.
Four regional seminars on the RETS were conducted by the staff during the last week of Hovember and the first week of December 1978.
The purpose of these'eminars was to clarify matters for the licensees but the seminars also convinced the staff that there was a
need for further guidance on ODCM preparation, a
need for guidance on the process control program (PCP),
and a
need for further revision 'of the model RETS.
Revisioti 2 of the model RETS and the additional guidance on the ODCH (Appendix B) and the PCP (Appendix C)'ere completed in February
- 1979, and were given to indiv'idual utilities in individual meetings.
Submittal s on the RETS were received frcm all but one licensee by December 1979.
Many of the submittals were 'reviewed by the'staff and individual dis-cussions were held with several licensees.
However, the process was halted in Harch
- 1979, by the accident at Three Mile Island (THI).
Since then the model RETS have been implemented at those plants undergoing operating license review but today none of the reactors'hat were operating before January
- 1979, (except TMI-1) have Tech Specs that are consistent with the model RETS.
~ I E ~ Reasons for Dela in Im lementation Possibly the single mbst important delay was the time required for the HRC staff to establish its position and provide guidance.
Re'v.
2 of the. model RETS was not published until nearly 4 years after the adoption of Appendix I.
Even today, after extensive review and formal approvals, staff members-dis-agree about RETS implementation.
The TNI accident was a second important factor in delaying implementation.
The people who had been working with the RETS were diverted to other tasks after the accident.
- Further, reactions to the accident created such a back-log of work that little staff time has been available for RETS work. Little progress was made with ORs (apart from internal discussions) until mid-1981.
Other important contributors to delay included the following.
1.
2.
3.
4.
5.
6.
" 7.
8.
Lack of explanation of deviations in licensee submittals Inflexible attitude of the staff in initial reviews Licensee. resi stance to,new requirements Staff manpower shortage Licensee manpower shortage Low priority; failure to implement the RETS generally is not seen as a thr.eat to public health and safety Midespread feeling that some model RETS requirements are-excessive for ORs Lack of a
common understanding of the meaning and purpose of RETS.
These contributors are not independent of each other and the relative import-ance of each has not been established.
Nevertheless, it is manifest that these factors must be taken into account if future implementation efforts are to be effective.
Renewed Efforts at Implementation Plans for rene~ed efforts to implement the RETS for ORs were auided by a desire to minimize delays and to limit impact on licensees.
Of course, failure to meet
.the objectives of the model RETS could not be accepted.
These considera-tions led to several key decisions.
First, the'existing model RETS (NUREG-0472 and -0473, Rev.~ 2) are to continue to be the basic guidance.
Developing new guidance would have taken consider-able time and would have needed new clarification.
Second, initial action is to be based on the 1979 submittals from licensees.
Requesting new submittals would have taken time and added to licensee work-load.
Furthermore, new submittal s might not have enhanced understanding or moved us closer to resolution of differences.
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=6-Third, where practicable, a flexible aproach is to be taken in. the review..
Licensee are being asked to meet the intent, rather than the letter, of the model RETS requirements.
This approach should minimize impacts on licensees..
It does, however, seriously complicate the reviewer's job.
Fourth, contractor support is being utilized.
This approach delayed the "
initial reviews but the workload did not permit the staff to handle the OR RETS in a timely manner.
In general, plants in the northeast are being re-viewed by Franklin Research Center while EG8G'Idaho is reviewing the others (Table 1).
Fifth, clarification will be achieved primarily through direct meetings with individual licensees.
Coordination with the AIF Task Force will be continued and papers'ill be presented at technical meetings but direct contacts are given highest priority.
Role of the Contractors The assignments of the two contracts differ only in the.plants they are.to revi ew.
These assi gnments,.for each plant, are as follows.
'eview the existing RETS and the RETS submittal and compare them to current requirements as experssed in the model RETS and in HVREG-0133.
- 2. 'btain the necessary additional information from the FSAR and frcm di scussions wi th licensees.
30 5.
6.
Identify deficiencies in the existing and proposed RETS in the sense of not meeting the intent of the requirements.
Discuss deficiencies with the licensee
- and, where practicable, work out mutually acceptable RETS.
Mhere differences cannot be resolved, recommend appropriate staff action.
Provide a brief written explanation of why the RETS meet (or do not meet) the intent of the guidance.
7.'pdate the relevant SAR on ccmpliance with Appendix I; the update is to ref1'ect recent release
- data, changes in radwaste systems and the RETS revi sions.
Schedule The current tentative schedule calls for the completion of all reviews by October 1982.
Experience is not yet.sufficient to provide a measure of our ability to meet this schedule.
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The plant-by-plant schedule is being developed.
This is complicated by con-.
flicting considerations.
First, it seems desirable to start with plants which seeat rel.atively close to compatibility with the model RETS to provide experience before proceeding to the more difficult plants.
On the other hand, those plants furthest from meeting current criteria merit prompt attention.
Furthermore, the staff is trying to give priority attention to those older plants in the systematic evaluation program (SEP).
Also, all the plants operated by a single lic'ensee should be evaluated at about the same time.
Clearly not all these objectives can'be met, but they'are being considered.
Implications of Flexibility in Review The model RETS and the supporting documents were developed primarily for plants not yet in operation.
The difficulties in meeting new requirements can be much greater for operating reactors.
Even the addition of a new flow meter or radia'tion monitor can be problematic in older plants.
Thus, it becomes ap-propriate to consider alternate approaches on a case by case basis.
Formats other than that of the model RETS are acceptable if appropriate.
Phraseology changes are acceptable if justified.
In particular, there's no objection to the addition of clarifying statements which the licensee feels may obviate problems with "over-zealous" inspectors.
Me are, of cours'e, striving for a common understanding of the'model RETS phrasiology.
Alternate ALARA "design objectives" based on RM 50-2 (rather than Appendix I)
.are acceptable provided that the full set from RM 50-.2 are used.
This offers the advantage
'of meeting the cost-benefit criterion by annual curie limits rather than by operating treatment equipment whenever doses'rom releases in one month are projected to exceed 1/48 the annual dose objective.
In the 00CH and in procedures, design objectives may be expressed in curies
- released, rather than in dose, if it is shown 'that the constraints on releases actually will keep offsite doses below intended levels.
This approach might alleviate some problems in training operators.
The process control program (PCP) for processing solid wastes need not be submitted to the staff for approval prior to. implementation at opera.ing reactors (only).
This should reduce problems in coping with changing burial ground requirements, as well as problems'in using or changing contractor's.
Conversly, there are numerous areas with little room for flexibility.
Specifically, the specified environmental moni.oring program is the minimum acceptable.
Also the lower limits of detecti'on in effluent samples are in-flexible; in particular, analysis for P-32 and Fe-55 is required, at least until it can be shown that these nuclidesare not important contributors to offsite doses.
~
~.Generally, all the requirements of the model RETS must be addressed but some items are more important than others.
The list of "esssential elements" (Table 2) indicates which items are considered most important.
Other deci-sions and considerations are discussed briefly in Appendix A.
Detailed guidance has not been developed for the flexible review process primarily because the radwaste systems and the problems with the RETS differ from plant to plant.
Current Status of Operatin Reactors The general objective of the effluent control program evidently is being met.
That is, radiation doses to the public from nuclear power plants (NPP) are low.
The conservative. calculations summarized in Table 3 indicate that in 1977 NPP effluents increas'ed doses less than 0.008" over natural background.
Subsequent improvement of the offgas treatment system of 4 BMRs 'has reduced the annual population dose since then,. probably from 700 to less than 300 person-rems.
Careful calculation of the doses free liquid effluents from six reactors would further reduce
- doses, probably to about 150 person-rems.
Available data (Tables 4 and
- 5) also suggest that the specific dose design objectives of Appendix I also are generally met.
Conclusions on this point are tentative because less than 'half the plants now report calculated 'off-.
site doses and because the methods for such calculations (OlKH) have not been approved.
One reason for implementing the RETS is to obtain this information-in an orderly manner.
While
. the off-site doses seem acceptably low, one cannot be so sangui ne about some other RETS i ssues.
As indicated in Table 6, almost al.l ORs have Tech Specs that address certain issues (specifically meeting Part 20 limits, monitoring releases and environmental monitoring) but other issues are only occasionally addressed.
For example, only 72 of the ORs have a Tech Spec on the land use census, even though it is explicitly called for by Appendix I.
Only 58 of the pla'nts have Tech Specs addressing a design objective on doses from liquid effluents, even though this is the hear't of Appendix I.
Thus, it is manif st that less than hal f the ORs meet the explicit ALARA requirements for Tech Specs.
Generally, the "non Appendix I" i.ems are less consistently addressed.
At this time it is not clear how many of the Tech Spec shortcomings represent real deficiencies and how many are merely paperwork problems.
It is known that in a number of plants some of the deficienci es are real.
The solid waste management programs generally need improvement.
Some monitoring equip-.
ment, particularly hydrogen-oxygen monitors, is missing.
Several radiochemical analysis programs exclude important nuclides, particularly P-32,and Fe-55.
The initial review process is expected to improve our understanding of the problem.
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TABLE. 1 OPERATING REACTOR. RETS ASSIGNMENTS E GAG 1.
Beaver 'al 1 ey (PA) l.
Arkansas 1
(AK) 2.
Big Rock Pt."*
3.
Brunswick 1-2 4.
Calvert Cliffs 1-2 5;
Cook 1-2 6..
Davis-Besse (MI)
.(HC (MD)
(NI)
(OH) 2.
3.
5.
6.
Arkansas 2
(AK)
Br owns Ferry 1-3 (AL)
Cooper (HB)
Crystal River (FL)
Dresden 1-3*+
{IL) 7.
Fitzpatrick 8.
Gi nna**
9.
Haddam Neck**
(HY)
(HY)
(CH) 7.
Duane Arnold (IO) 8.
Farl.ey 1
(AL) 9.
Ft. Calhoun~
(HB) 10.
Indian Pt.
2 (HY) 10.
Ft. St. Yrain
.(CO) 11.
Indian Pt.
3 12.
Maine Yankee 13.
Nil 1 stone 1**
14.
Nilstone 2
15.
Nine Mile Pt.
(
16.
North Anna 1-2
{HY)
.(ME)
(CH)
(CN)
(NY)
{YA) 11.
~ Hatch 1-2
.(GA) 12.
Kewaunee 13.
La Gros se*"
(MI) 14.
Nonticel l.o (NH) 15.,Point Beach 1-2 (MI) 16.. Prairie Island 1-2* (NN) 17.
Oconee 1-3 18.
Oyster Creek"*
(SC)
(NJ) 17.
18, guad Cities 1-2 (IL)
Rancho Seco (CA) 19.
?ali sades""
20.
Peach Bottom 21.
Pilgrim (NI)
(PA)
{MA) 19.
Robinson (SC) 21.
St. Luci e
{FL) 20; San Onofre 1**
(CA) 22.
Salem 1-2 23.. Surry 1-2.
(HJ)
(YA) 22.
23.
Trojan (OR)
Turkey Pt. 3-4*
(FL) 24.
Vermont Yankee (YT) 24.
Zion*
(IL) 25.
Yankee Rowe**
(NA)
"Plants participating in the source
.erm measurements program.
Systematic evaluation program (SEP) plants.
TABLE 2 ESSEHTIAL ELENEHTS OF THE RETS Technical Specifications must be included which address each of the following requirements 1.
All significant releases shall be m'onitored.*
2.
Off-site concentrations shall not exceed the 10 CFR 20 Table 2 values."
3.
Off-site doses shall be ALARA.*
5.
Equipment shall be maintained and used to keep doses ALARA."
Radwaste tank inventories shall be limited so failure would not cause off-site doses exceeding.the 10 CFR 20 limits.
e.
7.
8.
9.
Maste gas composition shall be controlled to prevent explosive mixtures.
Mastes shall be processed to.burial ground cri teri a under a documented program subject to gA verification.
An environmental monitoring program, including a land use census, shall be implemented.*
I The radwaste management program shall be subject to regular audit and revi ew.
10.
Procedures for control of effluents shall be followed.*
11.
Periodic and special reports on environmental monitoring and on releases shall be submitted.*
I 12.
Off-site dose calculations shall be performed using documented methods that are consistent with HRC methodology.
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TABI E 3
1977* TOTAL BODY POPULATION DOSES FROM NUCLEAR POWER PLANT EFFLUENTS (FROM.NUREG/CR-1498)
Site Population doses, 'rson-rems OL date Gaseous Li used Tota 1.
2 ~
3.
4 ~
5.
6.
7.
8.
9.
10..
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22'3.
24.-
45.
Millstone (lpga, 2C Dresden (GE)~
Subtotals
{
Pilgrim (GE)
Oys ter Creek (gE )
Oconee (BEY)
Ra ch (GE)~e)
Cook (W)
< e)
Zion (W)(e)
Davis.Besse (BEW)
Indian Point (W) (e)
Peach Bottom (GE )
Subtotals LaCrosse
{AC Brunswick (GE )
guad Cities (GE)
Surry (W)
Browns F'erry (GE)
Nine Mile Point (GE)
Big Rock Point (GE)
Haddam Heck (W)
TMI (BaW)
Kewaunee (W)
Calvert Cliffs (CE)
Arkansas (B8W)
Subtotals All Others (47)
TOTALS E) 70,75 60,70, 71 72 69 73,74,74 75
-. 75,78 73.73 77
- 62. 74,76 74,74 69
'75,77 72 $ 72 72 73 74,75,77 69
~
62 68
'74 74 75,77 74 220 (200) 180 {170)
ZKi g7Uj 52
(
7) 41 (110) 1(
I)
-(-)
'(
1) 9
(
7)'(-)
12
(
8) 5(
S)
~%397 1.6 (0.4) 6.3 (1.8) 1.3. (1.3).
1.3 (0.3) 2.7 (1.3) 0.1 (0.1),
0.3 (0.4) 2.2 (4.4) 1.7 (1.6)
( -
)
0.7 (0.7).
O;1 (O.l)
It3ll&7 1.7 (8.6) 540 (530)
-(-)
is)
- (.-)
-(-)
37
( ll) 35(-)
23
( 39) 13
( 16) 14 1
6.4
)
1) 10)
(5. 5) 2XBIKDT.
6.7(13.4) 160 (13O) 7.8 0.1 (0.5) 2.7 (6.0) 2.4 (1.1)
O.S (O.g) 3.O
( -
)
2.3 (2.2)':
0.2 (1.2) 0.3 (0.6) 1.9 (0.6) 1.2 (1.9) 1.5 (2.1) 35
(
23
(
22
(
'4*
13
(
11.4
)
4O) 23)
)
9) 1S) 9.4 6.4 4.0 3.7 3.2 3.1 2.6 2.4 2.0 1.9 1.9 1.6 (5. 9)
(2. 3)
(7.3)-
(1.4)
(2.2).
(0.1)
(2.6)
(5.6)
(2.2)-
(O.6)
(2.6)
(2.2) 8.4
( 22) 7OO (66O) 200 (200) 180 (187),
52
(
7) 41 (110) 38
( 12)
- 12. 3(11. 3) 15.9
( 14)
Mean, person-rems per site 3.6 (2.8)
Exposed Po'p'ulation, with 50 miles Total 92 million Average 2 million Maximum 16 million (Indi an Point)
Minimum 0.1 mil lion (Humbol dt Bay)
Average Dose 7.6 micro-rems (about:0.0076" of natural background) 2 Sites:
400/700
= 57~ (59~)
ll Sites:,650/700
= 93 (91
)
p p
p d
p from David A. Baker (PHL)
Offgas system augmented in 1978; 1979/80 releases were 2..7" of 1977/78 releases.
boresden-1 is shutdown; 1979/80 releases were 3.1 of '1977/78 releases.
""Augmentation now effecive; 1979 releases were 3.5 of 1977 releases.
Offgas system. augmentation no function'ing.
These calculated doses from liquid effluents appear unrealistically high.
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TABLE 4 CURIES OF RADIOACTIYITYRELEASED IH LIQUID EFFLUEHTS*
Plant*"
1977 1978 1979
~ '980 Oconee 1,
2, 3
Ha tch Cook 1, 2
Zion 1, 2
Davis Besse Peach Bottom 2, 3
4.5
- 0. 075 1.6(0.048)**+
0.95(0.003)
- 0. 026 2.2 6.5
- 0. 92 1.5
- 0. 032 1.5(0.7) 1.2
- 0. 036
- 0. 068 2.6(0.27) 1.4(0.004)
- 0. 84(0. 0076)
- 0. 47[0. 0012)
- 0. 21 5.1 19.
1.9
- 0. 090 (0. 082 )
- 0. 043(0. 082)
Dresden 1, 2, 3
(Dresden 1
- 0. 92(0. 018)
- 0. 60,
- 0. 81
- 0. 33
- 0. 26(0.13)
'0
- 0. 48
~ 0 Data rom licensee ef fluent reports.
- Listed in order of descending 1977 population doses.
- Numbers in parentheses are maximum individual total body doses from liquid effluents as reported by the licensees; the "worst" case shown here is Cook
'n 1978 where the dose of'0.7 mrem is a factor of 4 below the 3 mrem design objective.
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~h TABLE 5 KILO CURIES OF RADIOACTIYITY R LEASED IH GASEOUS EFFLUEHTS*
Plant lail l stone 1
Millstone 2
1977 620'.4 1978 570 0.36 (0.14) 1979 20'.
24 1980 12 (0. 24)~
1.3 (0.1)
Dresden 1, 2, 3
830 (7.4) 1330 31 (0.18) 36 (0.49)
(Dresden 1)
Pilgrim Oyster Creek Oconee 1, 2, 3
Cook 1,
2 Zion 1, 2
Davi s Besse 520 410 (5.3) 180 12 3.9 {0.07) 32 {1.0)1.3'50 33 (1.9) 1000 43 49 (1.4) 50 {.44) 2.0 14 (0.73) 1000 11 (0.04) 24 {0.17) 1.7 (0.04) 1.4 31 19 3..8 5.8 {0.007) 1,7 ata rom icensee effluent reports.
- Humbers in parentheses are maximum individual total body doses fran gaseous effluents as reported by the licensees.
The "worst" case shown here is Dresden in 1977 where the dose was 7;4 millirem.
This would be a factor of 2 below the design objective of 5 mrem per reactor if releases from the 3 units were equal.
Releases from Unit 1 may have produced some-4.6 mrem.
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TABLE 6 EXTENT TO WHICH ORs ADDRESS LCOs OF THE MODEL RETS>>
3.3.3.9 3,3.3.10 3.11.1.1 Limitin Conditions for 0 eration Liquid Effluent Monitor Gaseous Effluent Monitor Liquid Concentration Limit (Part 20) 3.11.1.'2
~
Liquid Dose Design Objective
- 3. 11. 1. 3
- 3. 11. 1..4 Liquid Waste Treatment Li.qui d Ta nk Cur ie. Limit 3.11.2.2
- 3. 11.2. 3 3.11.2.4 Noble Gas Dose Design Objective e
Iodine E Particle Design Objective Gas Waste Treatment 3.11.2.1 Gas Dose Rate Limit (Part 20) 40 29 60 32 56 Yes~
No Pet.
Yes 67 5
93 47 8
89 71 72 99 42 30 58 63 9
88 48 24 67
'2 72 100 48 24 67 3.11.2.5
- 3. 11. 2. 6 Gas Expl osive Hixture Gas Tank Curie Limit 11 '1.
15 43 29 60
- 3. 11. 3 3.11.4 3.12.1
- 3. 12:2.
Environmental Monitoring Land Use Census Interl ab.
Compari son 52 20 1
71 13 10 62 2
70 70 54 97 72 760 ~'488 Liquid Curie Design Objective'(RM 50-2) 45 27 61 63 Gaseous Curie Design Objective (RH 50-2) 28 44 39 "Based on an unpublished review by J. J.
Hayes (NRC).
<<"The existing Tech Specs have LCOs that roughly correspond to the LCOs of the model RETS; cc-...pli ance wi th the model PETS i s n'ot i...plied.
s
I e0'
14
~
~
Bibliogra A.
Regulations 10 CFR 50, Code of Federal Re ulations.
Titie 10, "Energy".
Chapter I, Huc ear Regu atory Commsssion".
Part 50, "Domestic Licensing of Production and Utilization Facilities".
Government Printing Office, Mashington, D.C.,
1981 (pp. 342-447).
2.
3.
10 CFR 50, Appendix I, "Humerical Guides for Design Objectives and Limiting Conditions for Operation to Neet the Criterion, "As Low As Is Reasonably Achievable'.for Radioactive Naterial in Light-Mater-Cooled Nuclear Power Reactor Effluents," Government Printing Office, Washington, D.C.,
(pp. 419-423).
10 CFR 20, "Standards'or Protection Against Radiation,"
Government Printing Office, Washington, D.C.,
(pp. 200-231).
B.
Topical Reports 4.
5.
6.
7.
8.
"Calculation of Releases of Radiaoctive Haterials in Gaseous and Liquid Effluents From Boiling Water Reactors (BWR-GALE Code),"
HUREG-0016 (April 1976).
"Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents From Pressurized Water Reactors
{PMR-GALE Code),"
HUREG-0017 (April, 1976).
J.
S. Boegli, et al, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power'Plants,"
HUREG-0133 (October 1978).
"XOQDOQ, Program for Neteorol.ogical Evaluation of Routine-Releases at Huclear Power Stations,"
HUREG-0324
{Septemb'er 1977).
F.
P. Cardile, et al, "Cost-Benefit Analysis Requirements of Appendix I to 10'CFR,Part 50; Their Application to Certain Huclear Power Plants Docketed Before January 2,
1971,"
HUREG-0389(January 1979)..
9..R.
Lo, et al, "Technical Report on Operating Experience With Boiling Water Reactor Offgas Systems,"
HUREG-0442 (April 1978).
10.
"Radiological Effluent Technical Specifications for PMRs,"
HUREG-0472, Rev.
2 (February 1, 1980).
11.
"Radiological Effluent'echnical Specifications for BMRs,"
HUREG-0473, Rev.
2 (February 1,.1980).
12.
E.
F. Conti, et al, "Radiological. Environmental Monitoring by HRC Licensees for Routine Operation of Huclear Facilities,".HUREG-0475 (October 1978).
0
~I
~
~
4 13.
F. J.
- Congel, "Methods for Demonstrating LMR Compliance with the EPA UraniumFuel Cycle Standard (40 CFR 190)," NUREG-0543 (January 1980).
1'4.
"Standard Review Plan for the Review of Safety An alysis Reports'or Nuclear Power Plants,"
NUREG-0800 (Formerly NUREG-75/087),
(1981).
C. 'egulatory Gui des 15.."Reporting Operating Information," Regulatory Guide 1.16, Rev.
4 (August 1975).
16.
17.
"Measuring, Evaluating and Reporting Radioactivity in Solid Mastes and Releases of Radiaoctive Materials in Liquid and Gaseous Effluents Fran Light'-Mater-Cooled Nu'clear Power Plants,"
Regulatory Guide 1.21 (June 1974).
"Calculation of Annual Doses to Man From Routine Releases
'of. Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Regualtory Guide 1.109, Rev.
1 (October 1977).
18.
"Cost-Benefit Analysis for Radwaste Systems for Light-Mater-Cooled Nuclear Power Reactors,"
Regulatory Guide 1.110 (March 1976).
19.
"Methods of Estimating Atmospheric Transport and Dispersion of Gaseous Effluents n Routine Releases from Light-Mater-Cooled Reactors,"
Regulatory Guide 1.111, Rev.
1 (July 1977)..
20.
"Calculation of Releases of Radioactive Materials in Gaseous and' Liquid Effluents from Light-Mater-Cooled Power Reactors,"
Regulatory Guide '1.112 (March 1976).
'21.
"Estimating Acquatic Di spersion of Effluents from Accidental and Routine Releases for the Purpose of Implementing Appendix I," Regulatory Guide 1.113, Rev.
1 (April 1977).
D.
Papers 22.
Collins, John T.,
"NRC Model Radiological Effluent Technical Specifications,"
Atomic Industrial Forum Conference (April 1979).
23.
Millis, Charles A., and Hayes, John J., Jr., "Radiological Effluent Technical Specificatjons:
Where We'e
- Been, Where We Are, and Where Me're Going," Health PHysics Society-Annual Meeting.(June 1981).
~
I s
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~
Appendix A OBJECTIONS TO THE HODEL RETS RAISED BY THE AIF AND OTHERS>>
General
(~ Zap~ (=y 1.
RETS are too voluminous.
"-C'- Z "-~ '/~
The model RSTS constitute about ~ of the Standard Tech Specs (STS).
This may not be excessive considering the public sensitivity to the issues.
Also the staff is considering dividing the Tech Specs to emphasize those issues of immediate importance to safety.
With this system the principal problem wi th volume will be eliminated.
2.
Clarification is needed.
3.
fs Considerable suppqrting documentation has been provided {see bibliography) as well as papers at technical society meetings, regular discussions with the AIF working group, and regional seminars.
This clarification effort is expected to continue but in the immediate future the staff plans to concentrate on meetings with individual licensees.
This affords the ad-vantages of dealing wi th the issues that are causing problems and of obtaining maximum feedback.
RETS should rely on existing equipment (no backfit).
I To the extent pr acticable, this is the staff.'s intent.
In. some instances,.
however, it is not clear how a licensee can show that the intent of a model RETS requirement is met in the absence of keg instrumentation or other equipment.
4..
The RETS go beyond the regulations..
I While the RFTS contain many "non-Appendix I" requirements, a regulatory basi s is given for each LCO.
Even the much-maligned PCP has a regula-tory basi s in 50.34a, 50.36a, and GDC-60.
The "surveillance" and "action" requirements generally are not addressed in the regulations, either for radwaste systems or for other parts of the plants, but they are important parts of the program.
4 5.
Tie THI "Action item" requirements are not included.
True; the "lessons learned" Tech Specs will be handled separately.
6.
STS format should not be required.
It is not required.
"Comments are part y based on personnal communications from J.
S. Boegli (NRC).
0
0 a
7.
Tech Specs based on dose necessitate operator training.
It is acceptable to express the Tech Specs in curies if the ODCN shows that these constraints are equivalent to, or more conservative 'than, the standard dose criteria.
This would not eliminate the need for calculating and reporting doses (Reg.
Guide 1.21).
Definitions 1,
Some standard definitions are inappropriate for older plants.
The problems with '"channel check", "calibration"., etc.
are recognized
, and the staff is prepared to accept other appropriate definitions on a case by'ase basis.
2.'any terms used in the model RETS need definitions..
Host of the important definitions have been presided though they are not all in one place.
Specifically, for a definition of "restricted area" see Part 20, for "exclusion area" see Part 100, for "continuous composite'ample".
see STS 1.6, etc.
Where a licensee sees potential problems in ambiguity, or other problems with definitions, the staff may accept r'easonable definitions pr oposed by an individual 'licensee.
One clarification here seems in. order.. The expression "on-site general public" is limited to those situations where' portion of the site is or may be occupied by members of the public for extended periods for recreational,
'occupational-or other purposes.
This does not include the visitor center or non-employees who occasionally come on site,to service Yendi ng machines, etc; Instrumentation 1.
Tech 'Specs should cover only the final release points.
.This contention seems based on the ccmmon misconceptioh that the RETS are limited to ensuring compliance wi th Appendix I.
In fact, the RETS are also intended to meet other requi rements including General Design Criterion (GDC) 61 which requires the control of radioactivity and of radioactive wastes and GDC 63 which requires monitoring.,of waste storage.
The need for process instrumentation is inescapable.
Deviations free the model RETS are acceptable where appropriately justified by the licensee.
.2.
Inoperable hydrogen and oxygen monitors should not require shutdown.-
~I
3.
Appropriate alternatives to the model RETS "actions" can be accepted.
In this case the "action" must provide incentive for timely repair of the.monitors and for compliance with GDC-3 (Fire Protection);
PWR turbine building sump often is difficult to monitor and generally should not be isolated.
4 ~
5.
The problems are recognized and plant-specific alternative means of monitoring and controlling these effluents can be accepted.
Pump curves should be adequate for flow determinatiqns.
The staff agrees where conditions are right.
Flow rate is irrelevant for some off-line radiaiion monitors.
The RETS requirement is for alarm on.loss. of flow, not for flow rate measurement.
Effluents 2.
P-32 analysis should not be required.
This analysis is needed because P-32 can be an important contributor" to dose due to 'its high b'ioconcentration factor.
This may be changed by.
HRC and AIF studies now.underway but the requirement stands until change is technically justified.
Ho crosscheck is available for P-32'or Fe-55.
The statement is true but it does not affect the need for the analysis of effluents for these.nuclides.
3.
Actions'on otal dose should be on a calendar year basis.
4 ~
This LCO is based on the EPA'regulation (40 CFR 190) so it follows the EPA time basis.
This Tech Spec should cause few practical problems because the staff has concluded that compliance with Appendix I provides reasonable assurance of compliance with 40 CFR 190 (see HUPEG-0543).
Why use 1/4 the dose objectives as a trigger for equipment use?
The staff has agreed to accept this trigger level as satisfying the cost-benefit criterion of Appendix I.
Two alternatives are available to licensees.
One option is to use the Design Objectives of RH 50-2 which i'nclude annual curie release criteria.
The other option is to provide a cost-benefit analysis justi ying some other level for triggering the use of the equipment.
The "trigger levels" may vary but the regulations
-explicitly require a Tech Spec on the maintenance and use of effluent
. treatment equipment.
l 4
4 5.
Mhy require liquid level monitors on outdoor tanks2 The primary purpose is,to detect leaks that might otherwise go unnoticed.
6.
How can curie limits be set for tanks of unknown size?
7.
Methods are described in HUREG-0133 and SRP 15.7.3.
General'ly the tank volume does not determine the limit.
The 'staff has accepted 10 curies as a default value in cases where substantial dilution ~ould occur before a spill could reach a potable water supply.
Mhy sample gaseous wastes daily following a power change if there is no ac'tivity "spike" in the reactor coolant2 This provision reflects the uncertainty about the "iodine spiking" phenomenon.
iodine concentrations have been observed reaching maximum values as much as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the transient.
Other "spikes" have persisted over several days and in some cases the concentrations have varied with time over a period of days.
Also, under some circumsunces the activity increase may be most readily observed in the waste gas system.
Mhere this requirement poses problems it may be practical to
'ustify other controls.
l'.
Sampling waste gas tanks should be minimized.
Sampling is not required.
The staff will accept the results of a gross gamma monitor calibrated to curies Xe-133 equivalent.
S..Hydrogen and oxygen monitors are unreliable and hard to calibrate'..
Redundant monitors provide some relief but this continues to be a
problem.
10.
Some PMR waste gas systems.(MGS) are designed for 3 hydrogen or oxygen.
Generally, BMR MGS can withstand an explosion while PMR MGS cannot.
SRP1.3 sets the limit at 2
for systems not designed to withstand an explosion.
The staff, however, has accepted deviations where justified on a case by case basis.
11.
Concurrent meteorology dose calcualtions are costly..
co Appropriate and conservative approximate methods are acceptable.
)
~ g
~
~ Offsite Dose Calculation Hanual (ODCH) l.
A.simplified ODCH should be accepted.
Simplicity is no barrier to acceptance so long as the methods are conservative and clearly explained.
2; The ODCH should contain only methodology.
Certain parameter values are also
- needed, as is the flow diagram.
In
- essence, we need enough information to permit a review of the substance of the document.
This document is the basis for accepting licensees' calculated offsite dose values so it should contain enough information to justify confidence.
3.
Hethods for 'determining setpoints shoul d suffice.
Specifics are needed for inspectability.
Fixed setpoints shoul d be given and justified.
For variable setpoints, the methodology should be suffi-ciently detailed and the parameter values so identified that (1) the reviewer can verify the adequacy of the approach and (2) an inspector can check actual setpoint calculations.
Process C'ontrol Program (PCP)'.
The "philosophy" of the PCP should be expounded.,
'he need for improved control of waste processing has been repeatedly demonstrated and the HRC has committed to making improvements.
The PCP approach was selected because it promised both reasonable assur-ance'of adequate protection and the flexibilityto accommodate diverse systems.
The essence of the PCP is the requirement to identify the important parameters in the process and specify their acceptable ranges (how much oil?; what temperature?;
etc.).
Other important aspects're committments to testing,
()A and compliance with procedures.
Thus, the PCP 'should provide a sound basis for the licensee's qantrol of his own activities, as well as a basis for inspection.
2.
The PCP is not needed because
'the burial grounds and the shipping r egulations establish the requirements.
The purpose of the PCP is preventative not punitive; the principal
.cern is proper impl.ementation not requirements.
con-'
I
~ ~
~l
)
- 6 3.
The PCP should not be required; plant procedures should be sufficient.
~
The PCP addresses issues not normally covered in procedures; viz. the the b'ackground bases and justi'fication for the. procedures.
The-PCP actually provides a basis for the review of procedures.
Mithout a PCP there is no assurance that the procedures even address the important
~ parameters.
4.,
There should be no need for the PCP to address dewatering of resins.
Unfortunately, the potential for deficiencies in this simple process has been demonstrated so dewatering cannot be ignored.
However, the poten-tial for wasted effort.in this area is small.
1f one khows what needs to be gone writing this part of the PCP is easy and if one does not know, the learning effort is justified.
~ 5.
Why require seismicly qualified pads for portable solidification equipment?
The staff has not required the application of Regulatory Guide 1.1'43 to such pads.
40>CFR 190 I+
p~iS I
1.
Some plants have a problem with direct radiation from stored solid wastes.
This is not a
RETS problem.
Compliance with 40 CFR 190 is requi red, whether or not it is specified in the Tech Specs.
I lj
APPENDIX 8 GENERAI CONTENTS OF THE OFFSITE DOSE CALCULATION MANUAL (ODCM*)
(Rev.
1, February 1979)
Section 1 - Set Points Provide the equations and methodology to be used at the station or unit for
~ each alarm and trip. set point on each effluent release point according to the Specifications 3.3.3.8 and 3.3.3.9.
The instrumentation for each alarm and trip set point, including radiation monitoring and sampling systems and effluent control features, should be identified by reference to the FSAR (or Final Hazard Summary).
This information should be.consistent with the recommendations of Section I of Standard Review Plan 11.5, NUREG-75/087, (Revision 1).
If the alarm and/or trip set point value is variable, provide the equation to determine the set point value to be used, based on actual release conditions, that will assure that the Specification is met at each release point; and provide the value to be used when releases are not in progress.
If dilution or dispersion is used, state the onsite equipment and measurement method used during release, the site related parameters and the set points used to assure that the Specification is met at each release..
point.
The fixed and variable set, points should consider the radioactive effluent to have a radionuclide distribution represented by normal and anticipated operational occurrences.
/a Section 2 - Li uid Effluent Concentration Provide the equations and methodology to be used at the station or un'it for each liquid release point according to the Specification 3..11.1.1.
For.
systems with continuous or batch releases, and for systems designed to moni.or and control both continuous and batch releases, provide the assump-tions and parameters to be used to compare the output of the monitor with
.the liquid concentration specified.
State the limitations for combined discharges to the same release point.
In addition, describe the method and assumptions for obtaining representative samples free each batch and use of previous post-releas~ a'naly~ or composiM sample analyses to meet the Specification; Section 3 - Gaseous Effluent Dose Rate Provide the equations and methodology to be used at the station or unit for each gaseous release point accordin'g to'Specification 3.1'1.2.1.
Consider the various
- pathways, release point elevations, site related parameters and radionuclide contribution to the dose impact limitation.
Provide'the "The format or the ODCM is left up to the licensee and may be simplified by tab'les and grid printout.
Each page should be numbered and indicate the facility approval and effective date.
~ I pO I
dose factors to be used for the identified radionuclides released.
Provide the annual average dispersion values (X/Q and D/Q), the site specific para-meters and release point elevations.
Sec ion 4 - Li uid Effluent Dose Provide the equations and methodology to be used at the station or unit for each liquid release point according to the dose objectives given in Speci-fication 3.11.1.2.
The section should describe how the dose contributions are to be calculated for the various pathways and release
- points, the equa-tions and assumptions to be used, the site specific parameters to be measured
'nd used, the receptor location by direction and.distance, and the method.'of
.estimating and updating cumulative doses due to liquid releases..
The dose
- factors, p'athway transfer factors, pathway usage factors, and dilution fac-tors for. the points of pathway origin, etc.,
should be given, as well as receptor age group, water and food consumption rate and other factors'ssumed or measured.
Provide the method of determining the dilution factor at the discharge during any liquid effluent release and any site specific parameters used in these determinations.
Section
.5 - Gaseous Effluent Dose Provide the equations and methodology to be used at the station or unit for each gaseous release point according 'to the dose objectives given in Specifications 3.11.2.2 and.3.11.2.3..
The section should describe how the dose contributions are to be calculated for the various pathways and.release
- points, the equations and assumptions to be used, the site specific parameters to be measured and used, the receptor location by direction and dist'ance, and the method to be used for estimating and updating cumulative doses due to gaseous releases.
The location, direction and distance to the nearest resi-
- dence, cow, goat, meat animal, g'arden, etc.,
should be given, as well as
- receptor, age group, crop yield, grazing time and other factors assumed or measured.
Provide the method of determining di-spersion values (X/Q and D/Q) for releases and any site specific parameters and release point elevations used in these determinations.
1 Sec tio n.6. -
P rojected Doses For liquid and gaseous radwaste treatment
- systems, provide the method of project'ing doses due to effluent releases for the normal and alternate pathways of treatment according to the specifications, describing the com-ponents and subsystems to be used.
~ I g
L'
~ ~
3 Section 7 - 0 erabilit of E ui ment Provide a flow diagram(s) defining the trea~~ent paths and the components of the radioactive liquid, gaseous and solid was.e management systems that are
'.o be maintain d and used, pursuant to 10 CFR 50.36a, to meet Technical Specifications 3.11.1.3, 3.11.2.4 and 3.11.3.1.
Subcomponents of packaged
..quipment can be identified by a list.
For operating reactors whose con-struction permit applications were filed prior to January 2, 1971, the flow diagram(s) shall be consistent with the information provided in conformance with Section V.B.1 of Appendix I to 10 CFR Part 50.
For OL applications whose construction permits were filed after January 2,
1971, the.flow diagram(s) shall be consistent with the information provided in Chapter 11 of the Final Safety Analysis Report (FSAR) or. amendments thereto.
Section 8 - Sample Locations Provide a map of the Radiological Environmental Monitoring ample Locations indicating the numbered sampling locations given in Table 3.12-1.. urther clarification on these numbered sampling locations can be provided by a list, indicating the direction and distance from the center of the buildi~g com-plex of the unit or station, and may include a discriptive name for identi-fication purposes.
J P
I
APPENDIX C
SOLID WASTE MANAGEMENT SYSTEMS
{Rev. 1, February 1979)
Standard Review Plan 11.4 and Branch Technical Position'TSB 11-3 re-(1)
(2) quire that each applicant for an operating license provide a detail descrip-tion of a Process Control Program
{PCP) to assure that the solid waste system will perform its intended function and that the product produced by this system contains no fre'e water" and i s a monolithic sol.id.
Specificytion 3.11.3.1 of the model Radiological Effluent Technical Specifi-cations{'~) require that the solid ragwyste system be maintained and used in accordance with the PCP.
NUREG-0133<4>
requires that at the time an applicant/
licensee submits proposed Radiological Effluent Technical Specifications that he submit the PCP for NRC review.**
NUREG-0133 futher requires that the PCP be documented in the plant'operating procedures.
To meet this commitment, the staff has prepared a general description of a PCP giving the essential points that should be covered by the applicant/
licensee in making this submittal.
Due to variations in system design and operation, the applicant/licensee should not interpret 'this outline to be all inclusive.
The PCP is plant specific and mist be established on a case-by-case basis since waste characteristics will vary from plant to plant.
PROCESS CONTROL PROGRAM A "Process Control Program" (PCP) for a solid radwaste system shall'be a
manual detailing the program of sampling analysis and'ormulation determina-tion by which solidification of radiaoctive wastes from liquid systems is assured.
The PCP shall provide assurance that the system is operated. as designed and produces a final product that contains no free water and has completely solidified all waste.
If properties of the final product have been determined by the manufacturer, the PCP shall also assure tha the soli-dified'waste products exhibit those physical and chemical properties (leach-ability, strength, flammability, etc.) that are characteristic of the product as demonstrated by the manufacturer for producing, an acceptable solidified waste product.
The PCP shall identify interfaces with other plant systems (e.g., liquid and gaseous radwsate systems),
identify equipment (interlocks,
- alarms, monitors, etc.) which are required to be functional before processing can
- commenCe, identify administrative cqntrols'r equipment features to.
.~ ree water is efined as uncombined water not bound by the solid matrix.
"Current (1981) position is that PCPs for operating reactors need not be submitted, for review prior to implementation.
4 I
assure that operating procedures will be followed, identify the sampling requirements prior to processing and identify the various processing steps and process parametrs which provide boundary conditions within which the solid radwaste system shall be operated.
Depending upon the type of waste (bead resins, powered resins, filter sludge, evaporator concentrates, sodium sulfate solutions, boric acid solutions, etc.) to be solidified and the"kind
.of solidification. agent (urea formaldehyde,
- cement, cement with sodium sil'icate, asphalt, polyester, etc.)
- employed, the process parameters shall include but are not limited to the type of waste, requirements for sampling prior to processing, pH, oil content, water content, tempera'ture, ratio of solidification agent to influent waste and the ratio of solidification agent to chemical a ddi tive.
NOTE:
For operating reactors which have systems instal led that are not capable, of solidifying the categories of "wet" ~aste as defined in SRP 11.4, BTP-ETSB 11-3 or HUREG-0133, the licensee shall define the limitations. of his present system and provide a Process Con rol Program to cover the waste that. can be.
processed by his existing system.
The licensee shall identify those wastes which cannot be solidified and indicate, the method of packaging currently being employed (dewatered resins, vermiculite, etc.).
In addition, the licensee shall provide a schedule for upgrading his solid waste system to provide the capability to process all types of "wet" wastes as defined in these reference documents.*
"for app >cation to operating reactors, delete this sentence.
- However, in anticipation of the promulgation oi 10 CFR Part 61, "Licensing Require-merits for the Land Disposal of Radioactive Mastes," efforts should be initiated by licensees to modify their equipment, procedures
.and PCP such that they will be in compliance with the new requirements for waste form and packaging when Part 61 goes 'into effect.
1
REFERENCES (1)
Standard Review Plan 11.4, Revision 1, Solid Waste Management
- Systems, HUREG-75/087.
('2)
Branch Technical Position - ETSB 11-3, Revision 1, "Design. Guidance for Solid Radioactive Was e Management Systems Installed in Light-Mater-..
Cooled Nuclear Power Reactor Plants,"
HUREG-75/087.
(3)
Draft Radiological Effluent Technical Specifications for PWRs and
- BWRs, HUREGs 0472 and 0473.
(4)
Preparation of Radiological Effluent Technical Specifications for.Nuclear Power Plants, NUREG-0133.
CONTROL OF EXPLOSIYE* GAS NIXTURES IH MASTE GAS SYSTEMS
~R The model radiological effluent technical specifications (RETS) include requirements for control of explosive gas mixtures in waste gas system (MGS) for both BMRs and PWRs.
The RETS requirements basically are for the normal precautions which are included in industrial systems when there is a potential for hydrogen and oxygen mixtures.
That is, the concentrations must be monitored and action mulct be taken to reduce the c'oncentrations when they become too high.
Controvers Even though the explosive gas mixture control requi rements have been approved i n both the RETS and the standard review plan (SRP),
some staff members and others continue to argue tha't these requirements should not be implemented.
The principal arguments against implementation are the following:
- l. '.This is, not an "Appendix I" issue and so should not be'ncluded 5n the RETS.
2.
Hydrogen and oxyg'en montiors are unreliable so the RETS requirements are onerous.
3.
The consequences of a waste gas explosion are too small to warrant attention in th'e Tech Specs.
The first argument is invalid because the RETS are the Tech Specs on the rad-
~aste and effluent tr eatment systems and on.environmental monitoring.
The RETS have never been limited to "Appendix I", issues.
"The term 'xp osion" is used here.to mean "detonation and/or deflagration".
)
l
The second argument is a cause for concern.
The unreliability of hydrogen and oxygen monitors is recognized.
In fact, this unreliablity is reflected in the requirement for dual and redundant monitors.
The potential impact is in having to take and analyze gas samples every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and in having to repair the monitor quickly, within 2 weeks in some cases.
Mhile this impact is limited, the staff is being flexible in its revi.ew in order
~o minimize the potential'mpact on operations of instrument unreliability.'.
The third argument, concerning the limited consequence of an offgas explosion, is discussed in the following section.
I
'General Design Criteria (GDC) 60, 61 and 64 require the control'f releases, the control of radioactivity and the monitoring of releases.
The failure to
'p'revent an offgas explosion is a violation of all three GDC.
On the other hand, an offgas explosion has little chance of causing a release that would'exceed the Part 100 dose criteria (but the same can be said of a core melt accident).
Furthermore, an offgas explosion probably will not prevent shutdown, interfere with core cooling, or violate containment integrity.
The possible effects on safety systems appear to be plant and accident specific.
Actually,'he possible arfd the probable consequences of an offgas explosion have not yet been delineated.
From the viewpoint of offgas explosions, BMRs are very different from PWRs.
An offgas explosion in a
BMR inherently has a high probabli ty of occurrence
3 but has a relatively low consequence potential.
For PWRs the situation is reversed.'hat is, the probability of a PWR offgas explosion is inherently low but the potential consequences are relatively high.
In BWRs, hydrogen and oxygen are continuously produced in stoichiometric ratio and in considerable quantity (about 70 SCFM/GWt).
Mithout controls, only an ignition source is needed to produce an explosion.
In the '70s, offgas explo-sions because endimic to BMRs.
The HRC study (Ref.
- 1) reports 29 incidents in 13 plants.
BMR offgas explosions are expected to be of little consequence'ecause continuous release precludes buildup of either explosive gas or radio-t active material.
This was indeed the case in the explosions of the seventies.
Workers were injured but luckily not killed.
A small building was blown up.
Rooms and pieces of equipment were damaged.
Power production was reduced and even interrupted.
Releases of radioactivity were slightly increased but offsite doses were not significantly increased..The consequences were far from catastrophic but they were serious enough to justify corrective action.
The'long succession of BMR offgas explosions seems to have been ended by new requirements a'nd correction of MGS augmentation problems.
One
- augment, the recombiner, reduces the probability of an explosion but another
- augment, the charcoal adsorber
- beds, increases the. possible con'sequences by collectng.the radioactivity (see Figures 1
& 2).
In PWRs, on the other ha'nd, the hydrogen is not the result of radiolysis but is deliberately added to the primary system to control oxygen.
The. hydrogen
J fy qt
-.4-is transferred to. the MGS* through letdown and degassing.
This hydrogen is in Small quantities (perhaps 20 SCF/day during normal operation) and is free of oxygen.
An explosion is not possible unless oxygen enters the system and usually oxygen is present only as a result of a failure such as a leak in a compressor,.
- Thus, the probability of a PMR offgas explosion is inherently low.
In PMRs, the waste gases are'collected in tanks and these tanks present the potential for explosions more serious than those in BMR MGS.
Without con-
A 1000 ft. tank't 100 psia could contain enough hydrogen and air. to release 3
the'same amount of energy as 160 lbs of TNT.
A MGS explosion would not.actually be the equivalent of 160 lbs of TNT, but the possible magnitude of such an explosion has not yet been established.
A PMR MGS explosion. does appear to have the potential for causing severe
- damage, including the rupture of the other.
MGS tanks.
The failure of all MGS tanks could result in the release of up to 67,000 Ci Xe-133 (for a 3.5 GMt plant with 1
failed fuel).
Thus, the possible consequences
'of MGS explosions.seem serious enough to wa r rant Tech Spec contr ol s.
"See
> gures, and 5.
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'he San Onofre Defla ration
~
On July 17,
- 1981, there was a deflag'r ation in a ~aste gas decay tank at San Onofre-1.
The incident has generated international interest and may be reported to Congress as an "abnormal occurence".
The incident is of particular interest in the present context because it is.
the only instance we know. about of an MGS explosion at a
PMR.
There have been "close calls" aplenty, where concentrations reached dangerous levels, but in the other cases, careful releases to the atmosphere were effected before things blew up.
These near-incidents span the period frcm the present (wi th Arkansas-One) back to at least 1970 {at Ginna) but generally they were not reported and are not documented.
Th'e inci dent at San Onofre {as described in Ref. 2-4) started with a "TNI Aetio'n item" requirement to improve the reliability of the instrument air system.
The instrument air was tied into the nitrogen system.
Unfortunately, the check valve between the two systems was impr operly installed, so inadvertently air, rather than nitrogen, was being used to dilute waste gas.
The San Onofre MGS normally operates in a hydrogen rich mode so the introduction of air quickly produced an unsafe situation.
Mhen installed, the MGS included an oxygen monitor to warn of such conditions, but,'n the absence of HRC require-
- ments, the monitor had fallen into disuse.'he MGS included a recombiner designed to remove small quantities of oxygen.
The high concentrations of oxygen caused the recombiner to over heat and.this is believed to be the
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ignition source.
The deflagration that occurred caused relatively..little J
damage and did'not propagate to the seven other tanks then containing ccm-bustible gas mixtures.
The consequences were slight for several reasons that appear
.to be simple matters of good fortune.
2 ~
Ho one was close enough to the tank to be injured by the blast.
The. e~xlosion did not propagate to the other tanks.
3.
The release was small because the gas had undergone sufficient decay and was ready to be released; also fuel had been good.
4.
There was no effect on operations because the plant was shut The down at the time of the explosion.
I incident does illustrate several points that are of interest here.
A MGS explosion need not have dire consequences (Murphy's law
'does not always hold).
2.
3p A MGS.explosion at a
PMR is.not so highly improbable as has been x
suggested.
Safety features that are not required tend not to b'e m'aintained.
4..The model RETS requirements (if implemented) are appropriate for prevention of such incidents.
Over all, the San, Onofre deflagration suggests that the model RETS provisions for the control of explosive gas mixtures are needed.
JJ
1 1
l I
~FI I
d Our understanding of MGS explosions s'eemed inadequate, so a co I ~
1' V
ntractor (EGSG; Idaho) has been asked to study the problem and to try to provide definitive answers to our questions.
Specifically the contractor was asked to do the following:
Evaluate the problem of potential flammable or explosive gas mixtures in nuclear power plant waste gas systems.
- Collect, summarize and evaluate existing data.of flammable and explosive concentrations of hydrogen, oxygen, nitrogen and other relevant gases.
Evaluate the re1iability and. dependability of commer-cially available instruments for montioring hydrogen and oxygen concentrations.
Evaluate the risks (probability an'd consequences)
I of gas explosions in reactor effluent treatment systems.
Evaluate and recommend appropriate control measures.
Document. this evaluation in a form suitable for publication as a
NUREG report.
This effort is just starting but the contractor already seems to have found a
reliable hydrogen monitor that coul d relieve some problems.
Conclusions It i s concluded that MGS explosions can be controlled, that they should be controlled and that the basic provisions of the model RETS constitute an appro-priate approach.
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It is further concluded that some details of the model RETS requirements may be unduly. onerous, especially for operating reactors and appropriate modifica-tions should be accepted where justified on a case by case basis.
References 1.
Lo, R., et al., "Technical Reports on Operating Experience Mith Boili.ng Mater Reactoi Offgas Systems",
NUREG-0442 (A'pril 1978).
2.,
Haynes, J.
G.
(SCE), Letter to Nuclear Regulatory Commission, July 31, 1981.
3.
- Ornstein, H. L., "Ignition of Gaseous Maste Decay Tank at San Onofre 1--
July 17, 1981,"
Memorandum for Carlyle Michelson, August 6, 1981.
4.
IE Information Notice No. 81-27, "Flammable Gas Mixtures in the Maste Gas Decay Tanks in PWR Plants,"
September 3,
1981.
STACK RELEASE MAINSTEAM CONDENSER OFFGAS STEAM JET AIR EJECTOR 30.MINUTE DELAYPIPE HEPA FILTER DRAIN SUMP LOOP SEAL DRAIN SUMP LOOP SEAL Figure I Schematic of a Typical BWR Offgas System
STACK
'ELEASE MAINSTEAM CONDENSER DELAYPIPE STEAM JET PREHEATER &
CONDENSER AIR EJECTOR HYDROGEN-OXYGEN R ECOMBINER CHARCOAL ADSORBER BEDS
.. LOOP SEAL i
Figure g, Schematic of a Typical BINR Augmented Offgas System
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